15.69. In the safe integral reactor (SIR), the entire PWR primary sys­tem, including the steam generator, is enclosed within the steel reactor vessel. Thus, we have a unit steam generating package with no possibility of a LOCA. In a sense, the characteristics of a PWR are combined with some BWR technology. Passive safety systems are included, such as safety injection accumulators and blowdown pressure suppression tanks. The ref­erence SIR design has an electrical generating capacity of 320 MW, al­though an extended SIR of 400 MW(el) is an option. Construction of a lead unit at Winfrith in the United Kingdom has been proposed.

9.23. Three general mechanisms are distinguished whereby heat is trans­ferred from one point to another, namely: (1) conduction, (2) convection, and (3) radiation. These three methods for heat removal will be reviewed and their general characteristics outlined. The subsequent discussion, how­ever, will be devoted to the first two, since they are of major importance in existing reactor designs. The preliminary treatment will refer to heat transmission in systems without internal sources, and subsequently the problems associated with internal sources, such as exist in reactor fuel elements, will be taken up.

9.132. There is often confusion regarding what is meant by the terms natural circulation and free convection. If we have a fluid being heated in a vertical tube, its density will decrease. Now, if the system is closed, that is, consists of a “loop” and the fluid is cooled before returning to our heated tube, we would have flow induced by the so-called chimney effect, which causes a pressure differential, or “driving pressure,” in the heated tube which would be resisted by frictional effects. Flow can be either laminar or turbulent, depending on the flow velocity achieved. This effect is known as natural circulation and can be analyzed by application of the mechanical energy equation (§9.111) to determine the flow rate. We would have the same effect, but less pronounced, if our heated tube was in a reservoir of fluid, as we have in a fireplace chimney.

Example 9.10. As an introduction to the role of the chimney in a natural circulation boiling-water reactor (§15.25), estimate the driving pres­sure resulting from a 10-m-high chimney above a boiling core that has an exit quality of 15 percent. At the dome pressure of 7.2 MPa, the specific volumes are 0.00136 m3/kg for the saturated liquid and 0.0265 m3/kg for the saturated vapor. Figure 9.18 shows the circulation path schematically. For purposes of this estimate, assume that the system is isothermal and that friction outside the core is negligible. Feedwater replaces the steam produced.

Specific volume of two-phase mixture = (0.85)(0.00136) + (0.15)(0.0265)

= 0.00513 m3/kg; p2_i — 195 kg/m3; p2_3 = 735 kg/m3;

then

p3 — px = 10(735 — 195)(9.8) = 53 kPa.

This compares with a core pressure loss in the SBWR of 44 kPa.

9.133. The terms free convection or natural convection normally apply to fluid motion caused by buoyancy effects when we have a heated surface immersed in a surrounding fluid. Under such conditions we have a so — called boundary layer problem rather than a chimney effect.

9.134. In designing reactor core cooling systems, it is very desirable to make the geometry and pressure losses such that there will be significant natural circulation cooling after shutdown in the event that the circulating pumps fail. Natural circulation cooling considerations are also important in the design of water-filled storage pools for spent-fuel assemblies (§11.34).

10.63. During the 1970s, concern about the long-term adequacy of nu­clear fuel resources led to substantial attention being given to ways of improving the utilization of fissile and fertile materials, independent of economic considerations. Associated with this effort, the matter of prolif­eration risk, which involved the diversion of fissile material for weapons purposes, was considered. As a result of the slowing of worldwide reactor construction, there is now less concern about resources and more emphasis given to enhancing safety, improving public acceptance, and minimizing investment risks. This change has affected the design objectives for fast reactors and other fuel-efficient systems. Nevertheless, we will briefly re­view here the principles of conversion and breeding since they may again become important.

10.64.

The efficiency with which fuel is being utilized in a reactor may be expressed by the ratio of the number of fissile nuclei (of all kinds) formed to the number destroyed. This ratio is called the conversion ratio (CR) when it is less than unity, as it is in thermal reactors based on uranium fuels, and the breeding ratio (BR) when it is larger than unity and breeding occurs. There are several different ways in which the breeding ratio is defined, but for the present purpose both the conversion and breeding ratio may be written as

where the numbers of fissile nuclei formed and destroyed are usually in­tegrated over an operating period of the reactor between fuel loadings. The fissile nuclei include uranium-235, plutonium-239 and plutonium-241 in a uranium-fueled reactor, and uranium-233 and uranium-235 when the fertile species is thorium-232.

10.65. In principle, the conversion or breeding ratio could be calculated in the same manner as fuel depletion (or burnup), along the lines described in §4.73 et seq., provided the reactor composition and neutron energy spectrum were known as functions of space and time. In practice, such calculations are very difficult, and a number of approximate approaches to the problem have been used. For example, the neutron spectrum may be simplified by considering three energy groups, namely, fast, resonance, and thermal. Furthermore, the local conversion or breeding ratio may be determined by dividing the reactor system into a number of regions of given composition. Allowance for changes with time can then be made as indicated in §4.75. The numbers of fissile nuclei formed and destroyed in each interval are then summed over the operating period to obtain the net conversion or breeding ratio for the given reactor system.

10.66.

In the subsequent treatment, no attempt will be made to calculate the overall (or net) conversion or breeding ratio. Some useful semiquan­titative results will be obtained, however, by considering the conversion or breeding ratio at a given time (or short time interval). For this purpose, equation (10.9) is modified by writing

where the rates of formation and destruction refer to a given time. The ratios calculated from this expression will generally vary with time during reactor operation; this point should be borne in mind in the following discussion. The net breeding or conversion ratio would be represented by integrating both numerator and denominator of equation (10.10) over the operating time of the reactor.

11.46. Approval for constructing an aboveground monitored retrievable storage (MRS) facility has been complicated by an unwillingness of state governments to accept it. Also, the 1987 amendment to the NWPA links the MRS to progress toward opening a permanent underground geological site, which has had numerous delays.

11.47. Over the years, various designs for a MRS facility have been proposed, varying from a scaled-up pool storage arrangement to a concrete storage cask array. There are two major considerations in the design. Decay heat must be managed and neutron and gamma-ray shielding provided. Heat generation and activity can be predicted accurately by codes such as ORIGIN (§2.213). However, as a rule of thumb, it is helpful to remember that during about 10 years of temporary storage the assemblies will have lost about two orders of magnitude in heat generation rate and activity from the discharge levels. Subsequent loss is fairly slow, being roughly proportional to the 0.2 power of time (§2.217).

11.48. An indication of retrievable storage cask designs is provided by those submitted by utilities for licensing by the NRC for air-cooled on-site storage to supplement pool capacity. A common design arrangement for a vertical cask accommodates 24 assemblies in a basket rack inside a con­crete shield. A typical cask is about 3.66 m (12 ft) in diameter and 6.1 m (20 ft) high.

12.72. Should there be a complete loss of both off-site and on-site power, all pumps would gradually coast to a stop and the result would be a loss- of-flow accident. However, within 30 s or so, emergency power should be available from the diesel generators to operate essential pumps. In the meantime, the reactor would have been tripped upon receipt of the loss — of-flow signal, and steam would be automatically dumped from the turbine. Since some onsite power would continue to be produced during the dump­ing process, the circulation pumps would normally be left connected to the main generator busbar for about 30 s. The circulation during coastdown together with some natural circulation of the coolant would normally be

sufficient to prevent the critical heat flux condition being attained following reactor trip.

12.171. In the United States, applicants for reactor licenses must pro­vide plans for coping with emergencies. Requirements are given in 10 CFR 50.47 and in its Appendix E. An Emergency Planning Zone (EPZ) must be specified related to a plume exposure pathway following a severe ac­cident. The radius of such a zone would depend on the typographical and access route features of a specific site but is likely to be on the order of 16 km (10 miles). However, an EPZ for a food ingestion pathway must also be considered, which may have a radius of about 80 km (50 miles).

12.172. Plans must be established for both on-site and off-site organi­zations for managing emergency situations, including protective measures, communication, transportation, and many other considerations. Lessons learned from the Chernobyl accident response (§12.187) have been helpful in identifying needs. The general trend of NRC regulations has been to favor new plant siting significantly away from population centers. However, in those European countries that are densely populated, it is not possible to have the luxury of remote siting. Therefore, reliance must be placed on a defense-in-depth design philosophy, which includes a very conservatively designed containment structure as backup.

Dr. Samuel Glasstone, the senior author of the previous editions of this book, was anxious to live until his ninetieth birthday, but passed away in 1986, a few months short of this milestone. I am grateful for the many years of stimulation received during our association, and in preparing this edition have attempted to maintain his approach.

Previous editions of this book were intended to serve as a text for students and a reference for practicing engineers. Emphasis was given to the broad perspective, particularly for topics important to reactor design and oper­ation, with basic coverage provided in such supporting areas as neutronics, thermal-hydraulics, and materials. This, the Fourth Edition, was prepared with these same general objectives in mind. However, during the past three decades, the nuclear industry and university educational programs have matured considerably, presenting some challenges in meeting the objec­tives of this book.

Nuclear power reactors have become much more complex, with an ac­companying growth in supporting technology. University programs now offer separate courses covering such basic topics as reactor physics, thermal — hydraulics, and materials. Finally, the general availability of inexpensive powerful micro — and minicomputers has transformed design and analysis procedures so that sophisticated methods are now commonly used instead of earlier, more approximate approaches.

In light of this picture, giving priority to needed perspective, even at the expense of some depth which is now available elsewhere, was considered appropriate. Also, since it was important to keep the length of the book about the same, necessary new material could only be accommodated by deleting some old material. Significant new material has been added, par­ticularly in the areas of reactor safety, fuel management, plant operations, and advanced systems. However, material that generally continues to meet the objectives of the book has been retained, both to preserve its flavor and to keep the revision effort within reasonable bounds. Also, after the passing of Dr. Glasstone, I felt it inappropriate to change the basic ap­proach of the book.

Readers of the book will want to use computer-based methods to sup­plement the text material, as appropriate. Space did not permit a mean­ingful presentation of the methodology required. All problems listed at the end of the chapters may be solved with hand calculations. Although I have continued the use of SI units, as begun in the Third Edition, complete adoption by industry has been slower than anticipated. Therefore, some problems utilize English units.

A two-volume format has been adopted for this edition to provide read­ers with some flexibility. The chapters have been rearranged somewhat to provide volume coherence, with basic material concentrated in the first volume. An Instructor’s Manual is also available for qualified instructors, to be ordered directly from the publisher.

The suggestions made by A. L. B. Ho, L. E. Hochreiter, В. K. Malaviya, V. H. Ransom, J. R. Redding, G. R. Odette, and T. G. Theofanous are gratefully acknowledged.

Thanks are due to the Chapman & Hall team that published the book, particularly Marielle Reiter for production administration and Barbara Zeiders for editorial assistance.

Finally, I wish to thank my wife for her help and encouragement during the preparation of the book.

Alexander Sesonske San Diego, California March 1994

INTRODUCTION

8.1. The availability of new, powerful digital computers in recent years has resulted not only in increased sophistication in nuclear reactor engi­neering but also in the development of other disciplines involving systems and optimal control theory. These subjects have potential applications in reactor engineering. In nuclear power plant design and operation, decisions must be made in applying engineering principles to the problems to be solved. Since plants are complex, aids to the decision-making process are essential. One aid is to recognize certain portions of the plant having a common function as a system, a term that we will explain further shortly. Decisions regarding dependencies between systems can be expedited by this type of representation.

8.2. The systems concept as a decision tool has its roots in a subject known as operations research [1], which evolved from the strategic planning needs of World War II. As powerful computers became available, sophis­ticated methods for systems analysis, modeling, and simulation were de­veloped for engineering and management applications. In this computer-
based information age, we are witnessing a trend toward greater and greater general use of such systems-related tools for decision making. Computer modeling of nuclear power plant behavior is essential for normal operation as well as for the analysis of postulated accidents to meet licensing re­quirements (§12.231 et seq.).

8.3. Space here does not permit even an introduction to operations research and modeling procedures. Reliability analysis applications will be mentioned in Chapter 12. However, we do wish to alert the reader to the possibility of using available operations research-based methods for the analysis and design of nuclear reactor systems. A first step here is to identify some typical systems.

8.4. In modern engineering practice which utilizes the power of the computer to accomplish most tasks, information of various kinds is a nec­essary tool for decision making. Nuclear engineers have available to them a variety of data bases, computer codes, and other information from many sources. An introduction to these resources is useful as a complement to the presentation of relevant reactor engineering topics in this book. Also, we have drawn attention to appropriate information sources in various other chapters.

9.67. Transient (or unsteady state) heat transfer is important in reactor safety analysis where the thermal behavior as a function of time is of major interest. Generally, the same principles apply as for steady-state heat trans­fer with the addition of time and heat capacity as parameters. In particular, consideration must be given to the heat release after shutdown (§2.215). A typical problem might be to determine the maximum temperature at­tained by the fuel cladding in a water-cooled reactor if the coolant flow is reduced. Even after the reactor is tripped, heat input to the cladding continues; this results from sensible heat stored in the fuel rods which may cause an initial heat-release rate approaching 50 percent of the full power value in a pressurized-water reactor, from fissions by delayed neutrons, and from radioactive decay of the fission products.

9.68. If the coolant-flow rate is decreased, not only will the heat-transfer coefficient decrease, but the thermal transport capacity of the coolant will
decrease since less mass will flow past a given point per unit time. There­fore, for the same heat flux (or heat-flow rate) from the cladding, the coolant will tend to attain a higher temperature. The temperature of the cladding as a function of time is then determined by a balance between the heat input from the fuel and the heat loss to the coolant. Since the temperature-difference driving force for each rate depends on the cladding temperature, the situation is complicated and numerical (computer) meth­ods are required for solution of the problem. Subchannel analysis ap­proaches (§9.135) may be incorporated in these codes to provide a detailed representation of core temperature behavior.

9.183. The peaking factor design approach for liquid-metal-cooled fast breeder reactors, i. e., LMFBRs (§15.51), is inherently similar to that used for water-cooled reactors, but differs in some respects [26]. In PWRs and BWRs, hot-channel factors are used primarily as part of the determination of design margins in relation to a boiling crisis. In LMFBRs, however, the factors and subfactors are used together with computer codes to identify operating margins in relation to a variety of other thermal and hydraulic design limits, e. g., those affecting cladding temperature and the structural integrity of the core internals. Fuel failure propagation is also of concern. A statistical treatment is generally used to analyze the entire core rather than to identify a single hot channel. An objective is to determine the number of channels that approach design limits within specified confidence ranges in a manner analogous to the PWR statistical approach described in §9.172 et seq.

INTRODUCTION

10.1. Reactor fuel management is concerned with those activities in­volved in the planning for and design of the fuel loading for a nuclear power plant. Planning activities include fuel procurement over a period of years as required for future reload batches and consideration of utility operational strategy that may affect the energy production of the given generating unit. Fuel loading design requires not only the development of specifications for a new fuel batch, but also the determination of the core location for fuel assemblies from previous batches which will be reinserted. The nature of these activities will become clearer to the reader as the relevant topics are presented.

10.2. The nuclear fuel cycle is the path followed by the fuel in its various states, from mining the ore to the disposal of the final wastes. At one time, the fuel “cycle” envisioned the recycling of recovered fissile and fertile material from spent fuel. However, in the United States, the prospect of this being implemented in the near future is remote. Therefore, from the viewpoint of an electric utility operating a nuclear power plant, major
interest is in the fuel burnup stage and associated fuel management. In fact, without plutonium recycle, as practiced in Europe, the objective in the United States is to obtain as much energy as possible at minimum cost using a once-through cycle. We will therefore cover the pre-reactor fuel cycle steps in only enough detail so that fuel procurement is understand­able. Post-reactor operations and waste management are treated in Chapter 11.

10.3. Economic considerations play a significant role in fuel manage­ment. Although a comprehensive discussion of nuclear power economics is beyond our scope, some background is essential for an understanding not only of core management but of many aspects of plant system design and operation. Nuclear energy-generating expenses normally include costs associated with the capital investment required, the cost of fuel, and op­eration and maintenance charges. Later in this chapter, a brief introduction to these cost categories will be given. In Chapter 14 we will show how the relative cost contributions affect plant operation strategy.