12.240. We are concerned here with an “evolutionary” or “advanced” plant as described in Chapters 13 and 15, as no new plants of current design are likely to be built. Since standardization is a major feature of all new plants, it is practical to certify a standard design based on a comprehensive safety analysis report. The amount of information required is formidable and is normally contained in many volumes. It is comparable to what has been required in a Final Safety Analysis Report in an application for an operating license. A complete description of the plant must be included. The safety analysis describes the response of the system to a wide range of accident situations. Details of operation plans, test programs, and tech­nical specifications are to be provided. The technical specifications list limits to be placed on all process variables. In particular, the core peaking factors as determined by emergency core cooling criteria are specified.

12.241. Review by the NRC staff of the Application for Design Cer­tification is a lengthy process which normally includes requests by the staff to the applicant for additional information. It may be necessary to resolve issues by amending the application. In addition, the application is evaluated by the Advisory Committee on Reactor Safety (ACRS), a statutory body of up to 15 specialists in reactor physics, engineering, materials sciences, and other areas related to reactor safety. Finally, after approval is rec­ommended, public hearings are to be held. In the past, an Atomic Safety and Licensing Board would be specially convened for this purpose.

Introduction

14.31. Expert systems and neural networks were introduced in Chapter 8 as information tools. Expert, or so-called “knowledge-based” systems play an important role in nuclear operator support systems and provide diagnostic aids in plant maintenance activities. Their role is to take ad­vantage of various knowledge sources and to provide guidance in solving problems. Neural networks, on the other hand, utilize highly intercon­nected elements that process information in a manner similar to that used in the human brain.

14.32. A feature of neural networks is their ability to recognize patterns, even when the incoming data are incomplete or distorted, which is not possible with normal computer programs. Therefore, the information­gathering power of expert systems, with the disadvantage of some input uncertainty or “fuzziness,” complements the processing power of neural networks, which can quickly yield results despite incomplete input. Al­though application of expert systems to nuclear power plants has become fairly common, neural network use has been mostly developmental in nature. Our primary purpose is to draw attention to the features of such applications.

15.65. The PIUS system features a very simple PWR design that is claimed to be so inherently safe that it does not require injection by passive systems to prevent a serious accident. A major feature is the submergence of the entire primary system, including the steam generator, in a very large pool of relatively cool, borated water. This pool is inside a large pre­stressed concrete vessel which serves as both the pressure boundary and containment.

15.66. As shown schematically in Fig. 15.7, during normal operation, primary coolant moves up through the core, through a tall “chimney,” and into a hydraulic interface region, from which it is pumped (recirculated) to the steam generator. On leaving, the coolant passes through a down­comer duct to a plenum under the core, below which is another hydraulic interface region.

15.67. The recirculation pumps are operated so that the pressure drops of the water flowing through the core results in a hydraulic balance pre­venting the pool water from mixing with the coolant water through the lower interface. If the circulating rate drops or the core temperature in­creases, the pressure balance will be upset and the borated pool water will flow automatically up into the core, shutting down the reactor and cooling it. Decay heat is removed by allowing the pool water to boil. Sufficient water is initially present so that the core will remain covered for about a week.

15.68. Since the design is still being developed, it is premature to pro­vide even preliminary specifications. Several design options are under con­sideration, including one in which three core modules are submerged in a single pool contained within a prestressed concrete vessel. Questions of cost, licensability, maintainability, and operability require resolution. For example, concern has been expressed regarding the effect on plant avail­ability factor of the time required to restart the plant after inadvertent shutdowns. A prototype plant would be needed to prove the feasibility of the concept.

13.22. The Combustion Engineering design evolved from that of the three-unit Palo Verde Generation Station in Arizona, which began com­mercial operation in 1986. A summary of design specifications for the proposed system is given in Table 13.2. The format follows that used in Table 13.1.

13.23. The fuel assemblies are in a 16 x 16 array with rod dimensions similar to those used in the Westinghouse core. However, as seen in Fig. 13.6, there are five large thimbles in each assembly. Four of these are for control rods and one for instrumentation. There are clusters of both four and 12 elements in the core. The 12-element cluster spans five assemblies. Each rod cluster is connected by a “spider” to the drive mechanism. Groups of these clusters are moved as units for power regulation. Burnable ab­sorber rods are substituted for fuel rods in the lattice, as needed, rather than inserted in the control rod thimbles.

13.24. The reactor cooling system arrangement is shown in Fig. 13.7. Four circulating pumps serve two large steam generators. The steam gen­erator secondary-side water inventory has been increased by 25 percent to improve the response to upset conditions and to extend the time available for countermeasures should the feedwater supply be interrupted. Similarly, the pressurizer volume has been increased by 33 percent to reduce the pressure changes during transients such as reactor trip and load rejection.

13.25. The reactor operating margins relative to previous designs were increased by the foregoing improvements as well as by a reduction in the hot-leg temperature and improved monitoring methods. Also, the new design provides for carrying out operating power level maneuvers using control rods only. Thus, the need for short-term boron-level adjustments is reduced.

13.26. Numerous other evolutionary design features contribute to sim­plification and improved safety margins. A probabilistic risk analysis in­dicated a reduction in the risk of severe accidents by two orders of mag­nitude compared with that for present systems.

General Thermal-Hydraulics Power Thermal 3800 MW Electrical (net) 1280 MW Specific power 37 kW(th)/kg U Power density 95.4 MW(th)/m3 Coolant Pressure 15.5 MPa(a) (2250 psia) Inlet temp. 292°C (558°F) Outlet/temp. 324°C (615°F) Flow rate 21.4 Mg/s (1.61 x 108 lb/ hr) Mass velocity 3.6 Mg/s • m2 (2.64 x 106 lb/hr) Rod surface heat flux Ave. 0.599 MW/m2 (1.90 x 105 Btu/hr-ft2) Max. 1.41 MW/m2 (4.46 x 105 Btu/hr-ft2) Steam pressure 6.90 MPa(a) (1000 psia) Core Length 3.81 m (12.5 ft) Diameter 3.66 m (12 ft) Fuel Rod, OD 9.7 mm (0.382 in.) Clad thickness 0.64 mm (0.025 in.) Pellet diameter 8.3 mm (0.325 in.) Rod lattice pitch 12.8 mm (0.506 in.) Rods per assembly 236 (16 x 16 array) Assembly overall 202 mm (7.97 in.) width Assemblies 241 Fuel loading, U02 116 x 103 kg (2.57 x 105 lb) Enrichment levels 3.3, 2.8, 1.9 percent by weight Control Control assemblies 68 full length, 25 part strength

13.27. A 61-m (200-ft)-diameter spherical containment for this system provides both safety and cost advantages. As shown in Fig. 13.8, a concrete shield building encloses an inner steel sphere which has a relatively large internal free volume of 9.5 x 104 m3 (3.4 x 106 ft3). The steel shell acts as a heat sink and the large volume provides energy-absorbing capability in the event of an accident. An in-containment refueling water storage tank (IRWST) combines the functions of a refueling water tank and a post — LOCA containment sump. Economic advantages also result from efficient space utilization and ease of construction.[27]

13.28. To control costs, this and other new reactor proposals feature extensive standardization, not only for the Nuclear Steam Supply System, but the entire plant. This is in accordance with NRC’s standardization rules given in 10 CFR 52, which also provides for design certification so that the plant can be pre-licensed prior to purchase. A 48-month construction period is estimated, which should help meet a construction cost goal for new plants established by EPRI of $1500 per kW(el).

15.8. During the years that have elapsed following the design and con­struction of present nuclear plants, much effort has been devoted to plan­ning the features of new plants. In fact, a major Electric Power Research Institute (EPRI) program was devoted to the development of a compre­hensive set of design requirements for advanced plants. Although there are several different advanced plant concepts, they all share features in­tended to simplify design, with an accompanying reduction in construction time and costs. Standardized design is emphasized for the entire plant, not just the nuclear portion. Also, all concepts utilize advanced instrumentation and control systems to improve safety and reduce costs.

15.9. Designers have specified standard components where possible and arranged these components into prefabricated modules. During recent years, the petroleum, shipbuilding, and power plant industries have developed successful construction practices featuring the manufacturing of very large component modules at a central location and then transporting them to the construction site. In this way, fabrication and inspection is more effi­cient and field labor costs are reduced. Clearly, cost benefits are increased if the manufacturing facility serves several plant sites. Modularization is carried an additional step with gas-cooled and liquid metal-cooled concepts in which several small steam generating units are coupled to a turbine — generator to form a power block. These may then be replicated at a given site to provide whatever station capacity needed.

15.10. The development of advanced instrumentation and control sys­tems has been coordinated by the Electric Power Research Institute. Con­trol rooms have been redesigned using human factors principles. Digital instrumentation and multiplex signal transmission have been featured. The level of automation of control tasks has been increased. Maintenance —

related instrumentation throughout the plant has received attention, par­ticularly that providing diagnostics to identify equipment that may be ap­proaching failure although still operationally functional.

12.203. Should an accident occur, what is to be done? Strategies need to be considered carefully since some adverse effects may accompany the apparent advantages of a given strategy. For example, should the core slump to the bottom of a PWR pressure vessel, a logical strategy is to flood the surrounding reactor cavity, which will submerge the lower vessel head. Hopefully, this would prevent vessel failure and help cool the contained debris. However, should the vessel breach, a steam explosion is now a possibility as a result of the reaction between the molten fuel and the water that has been introduced.

12.204. Assessment of severe accident management strategies by var­ious means is therefore a necessary feature of planning. Procedures based on decision trees (§12.212), an operations research method, can be used to evaluate systematically all of the options and to develop choices. Sen­sitivity analysis may also be included [38]. Then guidelines in a form useful for operators can be prepared.

13.60. Containment of the fuel bundles within individual pressure tubes localizes the consequences of possible fuel cladding failures and contributes to overall safety. Further, the combination of natural uranium fuel and on-line refueling results in a fuel burnup only about one fourth that of an LWR. Hence, the fission-product inventory is much lower.

13.61. In addition to the normal shutdown system, a backup system is provided for emergency shutdown. In earlier CANDU designs, the mod­erator could be dumped in an emergency. In the larger cores of recent design, injection of gadolinium nitrate solution into the moderator is preferred.

13.61. An emergency core-cooling system is available in the event of a loss-of-coolant accident. However, the separate moderator system provides a substantial heat sink should the coolant circuit be breached. Since there are many coolant inlet and outlet lines of relatively small diameter, com­pared with those in a LWR, the probability of a rapid depressurization accident is greatly reduced. A containment structure, consisting of a prestressed-concrete enclosure containing the nuclear steam supply system, combined with a water dousing arrangement, provides a pressure suppres­sion feature should there be a break in a coolant manifold.

15.41. The reactor core consists of stacks of hexagonal graphite prisms with holes for fuel rods, for coolant flow, and in some cases, for boron carbide control rods. As shown in Fig. 15.4, there are 10 prisms in a stacked column, arranged in an annulus. Unfueled graphite blocks serve as inner and outer radial and upper and lower axial reflectors. This reflected annular core arrangement is needed to permit passive decay heat removal while maintaining the maximum fuel temperature below 1600°C during a so — called conduction cooldown event during which there is no circulation of coolant. The core height is limited to 7.9 m to allow a maximum power rating while assuring axial power stability to xenon transients.

15.42. A set of 24 control rods located in channels in the inner ring of the annular reflector blocks are used for normal control and for emergency

shutdown. A set of six rods in the outer ring of the central reflector is inserted only during cold shutdown. A reserve shutdown capability is pro­vided by boronated graphite pellets, contained in hoppers above the core, which can be dropped into channels in selected core fuel blocks, if needed.

12.242. Site-related matters are evaluated separately. Documentation required includes such topics as a seismic description of the proposed site. An Environmental Report covers all aspects of the effects of the plant on the surroundings. In addition to radiological and thermal effects of plant operations, the effects on the environment of a spectrum of postulated accidents must be described. Also to be provided are economic and social effects as well as a cost-benefit analysis for the plant.

12.243. Emergency plans must be described. It should be recognized that requirements of the Environmental Protection Agency (EPA) and the Federal Emergency Management Agency (FEMA) also apply to the pro­posed site. Work is ongoing to minimize the impact on the nuclear industry due to the overlap of regulatory responsibilities among NRC, EPA, FEMA, and state safety and environmental agencies.

12.244. At any rate, after the site application has been approved pro­visionally, a public hearing is held to allow for additional input. After any issues that may be raised are resolved, an early site permit is issued.

14.33. The reactor operator must be able to react quickly to indications of abnormal conditions. Expert systems can help provide the information needed for effective action. For example, an expert system for PWR op­erator support would consist of several subsystems, each having a different function. A diagnosis subsystem identifies an abnormality. Without as­sistance, an operator confronted with various offnormal meter readings would have to infer the basic cause. Similarly, multiple alarms triggered by limit exceeding conditions may not be helpful in pinpointing the malfunction.

14.34. In one type of diagnosis subsystem or module, logic trees are prepared from a knowledge or “rule base” relating symptoms, such as those indicated by meters or alarms, to basic faults. A pattern search then seeks one or more matches. As a simplified example, consider leakage in the reactor coolant system (RCS) as the fault. If we had a high charging — rate flow alarm and an indication that the pressurizer level was decreasing, the search would conclude that RCS leakage was likely, which then could be verified by additional signals.

14.35. Once a fault has been identified by the diagnosis module, the computation moves to an operational guidance module which calls upon an expert knowledge base to provide corrective measures. In our simplified example, a load reduction and perhaps reactor trip would be indicated. However, for a more sophisticated case, a less obvious action might be recommended.