5. Cost reduction approach

Nuclear power plants are composed of the so-called “nuclear island” and a steam plant. With the exception of the steam generators, the steam plant of a nuclear power plant resembles the steam plant of a fossil fired plant. LWR steam plants use saturated or slightly superheated steam while the LMFBR steam conditions are more typical of a fossil-powered plant. Much experience has been accumulated in the power industry building economic steam plants which would carry over to any envisioned LMFBR plant, so any cost reduction initiative need only consider the nuclear island.

The prime factors driving the cost of the nuclear island are size and complexity. A third factor, congestion, can also become a major contributor if it is allowed to, and is best handled by avoiding design concepts that are over-integrated, i.e. that place excessive and sometimes competing demands on single components. Congestion lengthens construction adding carrying costs and it tends to occur late in the construction process when the carrying costs are the greatest. We shall see how this problem of congestion appears in certain LMFBR design approaches as we proceed with this discussion, and what steps can be taken to avoid it. A fourth factor is the extent of the plant that is safety related, i.e. necessary for safe shutdown. On CRBRP, the safety related portion of the plant extended to the steam system. Modern design approaches typically attempt to reduce the safety related envelope to the primary system and associated decay heat removal system(s). The two approaches shall be described. A fifth factor is the degree of required on-site fabrication. On-site fabrication has two adverse effects — it delays construction and is performed in a makeshift environment with typically lower skilled personnel. A sixth factor could occur if the size of fabricated components requires barge shipment.

Standards and regulation also significantly impact cost. Design organizations typically participate in standards setting committees in the interest of ensuring standards do not unnecessarily impose onerous requirements that will be unnecessarily expensive. Potential impacts of regulatory action will be treated in subsequent sections where CRBRP experience suggests that trouble may be imminent.

For the case of CRBRP, a 68 ft. long and 20 ft. diameter reactor vessel housed a reactor core that was just 5 ft. 4 in. in height and about 8 ft. in diameter. The vessel was centrally located in a cylindrical containment 186 ft. in diameter. The prime drivers impacting the reactor vessel height were the fuel element design and the refueling system design. The Primary Heat Transport System (PHTS) design was the driver for the containment design. The Steam Generating System (SGS) introduced complexity which can be eliminated in a straightforward way. This is related to the Decay Heat Removal System (DHRS) that was selected, for which simpler and possibly more reliable alternatives are available. The containment design approach was borrowed from PWR practice when a lower cost alternative is available and would be appropriate. The extensive 1E electric power system is also a candidate for elimination or at least, major cost reduction.

Before getting into the details, it is necessary to provide an overview of the LMFBR Heat Transport System. Figure 3 shows conceptually how heat is transported from the reactor to the Intermediate Heat Exchangers (IHX) via the Primary Heat Transport System (PHTS), then to the Steam Generator via the Intermediate Heat Transport System (IHTS) and ultimately to the turbine. In practice, there will be from two to four primary loops and an equal number of intermediate loops with the steam from all steam generators combining to a single turbine. The IHTS is made necessary by the requirement to eliminate the possibility of water leaking from failed steam generator tubes finding its way to the reactor. The IHTS, although adding complexity and cost in comparison to PWRs, does have the benefit of enabling the steam generators to be located outside containment, where they are more accessible, maintainable, and replaceable.

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Figure 3 Conceptual LMFBR Heat Transport System

Two sodium cooled submarine reactors, S1G and S2G, avoided intermediate loops by using double-walled steam generators that were located inside containment. It has been reported23 corrosion was experienced in the steam plant which had been accelerated by gamma radiation, which may or may not have credibility. Double-wall tubes have been deployed in certain liquid metal designs, extensively tested, and may find application in future LMFBRs, independently of whether or not they are used with or without an intermediate loop. Absent double-walled tubes, it is necessary to provide for potential tube failure in the plant design.

Design of a nuclear power plant must begin by setting requirements and objectives. As stated in Section 1, Superphénix parameters will be used, but the overriding objectives are to minimize cost – both capital and operating, while maximizing availability, complying with regulatory requirements and maintaining an acceptable level of safety.

A key objective of this design approach takes advantage of the breeding characteristic in such a way as to maximize the interval between refueling. Since the LMFBR produces more fuel than it consumes, it is possible to design the core so that the newly bred fuel occurs in regions of high importance, minimizing the reactivity swing attendant to burnup. That objective will be the prime driver for the proposed core design. This approach probably penalizes breeding ratio somewhat but maximizing time between refueling creates opportunities for economies in the refueling system, which is more important, particularly for the earlier plants.

It is necessary to next direct attention to the reactor core, proceeding outward to the reactor vessel and refueling system, heat transport system, containment, decay heat removal system, and auxiliaries. It is essential that the reactor core be designed first since it sets the requirements for subsequent systems, but the core design process is lengthy and complicated, even when treated in an overview fashion. Accordingly, it has been moved to the appendix where it can be consulted by those who have a particular interest in such matters. For the purposes of the following sections, the core is intended to have; 1) a ten year interval between refueling; 2) features that minimize core pressure drop to about 20 psi; 3) 1% per minute load following capability; 4) capability for power operations in the 15-100% range; 5) steam parameter consistency with Superphénix; 6) 95% capacity factor between refuelings; 7) a core diameter of approximately 22 ft. (including shield assemblies) and core height including the axial blankets of 5 ft. 10 in., which includes a 33% growth allowance (an extra foot in the fueled region) beyond minimum to accommodate the 10 year refueling interval.


23 R. G. Palmer, A. Platt, Fast Reactors, Temple Press, 1961