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Standardization of Reactor Designs

The nuclear power plants prior to Generation III were built (using Part 50 process) based on customized designs to meet the site-specific characteristics and plant-specific needs. Although the site investigation and design planning were conducted in the CP stage, detailed designs were often performed in parallel to the plant construction and safety issues identified during the design and construction should be resolved before the issuance of the OL. While the history of the safety operation of the current fleet of the U.S. commercial reactors serves as a testimony for the success of the Part 50 process, there exist uncertainties due to late identifications of safety issues during design and construction which could not be resolved among the stakeholders, thereby jeopardizing the future status of the plant. A textbook example to illustrate the detrimental impact of such uncertainty would be the Shoreham nuclear power plant [4]. The plant was located adjacent to the Long Island Sound in East Shoreham, New York. It was a General Electric (GE) BWR, which was begun construction in 1973 and was granted OL in 1989, but was decommissioned in 1994. The plant was never placed in service for generating electricity because of an impasse on issues raised during the construction associated with the plan for evacuation in events of accidents. The plant was eventually sold to the New York State for $1 for decommissioning and the total cost incurred to tax payers was $6 billion to end the two-decade saga.

A salient feature of Generation III and III+ reactors licensed under Part 52 process is the standardization of the reactor designs. An important lesson learned from Shoreham would be to streamline the licensing process to enable early resolution of safety issues, thereby ensuring predictable outcomes for design, construction and operation of a plant. The standardization based on Part 52 accomplishes this by requiring “essentially complete” design [5, 6] during a DC application. In essence, the essentially complete design need be sufficiently detailed to enable the NRC to judge the applicant’s proposed means of assuring that construction conforms to the design and to reach a final conclusion on all safety questions before the certification is granted.

For structures, the essentially complete design is often interpreted as the complete design of critical structural sections, i.e., these structural sections which provide primary load-carrying functions; these are structural elements found in direct load paths as essential blocks for ensuring the safety of the structure. The design of critical sections has evolved gradually over years. Experiences gained from past constructions of nuclear facilities have been used in some DC applications to identify critical sections [7]. Recently, in support of the U.S. EPR DC application, AREVA NP developed a selection criteria based on a three-tier, systematic approach [8].

The selection methodology described in the reference [8] consists of three-tiered criteria — qualitative, quantitative, and supplemental — applied in a sequential manner. The qualitative criterion is used to select critical sections of so-called Nuclear Island (NI) structures that perform safety-critical functions. The quantitative criterion is used, together with a numerical algorithm and the finite element (FE) analysis of the NI structures, to identify critical sections that are highly stressed but are not chosen under the qualitative criterion. The quantitative criterion ensures that the structural elements essential to the structural load paths are adequately identified and considered. The supplemental criterion is based on engineering judgment and is intended to capture critical sections of seismic Category I structures that are not screened by the other two criteria but are necessary to obtain an adequate representation of all types of structural elements. Preliminary estimates discussed in the reference [8] indicate that the critical sections selected using this methodology are representative of approximately 77% of NI structures and 84% of all seismic Category I structures.

Following the selection methodology described in reference [8], or similar systematic methodologies that combine numerical evaluations with engineering judgment, would greatly simplify the selection process and ensure a consistent set of critical sections adequate for an essentially complete design.

Another important feature of the standard design is that the design is typically based on a well established generic site which generally encompasses a wide range of site characteristics so that the standard design upon certification can be built at multiple sites. A COL applicant can incorporate a standard design by reference upon demonstration that the site characteristics are appropriately bounded by the site parameters postulated for the generic site. The advantage of the standardization is that the safety issues which have been resolved during the certification process are not subject to future litigations, thereby ensuring more predicable outcomes in the licensing process.

Other significant features of standardized designs include but not limited to:

  • • Simpler and more rugged design, making them easier to operate and less vulnerable to operational upsets
  • • High reliability and longer operating life — 60 years
  • • Enhanced safety by further reducing possibility of core damage accidents
  • • Protection against an aircraft impact
  • • Better fuel efficiency
  • • Expedited licensing, reduced capital cost and shortened construction time

With continued appreciation in the reactor safety, emphasizing on better understanding the systemic and phenomenological characteristics as well as severe accident managements, advanced technologies based on evolutionary and passive designs with enhanced safety features have been deployed in Generation III/ III+ reactors to provide better defense-in-depth and safety margins in accident preventions and mitigations than previous generations of reactors. Table 9.1 shows some of severe accident mitigating features seen in

TABLE 9.1

SEVERE ACCIDENT MITIGATING FEATURES AND RISK METRIC COMPARISON*

Feature

ABWR

ESBWR

AP1000

U.S. EPR

U.S. APWR

CDF/yr

LRF/yr

CCFP

(internal events at power)

2E - 7 <1E - 8 <0.1

2E - 8 1E - 9 0.08

2E - 7 2E - 8 0.08

3E - 7 2E - 8 0.08

1 E - 6 1E - 7 0.1

Core melt stabilization (CMSS) BiMAC or core catcher

core catcher passive core debris flooding

BiMAC, passive cooling of molten debris

IRWST flooding for cooling core debris in-vessel

core catcher passive cooling of molten debris

core debris trap

Reactor cavity flooding capability

AC-independent with alternate thermally actuated flooder system

GDCS with firewater as back-up

manual actuation

SAHRS and CMSS

dedicated cavity flooding pathway

Containment overpressure protection system (vent)

filtered vent

filtered vent

containment opening reserved for filter vent

Containment combustible gas control

inerted

containment

inerted

containment

manual igniters, 2 PARs for LOCAs

PARs

glow type igniters

In-containment refueling water storage tanks

included

included

included

RCS severe accident depressurization

normal ADS

normal ADS

ADS valves

2 trains, manually actuated

1 train manually actuated

Reactor cavity concrete floor

sacrificial layer

sacrificial layer

large floor area with thick concrete layer

sacrificial layer

large floor area

Information was taken from respective application FSAR or DCD.

various design certification applications in the United States. Also included in the table are comparisons of various risk metrics associated with different designs. The calculated core damage (CDF) frequencies for Generation III/III+ are typically in a range of 10-6 to 10-8 per year (internal events at full power) and corresponding large release frequencies (LRF) are between 10-7 and 10-8 per year, which are much lower than these seen in earlier reactors. For severe accident mitigations, new containment features include those such as: in-containment refueling water storage tanks (IRWST), core melt stabilization systems, containment overpressure protection systems, and containment combustible gas controls, etc. These features seen in standard designs provide much better measured safety performance than previous generations of reactors. In addition, all standard designs include evaluations to ensure that containments can withstand external impacts from commercial aircrafts in compliance with NRC regulation 10 CFR 50.150.

 
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