French electricity utility EDF has bought out US joint venture partner Constellation Energy's 50% stake in US nuclear new-build company UniStar for $140 million. The deal follows Constellation Energy's withdrawl from the US Department of Energy's loan guarantee programme, which it argued was prohibitively expensive.
In addition to sites for Calvert Cliffs 3 and a potential fourth reactor at Calvert Cliffs, which are both held by UniStar, Constellation will also transfer potential new nuclear sites at Nine Mile Point and R. E. Ginna in New York State to UniStar. With the sale of its share of UniStar, Constellation will no longer have responsibility for developing or financing a new nuclear plant at Calvert Cliffs 3.
Further to the terms of the agreement, EDF will transfer to Constellation 3.5 million of the shares that it owns in Constellation and will relinquish its seat on the Constellation board. The existing standstill agreement between the companies will be terminated. Constellation will terminate its rights under the existing put option and, as a result, will not sell any of its plants to EDF. The ownership structure of the companies’ existing Constellation Energy Nuclear Group partnership remains unchanged with Constellation holding 50.01 percent ownership and EDF maintaining 49.99 percent partner status.
The power purchase agreement between CENG and each of Constellation and EDF will be modified to be unit contingent through the end of its term in 2014, and includes commensurate changes to prospective monthly hedges. Pre-existing hedges under the PPA will remain in place as firm sales to Constellation and EDF at the specified price.
The deal came after an unusual public release of letters between executives at EDF and Constellation Energy in October in reaction to withdrawl from the loan guarantee. In the third letter in the exchange, Constellation Energy chief operating officer Michael Wallace offered to sell EDF its 50% stake in UniStar for $117 million.
|ACRS report excerpts|
The ESBWR design includes a boiling-water reactor (BWR) nuclear steam supply system (NSSS)...The ESBWR design utilizes a low-leakage containment vessel, which is comprised of the drywell and wetwell. The containment vessel is a cylindrical steel-lined reinforced concrete structure integrated with the reactor building. The DCD describes a nuclear plant with a NSSS thermal power rating of up to 4,500 megawatts thermal (MWt). Based on this reference design, the plant has a rated gross electrical power output of 1,594 megawatts electric (MWe) and a net electrical power output of approximately 1,535 MWe. The COL applicant will establish the rated electrical power output based on the turbine island design selected and site-specific conditions and may base the COL application on a lower rated thermal power output to satisfy site-specific environmental parameters. While the COL license period is for 40 years, GEH stated that the plant has a design life objective of 60 years without a replacement of the reactor vessel. Safety Enhancement Features
The ESBWR is a direct-cycle, natural circulation BWR and has passive safety features to cope with a range of design basis accidents (DBAs). Within the containment structure are the isolation condensers (IC), the elevated gravity-driven cooling system (GDCS) water pools, a passive containment cooling system (PCCS), and an elevated suppression pool. These systems can remove decay heat under all conditions. The ESBWR standard design includes a reactor building that surrounds the containment, as well as buildings dedicated exclusively or primarily to housing related systems and equipment.
The limiting ESBWR DBA is a Main Steam Line Break (MSLB). In this DBA, water and steam are initially discharged from the break into the drywell. As the drywell pressure increases, the horizontal vents between the drywell and wetwell clear. Subsequently, a steam-water mixture from the break flows through the vents into the wetwell suppression pool, where the steam is condensed, and the water is cooled to the pool temperature. As the primary system pressure falls to the drywell pressure, water makeup to the reactor vessel is provided by actuation of the GDCS; i.e., the GDCS squib valves open and water flows by gravity head into the vessel from the GDCS pools. This occurs approximately ten minutes after the initiation of the accident. The reactor core is never uncovered during the limiting DBA. Steam condensation in the suppression pool and pressure equilibration between the drywell and wetwell through the vacuum breakers reduce the drywell pressure causing the horizontal vents to close. The remaining noncondensible gases and steam in the drywell then flow up through the PCCS heat exchanger. The steam is condensed as it passes through the PCCS tubes. Water condensate is collected and returned to the GDCS pools, and the noncondensible gases flow into the wetwell gas space. This establishes a passive long-term recirculation cooling mode for over 72 hours. Non-safety-related recirculating fans are credited after 72 hours and result in a further reduction in the containment pressure. However, calculations show that even in a purely passive mode, the containment pressure remains below the design pressure for over 30 days.
Probabilistic Risk Assessment The ESBWR design certification application included a PRA in accordance with regulatory requirements. The ESBWR PRA is a Level 3 PRA that covers full power operation and shutdown conditions. The scope of initiating events includes internal events and assessments of internal plant fires and floods. The only quantified external events are high winds and tornadoes. A seismic margin analysis was performed, but the risk from seismic events and other possible external events was not quantified. Although many of the analysis elements are consistent with the ASME-RA-Sb-2005 Capability Category 2 Standard, those attributes were not consistently achieved at this stage of the PRA development. For example, some aspects of human performance, models for equipment testing and maintenance, and details of fire and flood damage cannot be analyzed in the absence of a physical plant, procedures, and operations staff. In these cases, surrogate analyses were performed and assumptions were applied to encompass potential plant configurations, operations and maintenance programs, and organizations. In addition, any analyses requiring site-specific characteristics were treated in a generic manner. Our review found that this PRA was acceptable for design certification purposes. The estimated frequencies of core damage and large releases provide confidence that the ESBWR design achieves NRC staff expectations for advanced plants. The PRA was an integral part of the ESBWR design process, and risk insights influenced a number of design changes throughout the review. This integrated risk perspective was an important contribution to achieving the estimated low risk. The limited scope, varying level of modeling detail, and lack of specificity with respect to â€œas- built, as-operatedâ€ plant conditions limit direct use of the current ESBWR PRA for risk-informed applications. Therefore, it is important that any future use of the PRA results during the COL process, such as the use of calculated risk importance measures for selection of SSCs for the Design Reliability Assurance Program, should be carefully examined and supplemented by appropriate engineering expertise.
On August 24, 2005, GEH submitted its application to the NRC for certification of the ESBWR design. This application was submitted in accordance with Subpart B, â€œStandard Design Certifications,â€ of 10 CFR Part 52. The NRC formally docketed the application for design certification (Docket No. 52-010) on December 1, 2005. The application consists of the ESBWR Design Control Document (DCD) and the ESBWR probabilistic risk assessment (PRA) report.
During these reviews, we issued 6 letters identifying issues of concern and areas for which we needed additional discussion. The applicant has submitted additional proposed revisions to the DCD to resolve all the open issues from the NRC staff and of interest to us. It is intended that these revisions be incorporated in Revision 8 of the DCD. Some of the issues included:
â€¢ Combustion control of flammable noncondensible gases in the PCCS: GEH revised the design of the IC and PCCS to address the potential for hydrogen detonations within the condenser tubes or the lower plenum. The IC system configuration was modified to isolate it from the ESBWR vessel for loss of coolant accident (LOCA) events and to vent it for non-LOCA events in order to address the possibility of combustion events in the IC. The primary structural material of the PCCS was changed to a high strength stainless steel, and component wall thicknesses were significantly increased so that the PCCS can withstand multiple combustion events under bounding conditions. In addition, a passive catalytic recombiner was added to the PCCS drain line to remove combustible gases from piping to the wetwell.
â€¢ Clarification and detailed explanation of digital instrumentation and control (DI&C) systems for ESBWR: GEH provided more detailed explanations and tabular information in the DCD revisions to give us confidence that the four fundamental principles are inherent in the hardware and software DI&C architectures, i.e., redundancy, independence, determinate behavior, and diversity and defense in depth. Finally, additional DAC/ITAAC were developed for the ESBWR to confirm that the final system design would meet these principles.
For the ESBWR, the proposed additional information to be included in Revision 8 of the DCD provides expanded detailed functional descriptions and DAC/ITAAC for the DI&C hardware and software architectures which support the conclusion that the design will meet requirements. However, there is a class of descriptive information, i.e., integrated system logic diagrams, that is not included. These diagrams would simplify the review and make the safety judgment more robust. Such functional descriptions would also aid in the inspection of DAC/ITAAC for final I&C qualification. Under current practice, the NRC staff does not require that such integrated system logic diagrams be included in the Tier 2 information. We suggest that staff consider requiring such information.
For a copy of the ACRS letter, please go to the following link: http://adamswebsearch2.nrc.gov/idmws/ViewDocByAccession.asp?AccessionNumber=ML102850376