Heat exchanger design is both an art and a science. The science part of it is premised on the fundamentals of convective and conductive heat transfer and the art part of it is based on determining how to effectively package the design to meet the process requirements with cost constraints. This paper is not concerned with the art portion of heat exchanger design. Rather, the paper presents a method of characterizing the performance of coupled heat exchangers with an intermediate heat load in process systems. In process plants, it is not uncommon to see heat exchangers that are coupled. Determining the performance of such heat exchanger arrangements can be daunting, but not insurmountable. Using first principles equations, engineers iteratively establish the performance of the heat exchanger arrangement. This process is time consuming. In this paper, we present a specific case of two coupled heat exchangers with an intermediate heat load in the intermediate system. The resultant model allows an engineer to perform sensitivity studies of the arrangement prior to investing resources in an actual design of the system. This method is expected to facilitate the process of examining the performance of similarly configured heat exchanger arrangements. Similar techniques can be used to simplify more complex heat exchanger arrangements.

As an advanced Gen III+ plant with passive safety systems, the AP1000® plant is uniquely equipped to handle an extended station blackout (SBO) event similar to what occurred at the Fukushima-Daiichi plants in March of 2011. These passive systems have been designed to maintain core cooling for up to 72 hours following all design basis events without the need for AC power or operator action. These core and containment cooling systems self-actuate such that even DC power is not required for their actuation. The Fukushima-Daiichi event demonstrated the effectiveness and desirability of the AP1000 systems. The AP1000 plant, like other pressurized water reactors (PWRs), is provided with defense-in-depth active systems, such as auxiliary feed water pumps, to remove decay heat using the steam generators in the event that offsite power is lost. During an SBO the diesel generators powering this active equipment would not be available. In the event of an SBO the safety-grade heat removal function would be accomplished by the passive residual heat removal (PRHR) heat exchanger (HX) located in the in-containment refueling water storage tank (IRWST). The PRHR HX is designed to remove decay heat from the reactor coolant system (RCS) to the water in the IRWST, which increases in temperature and eventually boils. Steam from the IRWST is vented to the containment atmosphere and actuates the passive containment cooling system (PCS), which is used to apply water to the outside of the steel containment vessel and passively remove heat via evaporation to the environment. Steam that is condensed on the inside surface of the containment vessel forms a water film that flows down the containment wall and is returned to the IRWST using a system of water collection gutters and piping. The PCS is sized to remove reactor decay heat for 72 hours without the need for operator action. Effective operation of the PRHR heat exchanger and PCS to remove decay heat from the reactor core to the environment depends on the ability to maintain water in the IRWST. Condensate that is not collected and returned to the IRWST is lost into the containment sump. There are several possible sources of loss. At the start of IRWST boiling, all containment structures will condense steam until their surface temperature approaches the steam temperature. This process is dependent on the heat capacity of these structures, and all condensation formed on these structures is considered lost. Since the containment wall is cooled by the PCS operation, condensation continues on the inside surface of the containment throughout the event. There are areas on the containment wall where condensate could be lost including the region at the top of the dome where the surface is nearly horizontal, and areas where weld seams and other obstructions could strip off some condensate film. To determine the coping time limits following an extended SBO, it is necessary to characterize these condensate losses. A Phenomena Identification and Ranking Table (PIRT) process was conducted to determine the important phenomena associated with the return of condensate to the IRWST. This PIRT process identified the need for further experimentation to quantify the losses. This paper describes the PIRT and the experimental facility design used to determine the condensate return losses arising from phenomena identified by the PIRT.

In September 2004, the Nuclear Regulatory Commission (NRC) issued Generic Letter GL2004-02 “Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors” to address Generic Safety Issue 191 (GSI-191) “Assessment of debris accumulation on PWR sump performance.” [1] GL2004-02 requested pressurized water reactor (PWR) licensees to perform a “downstream effects” evaluation of their emergency core cooling (ECCS) and containment spray systems (CSS). GL2004-02 also gave guidance on what analysis had to be completed in order to resolve GSI-191. These evaluations included a wear and plugging assessment of all ECCS and CSS components, including valves. During preliminary “downstream effects” analysis of a plant, it was determined that the positions of ECCS throttle valves could be such that the flow clearances through the valves would be too small to meet the criteria developed for component plugging or wear assessment. This suggested that a modification to the system needs to be made which allows the throttle valves to be more fully opened. In order to allow the throttle valves to be opened more fully, additional hydraulic resistance (i.e. pressure drop at the design flow rate) was added at another location. Several orifice designs were considered to provide the needed resistance. Since the required additional pressure drop was a substantial fraction of the total pressure drop, special design features of the orifice were necessary to preclude system instabilities due to cavitation, degassing and flow swirl. The purpose of this paper is to present a method for assessing the effectiveness of a multi-stage orifice that can be placed in the system to provide the required resistance, thus permitting the throttle valves to be used more efficiently. The paper presents the design aspects of the multi-stage breakdown orifice, CFD modeling used to select the design, and the system condition testing results.

Accurate calculation of pump performance margins relative to test acceptance criteria are driven by a variety of requirements and constraints. Inputs including design basis required performance, acceptance criteria assumptions, test conditions, and field versus vendor data are required. Test margin with respect to safety analysis limits may be more limiting than ASME Operations and Maintenance (OM) Code [1] inservice testing (IST) requirements. In response to component design basis inspection (CDBI) and design basis assurance inspection (DBAI) findings of potentially non-conservative equipment performance when operating at the extremes of Emergency Diesel Generator (EDG) Technical Specification (TS) limits of frequency and voltage, Westinghouse and the Pressurized Water Reactor Owner’s Group (PWROG) developed WCAP-17308-NP-A [2] to provide a simplified approach to incorporate these limits, by treating them as uncertainties, into design basis pump test acceptance criteria. The basic methodology provides a general approach to account for uncertainties by adjusting pump curves and test acceptance criteria. Depending on the magnitude of the EDG and instrument uncertainties, adjusted design basis related test acceptance criteria may challenge the tested performance of the pumps. Margin can be recovered by reducing uncertainties and taking credit for any available margin in the safety analyses.

Plant Technical Specifications are issued by the US NRC to ensure that safe nuclear power plant operation is maintained within the assumptions for parameters and Structures, Systems, and Components (SSCs) made in the plant safety analysis reports. The Technical Specifications are made up of Limiting Conditions for Operation (LCOs), which are the minimum set of requirements that must be met based on the assumptions of the safety analysis, Actions, which are the remedial or compensatory actions that must be taken if the LCO is not met, and Surveillance Requirements, that demonstrate that the LCO is met. The Technical Specification Actions contain Completion Times (CTs) which are the time within which remedial actions must be taken, in the event that the LCO is not met. The Improved Standard Technical Specifications (ISTS) for Westinghouse plants are contained in NUREG-1431, Revision 2. Condition A of Technical Specification 3.5.2 (ECCS- Operating) in NUREG-1431, Revision 2, allows components to be taken out of service for up to 72 hours, as long as 100% of the ECCS flow equivalent to a single Operable ECCS train exists. Condition A would allow, for example, the A train low head safety injection (LHSI) and the B train high head safety injection (HHSI) pumps to be taken out of service (for 72 hours) as long as it could be demonstrated that the remaining components could provide 100% train equivalent flow capacity. The “cross-training” allowed by this Condition in the ISTS provides flexibility when performing routine pre-planned preventive maintenance and testing, as well as during emergent corrective maintenance and testing associated with random component inoperabilities. Without this flexibility, a unit would have to initiate a plant shutdown within 1 hour, if component(s) were inoperable in different trains. In order to implement this flexibility, the various combinations of components in opposite trains must be evaluated to determine whether 100% of the ECCS flow equivalent to a single Operable ECCS train exists with those components out of service. This evaluation ensures that the safety analysis assumption associated with one train of emergency core cooling system (ECCS) is still preserved by various combinations of components in opposite trains. An ECCS train is inoperable if it is not capable of delivering design flow to the reactor coolant system (RCS). Individual components are inoperable of they are not capable of performing their design function, or support systems are not available. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of Condition A is to maintain a combination of components such that 100% of the ECCS flow equivalent to a single Operable ECCS train remains available. This allows increased flexibility in plant operations under circumstances when components in the required subsystem may be inoperable, but the ECCS remains capable of delivering 100% of the required flow equivalent. This paper presents a methodology for identifying the minimum set of components necessary for 100% of the ECCS flow equivalent to a single Operable ECCS train. An example of the implementation of this methodology is provided for a typical Westinghouse 3-loop ECCS design.