Feedwater system failures

Feedwater system failure Fuel sheath HT system boundary System thermal hydraulics Containment thermal hydraulics... 

The FWS, which provides feedwater via the startup feedwater system for heat removal from the reactor coolant system, in the event of a feedwater system failure. See subsection 6.6.5 for a description of the operation of the FWS. The startup feedwater is also automatically actuated on signals which indicate a loss of water inventory or heat sink in the secondary side of the steam generator (see subsection 7.7.1.8.2 of Reference 6.1). 

The probability of the auxiliary feedwater system failure during SBO, calculated with a... 

ECC is neither initiated nor required for a feedwater system failure. Large steam pipe breaks together with containment system impairments bound the containment behaviour for feedwater system failures. 

LEADIR-PS 200 has a graceful and safe response to all anticipated transients. For example, an overcooling event (as could be caused by loss of feedwater control or spurious opening of steam relief valves in combination with control system failure) causes the core inlet temperature (normally 350°C) to fall as the freezing point of 327°C is approached the coolant viscosity increases, coolant flow decreases, and in the absence of any control system action, the negative temperature coefficients of the fuel and moderator reduce reactor power. Heat removal is maintained by natural convection. 

The acceptance criterion for the resolution of GSI 124, is that the Emergency Feedwater System shall be designed so that its unavailability is no more than 1 x E-4 per demand after accounting for EFW support systems, common cause failures, and operational errors. Furthermore, this reliability goal should be demonstrated by PRA calculations consistent with the guis ance provided in SRP 10.4.9, Rev. 2. 

Redesign of the emergency feedwater system (EFW) to function in the event of a loss of offsite and onsite power. This design accommodates coincident failure of a single active mechanical or electrical component or the effects of a high or moderate-energy pipe rupture. 

Also, there is a requirement that the emergency feedwater system shall have sufficient inventory to permit operation at hot shutdown for at least 4 hours, followed by a cooldown to the conditions permitting operation of the RHR system. The inventory needed for cooldown, shall be based on the longest cooldown required with either onsite or offsite power available with an assumed single failure. 

After the Three Mile Island Unit 2 accident the NRC reviewed the auxiliary feedwater system for availability and reliability of components and decay heat removal capability. In particular, the EFW system was scrutinized with regard to the potential for failure under a variety of loss of main feedwater conditions. The safety concern was that a total loss of feedwater, i.e., loss of both main and emergency feedwater, could result in loss of core cooling. The NRC requested operating plants and plants under construction to review both the reliability and the capability of the EFW system to perform its intended safety function i.e., core decay heat removal. The evaluation by the plants was divided into three parts as discussed below. 

Part one consisted of a limited PRA to determine the potential for EFW system failure under various loss-of-main-feedwater transient conditions, with particular emphasis being placed on determining potential failures from human errors, common causes, single-point vulnerabilities, and test and maintenance outages. This evaluation applies to operating plants and plants under construction and not to advanced or future plants. Part two was composed of a deterministic review of the EFW system using the acceptance criteria of SRP Section 10.4.9 and the associated Branch Technical Position (BTP) ASB 10-1. Part three required a re-evaluation of the decay heat removal capability of the EFW system with respect to EFW system flowrate. Parts two and three apply to advanced or future plants. 

After the Three Mile Island Unit 2 accident, the NRC reviewed auxiliary feedwater system designs with respect to timely initiation, as described in 10 CFR 50, Appendix A, (GDC 20), (Reference 3). Upon completion of the review, the NRC determined that new guidance identified in NUREG-0737, (Reference 4) was necessary in order to assure a timely start of the AFW system after a design basis event (e.g., loss of main feedwater). Among this new guidance was automatic system initiation, environmental and seismic equipment qualification, and single failure criterion. 

Specifically, the auxiliary feedwater system shall incorporate such design features as automatic system initiation, protection from single failure, and environmental and seismic equipment qualification. 

This group of characteristics specifies components of a standby feedwater system that is started if the main feedwater system is not available because of a failure. These systems should have an independent power source available after loss-of-ofifsite-power events. The auxiliary feedwater pumps are usually connected to the main feedwater tank and may also be used as normal start-up feedwater pumps at low reactor power (see above). The emergency feedwater system is usually designed for use in case of a catastrophic failure of the main feedwater system, e.g. if a common feedwater pipeline ruptures, making both main and auxiliary feedwater pumps unavailable. The emergency feedwater pumps and flowpaths should be different to those of the main and auxiliary feedwater. 

Feedwater system malfunctions causing an increase in feedwater flow (two cases were modelled the accidental opening of one feedwater control valve with the reactor just critical at zero load conditions and the accidental opening of one feedwater control valve with the reactor in automatic control at full power). This fault models the failure of one protection division as the limiting single failure. This is fault 4.2.2 in the fault schedule. 

The operator returned to the auxiliary feedwater station expecting the AFWS to actuate and provide the much-neededfeedwater to the steam generators that were boiling dry. Instead, he first saw the No. 1 AFW pump, followed by the No. 2 AFW pump trip on overspeed -a second common-mode failure of the auxiliary feedwater system and abnormalities four and five. He returned to the SFRCS panel to find that he had pushed the wrong two buttons. 

Since the HTS boundary is preserved, the safety aspect is release of a portion of any radioactivity contained in the secondary side. Generally, the behaviour is bounded by steam and feedwater line failures. The secondary side controls are modelled in some detail to ensure that either their proper functioning, or lack of response, does not impair any safety system actions. Both normal and alternative modes of plant control are assessed. 

Abnormal transients Decrease in core coolant flow rate Partial loss of reactor coolant flow Loss of offsite power Abnormality in reactor pressure Loss of turbine load Isolation of main steam line Pressure control system failure Abnormality in reactivity Loss of feedwater heating Inadvertent startup of AFS Reactor coolant flow control system failure Uncontrolled CR withdrawal at normal operation Uncontrolled CR withdrawal at startup Accidents... 

The covering of all major abnormal transients by these proposed models are confirmed by comparing the results obtained by them with results obtained from detailed fuel rod analyses modeling each abnormal transient event. The following eight abnormal transient events are analyzed for confirmation inadvertent startup of the auxiliary feedwater system (AFS) loss of feedwater heating loss of load without turbine bypass withdrawal of control rods at normal operation main coolant flow control system failure pressure control system failure partial loss of reactor coolant flow and loss of offsite power. 

In the other types of abnormalities, the event classification follows those of LWRs because the components such as the valves and the control rod drives are expected to be similar to those of PWRs or BWRs. In the category of the reactivity abnormality, the incidents related to the control rods are taken from those of PWRs. The loss of feedwater heating is taken like BWRs. Most of the incidents of the pressure abnormality are taken from BWRs because the Super LWR also adopts the direct steam cycle. The reactor depressurization is taken from PWRs. The abnormalities categorized into the inadvertent start or malfunction of core cooling system are taken from those of PWRs or BWRs. The inadvertent startup of AFS of the Super LWR corresponds to the inadvertent startup of ECCS of PWRs. The core coolant flow control system failure is the same as the feedwater control system failure for the Super LWR while the two incidents are different in BWRs due to the recirculation system. All the accidents categorized into the loss... 

This is a typical flow increasing transient. The demand of the main coolant flow rate is assumed to rise stepwise up to 138% of the rated flow as is assumed in the feedwater control system failure of Japanese ABWRs. Since increase in the core coolant flow rate is mild in ABWRs due to the large recirculation flow, the feed-water flow rate is assumed to increase stepwise. This assumption is too conservative for the Super LWR. The main coolant flow rate is gradually increased by the control system in the safety analysis. The calculation results are shown in Fig. 6.31. The reactor power increases with the flow rate due to water density feedback. A scram signal is released when the reactor power reaches 120% of the rated power. The maximum power is 124% while the criterion is 182%. The increase in the pressure is small. The sensitivity analysis is summarized in Table 6.15. 

In the design of the steam and feedwater system in BWRs, provision should be made to allow the adequate isolation of the system in the event of its failure or the failure of the RCS. Examples of PIEs that would affect the steam and feedwater system are the loss of condenser vacuum, the closure of all steam... 

In a BWR, an isolation coohng system for the reactor core should be provided as a standby source of cooling water to provide a capability for feedwater supply whenever the main feedwater system is isolated from the reactor pressure vessel. Abnormal events that could cause such an isolation include the inadvertent isolation of the main steam lines, the loss of condenser vacuum, the failure of a pressure regulator, the loss of feedwater and the loss of off-site power. 

To maintain a sufficient inventory of reactor coolant for core cooling during and after design basis accidents that do not involve the failure of the RCS pressure boundary (for BWRs, this applies only to appropriate parts of the steam and feedwater systems) ... 

Conventional nuclear reactors and advanced breeder reactors were America s primary energy strategy since the 1950s to resolve the fossil fuel problem but when a reactor accident occurred in 1979 at Three Mile Island in Pennsylvania, public and investor confidence in nuclear fission dropped. The accident was triggered by the failure of a feedwater pump that supplied water to the steam generators. The backup feedwater pumps were not connected to the system as required, which caused the reactor to heat up. The safety valve then failed to act which allowed a radioactive water and gas leak. This was the worst nuclear power accident in the U.S., but in this accident no one was killed and no one was directly injured. At Three Mile Island faulty instrumentation gave incorrect readings for the... 

Chemical analysis of scale deposits present on the surface of the failed clamp by X-ray diffraction revealed the presence of predominantly sodium iron oxide, sodium carbonate sodium chloride ( 10%), iron oxide and iron sulfide. The scale composition was consistent with the evaporated residue from the 80% quality steam, which had been leaking from the joint prior to the failure. The high sodium concentration in the scale was attributed to the zeolite ion exchange system used to soften the boiler feedwater, while the chlorides and sulfides were naturally present in the feedwater. 

Other components, such as main steam lines or feedwater lines, whose dis-lodgement or failure might put in jeopardy the systems mentioned in items (1) and (2) above. 

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