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Epistemic

Aleatory. Epistemic.

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Epistemic

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  1. Aleatory Epistemic Despite over 50 years of nuclear plant operation, not all phenomenological processes to which a nuclear plant may be exposed are rigorously understood. Risk-related questions linger with regard to: (1) the likelihood of boric acid precipitation following a Large Break loss-of-coolant accident (LOCA), (2) the likelihood and impact of debris accumulation at the ECCS sump as at least a function of break size, (3) steam generator tube flaw growth, (4) reactor coolant pump seal performance under degraded conditions, (5) performance of pressure relief mechanisms following anticipated transients without scram and (6) issues associated with pressurized thermal shock. Nevertheless, a skilled PRA analyst is able to characterize relative risks of various design features and operating practices at the plant. Accurate risk insights (used to advocate an operating practice like increasing technical specification allowed outage times) must be based on a full understanding of the contributors to the PRA results and the impacts of the uncertainties. Epistemic uncertainty is the consequence of adding or removing features from the model. Modeling simplifications introduce biases and epistemic uncertainty into PRA calculations. But, even the most artful modeling will not eliminate all model uncertainty as increased model detail often relies on less precisely understood failure-modes. Raymond Schneider, Westinghouse;Steven Farkas, Hudson Global Resources

  2. Breaks can occur at any location around the typical RCS. While this is an obvious statement, many analyses used for setting success criteria are based on the limiting break location, i.e., at the bottom of the cold leg. • Large breaks can occur at • Hot Leg • Cold Leg • Crossover Leg (SG to RCP) • ECCS injection nozzles • Surge Line • Bolted Flanges • Shutdown Cooling Suction • RPV Head (Upper and Lower) • Smaller breaks can also appear at • RCP Seals • Pressurizer spray lines • Charging nozzles • SG manway seals • CRD Nozzle Penetrations • Safety relief valve seats • Drain lines • RPV instrument nozzles; RCS thermo-wells, and welded attachments • Pressurizer Heater sleeves • The likelihood of breaks is a function of the pipe and weld properties and dimensions. Westinghouse main coolant loops are constructed from stainless steel pipe. Hot leg, cross-over leg, and cold leg pipe are similar in most respects in Westinghouse plants. • CE PWRs employ stainless steel clad carbon steel main coolant loops. The CE two-loop design employs a single thick walled hot leg for each loop. As a result of the heavy walled hot leg pipe, hot leg LOCAs for CE PWRs are expected to be less likely than for similar (rated thermal power) sized WEC PWRs. Fort Calhoun Station has a unique design by having a CE design (configuration, dimensions) with an all stainless steel main coolant loop.

  3. Key Uncertainties • For the ECCS, the SI flow is directed into a common header, high or low pressure header (dependent on injection pump under consideration). The header accepts the pumped inventory and distributes the flow to the various RCS cold legs. This consideration is important when a valve located in the injection line is closed or a flow control valve (normally closed) does not open. WEC PWRs do not have flow control valves and the likelihood of an unavailable injection path is low. The header concept is important from several perspectives. • Failure or unavailability of a flowpath will increase the net system hydraulic resistance and reduce the SI flow to the vessel. Loss of injection flow may compromise event success. • Distribution of SI to all cold legs will assume that, for a cold leg LOCA, there are liquid inventory losses of injection water from the break. For example, during a large cold leg break LOCA in the ECCS headered plant with N cold legs, something greater than 1/N fractions of the injected inventory will spill directly into the containment (lower flow resistance into containment). Evolution of Mean LOCA Frequencies (Events per Plant Year) (Pipe break Contribution only) PSAs typically consider 3 or 4 break size designations, i.e., VSLOCA, SBLOCA, MBLOCA and LBLOCA. The largest break is associated with the DBA DEGB. VSLOCAs may be important to ice-condenser PWRs as a result of low containment pressure actuation setpoints for the containment-spray system as well as the need for high-pressure recirculation to mitigate the event [1] Median value [2] Median value [3] Median value [4] Includes Japanese plant data [5] Estimated values adjusted for mean (NOT FINAL). Results based on Calendar year. Lowered breaks impacted by inclusion of SGTR. Based on BWRs small piping should account for a ~ 5 x 10-4 per calendar year LOCA frequency. [6] Estimate includes SGTRs. WEC estimates that SGTR frequency to be between 0.007 to 0.019 per operating year. Variation reflects the number of SGs per plant. It is estimated that, per calendar year values decrease from SGTR. (See WCAP‑15955, “Steam Generator Tube Rupture PSA Notebook,” December 2002.

  4. Key Assumptions • The simplest PRA would estimate the frequency of any type of LOCA – lumping the very smallest hole in the RCS to the very largest. That is, make a frequency estimate of any event that directly or indirectly caused water to exit from the RCS faster than the pressurizer level control system could accommodate. • SGTRs and ISLOCAs are special classes of LOCA because instead of discharging RCS water into containment (available for recirculation), RCS water escapes into the atmosphere or into the auxiliary building respectively. • The PRA model would then have to determine the availability and reliability of the SSCs that can allow the operators to restore inventory control and establish long-term cooling. At a simple level, to achieve success (because the LOCA could be of any size), the event tree would have to include AFW and all the ECCS SSCs from HPSI to LPSI to CS to the safety-injection tanks, not to mention the support systems like air, cooling water, and electric power. • PRA analysts with even limited experience in estimating core-damage frequency (CDF) can see that this model could come up with a valid estimate of CDF, but that the model would assign nearly equal importance to all of the SSCs designed to help the operators restore inventory control and establish long-term RCS cooling. In fact, this is the situation in deterministic modeling that simplifies the PRA problem by employing “defense-in-depth” and guarding against “the worst single-active-failure.” • Some of the LOCAs are so small that the availability and reliability of accumulator tanks are not relevant to a large fraction of the LOCAs captured by the lumped LOCA frequency. A sophisticated PRA model will not penalize the results by requiring all SITs to be available and reliable for SBLOCAs and SGTRs. In fact, including SITs in the SSC success set for small LOCAs has the perverse effect of lowering the core-damage frequency estimate because the SITs are inherently reliable. • The PRA model can instead set up LOCA initiators that occur at a nearly infinite number of locations around the RCS. Fortunately for the PSA analyst, there relatively few permutations of AFW and ECCS equipment that can successfully restore RCS inventory and maintain long-term cooling. A rigorous model would be built by determining the frequency of LOCA break sizes that can be accommodated by each permutation of SSCs that can successfully restore RCS inventory and maintain long-term cooling. • The LOCA break sizes that can be accommodated by a particular set of SSCs depends on assumptions surrounding the behavior of the water contained in the RCS after the break occurs. There are a few key features of a thermo-hydraulic model that dictate whether or not an SSC will be helpful in restoring inventory control and establishing long-term cooling. For instance the size of the hole dictates how quickly RCS pressure falls below the shutoff head of HPSI pumps. The location of the break changes how fast the water level in the core barrel drops (e.g., cold-leg breaks empty the reactor pressure vessel faster than a hot-leg break). More subtle assumptions in the thermo-hydraulic model determine the mass of water in the lower plenum of the reactor vessel at the end of the blowdown phase. That amount of water determines how much water the ECCS has to put back into the RCS to recover the core. Of course, neutronics dictates how long the fuel can remain uncovered yet retain its structural integrity, and thus the flow rate ECCS needs to achieve in order to avoid core-damage. • Some of the LOCAs are a result of general transients that cause the primary safety-relief valves to lift. As those types of valves have a random chance of sticking open, the general transient can induce a LOCA putting demands on the same large set of SSCs mentioned above.

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