Probabilistic Calculations 22.39 Elements of Reactor Design, Operations, and Safety Lectures 10-11 Fall 2006 George E. Apostolakis Massachusetts Institute of Technology Department of Nuclear Science and Engineering 1 General Formulation XT = φ(X1,…Xn) ≡ φ(X) N N 1 1 X T = 1 − ∏ (1 − M i ) ≡ C M i N N −1 N N +1 N XT = ∑ Mi − ∑ ∑ Mi M j + ... + (−1) i =1 i =1 j =i +1 ∏ Mi i =1 XT : the TOP event indicator variable Mi : the ith minimal cut set or accident sequence Department of Nuclear Science and Engineering 2 TOP-event Probability N P (X T ) = ∑ P(M i ) + K + (− 1) N +1 1 ⎛N ⎞ P⎜ ∏ M i ⎟ ⎝ 1 ⎠ N P(XT ) ≅ ∑ P (Mi ) Rare-event approximation 1 The question is how to calculate the probability of Mi P ( M i ) = P ( X ik ... X im ) P (A B ) ≡ Conditional probability: P ( AB ) P (B ) In d e p e n d en t e v e n ts: P (A B ) = P (A ) P ( A B ) = P (A ) P ( B ) Department of Nuclear Science and Engineering 3 MinCutSet Probability P ( M ) = P ( X 1X 2 X 3 ) = P ( X 1 ) P ( X 2 X 3 / X 1 ) = = P ( X 1 ) P ( X 2 / X 1 ) P ( X 3 / X 1X 2 ) For independent events: P ( M ) = P ( X 1X 2 X 3 ) = P ( X 1 ) P ( X 2 ) P ( X 3 ) For accident sequences, we must include the initiating-event frequency per year: fr(M) = fr(IEX1X2 ) = fr(IE)P( X1X2 / IE) = fr(IE)P( X1 / IE)P( X2 / IEX1 ) N fr (X T ) ≡ CDF ≅ ∑ fr (M i ) 1 Department of Nuclear Science and Engineering 4 Example: 2-out-of-4 System 1 2 M 1 = X1 X 2 X 3 M2 = X2 X3 X4 M 3 = X3 X 4 X 1 M4 = X1 X2 X4 3 4 XT = 1 – (1 – M1) (1 – M2) (1 – M3) (1 – M4) XT = (X1 X2 X3 + X2 X3 X4 + X3 X4 X1 + X1 X2 X4) - 3X1 X2 X3X4 Department of Nuclear Science and Engineering 5 2-out-of-4 System (cont’d) P(XT = 1) = P(X1 X2 X3 + X2 X3 X4 + X3 X4 X1 + X1 X2 X4) – 3P(X1 X2 X3X4) Assume that the components are independent and nominally identical with failure probability q. Then, P(XT = 1) = 4q3 – 3q4 Rare-event approximation: P(XT = 1) ≅ 4q3 Department of Nuclear Science and Engineering 6 Overview • We need models for: ¾ The frequency of initiating events. ¾ The probability that a component will fail on demand. ¾ The probability that a component will run for a period of time given a successful start. Department of Nuclear Science and Engineering 7 Initiating Events: The Poisson Distribution • Used typically to model the occurrence of initiating events. • Discrete Random Variable: Number of events in (0, t) • The rate λ is assumed to be constant; the events are independent. • The probability of exactly k events in (0, t) is (pmf): k −λt Pr[k ] = e k! ≡ 1*2*…*(k-1)*k (λ t ) 0! = 1 k! 2 m = λt σ = λt Department of Nuclear Science and Engineering 8 Example of the Poisson Distribution • A component fails due to "shocks" that occur, on the average, once every 100 hours. What is the probability of exactly one replacement in 100 hours? Of no replacement? • λ t = 10-2*100 = 1 • Pr[1 repl.] = e-λt = e-1 = 0.37 = Pr[no replacement] • Expected number of replacements: 1 2 −1 1 e −1 = = 0.185 Pr[ 2repl] = e 2! 2 Pr[k≤2] = 0.37 + 0.37 + 0.185 = 0.925 Department of Nuclear Science and Engineering 9 Reliability and Availability • Reliability: Probability of successful operation over a period (0, t). • Availability: Probability the item is working at time t. • Note: ¾ In industrial applications, the term “reliability” includes the probability that a safety system will start successfully and operate for a period (0, t). ¾ The term “unavailability” usually refers to maintenance. Department of Nuclear Science and Engineering 10 Failure while running • T: the time to failure of a component (continuous random variable). • F(t) = P[T < t]: failure distribution (unreliability) • R(t) ≡ 1-F(t) = P[t < T]: reliability • m: mean time to failure (MTTF) • f(t): failure density, f(t)dt = P{failure occurs between t and t+dt} = P [t < T < t+dt] Department of Nuclear Science and Engineering 11 The Hazard Function or Failure Rate f (t ) f (t ) h(t ) ≡ = R (t ) 1 − F( t ) ⎞ ⎛ t F(t ) = 1 − exp⎜⎜ − ∫ h(s )ds ⎟⎟. ⎠ ⎝ 0 The distinction between h(t) and f(t) : f(t)dt: unconditional probability of failure in (t, t +dt), f(t)dt = P [t < T < t+dt] h(t)dt: conditional probability of failure in (t, t +dt) given that the component has survived up to t. h(t)dt = P [t < T < t+dt/{ t < T}] Department of Nuclear Science and Engineering 12 The “Bathtub” Curve h(t) I 0 II t1 III t2 I Infant Mortality II Useful Life III Aging (Wear-out) Department of Nuclear Science and Engineering t 13 The Exponential Distribution • f(t) = • λe − λt F( t ) = 1 − e λ> 0 t>0 − λt (failure density) R(t ) = e − λt • h(t) = λ constant (no memory; the only pdf with this property) ⇒ useful life on bathtub curve F(t) ≅ λt for λ t < 0.1 (another rare-event approximation) 1 m= =σ λ Department of Nuclear Science and Engineering 14 Example: 2-out-of-3 system Each sensor has a MTTF equal to 2,000 hours. What is the unreliability of the system for a period of 720 hours? • Step 1: System Logic. XT = (XAXB+XBXC+XCXA) - 2XAXBXC Department of Nuclear Science and Engineering 15 Example: 2-out-of-3 system (2) Step 2: Probabilistic Analysis. For nominally identical components: P(XT) = 3q2 – 2q3 q = F(t) = 1− e−λt System Unreliability: Rare event approximation: λ = 5 x10 −4 hr −1 ( FT ( t ) = 3 1 − e ) − λt 2 ( − 2 1− e ) − λt 3 FT ( t ) ≅ 3( λt ) 2 − 2( λt )3 Department of Nuclear Science and Engineering 16 A note on the calculation of the MTTF ∞ MTTF = ∫ R (t )dt 0 Proof ∞ ∞ ∞ dR )dt = − ∫ tdR = MTTF = ∫ tf (t )dt = ∫ t( − dt 0 0 0 ∞ 0 ∞ ∞ 0 0 = − tR + ∫ R(t )dt = ∫ R(t )dt Department of Nuclear Science and Engineering 17 A note on the calculation of the MTTF (cont.) since dF d(1 − R ) dR = f (t ) = =− dt dt dt and R(t → ∞ ) → 0 faster than Department of Nuclear Science and Engineering t→∞ 18 MTTF Examples ∞ Single exponential component: MTTF = ∫ e −λt 0 Series system: ∞ MTTF = ∫ dte− mλt = 0 1-out-of-2 system : ∞ ( 1 dt = λ 1 1 ≡ mλ λsystem MTTF = ∫ dt 2 e − λt 0 −e − 2 λt ) 3 = 2λ 2-out-of-3 system : ∞ ∞ 0 0 −λt 2 MTTF= ∫ RT (t )dt = ∫ [1 − 3(1 − e 5 ) ]dt = 6λ −λt 3 ) + 2(1 − e Department of Nuclear Science and Engineering 19 MTTF Examples: 2-out-of-3 System Using the result for FT(t) on slide 15, we get ∞ ∞ 0 0 MTTF = ∫ R T (t )dt = ∫ [1 − 3(1 − e −λt )2 + 2(1 − e −λt )3 ]dt 1 1 5 + = MTTF = 2λ 3λ 6λ 1 The MTTF for a single exponential component is: λ ⇒ The 2-out-of-3 system is slightly worse. Department of Nuclear Science and Engineering 20 The Weibull failure model Weibull Hazard Rate Curves Adjusting the value of b, we can model any part of the bathtub curve. 0.0035 b=0.8 b=1.5 0.003 b=1.0 b=2.0 0.0025 0.002 0.0015 0.001 0.0005 0 b b −1 0 200 400 600 800 1000 h ( t ) = bλ t R(t ) = e −(λt ) b For b = 1 ⇒ the exponential distribution. Department of Nuclear Science and Engineering 21 The Model of the World ¾ Deterministic, e.g., a mechanistic computer code ¾ Probabilistic (Aleatory), e.g., R(t/λ) = exp(- λt) ¾ Both deterministic and aleatory models of the world have assumptions and parameters. ¾ How confident are we about the validity of these assumptions and the numerical values of the parameters? Department of Nuclear Science and Engineering 22 The Epistemic (state-of-knowledge) Model • Uncertainties in assumptions are not handled routinely. If necessary, sensitivity studies are performed. • Parameter uncertainties are reflected on appropriate probability distributions. • For the failure rate: π(λ) dλ = Pr(the failure rate has a value in dλ about λ) Department of Nuclear Science and Engineering 23 Unconditional (predictive) probability R(t ) = ∫ R(t / λ )π (λ )dλ Department of Nuclear Science and Engineering 24 Communication of Epistemic Uncertainties: The discrete case Suppose that P(λ = 10-2) = 0.4 and P(λ = 10-3) = 0.6 Then, P(e-0.001t) = 0.6 and P(e-0.01t) = 0.4 R(t) = 0.6 e-0.001t + 0.4 e-0.001t 1.0 0.6 exp(-0.001t) 0.4 exp(-0.01t) t Department of Nuclear Science and Engineering 25 Communication of Epistemic Uncertainties: The continuous case Courtesy of US NRC. Department of Nuclear Science and Engineering 26 The lognormal distribution • It is very common to use the lognormal distribution as the epistemic distribution of failure rates. ⎡ (ln λ − μ )2 ⎤ 1 π( λ ) = exp⎢ − ⎥ 2 2πσλ 2σ ⎦ ⎣ ⎡ σ2 ⎤ m = exp ⎢μ + ⎥ 2⎦ ⎣ λ 95 = eμ +1.645σ median : λ 50 = eμ λ 05 = eμ −1.645σ λ 95 λ 95 λ 50 EF = = = λ 05 λ 50 λ 05 Y = ln λ Y is normally distributed with mean μ and standard deviation σ Department of Nuclear Science and Engineering 27 Courtesy of US NRC. 28 Department of Nuclear Science and Engineering Courtesy of US NRC. Department of Nuclear Science and Engineering 29 SIMPLIFIED SYSTEM DIAGRAM Courtesy of US NRC. Department of Nuclear Science and Engineering 30 HIGH PRESSURE INJECTION DURING LOOP 1-0F-3 TRAINS FOR SUCCESS Courtesy of US NRC. Department of Nuclear Science and Engineering 31 HPIS Analysis (1-out-of-3) • In the RSS HPIS, the three pump trains have a common suction line from the RWST. The South Texas Project design has separate suction lines for the three trains, as the fault tree shows. • Qtotal = Qsingles + QdoubleFail’s + Qtest&maint + QCCF • Representative single failures (single-element mcs): ¾ ¾ ¾ ¾ • Check valve SI 225 fails to open Check valve SI-25 fails to open RWST discharge line ruptures Other Qsingles = 1.1x10-3 (“point estimate”) Department of Nuclear Science and Engineering 32 HPIS: Double Failures • Representative double failures (double-element mcs): ¾ RWST supply MOVs 1115B and 1115D fail to open ¾ Born Injection Tank (BIT) inlet MOVs 1867A and 1867B fail to open ¾ BIT discharge MOVs 1867C and 1867D fail to open ¾ Service water pumps; cooling water pumps; BIT cooling system ¾ other • Q(MOVs 1867C and 1867D fail to open) = P(X1) P(X2) where P(Xi) is a lognormal with median 1.9x10-2 and EF = 3 • QdoubleFail’s = 2.5x10-3 (“point estimate”) Department of Nuclear Science and Engineering 33 HPIS: Other Contributions • Qtest&maint is negligible because of the 1-out-of-3 redundancy (if one train is out, double failures must occur for the system to fail). • QCCF(MOVs 1867C and 1867D fail to open) = βP(X1) = = 0.075x1.9x10-2 = 1.4x10-3 • Monte Carlo simulation yields (RSS): ¾ Qtotal,median = 8.6x10-3 ¾ Qtotal,upper = 2.7x10-2 ¾ Qtotal,lower = 4.4x10-3 Department of Nuclear Science and Engineering 34 In some important cases, ΔCDF and ΔLERF cannot be calculated. SSC Categories Based on Importance Measures Decision Options Impact on CDF and LERF Expert Panel Delibera tion Department of Nuclear Science and Engineering RiskInformed Decision 35 Fussell-Vesely Importance Measure FVi = R0 (i ) Mk R −i Pr[U Mk( i )] k R 0 −i 0 − R −i R R = = 1− 0 0 R R The base-case risk metric (CDF or LERF) = Pr[ U M k ] k The kth accident sequence containing event i The risk metric (CDF or LERF) with the ith component up (unavailability equal to zero) Department of Nuclear Science and Engineering 36 Risk Reduction Worth (RRW) R0 RRW i = − i R R0 − R −i R −i 1 FVi = = 1− 0 = 1− 0 RRWi R R •FVi is the fractional decrease in the risk metric when event i is always true (component i is always available; its unavailability is set equal to zero). •This importance measure is particularly useful for identifying improvements to the reliability of elements which can most reduce risk. Department of Nuclear Science and Engineering 37 F-V Ranking Loss Of Offsite Power Initiating Event DIESEL GENERATOR B FAILS DIESEL GENERATOR A FAILS COMMON CAUSE FAILURE OF DIESEL GENERATORS OPERATOR FAILS TO RECOVER OFFSITE POWER (SEAL LOCA) RCP SEALS FAIL W/O COOLING AND INJECTION OPERATOR FAILS TO RECOVER OFFSITE POWER BEFORE BATTERY DEPLETION 0.831 0.437 0.393 0.39 0.388 0.344 0.306 Department of Nuclear Science and Engineering 38 Risk Achievement Worth (RAW) RAW i = R+i R +i R 0 The risk metric (CDF or LERF) with the ith component always down (its unavailability is set equal to 1) RAW presents a measure of the “worth” of the basic event in “achieving” the present level of risk and indicates the importance of maintaining the current level of reliability for the basic event. Department of Nuclear Science and Engineering 39 RAW Ranking Loss Of Offsite Power Initiating Event 51,940 Steam Generator Tube Rupture Initiating Event 41,200 Small Loss Of Coolant Accident Initiating Event 40,300 CONTROL ROD ASSEMBLIES FAIL TO INSERT 3,050 COMMON CAUSE FAILURE OF DIESEL GENERATORS RPS BREAKERS FAIL TO OPEN Department of Nuclear Science and Engineering 271 202 40 Comments on Importance Measures • Importance measures are typically evaluated for individual SSCs, not groups. • The various categories of risk significance are determined by defining threshold values for the importance measures. For example, in some applications, a SSC is in the "high" risk-significant category when FV > 0.005 and RAW > 2.0. • Importance measures are strongly affected by the scope and quality of the PRA. For example, incomplete assessments of risk contributions from lowpower and shutdown operations, fires, and human performance will distort the importance measures. Department of Nuclear Science and Engineering 41