ML20063C926
| ML20063C926 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 01/25/1994 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9402070250 | |
| Download: ML20063C926 (4) | |
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GE Nuclear Energy
?- wnxec aut;w, t 75 cener Avenue %utu cs 951:5 January 25, 1994 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulatiori i
Subject:
Submittal Supporting Accelerated ABWR Schedule - Modelling Uncertainty in PRA Success Criteria
Dear Chet:
Enclosed are SSAR markups addressing modelling uncertainty in PRA success criteria.
These markups will be included in the next amendment.
Please provide a copy of this transmittal to Glenn Kelly.
Since cly, a 0 t)
Jatk Fox Advanced Reactor Programs cc:
Alan Beard (GE)
Norman Fletcher (DOE)
J Joe Quirk (GE)
Frank Paradiso (GE)
Jack Duncan (GE) l 1
.JNF94-006 Qo nnoa5 1s 9402070250 940125 i
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ABWR senadent saintyAnalysis Rapa A motor driven feedwater pump is combined in series with a condensate pump in order to provide a higher pressure system. Therefore, this option also depends on the availability of makeup water and electrical power. Sufficient makeup water is available to enable this series of pumps to maintain adequate core cooling for the small steam LOCA and -
transient events.
The fire protection oystem has two pumps which take suction from the firewater tanks and inject into the RPV through an RHR line. One pump is driven by an electric motor which requires AC power. The other is driven directly by a diesel engine. Once the reactor system has been depressurized, either pump can provide enough makeup water to resto i snd maintain the RPV water level following any transient (incluaing IORV) event. The analysis to support this conclusion assumes a full ADS blowdown begins within 15 minutes after the vessel water level has reached the level 1 setpoint. The subsequent reactor system depressurization allows injection from the fire protection system about 7 minutes after the start of the blowdown. The ability of the fire protection system to mitigate the consequences of LOCA events is g
conservatively ignored. For more information about the fire protection
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system refer to Subsection 5A.7.
(b) Containment Heat Removal Following the success of the core cooling function, heat must be removed from the containment. Containment heat removal is considered a success if the containment pressure is kept below the i
pressure at which loss of containment integrity is estimated to occur (Appendix 19F). Successful containment heat removal can be achieved by using the RHR System or, depending on the circumstances as defined in Table 19.S2, the normal heat removal path or the CUW System. The resultant ABWR longterm heat removal success criteria to prevent initial core damage for transient and Loss of Coolant Accident (LOCA) events with RPS scram are given in Table 19.S2.
The RHR has four major modes of operation and heat is removed from the containment in each of these modes. During the core cooling mode which is initiated automatically, the RHR heat exch9nger is in the loop and the heat removal process is established. If core cooling is accomplished without the use of an RHR System, and the suppression -
_i pool begins overheating, the suppression pool cooling rt. ode of the RHR will be automatically or manually initiated by the operator. Once initiated, an RHR System will begin removing heat from the containment and eventually terminate the pool heatup.
19 M IntemolEvent Analysis - Amendment 31
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. p-it is conservative to use the 22000F PCT ticensing limit as an acceptance criteria for the success criteria since tests have been performed which show that the core will remain in a coolable geometry with temperatures as high as l
27000F.
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A review of Table 19.3-2 shows that, for success, the inventory threatening events require the flow equivalent of only 1 RHR/LPFL or 1 HPCF pump available for large break cases and only 1 HPCF or 1 RHR/LPFL + 3 ADS.
l available for small break cases. The resulting PCTs fo_r the large break cases and transients were between 9000F and 11000F. For the small break cases with the flow equivalent of only 1 HPCF availabla the resulting PCTs were less L
than 10000F and with 1 RHR/LPFL + 3 ADS atadable the maximum PCT was 18000F.
Subsection 6.3.3.7.8 identifies the input parameters that significantly impact the LOCA results. If the above analyses were reanalyzed with these conservative input parameters, it is estimated thai only the resulting PCTs for the small break cases with 1 RHR/LPFL + 3 ADS available are above 18000F. For these cases the PCT is estimated to be about 23000F. However, even for these convervative LOCA calculations all the PCTs are less than 27000F which ir still acceptable and most LOCA cases and transients aro much less than 27000F.
Therefore, there is no need to include an uncertainty analysis in the generation of the success criteria.
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<t 19D.10 Data Uncertainty for ABWR PRA 19D.10.1 Introduction j
-i This analysis presents the results of a quantitative data uncertainty analysis for the.
Advanced Boiling Water Reactor (ABWR) Level 1 Probabilistic Risk Assessment (PRA).
1 M d= mdeFq; km'g' 7dmpleteness uncertain
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19D.10.2 Purposeqnd Summary a sah m.4 W35 maf-Modelh unce r
S JMpecfg (p, yjpg Conclusions The purpose of this study was to determine and propagate data uncertainty in the
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internal events analysis in the ABWR Level 1 PRA, to provide the probability distribution l
describing the uncertainty in the calculated core damage frequency (CDF).
Subsequent to the performance of this study, a more detailed assessment ofLevel 1 PRA importance measures revealed that the importance of the combustion turbine and each a
diesel generator had been understated. Results of these later analyses are presented in l Table 19K-1 and show theses components to rank number one and two, respectively, in Fussell-Vesely importance. These greater values ofimportance are not included in the uncertainty analyses which follow; and, these components would be expected to rank at.
the top of the list of contributors to uncertaintyin CDF presented in Table 19D.10-5. It is expected that revision of there analyses to reflect incorporation of these highervalues would not materially affect the conclusions.
The uncertainty analysis results show that the ABWR CDF has the distribution shown in Figure 19D.10-1, having a mean value of 1.56E-07 per reactor-year and an error factor of 4.2, (calculated as the 95th percentile divided by the median). The 95th percentile.
of the distribution is 2.9 times the mean value or 4.53E-07. The 5th percentile is 3.4Fe08 per reactor-year.
The top ten contributors to the uncertainty in the CDF were identified using the uncertainty importance measure. Nine of these are also in the top ten basic events ranked according to the Fussell-Vesely (F-V) importance measure. The basic event RCIMAINT (i.e., RCIC is down for test or maintenance) is the highest contributor to '
uncertainty in the CDF as well as to the mean value of the CDF. RCIC test and maintenance is part of the reliability assurance program (RAP), and is discussed in Subsection 19K.9. The remaining contributors are identified in Subsection 19D 10.6.1.
The results of the uncertainty show that the 95th percentile is only moderately sensitive to the error factors (EFs) of the basic events, and hence that lack of precise EF values has a rather small effect on the outcome. For example, doubling the EF values of each basic event simultaneously increases the 95th percentile of the CDF by only 12E When all EFs are set equal to 15, the 95th percentile increases by only 14E (Note 1 in '
Subsection 19D.10.8).
Dats Uncertaintyfor ASWR PRA-Amendmen 33 19D.10-1 a
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