ML073520401

License Amendment Request 256, One-Time Extension of Containment Integrated Leakage Test Interval, Impact on the Point Beach Nuclear Plant Large Early Release Frequency (LERF) Due to Level 2 Modeling Enhancements, Enclosure 4

ML073520401
Person / Time
Site:	Point Beach
Issue date:	12/06/2007
From:	Florida Power & Light Energy Point Beach
To:	Office of Nuclear Reactor Regulation
References
Download: ML073520401 (36)

Text

ENCLOSURE4 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 256 ONE-TIME EXTENSION OF CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL IMPACT ON THE POINT BEACH NUCLEAR PLANT LARGE EARLY RELEASE FREQUENCY (LERF) DUE TO LEVEL 2 MODELING ENHANCEMENTS 35 pages follow

White, Papr Impac on th Point Beach Nuclear Plant L.arge Early Release Frequency (LERF)

Due to Level 2 Model*ng.0 Enhancements Date Checked. by Date:

PointBeach Nuclear Plant - White Paper LERF Impact Due to Enhanced Level 2 Model

1. Introduction This white paper describes the impact on the Point Beach Nuclear Plant Large Early Release Frequency (LERF) due to Level 2 modeling enhancements.

1.1. Basis for Level 2 Change During a review of the Level 2 model, it was noted that there was an error in a Decomposition Event Tree (DET) that provides sorting for the Containment Event Tree

.(CET). The error caused sequences with successful Auxiliary Feedwater (AFW) or successful Main Feedwater (MFW) to be classified as "dry", and thus subject to a possible induced steam generator tube rupture (ISGTR).

The Plant Damage Diagram (PDD) questioned secondary cooling, and if it was successful it labeled the response to the heading as "Yes" and unsuccessful as "No". In one of the DETs that support the CET, the responses from the PDD were queried. However, the query asked whether the response was "S" for success or "F" for failure. These would be the appropriate responses from the Level 1.trees, but not from the PDD. Thus, the DET query did not find "Yes" or "No". Because the default DET response was "No", the CET incorrectly binned sequences with successful AFW or MFW as "dry". The correction was to havethe DET query for "Yes" and "No" versus "S" and "F". The correction of this error will reduce the frequency of sequences that are subject to possible ISGTR, and thus LERF.

1.2. Basis for Estimation of ISGTR The Point Beach Nuclear Plant Level 2 PRA analysis employs a value for the conditional probability of an ISGTR based upon the model developed in NUREG-1570. The conditional probability of an ISGTR is applied to frequencies of high RCS pressure/dry steam generator ("high/dry") sequences to arrive at the frequency of an ISGTR. An ISGTR is a containment bypass event, and is generally considered to contribute to the Large Early Release Frequency (LERF).

The Accident Progression Event Tree (APET) that was developed in NUREG-1570 to estimate the conditional probability of an ISGTR was converted into a fault tree format and enhanced to provide a more plant-specific result for the Point Beach Nuclear Plant. The plant-specific enhancements include: 1) The model was modified to account for plant-specific two steam generators, 2) The probability of high/dry sequences occurring concurrent with an RCP seal LOCA was made plant-specific, 3) The probability of pressure-induced steam generator tube ruptures (PI-SGTRs) was made more plant-specific, and 4) The probability of thermally-induced steam generator tube ruptures (TI-SGTRs) was made more plant-specific. These plant-specific enhancements will reduce the conditional probability of an ISGTR, and thus LERF.

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PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model The bases for, and the calculations of, the new estimates for these enhancements are provided in the sections below.

1.3. Level 2 Input from Extended Level I Model The Point Beach Unit 1 and Unit 2 extended event tree sequence models (see white paper discussing the extended event tree quantification) provide the input to the Unit 1 and Unit 2 Level 2 PRA models. The input takes the form of extended sequence frequencies and event tree heading failures and successes.

2. New ISGTR Variable Probabilities 2.1. High/Dry Sequences with Seal LOCA The APET that was developed in NUREG-1570 to estimate the conditional probability of an ISGTR included a top event that questioned whether a high/dry core damage sequence occurred concurrent with an RCP seal LOCA. If an RCP seal LOCA occurred, then the APET questioned whether loop seal clearing took place (APET branches B 1, B2, and B3).

The probability of a TI-SGTR is estimated as unity (1.0) in any steam generator attached to an RCS loop with a cleared loop seal.

The probability that a high/dry core damage sequence occurred concurrent with an RCP seal LOCA was estimated as 0.211 in NUREG-1570, based on the Surry Plant PRA core damage results, which are dominated by a Station Blackout (SBO) event.

The Point Beach Nuclear Plant PRA core damage results are dominated by a Loss of Service Water event. The RCP seals can be adequately cooled by one of three positive displacement charging pumps, which do not rely on component cooling water/service water for cooling. Thus, for Point Beach, the probability that a high/dry core damage sequence occurs concurrent with an RCP seal LOCA is significantly less than the probability calculated using the Surry results.

Table 1 lists the Point Beach Unit 2 Extended Level 1 Event Tree high/dry sequences with concurrent RCP seal LOCA. These events are SBO, Loss of Offsite Power (T1), and Loss of Service Water (TSW). Note that the Loss of Component Cooling Water (TCC) event also has sequences in which an RCP seal LOCA can occur concurrent with a dry steam generator. However, these TCC sequences have very low frequencies and were not included in this assessment.

For an SBO event, the Table 1 list includes all sequences with failure of auxiliary feedwater (AFW), all sequences wherein the core was uncovered at the time AC power was recovered, and all sequences where AC power was not recovered. All of these sequences Page 2 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model will exhibit an RCP seal LOCA and would most likely have dry steam generators. The sum of these frequencies is 1.22E-7/yr.

For a T1 event, the Table 1 list includes all sequences with failure of AFW. However, failure of RCP seal cooling is not guaranteed for these sequences, it is merely compromised. A plant-specific RCP seal cooling failure probability of 0.004 will be applied to the Table 1 TI frequencies that exhibit successful Emergency Diesel Generator operation to account for failure of RCP seal cooling, which would result in an RCP seal LOCA. A plant-specific RCP seal cooling failure probability of 0.02 will be applied to the Table 1 T1 frequencies that exhibit successful Gas Turbine operation to account for failure of RCP seal cooling. The sum of these frequencies, including the application of the 0.004 and 0.02 probabilities, is 1.83E-8/yr.

For a TSW event, the Table 1 list includes all sequences with failure of charging injection.

Failure of charging injection, together with the loss of service water, results in an RCP seal LOCA. Note that failure of AFW is not guaranteed for these sequences, although AFW may be compromised by the loss of service water. It was conservatively assumed, however, that all these sequences included failure of AFW for this analysis. The sum of these frequencies is 4.28E-8/yr.

The combined frequency of these high/dry sequences with concurrent RCP seal LOCA is 1.83E-7/yr.

The Point Beach Unit 2 Level 2 plant damage diagram provides a total high/dry frequency of 4.21E-5/yr.

Therefore, the probability of a high/dry sequence having a concurrent RCP seal LOCA is (1.83E-7)/(4;21E-5) = 0.004.

The Point Beach Unit 1 Level 1 and Level 2 results are very similar to the Unit 2 results.

Therefore, the estimated probability from above will also be applied in the Unit 1 ISGTR estimation.

2.2. PI-SGTR The APET that was developed in NUREG-1570 to estimate the conditional probability of an ISGTR utilized the probability of a pressure-induced SGTR (PI-SGTR) that was based on an NRC RES Branch-developed flaw distribution for steam generators with "moderate" degradation.

The probability of a PI-SGTR for a high/dry sequence was estimated in NUREG-1570 as 0.0549 per depressurized steam generator, based on ihe RES-developed flaw distribution.

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PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model The NRC recognized the sensitivity of the ISGTR results to the assumed flaw distribution.

Two analyses were conducted in NUREG-1570 to address this sensitivity.

The first sensitivity analysis assumed the steam generator tubes were in pristine condition (i.e., all steam generator flaws were eliminated or the steam generators were replaced).

The results of this case are contained in NUREG-1570 Table 5.8 as Case 8. The frequency of a PI-SGTR is shown to be 0/yr.

The second sensitivity analysis was based on an NRC NRR Branch-developed flaw distribution for steam generators for an "average" plant. NUREG-1570 Section 5.3.3 states that "the pressure-induced tube failure probabilities using the NRR distribution are an order of magnitude lower than those on the basis of the RES distribution."

The Point Beach Unit 1 steam generators were replaced in the mid-1 980s and the Unit 2 steam generators were replaced in the mid-1990s. None of the generators have experienced any significant degradation. The current number of tubes plugged is provided below:

Point Beach Unit 1 Steam Generators (SGs)

SG A (3214 total tubes) - 4 tubes plugged = 0.1245%

SG B (3214 total tubes) - 6 tubes plugged = 0.1867%

Point Beach Unit 2 SGs SG A (3499 total tubes) - 0 tubes plugged = 0%

SGB (3499 total tubes)- 4 tubes plugged = 0.1143%

The insignificant amount of degradation experienced by the Point Beach replacement steam generators suggests that the PI-SGTR probability can be reduced.

The sensitivity analyses presented in NUREG-1570 would indicate a percent reduction in the probability of PI-SGTR ranges between 100% reduction (Case 8) and 90% reduction (NRR-developed flaw distribution) for steam generators with insignificant amounts of degradation.

A reduction of 90% (i.e., a multiplier of 0.1) will be applied to the probability of a PI-SGTR in this evaluation.

2.3. TI-SGTR The APET that was developed in NUREG-1570 to estimate the conditional probability of an ISGTR utilized the probability of a thermally-induced SGTR (TI-SGTR) that was based on an NRC RES Branch-developed flaw distribution for steam generators with "moderate" degradation. The TI-SGTR probability varied depending on the condition of the RCS, Page 4 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model steam generators, and RCS loop seals. Note that the probability of a TI-SGTR is estimated as unity (1.0) in any steam generator attached to an RCS loop with a cleared loop seal.

The NRC recognized the sensitivity of the ISGTR results to the assumed flaw distribution.

Two analyses were conducted in NUREG-1570 to address this sensitivity.

The first sensitivity analysis assumed the steam generator tubes were in pristine condition (i.e., all steam generator flaws were eliminated or the steam generators were replaced).

The results of this case are contained in NUREG-1570 Table 5.8 as Case 8. The frequency of a TI-SGTR decreased approximately 50% for this case.

The second sensitivity analysis was based on an NRC NRR Branch-developed flaw distribution for steam generators for an "average" plant. NUREG-1570 Table 5.6 shows that the TI-SGTR probability was reduced by 35% (first Case 9R listing in Table 5.6) to 70% (Case 6N). Note that there was no reduction to the TI-SGTR probability for a cleared loop seal.

The insignificant amount of degradation experienced by the Point Beach replacement steam generators, as shown in Section 2.2 above, suggests that the TI-SGTR probability can be reduced.

The sensitivity analyses presented in NUREG-1570 would indicate a percent reduction in the probability of TI-SGTR ranges between 70% reduction and 35% reduction for steam generators with insignificant amounts of degradation.

A reduction of 33%. (i.e., a multiplier of 0.666) willbe applied to the probability of a TI-SGTR in this evaluation, except where the TI-SGTR probability is 1.0. The probability will not be reduced in this case.

3. APET Fault Tree Development The APET that was developed in NUREG-1570 was developed into a fault tree so that a new estimation of the ISGTR probability could be generated using the Point Beach plant-specific values generated in Sections 2.1, 2.2, and 2.3, above.

Figure 1 provides this fault tree. All APET RC-1 end states were captured in the fault tree.

The APET was modified to account for the Point Beach plant-specific two steam generators. The following items describe the changes made to the APET, based on the APET functional headings.

1. APET heading "A" is the high/dry frequency. For Point Beach, this is applied in the Level 2 analysis, so the value is not included in the fault tree.

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PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model

2. APET heading "B" accounts for stuck open pressurizer PORVs that would depressurize the RCS. The value used is based on Surry and was retained.

3. APET heading "C" accounts for RCP seal LOCA. Section 2.1, above, derived a plant-specific value for the conditional probability of having a concurrent RCP seal LOCA for a high/dry sequence of 0.004. This was applied to heading "C".

4. APET heading "D" accounts for all steam generators being depressurized due to stuck open MSSVs. The NUREG describes the bases for the value used, and generally used the Surry results. However, the NUREG also essentially duplicated the results by assuming a 40% probability that each steam generator could be depressurized due to a stuck open MSSV. This value was used to estimate a probability of 0.16 for two depressurized steam generators.

5. APET heading "E" accounts for one steam generator being depressurized due to a stuck open MSSV, given that all steam generators are not depressurized due to the same reason. The NTJREG continued to use the Surry results for this heading.

Since the NUREG values approximate a 50/50 split fraction, the NUREG values were retained for a two steam generator plant.

6. APET heading "F" accounts for late RCS depressurization. The NUREG basically states that they decided to use a 50/50 split fraction. This value was not changed.

7. APET heading "G" accounts for depressurized steam generators due to leaking MSIVs. The NUREG again states that they decided to use a 50% probability that 1 or more steam generators would be depressurized due to leaking MSIVs. The same split fractions were used for Point Beach.

8. APET heading "H" accounts for one of three depressurized steam generators, given that one or more are depressurized due to leaking MSIVs. For the original NUREG APET values, if there is a 50% probability that one or more (up to three) steam generators are depressurized, the probability of being depressurized is 0.206.3 per SG. So, the probability of one of three steam generators leaking is (0.2063)*(1-0.2063)A2*3 - 0.39. This is conditional given heading "G", so the value must be multiplied by the inverse of the heading "G" probability, so00.39

• 2 = 0.78 (this in fact matches the APET value for three steam generators). For two steam generators, given a 50% probability that one or more (up to two) steam generators are depressurized, the probability of being depressurized is 0.2929 per SG. So, the probability of one of two steam generators leaking is (0.2929)*(1-0.2929)*2 =

0.414. This is conditional given heading "G", so the value must be multiplied by the inverse of the heading "G" probability, so 0.414

• 2 = 0.83.

9. APET heading "I" continues the process to split out the two or three leaking steam generators. For a plant with only two steam generators, this heading is deleted.

10. Note that the description of items 7, 8, and 9 (for headings "G", "H", and "I")

applies to APET branches Al, B1, and Cl. For branches A2, B2, and C2, one steam generator is already depressurized due to heading "E". The changes for these branches are fairly intuitive, given the description of items 7, 8, and 9. For branches A3, B3, and C3, all steam generators are already depressurized due to heading "D", thus headings "G", "H", and "I"do not apply.

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Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model

11. APET heading "J" accounts for the probability of a pressure induced SGTR (PI-SGTR). Because of the earlier branching, the only changes required were for branches A3, B3, and C3. For these branches, all steam generators are depressurized due to heading "D". The NUREG APET used a probability of a PI-SGTR based on three steam generators. For a two steam generator plant, the probability of a PI-SGTR based on two steam generators is appropriate. (Note: The original two steam generator PI-SGTR probabilities were used in the fault tree. The Section 2.2 plant-specific reduction was made in a sensitivity run).

12. APET heading "K" accounts for loop seal clearing. A cleared loop seal in the loop with a depressurized steam generator results in a guaranteed thermally induced SGTR (TI-SGTR) (i.e., probability of TI-SGTR = 1.0). Loop seal clearing occurs for seal LOCA branches B 1, B2, and B3. The NUREG assumes one loop seal clearing will occur. Thus, if one of three steam generators is depressurized, the probability of the cleared loop seal occurring in that loop is 0.333. If two of three steam generators are depressurized, the probability of the cleared loop seal occurring in those loops is 0.667. For three depressurized steam generators, the probability is guaranteed (1.0). For a two steam generator plant, the probabilities are 50/50 or 1.0.

13. APET heading "L" accounts for the probability of a TI-SGTR. Because of the earlier branching, the only changes required were for branches A3, B3, and C3. For these branches, all steam generators are depressurized due to heading "D". The NUREG APET used a probability of a TI-SGTR based on three steam generators.

For a two steam generator plant, the probability of a TI-SGTR based on two steam generators is appropriate. Note that branch B3 used a TI-SGTR probability of 1.0, so this was not changed. (Note: The original two steam generator TI-SGTR probabilities were used in the fault tree. The Section 2.3 plant-specific.reduction was made in a sensitivity run).

14. APET heading "M" accounts for holdup of fission products. If they are held up, the release is mitigated. This heading is applied to branches with a seal LOCA and the primary is depressurized. It was not clear in the NUREG how.this value was calculated. However, the values are-essentially 50/50 split fractions and were retained.

3.1. Quantified Fault Tree When the enhanced fault tree is solved, the conditional probability of an ISGTR is calculated to be 0.1026. This should be compared to the NUREG value of 0.25.

Approximately 45% of the reduction is due to two versus three steam generators (or approximately 27% of the NUIREG-1570 probability of 0.25) and approximately 55% of the reduction is due to the plant-specific RCP seal LOCA split fraction (or approximately 32% of the NUJREG-1570 probability of 0.25).

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Point Beach Nuclear Plant - White Paper LERF Impact Due to Enhanced Level 2 Model 3.2. Point Beach Conditional ISGTR Probability A sensitivity analysis was conducted using the quantified fault tree described in Section 3 to develop the conditional ISGTR probability for Point Beach.

Figure 2 presents the inputs to, and results of, the sensitivity study.

A Point Beach plant-specific multiplier of 0.1, as determined in Section 2.2, above, was applied to the PI-SGTR probabilities, resulting in the substitute values shown in Figure 2.

The complement values were also substituted. Using a plant-specific value for PI-SGTR reduced the Section 3.1 calculated conditional ISGTR probability of 0.1026 by approximately 50%, or approximately 20% of the NUREG-1570 probability of 0.25.

A Point Beach plant-specific multiplier of 0.666, as determined in Section 2.3, above, was applied to the TI-SGTR probabilities, resulting in the substitute values shown in Figure 2.

Again, note that there was no reduction to the TI-SGTR probability (1.0) for a cleared loop seal. Using a plant-specific value for TI-SGTR reduced the Section 3.1 calculated conditional ISGTR probability of 0.1026 by another approximately 14%, or approximately 6% of the NUREG-1570 probability of 0.25.

Combined, the plant-specific PI-SGTR and TI-SGTR resulted in approximately a 64%

reduction in the Section 3.1 calculated conditional ISGTR probability of 0.1026 Figure 2 shows that the Point Beach plant-specific conditional ISGTR probability is 0.037.

This is approximately an 85% reduction in the NUREG-1570 conditional ISGTR probability.

4. Enhanced Level 2 PRA Model Quantification 4.1. Unit 1 As discussed in Section 1.1, the Level 2 decomposition event tree (DET) ISGTR.DET was modified to correct an error which caused sequences with successful Auxiliary Feedwater (AFW) or successful Main Feedwater (MFW) to be classified as "dry", and thus subject to a possible induced steam generator tube rupture (ISGTR). This correction reduced the frequency of high/dry sequences that are subject to possible ISGTR by approximately 10%.

An additional error correction was incorporated into the Level 2 DET HIWTRCAV.DET.

This error incorrectly classified Small and Medium LOCAs as having no water in the cavity. The correction of this error slightly increased the frequency of CET end point number 5, a containment liner failure that is considered an early containment failure. This correction produced a very minor increase (<<1%) in LERF.

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PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model The Level 2 DET ISGTR.,DET was modified to incorporate the Point Beach conditional ISGTR probability of 0.037, calculated in Section 3.2.

The Point Beach Unit 1 Level 2 model was then requantified. Figure 3 presents the requantified Unit 1 containment event tree (CET) results. Figure 4 presents the requantified Unit 1 source term category (STC) results.

4.2. Unit 2 The Point Beach Unit 2 Level 2 model was modified as described in Section 4.1 and then requantified.

Figure 5 presents the requantified Unit 2 CET results. Figure 6 presents the requantified Unit 2 STC results.

5. Enhanced Level 2 Results Discussion 5.1. Unit 1 Source Term Categories (STCs) 5, 6, 7, and 8, as developed in the Level 2 PRA and presented for Unit 1 in Figure 4, represent the Large Early Release categories. The sum of the frequencies of these STCs represents the Unit 1 Large Early Release Frequency (LERF).

The Point Beach Unit 1 internal events LERF, using the enhanced Level 2 model discussed in this white paper, is 2.1 IE-6/yr.

Table 2 provides a breakdown of the contributors to LERF for Unit 1. The first part of Table 2 presents the breakdown by STC category. Category 8 (early SGTRs) is further broken-down to distinguish the contributions by initiating event-related SGTRs and phenomenologically induced SGTRs. The largest contributor to LERF is due to ISGTRs, contributing 71.1%.

The second part of Table 2 lists all extended Level 1 sequences which contributed greater than 1E-7/yr to LERF STCs 5 to 8. The contribution of these sequences is broken-down by their contributions to each LERF STC. Extended Level 1 sequence ETSW-03 provides the largest contribution. This sequence is a Loss of All Service Water initiating event with failure of AFW. Feed and bleed cooling is not available for this sequence due to the dependence of the pressurizer PORVs on instrument air, which is dependent on service water. The overall contribution of sequence ETSW-03 to the Unit 1 LERF is approximately 22.6%.

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PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model 5.2. Unit 2 Source Term Categories (STCs) 5, 6, 7, and 8, as developed in the Level 2 PRA and presented for Unit 2 in Figure 6, represent the Large Early Release categories. The sum of the frequencies of these STCs represents the Unit 2 LERF.

The Point Beach Unit 2 internal events LERF, using the enhanced Level 2 model discussed in this white paper, is 2.19E-6/yr.

Table 3 provides a breakdown of the contributors to LERF for Unit 2. The first part of Table 3 presents the breakdown by STC category. Category 8 (early SGTRs) is further broken-down to distinguish the contributions by initiating event-related SGTRs and phenomenologically induced SGTRs. The largest contributor to LERF is due to ISGTRs, contributing 71.2%.

The second part of Table 3 lists all extended Level 1 sequences which contributed greater than 1E-7/yr to LERF STCs 5 to 8. The contribution of these sequences is broken-down by their contributions to each LERF STC. Extended Level 1 sequence ETSW-03 provides the largest contribution. This sequence is a Loss of All Service Water initiating event with failure of AFW. Feed and bleed cooling is not available for this sequence due to the dependence of the pressurizer PORVs on instrument air, which is dependent on service water. The overall contribution of sequence ETSW-03 to the Unit 2 LERF is approximately 19.3%.

6. ISGTR Model Conservatisms The Point Beach Nuclear Plant Level 2 PRA analysis employs a value for the conditional probability of an ISGTR based upon the model developed in NUREG-1570. The NUREG-1570 ISGTR model contains conservatisms that have been retained in the analysis documented herein. Two areas for which more realistic calculations would remove conservatisms are:

1. MSSV failure probability would decrease (based on the EPRI Topical for MSSV failing to reseat following steam challenges). Using the more realistic value would drop the combined steam generator depressurization probability to about 0.1.

2. PSV/PORV failure potential would increase (based on EPRI topical which is supported by experience on PSV/PORV failing stuck open based on repeated liquid and two phase challenges). This would result in an increased probability of a stuck open PSV/PORV.

The effect of using more realistic calculations/values for the above two items would be a reduction in the conditional ISGTR probability.

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PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Table 1 Point Beach Unit 2 Extended Level 1 Event Tree High/Dry Sequences with Concurrent RCP Seal LOCA Extended Level 1 Sequence Frequency ESBO-28 0.00E+00 ESBO-29 4.08E-10 ESBO-30 2.18E-10 ESBO-31 7.04E-08 ESBO-32 5.91E-09 ESBO-33 7.17E-10 ESBO-34 O.OOE+00 ESBO-35 0.OOE+00 ESBO-36 1.31E-08 ESBO-37 O.OOE+00 ESBO-38 2.21E-10 ESBO-45 0.00E+00 ESBO-46 0.OOE+00 ESBO-47 0.00E+00 ESBO-48 7.20E-10 ESBO-49 6.04E-11 ESBO-50 7.33E-12 ESBO-51 0.OOE+00 ESBO-52 0.00E+00 ESBO-53 1.91E-10 ESBO-54 0.00E+00 ESBO-55 3.96E-12 ESBO-76 0.OOE+00 ESBO-77 0.00E+00 ESBO-78 0.OOE+00 ESBO-79 1.65E-10 ESBO-80 0.OOE+00 ESBO-81 0.00E+00 ESBO-82 0.OOE+00 ESBO-83 2.91E-08 ESBO-84 0.OOE+00 ESBO-85 9.73E-10 ESBO-86 O.OOE+00 ET1-20 (successful EDG operation) 0.00E+00 ET 1-21 (successful EDG operation) 0.OOE+00 Page 11 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model ET1-22 (successful EDG operation) 1.69E-07 ETI-23 (successful EDG operation) O.OOE+00 ET1-24 (successful EDG operation) O.OOE+00 ET1-25 (successful EDG operation) O.OOE+00 ET1-26 (successful EDG operation) 7.18E-09 ET1-27 (successful EDG operation) O.OOE+00 ET1-28 (successful EDG operation) 1.19E-07 ET1-29 (successful EDG operation) 8.44E-09 ET1-30 (successful EDG operation) 1.01E-09 ET1-31 (successful EDG operation) 9.58E-09 ET1-32 (successful EDG operation) 1.93E-09 ET1-33 (successful EDG operation) 7.59E-07 ET1-34 (successful EDG operation) 4.35E-09 ET1-35 (successful EDG operation) 3.46E-06 ETI-72 (successful Gas Turbine operation) 3.23E-11 ET1-73 (successful Gas Turbine operation) 4.41E-10 ET1-74 (successful Gas Turbine operation) 1.21E-09 ET1-75 (successful Gas Turbine operation) O.OOE+00 ET1-76 (successful Gas Turbine operation) 1.34E-09 ET1-77 (successful Gas Turbine operation) 5.85E-12 ET1-78 (successful Gas Turbine operation) 1.80E-09 ET1-79 (successful Gas Turbine operation) 3.51E-11 ET1-80 (successful Gas Turbine operation) 2.22E-09 ETSW-1 1 5.42E-09 ETSW-12 6.16E-10 ETSW-13 O.OOE+00 ETSW-14 1.54E-09 ETSW-15 O.OOE+00 ETSW-16 4.97E-09 ETSW-17 O.OOE+00 ETSW-18 2.98E-08 ETSW-30 4.07E-10 Page 12 of 34

Point Beach NuclearPlant - White Paper LERF Impact Due to Enhanced Level 2 Model Table 2 Point Beach Unit 1 LERF Contributors Source Term Category Frequency (yr-1) Percent of LERF 5 (early liner failures) 2.94E-8 1.4%

.6 (cntmt isolation failures) 1.50E-8 0.7%

7 (ISLOCAs) 2.37E-7 11.2%

8 (initiating event SGTRs 3.33E-7 15.8%

and main steam and feed line break-initiated SGTRs; with early release) 8 (ISGTRs) 1.50E-6 71.1%

TOTAL 2.1 1E-6 100% (accounting for round off)

STC 5 STC 6 STC 7 STC 8 Extended Frequency Frequency Frequency Frequency Level 1 Contribution Contribution Contribution Contribution Sequence (yr-1/%) (yr 1 /% (yr'I/%) (yr-1/%)

ETSW-03 1.18E-8/40.3`% 3.98e-9/26.4% 0/0% 4.60E-7/25.1%

EISL-02 0/0% 0/0% 2.35E-7/99.0% 0/0%

ET2-11 0/0% 1.63E-9/10.9% 0/0% 1.89E-7/10.3%

ETD1-16 0/0% 9.80E-10/6.5% 0/0% 1.13E-7/6.2%

ETD2-17 0/0% 9.80E-10/6.5% 0/0% 1.13E-7/6.2%

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Point Beach NuclearPlant - White Paper LERF Impact Due to Enhanced Level 2 Model Table 3 Point Beach Unit 2 LERF Contributors Source Term Category Frequency (yr-) Percent of LERF 5 (early liner failures) 2.70E-8 1.2%

6 (cntmt isolation failures) 1.55E-8 0.7%

7 (ISLOCAs) 2.37E-7 10.8%

8 (initiating event SGTRs 3.52E-7 16.1%

and main steam and feed line break-initiated SGTRs; with early release) 8 (ISGTRs) 1.56E-6 71.2%

TOTAL 2.19E-6 100% (accounting for round off)

STC 5 STC 6 STC 7 STC 8 Extended Frequency Frequency Frequency Frequency Level 1 Contribution Contribution Contribution Contribution Sequence (yr'I/%) (yr'/% (yr 1 %) (yr-1/%)

ETSW-03 1.05E-8/39.0% 3.54e-9/22.8% 0/0% 4.09E-7/21.4%

EISL-02 0/0% 0/0% 2.35E-7/99.0% 0/0%

ET2-11 0/0% 1.46E-9/9.4% 0/0% 1.69E-7/8.8%

ET1-35 0/0% 1.11E-9/7.1% 0/0% 1.28E-7/6.7%

ETD2-17 0/0% 1.06E-9/6.8% 0/0% 1.22E-7/6.4%

ETSW-06 2.97E-9/11.0% 9.97E-10/6.4% 0/0% 1.15E-7/6.0%

Page 14 of 34

PointBeach NuclearPlant - White Paper LERF Impact Due to Enhanced Level 2 Model Figure 1 Fault Tree Model of NUREG-1570 APET NUREG-15670ISGTRAPET NC-i WITH2 SG FOR POINTBEACH jAnayst: MDW lCretion Date:08.20-2007 Revision:

11-27-2027 NOGTRN LGC(R-. 27)W4MNPRA 34 Prod444Oh, 2

4 5

6.

7 8

A C

Page 15 of 34

PointBeach NuclearPlant - White Paper LERF Impact Due to Enhanced Level 2 Model 0 I 1 2 1 24 1 5 NUREO.IS7GISGTRAPETRC-l G 2 SOFORPOINT WITH BEACH ISS 2 IATO MOW M~alyst: lCreationDate:08-20-2007 Revision:11-27-2007 CGI03OLC(R.,. 2T)3 k4ONfPRA 3.0Pr03dinn

¶ 0 CLo 03 Page 16 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Q 1 2 3 4 NUREG-1570 WITH JSGTR 2SGFOR jMalyst:MDW APETRC-1MG3 POINTBEACH lCreationDate:08-20-2007 f

Recismo:

IG 11-27.2007 (R-~ 27)WrIN&PRA

,ISGTRNIOC 3 0 Produdic Page 17 of 34

Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model I 1 2 3 1 45 NUREG-1570ISGTR APETRC-1 UEACH JG 0 WITH2 SG FORPOINT Imalyst: MOIW ICI-ationD.t.: OKO.2002f7 Rmai-in 11-27.2007 (RY. 27)vWVINPIA ISOTRN-LOC 300Potý 2

4 5

6 7

B 9

A a

C Page 18 of 34

Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model U 1 2 1 3 1 4 NUREC-1570ISGTRAPETRC.I WrHl2 SGFORPOINTBEACH IG MDW lAanlyst: Date.08-:l~0 IC~aetion Rsin 12-10J7 Rev.27tV*MPRA 10 Proawctý (0R.0 APET PdthCL I SGL5k Page 19 of 34

Point Beach Nuclear Plant" White Paper LERF Impact Due to Enhanced Level 2 Model a 1 2 3 1 4 1 5 NUREG-1570 ISGTR APET RC-1 ISG, 3 WTH 2 SG FOR POINT REACH Analyst: MDW lCttion ODte:08-20-2007 Reeision. 11.27-2007 ISGTRN.LOC (Re,. 27)V*rPnNPRA3.0Proction Page 20 of 34

Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Page 21 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model U 1 2 1 3 4 NUREG-1570ISGTR APET RC,1 BEACH ISG, 8 WITH2 SG FOR POINT Analyst: MDW lCreation Date: 08-20-2007 Reaision: 11-27-2037 ILGC (Rev 27)WVnNPRA SOTRN 30 Prodctian 1.21o1-00 1.2440-51 Page 22 of 34

Point Beach NuclearPlant - White Paper LERF Impact Due to Enhanced Level 2 Model 0 1 1 2 1 3 I 4 5 NUREG-1570ISGTR APET RC-I WITH 2 SG FOR POINT BEACH AnIyst: MOW lCreation Data: 08-20-2007 Revision: 11-27-2007 WGTRNLOC CRv.27)TVonNLPRA 3.0 Prodetion Page 23 of 34

Point Beach NuclearPlant- White Paper LERF Impact Due to Enhanced Level 2 Model Page 24 of 34

Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Page 25 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model U 2 .14 4 AFETPýtA NUJREG-15"BISGTR APET RC-1 S,1 wITH2 SG FOR POINT SEACH TFRTo AnaIyst: MDW IC,.tion DTtN 56-25-2W? RelisioT: 11-27-2007 LGC(Re,. 27) *"nJPRA3.0Productlon ISGTRN 1 [fi~

P1orNoP1SGTW G1501240 NOPt 5072 TI . N TI SGIR I5 TI-lOTS ndNoHoldup A

C Page 26 of 34

Point Beach Nuclear Plant- White Paper LERF Inpact Due to Enhanced Level 2 Model Page 27 of 34

Point Beach Nuclear Plant - White Paper LERF Impact Due to Enhanced Level 2 Model NUREG-1570 ISGTRAPET RC-1 ISG. 14 0 WITH2 SG FOR POINT BEACH Analyst: MDW lCreation Date: 0B-20-200X7 RResion: 11-27-2007 G*MTNLGC(11-. 27) V&M"RA 3 0 Produtbn 2

3 4

5 6

7 9

A a

C Page 28 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model 0 1 1 3 -1 4 - _ _

1NUREGT-1 570ISGTRAPETRC-I 0 WITH2 SGFORPOINT BEACH S,1 M~nIyst:MO5W ICro.tionOte: 08-20-2007 Revision:

11.27.2007 LOC(R-~271Wa,NIPRA tSGTRN 30 Pro.ddtn APET Pdh Al I SGLe.k 2

3 4

5 6

7 6

9 A

B C

Page 29 of 34

Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Figure 2 Quantified Fault Tree with PI-SGTR and TI-SGTR Sensitivity WinNUPRA 3.0 Production (SR-3) Licensed to: Mark Walz Sensitivity Analysis EQN File: C:\TEMP\ISGTRN.EQN BED File: C:\TEMP\ISGTRN.BED Data Source: BED File Values Orig. Top Event Unavailability: 1.026e-001 New Top Event Unavailability: 3.707e-002 Change: -63.87 %

Event Name Probability New Prob.

0 OF 2 SG LEAK S.000e-001 0 SG DEP 4.050e-001 0 SG P TI-SGTR 1.730e-002 1 OF 2 SG LEAK 8.300e-001 1 SG D TI-SGTR 1.840e-002 1.225e-002 1 SG DEP 5.950e-001 1 SG LP SEAL CLE 5.000e-001 1 SG NO SEAL CLE 5.000e-001 I SG P TI-SGTR 7.910e-002 5.268e-002 1 SG PI-SGTR 5.490e-002 5.490e-003 1 SG S L TI-SGTR 1.000e+000 1 SG S TI-SGTR 1.840e-001 1.225e-001 2 OF 2 SG LEAK 1.700e-001 2 SG D TI-SGTR 3.650e-002 2.431e-002 2 SG DEP 1.600e-001 2 SG LP SEAL CLE 1.000e+000 2 SG P TI-SGTR 9.700e-002 6.460e-002 2 SG PI-SGTR 1.070e-001 1.070e-002 2 SG S L TI-SGTR 1.000e+000 C1 VALVE LEAKAGE 5.000e-001 HOLDUP 1 5.050e-001 HOLDUP 2 5.140e-001 INTACT 8.640e-001 NO 1 SG PI-SGTR 9.450e-001 9.945e-001 NO 2 SG DEP 8.400e-001 NO 2 SG PI-SGTR 8.930e-001 9.893e-001 NO HOLDUP 100 5.140e-001 NO HOLDUP 72 5.000e-001 NO HOLDUP 74 5.140e-001 NO HOLDUP 98 5.050e-001 NO SEAL LOCA 9.960e-001 RCS DEPRESSURIZE 5.000e-001 RCS PRESSURIZED 5.000e-001 SEAL LOCA 4.000e-003 SORV 1.360e-001 VALVE LEAKAGE 5.000e-001 Page 30 of 34

Point Beach NuclearPlant- White Paper LERF Impact Due to Enhanced Level 2 Model Figure 3 Point Beach Unit 1 Quantified Containment Event Tree CRITERIA> BYPASS ISOLATE PRES ISGTR HIWTRCAV RPVFAIL CFVB DEBRIS CHR CFMODE p EGtUyfrom CONTAINMENTCONTAINMENTRCS PRESSURE ISGTR HIGHWATER RPV CONTAINMENT DEBRIS CHR MODE Fleouny.

Plant Danmage BYPASS ISOLATION AT OR LEVELIN FAILURE FAILURE IN OPERATES ANDTIME D States STATUS COREMELT HOTLEG CAVITY LOCATION AT COOLABLE FOR OF fromn SURGELINE FAILURE AT VESSEL BREACH VESSEL BREACH GEOMETRY EX-VESSEL 24 NRS CONTAINMENT S FAILURE NoRule delned Ro Rule denlled No Rule deflNed No Rule deflned NoRule dellled No Rule dellted NoRule deslted No Rule deflned NoRuletfolned No Rule eflned NoRtde deflned

{

NOFAIL NONE 1 1.080.-008 1.0000-000 CHR NONE 2 1.004e.006 COOLABLE_ 104.0 HIGH

1. 0900.006N O CHR RUPTURE 3 5.2 5e. .008 NOCF 5.285a.008 1.0688-006 UNCOOLABLE BASEMAT 4 1.0AA.A00 l.AA0e.AAS 1.068o-008 LOW LINER EARLY 5 1.069.-009 1.198.006. e CHR NONE 6 1.118oe007 11.176e.007 N NOCHR RUPTURE 7 5.882e-.09 NOTHIGHE ABOVEWTR 5,.882.o009

.188ae.007 UNCOOLABLE BASEMAT B N.188.-00E 1.188e-009 NOONE A 2.859e.007 2.859e-007 ISOLATE CHR NONE 10 . 27-38-.05 4.698e-005 2.382..005 HIGH EOOL.ABLE NOCHR RUPTURE 11 4.173e.006 NOCF 4.173e.006

2.828e-005

,UNDERWTR UNCOOLABLE BASEMAT 12 2.828e.007 2.9-leOOSI.2.._8e007 HOTLEG LINER

• EARLY 13 2,831eUUS 4.429e.005 2.831a.00813 28a.8 NOBYPASS 4.700..005 COLBE CHR 1.087e-005 NONE 14 1.087e.005 SHIGH 11.554o005 NOCHR RUPTURE 15 4.666ae.006 4.578e.005 NOTHIGH ABOVEWTR NOCF 4.666e.006 UNCOOLABLE RUPTURE 16 1.569e.007 1.569e071 ISGTR EARLY 17 1.4980e.000 1.498e-006 UNISOL-E EARLY TB 1.504e.008 t.5040-008 EARLYSGTR EARLY 19 3.327e.007 3.327e-007 LATESGTR LATE 20 2622e-006 2.622e-006 ISLOCA EARLY 21 2368e.OO7 2.368e-007 POINTBEACH- CONTAINMENT EVENTTREEDIAGRAM C-Pobn BeoautSLonl2AnIBI 1ISGTRSenkCETConmU-.PAD Last Sa-1, Tuesday, No-ember 27, 2007 Last Updated:Thursday, JOruraey 2.,1987 WIRIUCAP1.0 Ucensed to; Mirk Walz Page 31 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Figure 4 Point Beach Unit 1 Quantified Source Term Diagram

-CFTIME CFMODE S Frqec ENTRY TIMEOF MODEOFFrucy FROM CONTNMT CONTNINT T CET FAILURE FAILURE from ENDPOINTS C NOPRleidefined Rl eie tl eie # POBASIC MN*aB II*

0 3.611G.005 BASESIAT 0 2.947..007 2.947e.007 LATE -SG06.LT 0 2.6224.006 1,197.-005 2.622e-006 5.019e.30S RIDTIIR-a 1 9.055-0006 9.055e-006 LINER 2.938e.008 2.938e-008 1.504e.S008 1.5040.600 EARLY Z1120-006 S Z36..00.07 2.368e0007 SGTR-E 1.520n.000 a

1.830e-006 POINTBEACHUNIT2 2*nit I II4SGTRSenIBBASICSTD C.tPob Beachl.eel'n LOstSaved: Tuesday, August 14,2007 SOURCETERMDIAGRAM Last UpdaedIMTue y, Nomen 27,2007 WinNUCAP1.0 Licensed to: MarkWaLz Page 32 of 34

PointBeach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Figure 5 Point Beach Unit 2 Quantified Containment Event Tree CRITERIA> BYPASS ISOLATE PRES ISGTR HIWTRCAV RPVFAIL CFVB DEBRIS CAR CFMOOE p EItryffoin CONTAINMENTCONTAINMENTRCS PRESSURE ISGTR HIGHWATER RPV CONTAINMENT DEBRIS CHR MODE Plant Damage BYPASS ISOLATION AT OR LEVELIN FAILURE FAILURE IN OPERATES ANDTIME D States STATUS COREMELT HOTLEG CAVITY LOCATION AT COOLABLE FOR OF Sroln NoRule defined No Rule defined No Ruledef0ne SURGELINE FAILURE AT VESSEL BREACH VESSEL BREACH GEOMETRY EX-VESSEL 24 HRS NoRole doened No Rule defined No Rule defined NoRole donned No Ruledefined NoRule defi1d No R01ule CONTAINMENT S FAILURE defned NoRule defned "

{

NOFAIL NONE 1 1.063e.008 1.0630o-008 CHR NONE 2 0.8880.007 9.808a-007 HIGH COOLABLE 1.063e-000 1.04 1e-00A NOCHR RUPTURE 5.204e.008 NOCF 5.204e.008 t.051..006 UDRWRUNCOOLABLE BASEMAT TOUT51.0008 LOW I NrR FARLY 1.1030.006 1.OG52.O00 1,O.02.409 CHR NONE 1.125o.007 1.15e.007 COOLOBLE NOCHR RUPTURE 5.922..009 NOTHIGH ABOVEWIR 5.922e.009 1.196e.007 UNCOOLABLE BASETAT 11.1964,009 TI1ABA-TOO Nn Fell NONE Z625o-RE7 2.625e.007 ISOLATE CHR NONE 4.0550-000 2.056..005 HIGH COOLABLE 1 2.11560.005 2.0250.005 I NOCHR RUPTURE 5.144e.006 NOCIF 5.144e-006 2.590o.005 UNCOOLABLE BASEMAT 2.596o.0107 UNDERWIR 2.599o-005 2.596eý007 HOTLEG LINER EARLY 2.599.-GO8 4.581,0-05 2.599e-008 NGEYoASS CHR NONE D.2A00.006 14 260e.006U 4.857a.9105 F COOLABLE 1.9360-O0S HIGH NOCHR RUPTURE Is 1.Tloe.EBA 4.7370-005 NOTH0GH ABOVEWTR NOCF 1.010e.005 1.9560.005 ONCOOLABLE RUPTURE 16 1.EU56..00 1.E56e.07 IC*TO 1.SGUR EDO 17 1.559e.006 PDS 5.181e-005 UNISOLE EARLY 1.554"8TE Is 1.5540.008 I::ARIY

.... ...... .. 19 3.516a.007 3.5tN0.007 LATESGTR LATE 20 2.655e-0TE 2.655.0006 ISLOCA EARLY ,I 2.368e.007 2.368-.007 POINTBEACH-CONTAINMENT EVENTTREEDIAGRAM Cr"Po.lBech'Level 21.1.1t2 oISGTRS.nCETCompU2.PDD Last Saved. Tuesday, NOvember 27. 2007 LaNt Updated: Thursday, Janfatuy29,1987 WinNUCAP1.0 Llcensed to; MaIStWait Page 33 of 34

Point Beach Nuclear Plant- White Paper LERF Impact Due to Enhanced Level 2 Model Figure 6 Point Beach Unit 2 Quantified Source Term Diagram CFFIME CFMODE S FFrequericy ENTRY TIMEOF MODEOF FROM CONTHMT CONTNMNT T CET FAILURE FAILURE fron ENDPOINTS C

1. PBSASIC No Role defined Rl elldRl eie NO FAILURE NO FAILURE 0 3.119e.005 BASEMAT 2.7130.007 0

2.713e.007 LATE SGFR1.T 0 1 2.6550.006 1.843.-005 2.655a.006 S1t81.0O5 RUPTURE 1.550e-005 0

1.550e.005 LINER 0 1 2.704e.008 2.704e-000 UNISOLATED 1.554e-009 0

1.554e400

-,RV 2.190o-006 ISLOCA 0 2.368-007 2.3680-007 OGTR.F 0 1.9100.006 1.10oe.0006 2

POINTBEACHUN0T C.Point Beuch;Lesel 2U0t 2 nJSGTR SenPBBASIC.STD Last Saved: Tuesday, Auoust 14,2007 SOURCETERMDIAGRAM Last Updated: Tuesday, Noveemer 27,2007 IA/InNUCAP 1.0 Licensed to: Mark Waiz Page 34 of 34