ML20035B779
| ML20035B779 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/29/1993 |
| From: | Abraham P, Barrett M, Brewer H DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20035B760 | List: |
| References | |
| NUDOCS 9304050103 | |
| Download: ML20035B779 (8) | |
Text
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Severe Accident Analysis Report
Title:
Probabilistic Risk Perspective of the Steam Generator Tube Structural Inteerity Issue Related to--
the Interim Steam Generator Tube Pluccine Criteria for the Catawba Nuclear Station i
SAAG File:
87 Project :
Catawba Plant Support i
Prepared By :
7 Date :
29 3
a4 Date :
3/29/93 Checked By:
f1kk Approved By :
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3/A9/93 l
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i Probabilistic Risk Perspective of the Steam Generator Tube Structural Integrity Issue Related to the Interim Steam j
Generator Tube Plugging Criteria for the Catawba Nuclear Station i
I. Introduction The steam generator tubes of a pressurized water reactor are an integral part of the primary system boundary. Technical specifications provide repair limits to ensure that degraded tubes which cannot adequately perform this function are either sleeved or plugged. The interim tube repair limits for Catawba Unit I consist of voltage amplitude criteria rather than the traditional depth-based criteria. Thus, the interim I
repair criterion represents a departure from the past practice of explicitly enforcing a minimum wall thickness requirement.
This change in practice has raised some concerns about the structural integrity of the steam generator tubes. This probabilistic risk study evaluates the implications of allowing Catawba Unit 1 to run a full cycle after using the interim repair criteria.
II. Method Since the publication of WASH-1400, it has been recognized that the primary risk to the public from nuclear power plants results from low probability high consequence events invohing severe overheating and damage of the reactor core. ~ Other potential accidents that do not involve core damage result in very low risk. Changes to the plant, such as system modifications or procedure revisions, can be evaluated to determine their impact on the core damage frequency and the public health risk.: These quantifiable characteristics can be compared to the normal plant risk to determine the j
safety significance of a change.
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NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety 1
1ssues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity," also found that core damage risk following a tube rupture is far more significant than the risk 1
associated with the release of coolant activity. The NUREG found the public health j
risk of non-core melt releases to be more than'a factor of 1000 lower than the risk of..
~j tube rupture sequences invohing core melting.
Therefore, a probabilistic risk j
assessment method which evaluates the change in core damage risk in a manner i
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i consistent with the Catawba PRA, can be used here to evaluate the safety significance of the interim steam generator tube plugging criteria.
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III. Calculations I
The interim plugging criteria may impact the structural integrity of the steam generator tubes. Therefore, the risk implications of the interim plugging criteria are evaluated by assessing the impact on plant risk that might result from changes in the structural integrity of the steam generator tubes. Changes in the structural integdty of the tubes could impact plant risk by means of the following:
Increased potential for random tube rupture which results in core damage, Increased potential for a transient to induce a tube rupture which results in core damage, For the two concerns listed here, the risk associated with each one is a product of the frequency of the initiating event, the probability of core damage, and the consequences t
of core damage. The risi calculations all follow the same general form:
Risk = (Initiator Freq.)x(Conditional Prob. of Core Damage)x(Public Health Conseq.)
If the interim tube plugging criteria has an effect on any part of this three part equation, then it has an impact on the risk. The review of each concern is discussed below.
A. Random Tube Rupture The first concern is that the interim tube plugging criteria may significantly increase the chance of a random tube rupture. Of the three parts of the risk calculation: (1) initiator frequency, (2) conditional probability of core damage, and (3) consequences, only the initiator frequency could be affected by the interim plugging criteria. The Catawba PRA assumed a steam generator tube mpture initiator frequency of 9dO3RY.
This was based on industry data that involved seven tube ruptures in the history of operating PWRs. The Catawba PRA did not explicitly consider the existence of degraded tubes in the analysis. To determine the effect on random tube rupture risk, it i
must be determined if the interim tube plugging criteria would significantly change the j
Catawba tube mpture frequency.
The type of tube degradation experienced at Catawba causes cracks which are confined to the part of the tubes surrounded by the tube support plates. Unless these cracks are exposed by movement or deflection of the support plate, they will not f
mpture. Since the support plate does not move under normal operation, there is
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essentially no risk of a random tube rupture associated w:th the interim tube repair criteria. In fact, the full length inspection of all the Catawba tubes during the last outage leads to additional confidence that the probability of a random tube rupture at Catawba is less than the value assumed in the Catawba PRA Study.
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The PRA analyzed random tube ruptures as a possible way to cause a core melt i
accident and concluded that SGTRs contribute an insignificant amount to the overall core damage frequency and risk. In other words, the probability of a random tube' i
rupture leading to core damage was less than IE-10 per reactor-year (RY). Since the frequency of random tube mpture associated with the interim tube repair criteria is likely lower than the value assumed in the Catawba PRA study, it can be concluded that the interim tube plugging criteria contributes an insignificant fraction of the total core damage and public health risk.
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i B. Transient Induced Rupture Another potential concern about the interim tube plugging criteria is that if a large differential pressure transient were to occur, the steam generator tubes may be stressed to the point that one or more tubes may rupture. If the transient were to proceed to core damage, fission products would be released to the atmosphere without the benefit of the containment barrier. Also, the loss of primary coolant through the failed tubes would increase the likelihood of core damage.
For this concern, the three component equation must be modified slightly to the following; Risk = (Init. Freq.)x(Prob. of Tube Failures)x( Prob. Core Damage)x(Conseq.)
For the equations above, the probability of a single tube rupturing is likely very
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different from the probability of multiple tube ruptures.
Also, the conditional probability of core damage and the consequences would be different for multiple tube mptures than for a single tube rupture.
However, to simplify the-calculation, I
conservative values will be used for each parameter in estimating the potential risk.
The first component is the frequency of transients that produce large differential pressures. Table 1 presen;s a screening analysis to identify the types and frequencies l
of transients that cause large differential pressures. Table 2 presents some additional information concerning estimated sequence frequencies associated with the most likely transient to result in large differential pressures. These sequences can be characterized as a steam line break with a failure to terminate safety injection prior to the pressurizer l
becoming water solid. The sequence frequency is approximately 3.3E-4/RY for sequences with approximately 2335 psi differential pressure.
Westinghouse has calculated that if the tubes are exposed to a differential pressure of 2335 psi at the end of the cycle, the conditional probability for a single tube failure is approximately 7E-5.
The probability of multiple tube ruptures is limited by the inspection of all tubes; however, for conservatism, we can also assume that this is the i
probability for multiple tube failures and that it applies for the full cycle.
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i Next the probability of core damage is estimated, given that there has been a transient i
that resulted in a large differential pressure, and that one or more steam generator
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tubes have failed. NUREG-0844 estimated a probability of IE-2 for sequences of this l
type and the Catawba PRA estimated a probability less than IE-6. For conservatism, a j
value of 1E-2 will be used for the conditional probability of core damage.
The consequences from a tube rupture core melt sequence have been taken from the Catawba PRA, release categories 1.02 and 1.04. The whole-body person-rem risk calculated for these release categories is -1.37E+07 per event.
From the combination of the above; Risk = (Init. Freq.) ( Prob. of Tube Failure) (Prob. Core Damage) (Conseq.)
Risk = (3.3E-4) (7E-5) (1E-2) (1.37E+07) = 3.2E-03 wbpr The Catawba PRA has calculated the normal plant risk from internal events to be 2.24E+01 Person-Rem per year. In comparison, the risk of 3.2E-03 Person-Rem per year associated with steam generator tube rupture following a large delta pressure transient is a very small fraction of the total plar'.isk.
V. Comparison to NUREG -0844
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The PRA study done here has estimated that the public health risk associated with the l
interim steam generator tube repair criteria is on the order of 3.2E-03 Person-
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Rem /RY. NUREG-0844 also analyzed the public health risk of SGTR. A comparison of the Catawba specific results and the NUREG results can be made by convening the NUREG results to equivalent Whole-Body Person-Rem exposure. Using a conversion factor of I latent fatality /15,000 Person-Rem, the NUREG latent fatality total of 1.7Er3 latent fatalities /RY, is equivalent to approximately 25~ Person-Rem. The Catawba specific results are nearly five orders of magnitude lower than the NUREG value, for which it was concluded that, " risk from steam generator tube mpture events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed."
l IV. Conclusions
-i The Catawba interim repair criterion represents a departure from the past practice of explicitly enforcing a inimum wall thickness requirement for the steam generator m
tubes. A probabilistic risk analysis has been performed to assess the impact of running I
a full cycle following the use of the interim steam generator tube plugging criteria.
Two accident scenarios were identified that could be affected by steam generator tube 4
structural integrity concerns. This analysis has determined that the core damage-frequency and public health risk are low, when compared to the normal risk associated with plant operation as estimated in the Catawba PRA. Since the Catawba PRA did not explicitly consider the existence of degraded tubes, the risk value calculated in this analysis could conservatively be considered as the change in calculated risk due to the existence of tubes that are degraded within the limit allowed by the interim tube plugging criteria; even by such a conservative criterion the " combined risk" is less than that found to be acceptable in NUREG-0844. In actuality, tubes exhibiting some level of degradation have always existed. Therefore, it can not be concluded that the actual l
operating risk has increased by the implementation of the IPC, it is concluded that the contribution to the calculated risk due to the existence ofdegraded tubes, as permitted by the IPC, is insignificant.
Also, these risk values have been shown to be much less than the risk values calculated in NUREG-0844, which the NRC has concluded that, " risk from steam generator tube rupture events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed."
The quantitative result are summarized in Figure 1.
l VI. References
- 1. Catawba Nuclear Station Unit 1, Probabilistic Risk Assessment Study, Duke Power Corporation, September 1992.
- 2. Catawba Unit 1, Technical Support for Steam Generator Interim Tube Plugging Criteria for indications at Tube Support Plates, Westinghouse Electric i
Corporation, WCAP-13494, Revision 1,1993.
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- 3. NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity, U. S. Nuclear Regulatory Commission, NUREG-0844, September 1988.
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Table 1 - Seonences Which May Result in Substantial Pressure Differentials Across the Steam Generator Tubes (Chanter 15 Seouences)
Differential Pressure FSAR Chapter 15 Estimated Sequence
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Across SG Tubes Sequence Description Frequency 2485 psid (pressure is limited 15.1.5 Steamline Break 4.3E-8/RY (See Table 2) by the three pressurizer safety Outside Containment and valves) 15.2.8 Feedline Break 15.1.4 Spurious Secondary Bounded by SLB Frequency System Valve Opening Total Category 4.3 E-8/RY 2335 psid (pressure is limited 15.1.5 Steamline Break 3.3E-4/RY (See Table 2) by the three pressurizer Outside Containment and PORVs) 15.2.8 Feedline Break 15.1.4 Spurious Secondary Bounded by SLB Frequency System Valve Opening i
Total Categorv 3.3D04/RY l
2150 - 2335 paid (pressure 15.1.5 Steamline Break 6.7E-4/RY (Sec Tabic 2) dependent on time of SI Outside Containment and termination) 15.2.8 Fecdline Break 15.1.4 Spurious Secondary Bouried by SLB Frcquency System Valve Opening 15.8 ATWS Does not apply to period under consideration.
Total Categorv 6.7L3/RY 1800 - 2150 psid 15.8 ATWS Does not apply to period under consideration.
1450 - 1800 psid 15.8 ATWS 1.V L5/RY 1400 - 1450 psid 15.4.1 Zero Power Control Screening value IE-2/RY Bank Withdrawal 15.4.2 Control Bank Screening value IE-2/RY Withdrawal at Power 15.4.8 Rod Ejection Some very low value - e l,,
Total Categorv 262/RY 6
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k Comparison of Risk for Interim Plugging Criteria 1.00E+02 __
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l Interim -
Cotowba -
NUREG -
Midcycle Plugging IPE 0844 Outage Criterlo Exposure l-i Figure 1
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