ML20153B581
| ML20153B581 | |
| Person / Time | |
|---|---|
| Issue date: | 09/15/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20153B570 | List: |
| References | |
| NUDOCS 9809230183 | |
| Download: ML20153B581 (8) | |
Text
m U S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION l SAFETY EVALUATION OF EPRI REPORT TR-105707. OCTOBER 1996.
"BWR VESSEL AND INTERNALS PROJECT. SAFETY ASSESSMENT OF BWR REACTOR INTERNALS (BWRVIP-06)"
1.0 INTRODUCTION
By letter dated October 5,1995, the BWR Vessel and Intemals Project (BWRVIP) submitted the Electric Power Research Institute (EPRI) Proprietary Report TR 105707, October 1995, "BWR Vessel and Internals Project, Safety Assessment of BWR Reactor Intemals (BWRVIP-06),"
(Reference 1) for NRC staff review and approval. The BWRVIP 06 report was supplemented by letters dated December 20,1996, (Reference 2) and June 16,1997 (Reference 3).
The BWRVIP-06 report provides a generic safety assessment of BWR/2 6 reactor Internals to determine the short and long term actions required to assure safe operation with the potential for component cracking. The assessment considers the reactor internal components function during normal, transient, seismic, and design basis accident (DBA) conditions. The results of the BWRVIP-06 report wa; Intended to provide utilities with a generic reactor intemals management plan which can be tailored to meet the needs of the Individual utilities. Additionally, the BWRVIP-06 report was intended to provide the NRC with information needed to evaluate future cracking in BWR intemal components.
increased occurrence of identified intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) internals prompted the US BWR executives to form the BWR Vessel and Intemals Project (BWRVIP) in June 1994, to address integrity issues arising from service related degradation of these important components. It is apparent to the BWRVIP and the NRC staff that as inspection techniques improve and as more inspections are performed, additional IGSCC related cracking in welded and bolted locations of reactor internals will be identified. On this basis, the BWRVIP submitted this document as a means of exchanging information with the NRC for the purpose of supporting generic regulatory efforts related to assessing the safety consequences of potential cracking of BWR/2 6 reactor intemais. The BWRVIP-06 report generically evaluates postulated failures caused by IGSCC in welded and bolted locations of vesselInternal components and establishes long term actions which the BWRVIP stated are -
appropriate to ensure continued safe operation.
2.0
SUMMARY
OF BWRVIP-06 REPORT 2.1 Safety Assessment The objectives of the BWRVIP-06 report were to perform a qualitative safety assessment of BWR reactor internals and attachments to assure continuing safe, operation of BWRs with assumed single component failures, to define short-term and long-term actions needed to ensure safe operation, and to develop an overall prioritized list of components. The staff notes that the prioritized list was used as a guide by the BWRVIP to establish which components require the development of inspection and evaluation guidelines first. The safety assessment ATTACHMENT 9809230183 990915 PDR TOPRP EXIEPRI C PDR ,
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evaluated the consequences of fully failed component lochitions of safety related and non-safety- l related components. The components addressed in BW3 VIP-06 are as follows and are not listedin orderof priority: l l
l Safetv-related l
Control Rod Guide Tube, Control Rod Drive (CRD) Housing and Stub Tube Core Plate dP/ Standby Liquid Control System (SLCS) Line Core Plate
, Core Spray Piping i
Core Spray Sparger Jet Pump Assembly Low Pressure Coolant injection (LPCI) Coupling
) Incore Housing and Dry Tube Orificed Fuel Support Core Shroud L Core Shrcud Support Access Hole Cover Top Guide / Grid Vessel Instrumentation Non-safetv related Steam Dryer
- Core Shroud Head and Separators Feedwater Spargers Surveillance Capsule Holder
. Given these components, the BWRVIP-06 report identified the welded and bolted locations of each component where IGSCC would be likely to occur. The BWRVIP-06 report also identifies the different bolted and welded locations for the same component for the several BWR product lines, when applicable. As an example, the core spray piping in a BWR/2 has a different layout and configuration than the core spray piping of BWR/3-6. However, the BWRVIP-06 report does not account for any plant-spee!fic modifications that may have been made to the components during construction or operation.
The BWRVIP considers the BWRVIP-06 report to be a bounding assessment of postulated failures of BWR reactor intemals. The worst case assumption of a complete failure at each location of the components was assumed in the BWRVIP-06 report. The safety consequences of the complete failure were evaluated at each location. The staff notes that one complete failure was assumed while all other components susceptible to IGSCC were assumed to be intact (i.e., neither common mode failure nor consequential failures of multiple components due to IGSCC were considered in the BWRVIP-06 report). In some cases, failure of multiple welds of a particular component was considered. However, these cases were generally considered to be non-credible events. The staff further notes that no other consequential failures, such as l ' those due to jet impingement and pipe whip, were considered in the BWRVIP-06 report.
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3 The postulated failures were evaluated for normal operations and design basis accidents (DBAs) l such as main steam line break, recirculation line break, and safe shutdown earthquake. The BWRVIP-06 report stated that these evaluated failures are beyond the current design or licensing basis of operating BWRs and that no new design bases are implied by the failures considered in the BWRVIP-06 report. It is important to note that the BWRVIP-06 report's acceptance criterion for the consequence analysis was to achieve a safe shutdown, not to l maintain original design margins or maintain long-term core cooling.
Examples of the deterministic evaluation of some safety-related intemals follow:
- 1. Standby Liquid Control System (SLCS) / Core dP: T he BWRVIP-06 reoort concluded l that there were no safety consequences due to failure of any portion of SLCS as long as the Emergency Procedures Guidelines (EPGs) for an Anticipated Transient Without Scram (ATWS) are followed.
- 2. Core Plate: A seismic event in combination with aligner pin weld failures for plants without restralning wedges could result in limited horizontal movement of the core plate.
Limited movement below the amount specified in the BWRVIP-06 report would not result in slower scram times; however, movement above the amount specified could result in failure to insert control rods. In this case, SLCS is available to shutdown the reactor.
- 3. Core Spray Piping: The BWRVIP-06 report stated that, since the core spray piping is inspected every refueling outage, there are no short term actions required.
- 4. Cure Spray Sparger: Failure of these welds is not detectable during normal operations; however, the spargers are inspected each refueling outage.
- 5. Jet Pump Assembly: Failure of jet pump welds which cause separation of the jet pump are detectable dt. ring normal operations; however, jet pump disassembly during a recirculation line break could result in safety consequences.
Based on the results of the deterministic BWRVIP 06 report, the BWRVIP concluded that no short term actions, beyond planned inspections and possible implementation of new monitoring procedures, were required. The BWRVIP-06 report states that all BWR product lines have sufficient level of safety based on the following:
- 1. Detectability of component failure by online instrumentation The BWRVIP 06 report stated that "..if a location failure which could ir ~.iore with safe shutdown during an accident scenario can be detected during plant operation, a safe shutdown can be achieved before the component is challenged by the accident scenario." The BWRVIP-06 report also notod that in some cases, potential detection may require implementation of new procedures to monitor reactor operating conditions.
- 2. Structural and/or functional redundancy in this case, the BWRVIP-06 report assumed that '...lf failure of a component location results in the loads being redistributed to other components or locations on the same component which have adequate margin to accommodate the additional loads, there will
4 be no adverse impact on the ability of that component to function in achieving safe shutdown."
- 3. Detectability of component failure by current inspections The BWRVlP stated that "...if inspection is being performed with r.ufficient frequency and l detail, the possibility of significant undetected cracking can be axcluded when considering the short term significance of potential cracking."
l 4. Low probability of challenging event In some cases, postulated failures and a DBA are required to pose any safety consequences. Since DBAs are low frequency events, no short-term actions are i required.
i The staff notes that all long-term actions generally consist of development of inspection and l evaluation guidelines and repair or replacement criteria for the specific components. This is based on the underlying assumption of the BWRVIP 06 report that the developed inspection and evaluation guidelines will assist utilities in identifying potential cracking locations and therefore, fixing the cracked location before the component falls.
2.2 Priority List The NRC staff met with membere of the BWRVIP on April 29 and 30,1997. During the meeting the BWRVIP presented the current prioritized list of reactor intemals. The list is as follows:
Hiah Priority Ccmoonents Shroud Core Spray Piping and Sparger :
Shroud Support
. Top Guide Core Plate Standby Liquid Control System Medium Priority Comoonents Jet Pump Assembly Low Priority Componenta Control Rod Drive Guide Tube Control Rod Drive Stub Tube Incore Housing Dry Tube instrument Penetrations i Vesselinside Diameter Brackets Low Pressure Coolant injection Coupling l 1 l l o'
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J The BWRVIP's intent of the prioritized list is to address the high priority components first by
! developing inspection and evaluation guidelines and repair or replacement criteria. The staff notes that the list has changed slightly since its development in 1904. These changes were i based on inspection findings and resulted in a few components moving to a higher priority on the i list. The staff also notes that most of the components on the prioritized list do not have j
regulatory requirements for inspection. However, some components are inspected every j refueling outage or every other refueling outage. For example, NRC Bulletin 80-13 (Reference 4), " Cracking in Core Spray Spargers," requested that utilities perform a visual inspection of the
' Core Spray Spargers and the segment of piping between the inlet nezzle and the vessel shroud every refueling outage. On the basis of this list, the BWRVIP has and will continue to develop inspection and evaluation guidelines and replacement and repair criteria for the high and 4
j medium priority components.
l 2.3 Quantitative Safety Assessment of BWR Reactor Intemals (BWRVIP-09) s j
The BWRVIP performed a probabilistic risk assessment (PRA) of failure of intemal components due to IGSCC. This proprietary PRA was submitted to staff by letter dated June 16,1997, and is referred to as BWRVIP-09. The purpose of the PRA was to provide additional confidence in the i conclusions of the BWRVIP 06 report prioritized list. The PRA evaluated eight components
{ whose failure and occurrr.ce of a low probability event could result in increased core damage frequency. The eight components included the control rod guide tube / housing, core plate, core spray olping, core spray sparger, jet pump assembly, low pressure coolant injection (LPCl) couphng, shroud support access hole covers, and top guide. Access hole covers are located in
) the shroud support plate of BWR Jet pump plants, approximately 180 degrees apart, and are
] used to cover the access holes used during construction. The staff notes that the analysis
! evaluated the failure of each of the eight components in conjunction with a loss of-coolant
! accident (LOCA) and seismic events. Based on this study, the BWRVIP concluded the j following:
1 l 1. All BWR product lines possess a sufficient level of safety based on detection of l component failure, structural redundancy and low probability.
{ 2. Core damage frequencies are below levels of concem.
- 3. Analysis supports BWRVIP work prioritization.
The staff reviewed the results of the BWRVIP-09 report but did not evaluate the adequacy of the event tree development or the assigned probabilities. As the basis of the review, the staff used i the approach described in Draft Regulatory Guide (DG) 1061 (Reference 5). DG 1061 is 4
intended to improve consistency in regulatory decisions in areas in which the results of risk 4
analyses are used to help justify regulatory action. The principles, process, and approach discussed within DG-1061 also provide useful guidance for the application of risk information to a broader cet of activities than plant specific changes to a plant's current licensing basis, i.e.,
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generic activities. As such, DG 1061 was used to help evaluate the change in core damage frequency as a result of postulated failurcs of individual components, in implementing risk-informed decision making, changes are expoeted to meet the following set of key principles.
f 1. The proposed change meets the current regulations. This principle applies unless the i proposed change is explicitly related to a requested exemption or rule change.
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- 2. Defense in-depth is maintained.
- 3. Sufficient safety margins are maintained. ;
- 4. Proposed increases in risk, and their cumulative effect, are small and do not cause the !
NRC Safety Goals to be exceeded. 1
- 5. Performance-based implementation and monitoring strategies are proposed that address uncertainties in analysis models and data and provide for timely feedback and corrective action.
Since the PRA was performed for a general set of reactors, the mean core damage frequency of a particular plant to which the BWRVIP-06 report would apply was not known. However, the staff assumed that any particular plant subject to this analysis had a mean core damage j 4
frequency less than 1 x 10 per reactor year. This assumption was based on the fact that the 1 purpose of the PRA was to verify the prioritized list and conclusions of the BWRVIP 06 report.
4 In addition, during the meeting with the staff on April 29,1997, the BWRVIP stated that the average core damage frequency from BWR IPEs was 2 x 10'5 per reactor year. Using this assumption, DG-1061 allows for increases in calculated core damage frequency that are very '
small (e.g., core damage frequencies of less than 1 x 104per reactor year) when combined with '
the applicable large early release frequency guidelines that are also described in DG 1061.
The staff notes that the BWRVlP started with a bounding assessment approach. in this case, the quantification of models was performed with the probability of IGSCC degraded component failures set to 1. The success criteria for this bounding analysis was that frequencies from all 3 accident sequences were less than 1 x 10 per reactor year, if the success criteria was met, no further analysis was necessary, if the bounding case resulted in frequencies greater than 1 x 104 per reactor year, then the crack growth model was used to calculate component failure probabilities, the PRA models were re-evaluated, and sensitivity calculations were performed.
Based on this procedure, the BWRVIP was able to identify which components are more important to safety based on prevention of core damage and was able to limit all increases of
) core damage frequency to less than 1 x 104 per reactor year. The staff notes that the PRA results verified the priority lists and the conclusions of the deterministic BWRVIP-06 report.
3.0 STAFF EVALUATION The staff has reviewed the BWRVIP-06 report which states that, in some cases, potential detection of cracking internals may require implementation of new procedures. On the basis of this review, the staff has determined that establishment of any new procedures or changes to procedures for the purpose of detecting potential failures of reactor intemals that may be required shculd be a near term plant-specific action.
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As discussed above, the BWRVIP-06 report evaluated complete failures of one component location while all other components susceptible to IGSCC were assumed to be intact (i.e.,
neither common mode failure nor consequential failure of multiple components due to IGSCC were considered in the BWRVIP 06 report). The staff has determined that it would have been more appropriate to evaluate the safety consequence; of partial cracking in several components i
during normal operations and DBAs. This is besed on the fact that many of the components are
! subject to the same environmental conditions which are conducive to IGSCC.
i The staff notes that this analysis only discusses the capability to achieve a safe shutdown with one component location failed. Long term cooling is not evaluated in this analysis.
Acknowledging this fact, the staff continues to have unresolved questions about the BWRVIP's l conclusions. For examplo, the BWRVIP stated that there were no safety consequences due to l failure of any portion of SLCS or the core spray spargers. With respect to SLCS, the staff cannot conclude that SLCS will be available if the pipe is crushed and not severed. Additionally,
' the BWRVIP is recommending reduced inspections of core spray spargers for all BWR/3s and BWR/4s (except Hope Creek and Limerick) which appears to be in conflict with the BWRVIP's i
' conclusions discussed above. The staff has determined that core spray is an important safety system because it provides a uniform distribution of spray to assure cooling when the core cannot be fully reflooded, it helps protect some fuel designs from exceeding their fuel safety limits, and in some casos the high pressure core spray system provides a flow path for injection of boron for SLCS.
The staff is reviewing the SLC issue further in its evaluation of the BWRVIP 27 report. Further, l In its review of the BWRVIP-18 report, the staff found the reduced inspection frequency for i 304L/316L non-creviced welds acceptable because no cracking of these welds have been
- reported. However, the staff did not agree that the proposed schedule for non-creviced welds in susceptible materialis appropriate since much of the sparger and intemal piping that has cracked is non-creviced. This issue is being discussed with the BWRVIP.
However, the Industry has implemented effective inspection and repair programs to reduce the likelihood of component tallure as a result of IGSCC. Although the consequences of multiple failures have not been specifically addressed in the report, inspections, evaluation and repairs of components to date are sufficient to provide the staff confidence that multiple, undetected component failures are unlikely and to ensure component integrity for the components evaluated in the BWRVIP-06 report. In addition, this SER finds that the BWRVIP-06 report is acceptable for ranking intemal components for the development of inspection and flaw evaluation guidelines.
The NRC's Office of Nuclear Regulatory Hesearch (RES)is presently performing confirmatory research into the possible consequences of multip's, cascading component failures. Should the results from this confirmatory research program identify any significant issues, they will be addressed separately with the BWRVIP.
Based on this valuation, the staff has determined that the BWRVIP 06 report was useful for gaining insights which established a priority list for the evaluation of postulated cracked components. The staff agrees with the BWRVIP that the priority list discussed in Section 2.2 of this SER is reasonable considering the results of the BWRVIP PRA and the staff's engineedng
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l 4.0 CONC 1.USIONS The NRC staff has reviewed the BWRVIP 06 report of operation with lGSCC degraded intemals and reviewed the results of the BWRVIP-09 PRA with respect to DG 1061 and finds, in the enclosed Safety Evaluation Report (SER), that: (1) if new procedures and/or changes to orocedures are needed to detect potential failures in some cases, these items need to be considered to be near-term plant-specific action items, to be implemented as soon as possible, and (2) common mode failure of reactor intemals due to IGSCC should have been evaluated in the BWRVIP 06 report because the BWRVIP did not provide any basis for the assumption that !
only a single failure should be ' considered. l However, since the issuance of the report, the industry has implemented effective inspection and ;
repair programs to reduce the likelihood of component failure as a result of IGSCC. Although the consequences of multiple failures have not been specifically addressed in the report, inspections, evaluation and repairs of components to date are sufficient to provide the staff with a high degree of confidence that multiple, undetected component failures are considered unlikely and to ensure component integrity for the components evaluated in the BWRVIP 06
- report. In addition, this SER finds that the BWRVIP-06 report is acceptable for ranking Intemal components for the development of inspection and flaw evaluation guidelines.
4 The NRC's Office of Nuclear Regulatory Research (RES) is presently performing confirmatory research into the possible consequences of multiple, cascading component failures. Should the results from this confirmatory research program identify any significant issues, they will be addressed separately with the BWRVIP.
5.0 REFERENCES
- 1. Beckham, J. T. Jr., BWRVIP, to USNRC, " Safety Assessment of BWR Reactor Intemals (BWRVIP-06), EPRI TR 105707, October 1995," October 5,1995.
- 2. Dyfe, Robin, BWRVIP, to USNRC, "BWRVIP Response to NRC Request for Additional Information on BWRVIP-06," December 20,1996.
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- 3. Terry, Carl. BWRVIP, to USNRC, "BWR Vessel and Intemals Project Quantitative Safety Assessment of GWR Reactor intemals (BWRVIP-09)," June 16,1997.
4.' IE Bulletin 8013, " Cracking in Core Spray Soargers," USNRC, May 12,1980.
- 5. Draft Regulatory Guide 1061, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Current Licensing Basis," USNRC, Offica of Nuclear Regulatory Research, March 1997.
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