ML20207B024
| ML20207B024 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/18/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20207B023 | List: |
| References | |
| NUDOCS 9905280099 | |
| Download: ML20207B024 (8) | |
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t U.S. NUCLEAR REGULATORY COMMISSION l l OFFICE OF NUCLEAR REACTOR REGULATION SAFETY EVALUATION OF l *BWR VESSEL AND INTERNALS PROJECT. BWR TOP GUIDE INSPECTION l
l AND FLAW EVALUATION GUIDELINES (BWRVIP-26)."
l EPRI REPORT TR-107285. DECEMBER 1996 j
1.0 INTRODUCTION
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1.1 Background
By letter dated December 27,1996, the Boiling Water Reactor Vessel and Intemra!s Project (BWRVIP) submitted the Electric Power Research Institute (EPRI) proprietary Repon TR-107285, "BWR Vessel and Intemals Project, BWR Top Gu% inspection and Flaw Evaluation Guidelines (BWRVIP-26)," December 1996. This report was supplemented by letter dated December 19,1997, which was in response to the staff's request for additional information (RAl), dated March 14,1997.
The BWRVIP-26 report provides generic guidelines intended to present the appropriate inspection recommendations to assure safety function integrity of the subject safety related RPV intemal components. It also provides design information on the top guide, geometries, weld locaUons, and potential failure locations for the several categories of boiling water reactcis (BWR/2 through BWR/6).
1.2 Purpose The staff reviewed the BWRVIP-26 report to determine whether its guidance will provide acceptable levels of quality for inspection and flaw evaluation (l&E) of the subject safety-related RPV intemal components. The review considered the consequences of component failures, potential degradation mechanisms and past service experience, and the ability of the proposed inspections to detect degradation in a timely manner.
1.3 Organization of this Report Because the BWRVIP-26 report is proprietary, this safety evaluation (SE) was written so as not to repeat proprietary information contained in the report. The staff does not discuss in any detail the provisions of the guidelines nor the parts of the guidelines it finds acceptable, except l foritem 4.0. A brief summary of the contents of the BWRVIP-26 report is give's in Section 2 of i this SE, with a detailed evaluation in Section 3. The conclusions are summar'. zed in Section 4.
The presentation of the evaluation is structured according to the organizatiori of the BWRVIP-26 report. l l
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( ENCLOSURE '
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2 2.0
SUMMARY
OF BWRVIP-26 REPORT The BWRVIP-26 report addresses the following topics:
1 Component Description and Funr*!an - The various top guide configurations are described in detail by a series of illustrations along with brief descriptions of each configuration's function and characteristics. Differences among the various models of BWRs (BWR/2, BWR/3-5 and BWR/6) are identified.
Suscesiianty Factors - The various types of material degradation mechanisms (fatigue,
. stress corrosion cracking, age embrittlement) that could impact lower plenum components are described. Materials, stress, and environmental factors are described in general terms, and followed by specific references to actual occurrences for each degradation mechanism relative to plant operating experience for particular mechanisms and components.
Potential Failure Locations and Safety Conseauences - Each of the top guide configurations are addressed from the standpoint of inspection history, future susceptibility to degradation, and consequences of failures in terms of component functions and plant safety. Based in these qualitatP/e considerations, the BWRVIP-26 report makes recommendations as to the need for inspe,deas !cr each of the top guide configurations.
Backaround and insoe': tion Historv- Data on service related failures of components are summarized. The major sources of such data are the various GE SILs and Rapid Information Communication Service Information Letters (NCSILs). Inspection requirements are evaluated according to the following four criteria: 1) the potential consequences of a failure to plant safety,2) the ability to detect degradation,3) field cracking history as a means to identify the most likely locations for material degradation, and 4) the extent to which results from prior inspections provide a high level of confidence that no degradation mechanisms are active for the components of .
concem. l BWRVIP Insoection Guidelines -The guidelines recommend the specific locations, NDE methods, and inspection frequencies for examinations of top guide configurations.
The recommended NDE methods are limited to visual examinations, with reference
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made to the BWRVIP-03 report for detailed requirements for implementing these visual examinations. The BWRVIP-26 report recommends only a limited number of j inspections for the top guide, based mainly on the relatively good service experience to ;
date that Indicates no evidence of generic cracking. The relatively small safety I consequences of structural failures is cited to justify the recommended level of !
inspection. l Loads - This section describes the loads used in fracture mechanics evaluations to i address the effects of detected flaws on structuralintegrity. The various types of loads (e.g., pressures, seismic, etc.) of concem are listed.
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3.0 STAFF EVALUATION The inspection guidelines provided by the BWRVlP-26 report are comprehensive in dealing with the top guide configurations. Differences in the design and operating experience for the various types of BWR reactors (BWR/2, BWR/3 5 and BWR/6) are reviewed and taken lato account as part of the inspection strategy. The guidelines limit the inspections to visual methods, but take steps to ensure that future visual examinations will be performed with an enhanced level of effectiveness compared to examinations performed in the past.
Implementation of the recommendations are intended to ensure an enhanced inspection relative to the examinations now required by ASME Section XI, but will replace or eliminate examinations recommended by GE in past SILs or RICSILs. Industry wide implementation is intended to provide operators of BWR plants with improved knowledge of the initiation and progress of materials degradation within the top guide region.
3.1 Analysis inputs and Loads inspection and evaluation guidelines and potential failure locations in BWR/2-6 top guide components were provided in this report. The BWRVIP 26 guidelines also contain a discussion of susceptibility considerations which indicate that all top guide systems may be subject to cracking, although much less so for BWR/6. However, in evaluating the consequences of potential cracking, the conclusion is that, for many locations, significant cracking can be tolerated without loss of essential top guide safety functions.
Analyses of various top guide locations with generally conservative loading were used to determine consequences of some of the postulated cracking scenarios. These analyses are included as representative examples of the type of plant-specific analyses that can be j performed to quantify available margin. The BWRVIP believes that such plant specific l analyses would, in many cases, demonstrate that more margin exists than is shown in the example analyses included in the report.
3.1.1 Development of Loads The applied loads on the top guide include those due to deadweight, seismic, fluid drag, safety relief valve (SRV), LOCA and fuel lift. The seismic loads are treated as primary loads f,or flaw evaluation purposes. These loads are imparted to the top guide at the connection to the shroud and are due to seismic excitation of the RPV, the core shroud, and the fuel. The fluid drag loads are imparted by the fluid flow past the top guide. The faulted condition loading corresponds to a postulated doubled-ended break of either the recirculation line or the main steam line, whichever is more limiting. For the faulted condition, the drag loads on the top guide are expected to be significant. The SRV air clearing loads are induced by SRV actuations which produce a rapid compression of the air mass in the interior of the SRV i
discharge pipes. The intemal pressure drives the water out of the submerged SRV discharge device (rams head or quencher) and ejects a high-pressure air bubble into the suppression pool below the water surface, causing osci'lating pressures on the suppression pool boundary.
The oscillating pressures impart structural motions which may cause dynamic excitations of the l structure and contained equipment.
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The various low probability LOCA conditions postulated to occur are small break accident (SBA), intermediate-break accident (IBA) and design basis LOCA (DBA). The various loads considered with these accident conditions include those due to annulus pressurization, fuel lift, and SRV discharge.
Annulus pressurization (AP) refers to the loading on the biological shield and the reactor vessel following a postulated pipe rupture. The pipe break is assumed to be an instantaneous i guillotine rupture occurring at the vessel nozzle safe end to pipe weld. The rupture allows a l
rapid mass and energy release into the small annular region between the biological shield wall and the RPV. The mass and energy released during this postulated pipe rupture causes a transient asymmetric differential pressure within the annular region between the biological shield wall and the RPV. In addition, there are Jet loads and pipe whip loads due to the ruptured pipe.
Annulus pressurization leads to interactions between the shield wall, reactor vessel, and the reactor pedestal. These inteactions transmit loads to the vesselintemals. In addition, pressure oscillations due to pool swell, condensation and main vent clogging may cause dynamic excitations of the structure and the contained equipment.
The fuel bundles in the reactor core are aupported horizontally by the fuel-support casting and core plate at the lower end and by the top guide at their upper end. In the vertical direction, the fuel bundle weight is supported by a system composed of the fuel support casting, control-rod guide tube, upper portion of control rod drive housing and RPV bottom head. In the vertical direction, the bundle weight resists the upward forces caused by the flow of reactor coolant inside the vessel and any additional dynamic forces due to postulated accident loads which could potentially result in relative motion between the fuel bundle ant its supports. Under postulated dynamic conditions, the bundle could move relative to its support, potentially unseat fuel supports and produce impact forces on the top guide.
The fuel lift loads are due to hydrodynamic effects, where the fuel transfers frictional forces to the top guide grid beams in the upward and downward vertical directions. The most severe loading combination that contributes to fuel lift loads include response from natural phenomena [ safe shutdown earthquake (SSE)), and f rom other events, such as LOCA and
' SRV discharge.
In general, hydrodynamic loads incurred due to SRV discharge, annulus pressurization, and LOCA phenomenon are applicable to Mark ll and lli containments. However, these loads are not significant for the vessel and intemals in Mark I containments where the torus and drywell are not dynamically coupled to a substantial degree. Therefore, the structural analysis results for a Mark l containment plant would show more margin than the example presented in the report. Based on its review of the loads development for analyzing BWR/2-6 top guide components as discussed above, the staff finds they are reasonable and acceptable.
3.1.2 Load Combinations For a plant specific analysis, BWRVIP recommends that the loads be combined according to the plant licensing basis. In the absence of such a basis, the combinations provided in the report can be considered. The totalload on the top guide is determined by combining the
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- dynamic loads using the Square-Root-Sum-of Squares (SRSS) method, and adding this
- dynamic term to the static loads using an absolute sum approach. The loads on the top guide are determined by evaluating the various possibic load combination cases to select the upset and faulted cases which would be most limiting. For the conservative example here, the limiting combinations were the following
Case 1: Upset Load Combination, OBE + SRV + FL Case 2: Faulted Load Combination, SSE + SRV + LOCA + FL Case 3: Faulted Load Combination, SSE + LOCA + FL The load combinations include the effects of deadweight and upset faulted pressure differentials. The main contributor to the estimate of LOCA loads is chugging in Case 2 and AP in Case 3.
Once the limiting load combinations are determined, those loads are applied to the model in three directions corresponding'to the following. !
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' Horizontal seismic and new loads are applied to the model in the direction of interest (North / South or East / West) o Seismic loads alone are applied to the model in the horizontal direction o Vertical seismic + new loads are applied in the vertical direction I The staff finds the recommended load combinations in accordance with SRP guidelines and therefore acceptable.
3.1.3 Example Stress Analysis Sample top guide analyses are presented in Section 4.3. These results were used to establish the inspection recommendations discussed in Section 3.0 of the report.
For any particular load source, the load magnitudes at various locations in the top guide are typically determined through finite element analysis. The finite element model used in the example evaluation consists of the top guide, including grid beams and rim, and a section of the shroud, each of which were modeled using elastic quadrilateral shell elements. For some cases, wedges are also added to the model, using one-dimensional compression-only spar elements. Cracking can be simulated in the model by removing coincident nodes or couplings at the crack locations. In order to create a maximum load scenario for the aligner and lateral support hardware, locations such as the rim weld were modeled as completely cracked. In '
other cases, the most conservative modeling was with the maximum restraint, meaning no cracks but at the location of interest. -
The analysis is based on the loads for a BWR/4 with a large core, high seismic accelerations and Mark ll plant hydrodynamic loads. There are two types of loading, as illustrated in Table 4-1 of the report. The first type consists of the horizontal and vertical accelerations that the loads induce on the entire structure. These are applied to the finite element model as accelerations of the model mass. The second loading type consists of the horizontal and
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vertical shear forces caused by the fuel impacting the grid beams. The fuel loads are applied as line loads on the grid beams in the model.
It was determined by analyses that the Case 3 faulted load combination bounded Case 2, so Cases 1 and 3 form the basis for the top guide load evaluations.
3.2 Analysis of Inspection Scope and Strategy 3.2.1 Inspection History ,
The BWRVIP-26 report provides design information on the top guide geometries and weld locations for BWR/2-6 plants. The guidelines address all welded and bolted locations identified from design drawings of the top guide lateral supports, aligner pins, hold-down bolts, grid beams, and reinforcement blocks. In addition, shroud repair were considered in the BWRVIP-26 report development, and it was determined that the BWRVIP-26 report is applicable whether or not a shroud repair has been N led. Typical top guide configurations are shown schematically, and these figures identify u.i welded and bolted top guide locations.
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The BWRVIP 26 report presents inspection approaches which vary depending on the type of plant and its associated combination of lateral support hardware. Inspection options are also presented considering implementation of repairs and performance of plant-specific analyses.
The top guide beams and lateral support components are generally accessible for visual inspection, and are included in most BWRs' augmented in-vessel visual inspection (IVVI) program. In 1991, through-thickness cracks at the bottom of unnotched areas of the 304 stainless steel top guide grid beams were discovered in the Oyster Creek top guide.
Subsequent inspections at the plarit's next refueling outage revealed two more cracks similiar to the first. In response, General Electric (GE) recommended that owners of GE BWR/2-5 plants with top guide fluence levels above 1x102' n/cm' visually inspect the top guides at grid locations where fuel and blade guides have been removed, and if cracking is found, perform an ultrasonic inspection (UT) of the top guide beam intersections which have the highest fluence. Further UT of top guide beam intersections which have an accumulated dose exceeding 2x102 ' n/cm2 were recommended.
To date, no domestic BWR, other than Oyster Creek, has reported any significant top guide cracking. However, there was cracking visually observed in the heat affected zone (HAZ) of the rim assembly of the top guide of a non-GE BWR, located outside of the U.S., with a top guide design similiar to those of GE BWR plants. Since similar welds exist in most GE BWRs, it is possible that similar cracking may occur in GE BWRs and, as such, visualinspections of BWR/2-5 plants have been performed to verify component integrity. Due to their designs, BWR/6 plants have not had to perform this inspection.
3.2.2 Inspection Strategy The specific inspection methods recommended in the BWRVIP-26 report relies on the UT and visual (VT) methodologies described in the BWRVIP-03 report, dated October 1995, with the conclusions and exceptions as stated in the staff's SE dated June 8,1998. The staff has received Revision 1 to the BWRVlP-03 report, dated March 31,1999, which is intended to address the issues raised in the staff's SE, and it is presently under staff review. I l
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7 3.2.3 BWR Inspection Guidelines The BWRVIP-26 report provides flexible options for inspection while assuring that structural integrity and/or function of the top guide are adequately maintained. The guidelines are generic in nature, based on the overall understanding of the various designs of the top guide.
These designs include the basic top guide configurations, which are:
o- BWR/2 o BWR/3,4 without wedges ,
o BWR/4,5 with Aligner Pin Assemblies Plus Wedges o BWR/6 Table 3 2 in the BWRVIP-26 report summarizes the inspection recommendations for each top guide location. The recommendations are based on the results of conservative structural analyses of a representative geometry and the safety consequences of component failure.
The table also provides the option of reduced inspection for plants that have done, or may consider, plant specific analyses, evaluations and/or repair modifications. The inspection guidelines are based on the function of each location, and how that function contributes to providing lateral support for the fuel.
3.3 Open issues With the exception of issues described below, this review finds that the guidance provided in the subject report to be acceptable.
Issue 3.1.2 Examination Methods The BWRVIP informed the staff during a public meeting on December 17,1998, that committee members are working on revisions to the BWRVIP-03 report, and plan to make the enhanced visual inspection method (EVT-1) a visual exam capable of achieving a % mil wire resolution. The BWRVIP 26 report currently requires only a VT-1 examination, which is less capable than the EVT-1 in detecting cracking.- The NRC staff finds that allInspections recommended in the BWRVIP-26 report should be performed to at least EVT-1 standards.
Issue 3.2.2 Insoection Recommendations by Locations The BWRVIP-26 report does not require an initial one time baseline inspection of the top guide at each BWR plant. The NRC staff finds that an initial baseline inspection of the top guide, inclusive of all locations, should be performed, and that a plant-specific analysis which determines that the BWRVIP 26 recommended inspection frequency is adequately dispositioned, be completed and maintained by each licensee.
4.0 CONCLUSION
S The staff has reviewed the BWRVIP-26 report and finds that the guidance of the report is acceptable for inspection and flaw evaluation of the subject safety-related core intemal components, except where the staff's conclusions differ from the proposed guidance, as discussed above. The staff requests that the BWRVIP review and resolve the issues raised
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8 above, and incorporate the staff's conclusions into a revised BWRVIP-26 report. Please I inform the staff in writing as to this resolution.
5.0 REFERENCES
! 1. Cari Terry, BWRVIP, to USNRC, "BWR Vessel and intemals Project, Top Guide I
Inspection and Flaw Evaluation Guidelines (BWRVIP-25)," EPRI Report TR-107285, December 1996, dated December 27,1996.
- 2. C. E. Carpenter, USNRC, to Cari Terry, BWRVIP, " Proprietary Request for Additional Information - Review of BWR Vessel and Intemals Project Reports, 'BWR Core Plate l inspection and Flaw Evaluation Guidelin3s (BWRVIP 25),' and ' Top Guide Inspection I
and Flaw Evaluation Guidelines (BWRVIP-26),'(TAC Nos. M97802 and M97803),"
dated March 14,1997.
- 3. Vaughn Wagoner, BWRViP, to USNRC, "BWRVIP Response to NRC Request for i Additional information on BWRVIP-25 and BWRVIP-26," dated December 19,1997. j Principal Contributors: C. E. Carpenter J. R. Rajan I
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