ML20237D569

From kanterella
Revision as of 07:58, 2 December 2024 by StriderTol (talk | contribs) (StriderTol Bot change)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Rev 0 to MPR-1957, Design Submittal for Oyster Creek Core Plate Wedge Modification
ML20237D569
Person / Time
Site: Oyster Creek
Issue date: 08/31/1998
From: Leshnoff S, Lipford B, Mccurdy H
MPR ASSOCIATES, INC.
To:
Shared Package
ML20237D567 List:
References
MPR-1957, MPR-1957-R, MPR-1957-R00, NUDOCS 9808270006
Download: ML20237D569 (46)
v • d • e


Text

{{#Wiki_filter:. - _ _ - _ _ _ lliEMPR Asso ci AT E S IN C. ENG1NEER$ ~~ MPR-1957 August 1998 Design Submittai for Oyster Creek l Core Plate Wedge Modification l 1 l i i l i Prepared For ) i GPU Nuclear Corporation 1 One Upper Pond Road Parsippany, NJ 07054-1095 9808270006 900825 PDR ADOCK 05000219 P PDR rp g,. 22tc 0 o b l

l lMMPR A S S O C I A T E S I N C. ENG1NEER$ Design Submittai for Oyster Creek Core Plate Wedge Modification MPR-1957 Revision 0 August 1998 Principal Contributors H. William McCurdy, MPR Associates Stephen Leshnoff, GPU Nuclear Brian Lipford, Key Technologies I 320 KING STREET ALE XANDRI A, VA 22314 3230 703-519 0200 FAX: 703-519 0224

Table of Contents 1 introduction and Summary................................ 1-1 1.1 In t ro d u ctio n.............................................. 1 - 1 1.2 S u m m a ry................................................. 1 - 1 1.2.1 Modification Overview................................ 1-1 1.2.2 Structural and Design Evaluations...... ............... 1-2 1.2.3 System Evaluations................................... 1-3 1.2.4 Materials and Fabrication.............................. 1-3 \\ 1.2.5 Pre-Modification and Post-Modification Inspections........ 1-3 2 B a c k g ro u n d............................................ 2-1 2.1 Requirement of BWRVIP-25 (Core Plate Inspection and Flaw Evaluation Guidelines) to Inspect Core Plate Holddown Bolts or Install Core Plate Wedges................................... 2-1 2.2 Oyster Creek Core Plate Wedge Installation Per the Requirements of BWRVIP-50 (Top Guide / Core Plate Repair Design Criteria)... 2-2 3 Description of Modification................................ 3-1 3.1 Design Objectives.......................................... 3-1 3.2 De sign Crit e ria............................................ 3-1 3.3 Description of Modification Components and Design Features.... 3-2 4 Structural and Design Evaluation........................... 4-1 4.1 Design Loads and Load Combinations......................... 4-1 4.2 Modification Hardware Evaluations........................... 4-1 4.2.1 Modification Hardware Structural Evaluation............. 4-1 l MPR 1957 (( Revision 0

Table of Contents (cont.) I l 4.2.2 Flow Induced Vibration................................ 4-2 4.2.3 Radiation Effects..................................... 4-2 4.2.4 Loose Parts Considerations............................. 4-2 4.2.4.1 Design Features to Preclude Loose Parts.......... 4-2 4.2.4.2 Effects of Postulated Modification Assembly Fail u re s...................................... 4-3 4.3 S h ro u d Evalu a tions......................................... 4-3 4.3.1 Flawed Shroud Evaluations............................. 4-3 4.3.1.1 Analysis Model................................ 4-3 4.3.1.2 Load Application............................. 4-4 4.3.1.3 Analysis Results............................... 4-4 4.3.2 Intact Shroud Evaluations.............................. 4-5 4.3.2.1 Analysis Model................................ 4-5 4.3.2.2 Load Application.............................. 4-5 4.3.2.3 Analysis Results............................... 4-5 4.4 Core Plate Evaluations...................................... 4-6 4.4.1 Analysis Model................................ 4-6 4.4.2 Imad Application.............................. 4-7 4.4.3 Analysis Results............................... 4-7 4.4.4 Flawed Core Plate Evaluations................... 4-8 4.5 Control Rod Guide Tube Evaluations......................... 4-9 4.5.1 Load Discussion..................................... 4-9 4.5.2 Structural Adequacy per BWRVIP-25 and Oyster Creek Am e n dm e n t 37....................................... 4-9 MPR 1957 jjj Rcvision 0

l Table of Contents (cont.) l 5 Seismic Analysis........................................ 5-1 5.1 Effect on Wedge Installation on Seismic Model................. 5-1 5.2 Seismic Evaluations........................................ 5-1 5.2.1 Spring Constant for Load Path With Wedges in Place....... 5-1 5.2.2 Seismic Loads With Wedges in Place..................... 5-2 6 Systems Evaluation...................................... 6-1 6.1 Core Plate Displacement.................................... 6-1 6.1.1 Allowable Displacement for Control Rod Insertion......... 6-1 6.1.2 Calculated Displacement With Wedges in Place and Comparison With Allowable............................ 6-1 6.2 Core Bypass Flow.......................................... 6-1 6.2.1 Bypass Flow Around Core Plate and Comparison With Allowable....................................... 6-1 6.2.2 Bypass Flow Through Flawed Shroud Vertical Welds V-11 and V-12 and Comparison With Allowable........... 6-2 7 Materials and Fabrication................................. 7-1 7.1 Material Selection.......................................... 7-1 7.2 Material Procurement Specifications.......................... 7-1 7.3 Material Fabrication........................................ 7-2 7.3.1 Cutting, Forming and Cleaning......................... 7-3 MPR 1957 jy Revision 0

r Table of Contents (cont.) f 8 Pre-Modification and Post-Modification inspection ............8-1 t 8.1 Pre-Modification Inspection................................. 8-1 8.2 Post-Modification Inspection................................. 8-1 8.2.1 Prior to RPV Reassembly.............................. 8-1 8.2.2 During Subsequent Refueling Outages................... 8-2 9 R e fe re n c e s............................................. 9-1 MPR-1957 y Revision 0

Tables 4-1 Load Combination s..................................... 4-11 4-2 Loading Conditions for Flawed Shroud Evaluations........... 4-11 4-3 Summary of Maximum Stresses in the Shroud Shell Between Circumferential Welds H5 and H6A for Flawed Shroud Evalu atio n s............................................ 4-12 4-4 Summary of Maximum Stresses in the Shroud Shell for intact S h ro u d Eval u ation s..................................... 4-12 4-5 Maximum Individual Wedge Loads........................ 4-1 3 Core Plate - Top Plate Stresses........................... 4-13 4-6 4-7 Core Plate - Ring Stresses............................... 4-14 4-8 Core Plate - Stiffener Beam Stresses....................... 4-14 5-1 Spring Constants at Bottom Radial Restraint................. 5-3 5-2 Summary of Results From Dynamic Seismic Analyses for Oyster Creek Core Shroud Assembly....................... 5-4 MPR 1957 yj Revision 0

) Figures 1-1 Core Plate Wedge Modification Arrangement................. 1-4 3-1 Azimuthal Location of Wedges............................. 3-3 4-1 Shroud Stress Contours at Wedge and Radial Restraint Location s............................................. 4-15 5-1 MPR Dynamic Model of Oyster Creek Reactor internals........ 5-5 l MPR-1957 yjj Revision 0

a - I' \\ I [ \\ I I E I I I I I E E I I I I I I

I I Et, i n g !i

l g,

I I I I I I I I I I l I L l I I I I I

E I 1 I introduction and Summary 5 1.1 Introduction I This report documents the design for the Oyster Creek core plate wedge modification. The report content is formatted to address the design criteria developed by the BWR Vessel and Internals Project for top guide / core plate g modifications, BWRVIP-50 [1], including the specific design criteria and requirements for wedge-type modifications. I 1.2 Summary The design for the Oyster Creek core plate wedge modification addresses the potential degraded conditions of the core plate which affect the core plate lateral alignment (a3 defined in BWRVIP-25 [2]). The modification is not included under the ASME Boiler and Pressure Vessel Code Section XI definition for modification and replacement. Rather, the modification is developed as an alternative modification pursuant to 10 CFR 50.55a (a)(3). As summarized below, the modification satisfies the requirements specified in the Oyster Creek specification (Reference 3) for the core plate wedges and the BWR Vessel and Internals Project " Top Guide / Core Plate Repair Design Criteria" (Reference 1). The modification is consistent with the current plant licensing basis and ensures that the core plate will be maintained in the lateral position and its operational and safety functions are satisfied even if the core plate lateral restraint components fail (i.e., aligner pins and holddown bolts). The modification design maintains the safety margin and functional capability of the core shroud (i.e., to withstand the localized wedge loading conditions). 1.2.1 Modification Overview As shown in Figure 1-1, the modification consists of a set of eight core plate wedge assemblies which are used to laterally restrain the core plate. The eight wedge assemblies are located adjacent to eight of the shield angles attached to the shroud in the annulus between the core plate and the shroud. MPR 1957 }.} Revision 0

't.2.2 Structuraland Design Evaluations ) The core plate wedges limit the lateral displacement of the core plate by transmitting the lateral fuelload from the core plate directly into the shroud. The shroud tie rod modification radial restraints transmit these loads between the shroud and the reactor vessel. Modification Assembly The load carrying capability of the wedges is sufficient to maintain the lateral position of the core plate. Shroud (Intact and Flawed Welds) The loads on the shroud were evaluated for intact and flawed shroud conditions. The flawed conditions evaluated included the case with all circumferential welds and vertical welds in the H5/H6A shroud section assumed to be completely failed, and the cases where only the vertical welds in H5/H6A were failed (i.e., circumferential welds intact). The shroud and shroud tie rod modification radial restraints are capable of transmitting the loads between the core plate wedges and reactor vessel. The stresses in the shroud resulting from the modification will be within the stress allowables of Section III, Subsection NG of the ASME Boiler and Pressure Vessel Code. Core Plate Loads on the core plate were evaluated for intact shroud conditions, which result in the highest fuel loads being transmitted through the core plate structure. The core plate assembly is capable of transmitting design loads through the core plate and into the shroud. Stresses within the core plate will be within stress allowables of Subsection NG. Margm against fatigue and buciding were also evaluated and found to be acceptable. Control Rod Guide Tubes The control rod gum 3 tubes limit the vertical displacement of the core plate to 0.5 inch in th event of failure of the majority of the core plate holddown bolts. MPR 1957 1-2 Revision 0

1.2.3 System Evaluations l The impact on plant operations with the core plate wedges installed was evaluated. These evaluations showed that there would be no impact on normal plant operations. The parameters considered in the evaluation include core plate displacement and core bypass flow. See Section 6 of this report for additional information on these evaluations. 1.2.4 Materials and Fabrication The materials specified for use in the wedge modification assemblies are resistant to stress corrosion cracking and have been used successfully in the BWR reactor coolant system environment. The wedge assemblies are fabricated from solution annealed 300 series stainless steel and Alloy X-750. No welding is permitted in the fabrication or installation of the modification and special controls and process qualifications are imposed in the fabrication of the modification to assure acceptable material surface conditions after machining. See Section 7 of this report for additional information on modification hardware materials and fabrication. 1.2.5 Pre Modification and Pnst Modification Inspections The inspections to be performed to support the modification are summarized below: Pre-Modification Inspection: Visual examinations will be performed prior to installation to confirm that the area around each installation site is free of obstructions and debris and that the core plate (top plate) and shroud are structurally sound. In addition, the as-built gaps between the core plate and shroud will be measured to pennit final wedge machining to obtain an installation gap of 0.02 to 0.03 inches. Post-Modification Inspection: Prior to vessel reassembly, visual inspections will be performed to confirm proper installation of each wedge. During Subsequent Refueling Outages: During the next tefueling outage, a baseline visual inspection (VT-3) will be performed on accessible areas of the wedges to confirm the post-modification inspections. During subsequent outages, inspections are expected to be limited in scope. MPR-1957 1-3 Revision 0

a BOLT RETAINCR $PRINC SHRoua .L IS b\\*' G I ~ \\ sex sett ~ swicto suppent ^"C" coac awc y kanse iit b -m l--['lJ I l fsanovo l '"' " '"" " "'"* .x, ,L,- View of Single Wedge Assembly interface with Shield Support Angle Figure 1-1 Core Plate Wedge Modification Arrangement MPR 1957 1-4 Revision 0 -

f 2

Background

I t l 2.1 Requirement of BWRVIP-25 (Core Plate Inspection and Evaluation Guidelines) to inspect Core Plate Holddown Bolts or install Core f Plate Wedges BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25 [2]) document the results of conservative structural analyses of individual sub-assemblies of the core support plate, such as the core plate to rim weld, stiffener beam to rim ) weld, stiffener beam to core plate welds, and rim holddown bolts. The safety consequence of these and other sub-assembly failures is also addressed. Based on these analyses, the rim holddown bolts are found to be the only core plate location which need to be addressed with a plant-specific inspection strategy. The inspection recommendation options that are provided by BWRVIP-25 [2] are: 1. Install core plate wedges to structurally replace the lateral load resistance provided by the rim holddown bolts and perform no inspections. 2. Do not install wedges and inspect the core plate bolts to assure an adequate number are intact to prevent lateral displacement of the core plate. The latter option to inspect the bolts is considered difficult. UT equipment has not yet been developed and visual exams, if performed, must be conducted both above and below the core plate. Access below the core plate is limited and difficult to achieve. It would require fuel and component removal and excessive in-vessel time. The damage mechanism that can potentially degrade the structural reliability of the holddown bolts is crevice induced IGSCC. The holddown bolts are the one location on the core plate which, in the absence of wedges, must be intact to assure the design safety function of the core support plate. Wedges structurally replace the lateral load resistance provided by the rim holddown bolts, making bolt inspection unnecessary. One of the two options for addressing the IGSCC susceptibility of the holddown bolts is to install wedges. If wedges are added, it is concluded in BWRVIP-25 [2], lateral support for the core to prevent misalignment that could hinder insertion of control rods is assured even if all holddown bolts fail (core plate lift is limited by guide tubes). Therefore, holddown bolt inspection is not required. MPR-1957 2-1 Revision 0

2.2 Oyster Creek Core Plate Wedge Installation Per the Requirements of BWRVIP-50 (Top Guide / Core Plate Repair Design Criteria) BWRVIP-50 [1] provides general design criteria for modifications to the top guide and core plate structures in BWR plants. These criteria are applicable to any type of modification to the structures either to address specific degradation found during inspections or on a preemptive basis. In addition, Appendix A to this document provides additional design criteria for a specific modification which adds wedge-type structures in the annular space between the top guide or core plate and the core shroud. Although the design and analysis of the Oyster Creek core plate wedge modification was substantially completed prior to the issuance of BWRVIP-50 [1], the wedge modification design and analysis are in accordance with all pertinent requirements provided in BWRVIP-50 [1]. MPR 1957 2-2 Revision 0

) l 3 Description of Modification r 3.1 Design Objectives The specific design objectives for this modification are as follows: Provide a fully redundant lateral support mechanism for the core plate assembly, such that the structural integrity of the existing holddown bolts (and jacking screws and positioning cams) need not be verified. Provide a design life up to 40 years, such that the modification will remain functional for the plant's remaining life, as well as any possible extensions. Ensure that the modification,in conjunction with the control rod guide tubes, provides fully redundant support and positioning control for the core plate assembly such that all design safety functions continue to be performed. Ensure that the modification is properly integrated with existing reactor internals, including the shroud repair hardware. 3.2. Design Criteria The fundamental design criteria applicable to the proposed modification are as follows: The proposed wedge modification is not included under the ASME Boiler and Pressure Vessel Code, Section XI, but is developed as an alternative to the requirements of the ASME code pursuant to 10CFR50.55a(a)(3). The modification satisfies the requirements specified in the design specification (Reference 3) and the criteria specified in BWRVIP-50, Top Guide / Core Plate Repair Design Criteria (Reference 1). The modification is evaluated in accordance with the ASME Boiler and Pressure Vessel Code, Section III, Subsection NG (Reference 6), although the components are not code stamped. This is consistent with the original design standards for the reactor internals which took guidance from the ASME code, but did not classify the internals as code components. MPR-1957 3-1 Rcvision 0

The modification will function and accommodate all reactor design conditions and fuel configurations currently specified in the plant's licensing basis. Structural integrity for all affected components will be demonstrated. 3.3. Description of Modification Components and Design Features 7 The proposed modification consists of eight-wedge assemblies to be installed around the periphery of the core plate assembly,in the annulus between the core plate and the shroud. The proposed installation arragement is shown on Figure 3-1. Each wedge assembly consists of one wedge, one base, one jack bolt, and two jack bolt retainer springs, as shown on Figure 1-1. The base of the wedge rests on the shroud lower ledge, with large gaps between the shroud and the core plate (such that the base cannot transmit radial loads). The wedge portion (above the base) will sit at the top core plate elevation, and will be machined to close tolerances to provide a tight fit between the top plate and the shroud (installation gaps of approximately.02 to.03 inches).' Thus, only the wedge portion will transmit loads between the top core plate and the shroud. During installation of each wedge, the jack bolt is used to raise and engage the locking arm of the wedge on the existing shield angle.2 The jack bolt is rotated until the locking arm contacts the shield angle, and is then lightly torqued to ensure proper engagement. Once the locking arm is engaged, the jack bolt is turned slightly until the two locking springs engage in slots on the lock bolt. This ensures that the assembly cannot loosen and disengage during operation or accident conditions. The wedges will be installed with long-handled tools from the refuel floor. The installation will not require any modifications or alterations to existing reactor internals. The wedges can also be removed in the future (for whatever reason) without damage to any reactor internals or the wedges themselves. 'In-reactor measurements will be taken at each proposed wedge site. The wedges will be marked accordingly and " final" machined specific to each installation site. 2The shield angles are existing structural angles, welded to the inside of the shroud to i support shield plates at each recirculation line nozzle. i i MPR 1957 3-2 Reesion 0

i o 0* 24* / /// o 312* iik", .- b _' 60* ~ ,/ TYPICAL WEDGE 276*- / 270* / .,9 a ~ / + 90* l / e~ \\ i, ~ 96* / s 240' i I ~, \\ ~ ]"# 132' SHROUD /// / 204* 180* Figure 3-1 Azimuthal Location of Wedges MPR-1957 3-3 Revision 0

f 4 7 Structural and Design Evaluation 4.1 Design Loads and Load Combinations 2 The following applied loads were considered in the evaluation of the shroud for the wedge modification: ~ DifferentialPressure Loads: The pressure difference across the shroud creates a hoop load (stress)in the shroud which must be reacted through vertical welds in each shroud cylinder. RadialRestraint Loads: The reactor vessel, shroud, and fuel are excited in the = x horizontal direction by an earthquake. Relative motion between the shroud and reactor vessel causes the shroud modification radial restraints to contact the shroud resulting in lateralloads on the shroud shell. Seismic FuelLoads: The reactor vessel, shroud, and fuel are excited in the horizontal direction by an earthquake. Motion of the fuel causes lateral loads which are reacted through the core plate and the top guide. LateralRLB Loads: A recirculation line break (RLB) causes lateral loads on the shroud shell. Lateralloads are reacted into the shroud when it contacts the radial restraints. The load combinations used for the wedge modification design and analysis are provided in Table 4-1. 4.2 Modification Hardware Evaluations 4.2.1 Modification Hardware Structural Evaluation The proposed wedge hardware is designed to react radial compressive loads between the core plate and shroud. The wedges are not able to transmit or react tensile or shear loads across the wedges. Consistent with BWRVIP-50 (Reference 1), the structural components of the wedges have been evaluated in accordance with Subsection NG of the ASME Boiler and Pressure Vessel Code (Reference 6). MPR 1957 4.} Rcvision 0

l The maximum calculated wedge reaction loads were determined and shown in Table 4-4 for the proposed eight-wedge configuration. The maximum ( load at any single wedge was found to be 79 kips (from Reference 15). The i stress in the wedge for this load is about 13 ksi (allowable of 42 ksi; 2.4S ).2 m j 4.2.2 FlowInduced Vibration The potential for flow induced vibration is not considered a concern for this design. The proposed wedges will be installed in the annulus between the shroud and the core plate. Flow in this region is low and should not affect the wedge design. i 4.2.3 Radiation Effects The proposed wedges are benign steel components, constrained by mechanicalinterference and locking devices. They are not considered susceptible to degradation from radiation. 4.2.4 Loose Parts Consideration 4.2.4.1 Design Features to Preclude Loose Parts Each wedge assembly includes one wedge, one base, one jack bolt, and two jack bolt retainer springs. The wedges are located at the core plate elevation and are restrained in the vertical and circumferential directions by an interference fit with existing shield support angles. The wedges are positioned underneath the shield support angles and jacked into position with a jack bolt threaded into the wedge assembly base. The jack bolt is constrained by two retainer springs which latch into slots in the jack bolt head. The base of each wedge rests on the shroud lower ledge, providing vertical support. The top of the wedge, adjacent to the core plate, is machined to provide horizontal restraint between the core plate and the shroud. The dual retaining springs are used to ensure that the jack bolts do not loosen. The springs maintain the jack bolt position under the existing shield angle to preclude any loose parts from the wedge. Failure of the wedge assembly (which could result in loose parts concern) is not considered likely due to: low / negligible operating stresses in the wedge components, a dual locking springs that ensure wedge retention and position. ' Based on a wedge compressive force of 79 kips on s wedge face that is 1.5" high by 4" wide. MPR 1957 4-2 Revision 0

[ 4.2.4.2 Effects of Postulated Modification Assembly Failures l A postulated failure of a wedge assembly is not considered to have adverse affects on plant operation or safety. The wedges are installed at the core plate periphery in a low-flow regime. Failure l of a wedge during normal operation is expected to result in the wedge or its piece parts falling into the annulus between the core plate and shroud. This would not create safety or operating concerns for the reactor or other primary systems. 4.3 Shroud Evaluations 4.3.1 Flawed Shroud Evaluations Evaluations were performed to determine the shroud stresses in the vicinity of the wedge locations for the case where the shroud horizontal welds are cracked. These evaluations are documented in Reference 16. 4.3.1.1 Analysis Model The analysis uses a finite element model to analyze the shroud cylinder between welds H5 and H6A with the core plate wedges installed. In addition, this model addresses the effect of potential cracking in the vertical welds on the structural integrity of the shroud cylinder between H5 and H6A. Note that vertical weld cracking is included in the evaluation to consider its effect on the acceptability of the wedge modification as well as to support the development of flaw tolerance criteria for the vertical weld inspections to be performed during the Fall 1998 Oyster Creek plant outage. In all evaluations, both H5 and H6A are assumed to be cracked through wall, all the way around the shroud. This is the most limiting condition for the vertical welds because intact ligament in the horizontal weld provides an alternate load path around the vertical welds for hoop loads in the shroud cylinder. To evaluate the vertical welds in the H5/H6A shroud cylinder, the shroud sections between horizontal welds H4 and H6B are modeled. The additional sections are modeled to include the effect of compression across horizontal welds from tie rod preload in the evaluation of the vertical welds. The core plate wedges and the radial restraints are also included in the finite element model. MPR 1957 4-3 Revision 0

4.3.1.2 Load Application The analysis considers the design loads and load combinations identified in Section 4.1 above. The controlling service loadings for comparison with the stress limits can be determined by examining the load components for each service condition. Loading conditions for the evaluation of the H5/H6A shroud cylinder are summarized in Table 4-2. The following model boundary conditions are applied to all the models: The vertical welds in the H5/H6A shroud cylinder are assumed to be completely failed. All circumferential welds are considered completely failed. The tie rod preload keeps the entire shroud in compression during normal operation. The differential pressure during an RLB is bounded by normal operating conditions. Consequently, there is compression across all the failed circumferential welds during the RLB + SSE load case. During a MSLB accident the differential pressure across the shroud is significantly increased. This results in a large upload applied to the core plate due to differential pressure. Since it is assumed a weld below the core support ring is failed, the large core plate upload and the restraining loads from the tie rods keep all the circumferential welds above the core support ring in compression. Since there is compression across each of the failed circumferential welds, H5 and H6A are modeled as pinned joints, i.e., they can carry shear but not moment. The outside surfaces of the radial restraints are restrained from motion in the radial direction by virtue of their constraint by the reactor vessel. 4.3.1.3 Analysis Results The maximum stresses in the shroud section between circumferential welds H5 and H6A are summarized for the limiting load cases in Table 4-3. Typical stress contours are shown in Figure 4-1. As shown, these stresses meet the requirements of Subsection NB of the ASME Boiler and Pressure Vessel Code, 1989 Edition [4]. MPR 1957 44 Revision 0

4.3.2 Intact Shroud Evaluations j Additional evaluations were performed for a shroud cylinder with intact i horizontal welds to confirm that the shroud stresses in the vicinity of the wedge locations are acceptable. This evaluation is required since the fuel and wedge loads are substantially higher for the intact shroud seismic case than for the seismic cases with broken horizontal welds. Tliese evaluations are documented in Reference 15. 4.3.2.1 Analysis Model The analysis uses a finite element of the shroud between horizontal welds H5 and the vessel attachment weld and includes the core plate wedges. As in the flawed shroud model described in Section 4.3.1 above, the model also addresses the effect of potentia! cracking of the vertical welds on the structural integrity of the shroud cylinder between horizontal welds H5 and H6A. 4.3.2.2 Load Application The analysis censiders the seismic fuelload of 143 kips which is given in Reference 9. This load is applied at the wedge locations. Th;re is no load from the radial restraints. The following boundary conditions are applied: The vertical welds in the H5/H6A shroud are either intact or flawed over their entire length. At horizontal weld H5 (the top of the model), a free boundary condition is used. This is reasonable since the load path is downward through the shroud to the reactor vessel and the V9 weld is assumed to be intact. At the vessel attachment weld (the bottom of the model), a fixed boundary condition is used. This is reasonable based on the stiffness of the reactor vessel. 4.3.2.3 Analysis Results The maximum stresses in the H5/H6A shroud cylinder at the wedge locations are summarized in Table 4-4. As shown, these stresses meet the requirements of Subsection NB of the ASME Boiler and Pressuu Vessel Code,1989 Edition [4]. MPR-1957 4-5 Revision 0

4.4 Core Plate Evaluations 4.4.1 Analysis Model As described in Reference 5, the core plate assembly was explicitly modeled including the top plate, outer ring, stiffener beans, and stabilizer bars. Shell elements were used to model the core plate, the ring, and the stiffener beams under the core plate. Beam elements were used to model the bars between the stiffener beams. Pipe elements were used to model the vertical pipe connections between the upper and lower bars. Compressive only elements were used to model the wedges. Each wedge element was modeled to function as a load carrying member when loaded in compression, and as an open gap otherwise. A 360" full model was developed for the core plate assembly to properly address the asymmetty of the wedge and core plate assembly. The core plate was restrained vertically at the 36 core plate hold-down bolt locations; no lateral support was modeled at these bolt locations, nor were friction effects considered between the core plate and the shroud lower ledge. For purposes of maximizing reaction loads and stresses in the core plate assembly, the shroud assembly welds (the circumferential and vertical welds) were assumed to be intact sufficient to transmit design loads and moments. This results in the maximum fuel shear loads being transmitted from the fuel into the core plate, wedges and shroud (143 kips). Several load cases were evaluated using different boundary conditions and wedge arrangements as summarized in the table below: Load Case Description Reference 10 Wedge support configuration with conservative A 5 boundary conditions 1 Wedge support configuration with realistic B 5 boundary conditions 8 Wedge support configuration with realistic C boundary conditions 15 Cases A and B were evaluated to determine maximum reaction loads and streses in the core plate with a ten-wedge support configuration (i.e., ten wedges equally spread around the core plate perimeter). Case C was evaluated to determine loads an

resses for an eight-wedge support configuration. The boundary coiditions modeled for Case A were recognized as being conservative; the wedges were rigidly constrained in all MPR 1957 46 Revision 0

directions (which ignores the shroud radial stiffness) and zero gaps were input between the core plate, wedges, and shroud. Case B was similar to Case A, but the shroud radial stiffness and initialinstallation gaps were modeled. Case C was similar to Case B, but an eight-wedge support configuration was used versus ten. This last case was examined since it was l found that two of the proposed ten wedges would be located at vertical weld seams in the shroud (at V11 and V12). To avoid applying high radialloads at these weld seams, GPUN examined Case C to confirm the acceptability of the design with only eight wedges.2 4.4.2 Load Application The core plate assembly weight was included in the model by specifying a vertical gravity acceleration reduced by a ten percent buoyancy effect. The normal operation and main steam LOCA pressures on the core plate were applied as uniformly distributed upward pressure loads. The recirculation line LOCA pressure was applied on the core plate as a uniformly distributed downward pressure load. The inertia load on the core plate assembly during OBE and SSE events were applied as lateral accelerations on the core plate assembly. Two bounding seismic cases with different seismic directions were evaluated. One case input the seismic loads approximately parallel to the stiffener beams under the core plate. The second case input the seismic 1,oading in the direction approximately perpendicular to the stiffener beams of the core plate. The fuel shear load was included by increasing the lateral accelerations to the core plate assembly. 4.4.3 Analysis Results A summary of the calculated wedge reaction loads into the core plate are provided in Table 4-5. As expected, the reaction loads into the core plate for an eight-wedge configuration (Case C) are higher than a ten-wedge configuration (Case B) for similar boundary conditions. However, the bounding (highest) reaction ' Based on analysis results from Case B, it was discovered that peak wedge reaction loads were not dependent on seismic orientation (relative to the orientation of core plate and stiffener beams). As such, the core plate model for Case C was simplified to provide a stiff radial structure (relative to the shroud). MPR-1957 47 Revision 0

I loads are for the ten-wedged configuration (Case A) with conservative boundaiy conditions.' With regards to stresses in the core plate assembly, the top plate is primarily affected by the addition of the wedges. Local stress regions are created in the I top plate, adjacent to each wedge, which are compressive and membrane in nature. I For conservatism, stresses in the core plate were only determined for load Case A, with the ten-wedge configuration and conservative boundary conditions. This load case produces the highest wedge reaction loads and is l considered to bound the other evaluation cases, including the proposed eight-wedge configuration (Case C). Maximum calculated stresses in the core plate for load Case A are summarized in the Tables 4-6,4-7 and 4-8 (for the top plate, the core plate ring and the core plate stiffener beams). Fatigue usage and buckling were evaluated for the core plate assembly with the proposed wedge installation and found to be acceptable (Reference 5). 3 Fatigue usage was found to be negligible. Buckling was evaluated as a potential failure mechanism in the top core plate and the stiffener beams, both of which could see compressive loads during design conditions. Results from this work indicate that buckling is not considered a failure mechanism, with or without the wedge installation. 4.4.4 Flawed Core Plate Evaluations BWRVIP-25 [2] indicates that the various assembly welds for the core plate are not design reliant welds since failure of the welds will not adversely affect the safety functions of the assembly. GPU Nuclear agrees with the assessments in this document. As an additional evaluation, GPUN evaluated the flaw tolerance of the core plate assembly, with the proposed wedges, to determine the extent to which the core plate assembly welds could crack and fail. As reported in Reference 12, three cases were examined with regards to various assumed weld failures; case 1 considered a 10 percent failure for all of the welds ofinterest (i.e.,10 % failure of the stiffener beam-to-core plate welds,10% failure of the stiffener beam-to-ring welds, and 10% failure of the core plate-to-ring welds). The failures were uniformly distributed, within the limitations of the refinement of the finite element model. The other cases considered were with 20% and 30% degradation of the welds. Each of these ' Radial wedge reactions loads were generated for this load case due to vertical differential pressures across the core plate. The reactions produced by the pressure loads are secondary in nature (non-equilibrium), caused by out-of-plane deflections of the core plate. The loads were created by the conservative be,andary conditions that assumed a rigid shroud and no installation gaps. MPR 1957 4.g Revision 0

cases were evaluated with the ten-wedge configuration and conservative boundary conditions (Case A) which is considered bounding. The stresses j for comparison with allowable values were conservatively determined by picking the maximum stress rather than linearizing the stresses through a section. The stresses in the core plate and ring were found to be acceptable f for each of the three cases evaluated. The 30% cracking case showed the least margin (safety factor equal to 1.03) at the core plate where the applied loading is reacted on one of the wedges. Since the evaluations were done for the conservative loading condition (Case A),it is expected that additional flaw tolerance beyond 30% would be acceptable. All of these analyses assumed that vertical support was provided only by the holddown bolts, at the i periphery of the core plate assembly. If degradation of the assembly welds became excessive, it is expected that vertical support would be provided by the CRD guide tubes. l Potential local buckling of the stiffener beams may be a concern with { significant levels of weld degradation. The likely result of any such j degradation would result in an increased upward deformation of the core i support plate. In such a scenario, the core support plate would come in contact with the CRD guide tube assemblies after about a half-inch of j displacement, and a stable configuration would result. An analysis was conducted to evaluate such a scenario by postulating the failure of all the stitch welds on the stiffener beams. This analysis showed that the CRD guide tubes are capable of supporting the design verticalloads under such an extreme condition. i 4.5 Control Rod Guide Tube Evaluations i 4.5.1 Load Discussion As discussed in BWRVIP-25 (2), the core plate wedges provide lateral i support for the fuel and core plate but do not provide restraint for vertical j loads acting on the core plate (e.g, due to differential pressure). 4.5.2 Structural Adequacy Per BWRVIP-25 and Oyster Creek l Amendment 37 With regard to vertical welds, BWRVIP-25 [2] states that core plate lift is limited by the control rod guide tubes and also states that a vertical displacement of the core plate would lead to a detectable change in power performance which would allow for safe shutdown. The conclusion in BWRVIP-25 [2] that the control rod guide tubes are structurally adequate to react the core plate vertical loads is supported by the MPR-1957 4-9 Revision 0

discussion in Reference 7 (Amendment 37 to Oyster Creek License Application) which states the following: "The guide tube is restrainedfrom lifting by means of a bayonet coupling, which locks each guide tube to its associated control rod drive housing. Two features of the guide tubeprevent itfrom beingpulled through the hole in the coreplate. First, there are two 1/4"x 1-1/4" cross-sectionallugs on each guide tube which must be sheared off Ifit ispostulated that these lugs sheared off then the coreplate would contact a jlange whose diametrical dimension is over 0.2 inches larger than the hole in the coreplate. Preventing the guide tubefrom being drawn through sne coreplate hole is thefuel support casting which slips inside the guide tube andfits with an average radial clearance of 0.006 inches. This castingprovides a substantial backing support to the inside surface of the guide tube. The guide tube is capable ofsustaining an axialpullin excess of 25,000lbs." Therefore, based on a core plate maximum verticalload for a main stream line break of 667 kips from Reference 8, only 27 of the 129 guide tubes are needed to react the applied upload on the core plate. For the core plate vertical load of 199 kips during normal operation, only eight guide tubes are needed to react the upload. MPR 1957 4-10 Revision 0

I Table 4-1 Load Combinations ^ Load Case Load Combination I t Normal Level A Normal differential pressure (AP) Upset Level B Upset AP OBE Level B Operating basis earthquake (OBE) loads plus normal AP I SSE + MSLB Level D Main steam line break (MSLB) AP plus safe shutdown earthquake (SSE) loads SSE + RLB LevelD Recirculation line break (RLB) loads plus normal AP plus SSE loads Table 4-2 Loading Conditions for Flawed Shroud Evaluations Loading Value Note MSLB + SSE: - Steam line break differential pressure across shroud wall 19.0 psi 1 - SSE seismic bumper load 74 kips 2 - SSE seismic fuelload 40 kips 2 RLB + SSE: - Normal operating differential pressure across shroud wall 4.34 psi 1 - SSE seismic bumper load 74 kips 2 - SSE seismic fuelload 40 kips 2 - Lateral recirculation line break load 41,3 kips 1 Notes:

1. Load determined in Reference 8.
2. See Section 5 of this report. Loads determined in Reference 14.

I MPR 1957 4 11 Revision 0

L Table 4-3 Summary of Maximum Stresses in the Shroud Shell Between Circumferential Welds H5 and H6A for Flawed Shroud Evaluations") Number of AH wane Stress Calculated Stress Stress Type Wedges Stress U mit Stress (ksi) Ratio Installed (ksi) Service Level D - Main Steam Line Break plus Safe Shutdown Earthquake Primary Membrane plus Bending 8 3.6Sm 31.7 60.0 0.53 (Pm+ Pb) Service Level D - Recirculation Line Break plus Safe Shutdown Earthquake Primary Membrane plus Bending 8 3.6Sm 44.5 60.0 0.74 (Pm+Pb) Notes: 1. From Reference 16. Table 4-4 Summa:y of Maximum Stresses in the Shroud Shell for Intact Shroud Evaluations") Shroud Stress at Wedge Maximum Location (ksi) Load Case Wedge Load (kips) Calcu:ated Allowable

  • Vertical welds intact 79'0 53.2 60.0

- Load directed at wedge Vertical welds failed 78'9 53.2 60.0 - Load directed at wedge Vertical welds failed 76.5 51.9 60.0 Load directed at weld Notes:

1. From Reference 15.
2. From Reference 16.

MPR-1957 4-12 Revision 0

Table 4-5 Maximum Individual Wedge Loads Loading Condition Case A") Case BW Case CW Normal Pressure (aP) 28 kips 0 kips 0 kips OBE + Normal Pressure (aP) 43 kips 36 kips 40 kipsW SSE + MSLB 107 kips 67 kips 79 kips SSE + RLB 87 kios 64 kios 79 kios Notes: 1 1. From Reference 5. 2. From Reference 15. 3. The OBE load is assumed to be half the SSE load. 1 i Tab'e 4-6 Core Plate -Top Plate Stresses (0 Load Stress Maximum Acceptance Allowable Safety Combination Category Stress Basis Stress Factor (psi) Normal (n) Pm 6793 Sm 16950 2.4 Operation Pm +Pb 6807 1.5 S m 25425 3.7 OBE+ Normal (u) Pm 10978 Sm 16950 1.54 Pm+Pb 10985 1.5 Sm 25425 2.3 SSE+MS (f) Pm 26980 2.4Sm 40680 1.51 LOCA Pm+Pb 27G47 3.6Sm 61020 2.2 SSE+RL LOCA (f) Pm 25409 2.4 S m 40680 1.60 Pm+ Pb 28108 3.6Sm 61020

2.1 Notes

1. From Reference 5. 1 i l MPR.1957 4 13 Revision 0

Table 4-7 Core Plate - Ring Stresses (D Load Stress Maximum Acceptance Allowable Safety Comb! nation Category Stress Basis Stress Factor (psi) Normal (n) Pm 2363 Sm 16950 7.1 Operation Pm+Pb 6719 1.5 Sm 25425 3.7 OBE+ Normal (u) Pm 2502 Sm 16950 6.7 Pm+ Pb 11366 1.5 S m 25425 2.24 SSE+MS (f) Pm 6736 2.4Sm 40680 6.0 LOCA Pm+Pb 27138 3.6Sm 61020 2.25 SSE+RL (f) Pm 18438 2.4Sm 40680 2.21 LOCA Pm+Pb 42821 9.6 S m 61020 1.43 Notes: \\ 1. From Reference 5. Table 4-8 Core Plate - Stiffener Beam StressesW Load Stress Maximum Acceptance Allowable Safety Combination Category Stress Basis Stress Factor (psi) Normal (n) Pm 4627 Sm 16950 3.6 Operation Pm+ Pb 4703 1.5 S m 25425 5.4 OBE+ Normal (u) Pm 4961 Sm 16950 3.4 Pm+Pb 5327 1.5 S m 25425 4.7 SSE+MS (f) Pm 13326 2.4 S m 40680 3.0 LOCA Pm+Pb 14064 3.6Sm 61020 4.3 SSE+RL LOCA (f) Pm 34436 2.4 S m 40680 1.18 Pm+ Pb 35101 3.6Sm 61020 1.74 Notes: 1. From Reference 5. MPR 1957 4 14 Revision 0

Wedge Locations \\ h Radial Restraint Location Figure 4-1 Shroud Stress Contours at Wedge and Radial Restraint Locations n[vIsIEn'o 4-15

l 5 7 Seismic Analysis [ 5.1 Effect of Wedge Installation on Seismic Model [ Dynamic seismic analyses of the Oyster Creek reactor internals were performed in l Reference 9 (MPR-1579) to determine the seismic forces and displacements with intact and failed horizontal shroud welds. Results of the seismic analysis are provided in Table 1 and Appendix C of Reference 9. The dynamic model of the reactor internals is shown in Figure 5-1 (same as Figure 1 i in Reference 9). Since the wedge installation will change the load path between the fuel and the bottom seismic radial restraints, the spring constant for this load path (K ) is affected in the seismic model. To determine the new spring constant, finite-I s5 element analyses were performed with the wedges modeled similar to the analyses i performed to develop the spring constants used in the seismic evaluations in Reference 9. I 5.2 Seismic Evaluations 5.2.1 Spring Constant for Load Path With Wedges In Place The spring constants for the original load path and the modified load path with the installed wedges are summarized in Table 5-1 which indicates that the spring constant is reduced with the installed wedges. This is the result of the different es in the load paths as follows: The originalload path between the bottom of the fuel and the reactor vessel is through the core plate to the lower shroud support ring to the seismic radial restraints. The modified load path is through the wedges and core shroud to the seismic radial restraints. This load path results in a lower stiffness since the shroud is a less-stiff structure than the thick shroud support ring. This evaluation is documented in Reference 13. MPR 1957 5-1 Revision 0

5.2.2 Selsmic Loads With Wedges In Place } Time-history analyses were performed using the ANSYS computer i program to determine the loads and displacements with the model modified to represent the modified load path at the bottom radial ) restraints. The results of the seismic analyses are given in Table 5-2 which lists the fuel and radial restraint loads and fuel displacement for the original and modified shroud seismic load paths. Note the following: The fuel and radial restraint loads are generally lower with the modified load path as the result of the lower spring constant. The fuel displacements are slightly higher with the modified load path but less than the allowable of 0.75 inch identified in Section 6.1 of this report. The seismic loads for the fuel and radial restraints given in Table 5-2 were used in the structural evaluation of the shroud in Section 4.3.2 of this report. This evaluation is documented in Reference 14. MPR 1957 5-2 Revision 0

Table 5-1 Spring Constants at Bottom Radial Restraint Spring Constant (10'Ib/in) Seismic Case for Load Path Kf Multiple Breaks" - Original 0.90W - w/ Wedges 0.48W Notes:

1. Also considered flawed vertical welds in the HS-H6A shroud cylinder.

\\

2. Sliding connection at shreud weld H68, pinned connections at welds H1, H2, H3, H4 H5 and H6.
3. Taken from Reference 9.
4. Taken from Reference 13.

MPR 1957 5-3 Revision 0

1 Table 5-2 Summary of Results From Dynamic Seismic Analyses I for Oyster Creek Core Shroud Assembly 1 i Seismic Case - Multiple BreaksW I h \\ Parameter W Original

  • w/ Wedges
  • W t

Fuelloads L Shear (kips) -Top 23 24 - Bonom 35 40 Radial Restraint Loads -Top (kips) 165 112 - Bottom (kips) 61 74 W Fuel Displacements - Top (in.) .575 .558 j - Bottom (in.) .441 .529 ) l Notes: t

1. Sliding connection at shroud weld H6B, pinned connections at welds H1, H2, H3, H4, H5 and H6A.
2. Original load path from the core plate through the lower shroud support ring to the radial restraints.

3. Modified load path from the core plate through the wedges and through the shroud wall to the radial restraints.

4. This is the transient fuel lateral displacement during the seismic event. The permanent l

displacement is less than 0.375 in. ] i 5. Maximum loads and displacements for limiting normal, compressed or expanded time j scale runs and north-south or east-west tima histories. ~ i

6. All parameters values from Reference 14.

MPR 1957 5-4 Revision 0

K m - Axict Stiffness of Tie Rods Ic3 - Inertio of Core Support o K m - Rotational Stiffness of Tie Rods Irc - I"ertic of Top Guide r Kg - Bumper Stiffness WCS - Moss of Core Support h - Numbec in C;rcle Indicates Node ~ ~ S / ^g y ru:-u -u. u u us / / 4' / ^ ^ ^ f. / b . I. ~ / X e / t JL JL _JL JL.JL _JL, ^ / Z d'08 / = = = = =: = / e w a w ri e i ~~"' ,,, n / K r rm K, g _/ @X x / nnnonnnnnnnn / Tc gy] ': L g w w X @ x.m s2 x/-tu ) M h 0 MH !/ !/N H1 lS N i H d/ E j j y/ {U1 7 ~ H H(M] 20 u Q g j g h 6 M t g p / / Ks4 /-{u,] %u,H h / / / I / / -[u,) g v i / / /-[M ] g 3 / L ? j /////s o Figure 5-1 MPR Dynamic Model of Oyster Creek Reactor Internals MPR-1957 5-5 Revision 0

6 System Evaluations 6.1 Core Plate Displacement This section addresses the horizontal displacement of the core plate and the acceptability regarding control rod insertion. Note that there are no limits to address for the vertical displacement of the core plate as discussed in Reference 10 (since the fuel assemblies and control rod guide surfaces are supported in the vertical direction by the control rod guide tubes). 6.1.1 Allowable Displacement for Control Rod Insertion The allowable horizontal displacement of the core plate is identified to be 0.75 inch in Reference 10. The basis for this allowable displacement is the testing of CRD scram behavior at various reactor pressures which showed acceptable scram times and minimal effect on fuelintegrity. 6.1.2 Calculated Displacement With Wedges In Place and Comparison With Allowable For the modified core plate support configuration with installed wedges, the seismic analysis discussed in Section 5 of this report determined the maximum core plate transient displacement. The results of these seismic analyses in Table 5-2 show that the maximum calculated fuel / core plate displacement is 0.529 inch which is less than the allowable displacement of 0.75 inch. Also note that the permanent core plate lateral displacement is essentially equal to the seismic bumper clearance of 0.375 inch. Note that the permanent displacement of 0.375 inch is concluded to be acceptable in Section 2.3 of Reference 17. 6.2 Core Bypass Flow 6.2.1 Bypass Flow Around Core Plate and Comparison with Allowable As discussed in Section 4.5, core plate restraint in the vertical upward direction is provided by the control rod guide tubes in the event of the failure of most of the core plate holddown bolts (only 12 of the 32 bolts are required to react the normal core plate upload). Since the guide tubes MPR-1957 61 Revision 0

permit a vertical core plate displacement of about 0.5 inch, a core flow bypass path would develop around the core plate. As stated in Table 3-2 of Reference 2 (for failure Location 10), this would lead to a detectable change in power performance and would allow for safe shutdown. 6.2.2 Bypass Flow Through Flawed Shroud Vertical Weids V11 and V12 and Comparison with Allowable The maximum leakage path flow area through a flawed vertical weid in the H5/H6A shroud segment is determined using the finite element model discussed in Section 4.3 above. As in Section 4.3, the vertical weld is assumed to be completely cracked. During normal operating conditions, the only load that opens the cracked weld is differential pressure. The normal operating differential pressure is 4.34 psi-(Reference 8). Also during normal operation, the tie rod preload is sufficient to maintain compression across all failed circumferential welds, allowing them to be modeled as pinned joints. The finite element model for the ten wedge case is re-solved using these loads and boundary conditions to examine the displaced shape at the crack. (Note that the finite element model with eight wedges provides essentially the same results for the leakage path flow area.) The total leakage path flow area can be determined by examining the circumferential displacement of selected nodes relative to the corresponding nodes on the opposite side of the crack face. The maximum leakage path flow area through a fully-cracked vertical weld in the H5/ z H6A shroud segment during normal operating conditions is 0.495 in (Reference 16). This flow area was used to evaluate the effect of reactor coolant flow that bypasses the core through the cracked vertical weld. The volumetric flow rate is obtaineo by first determining the mean flow velocity based on the total equivalent flow resistance for the cracked weld flow path. From the mean velocity, the volumetric flow rate can be calculated. The total leakage through the vertical welds in the H5/H6A shroud segment is estimated to be approximately 30 gpm. This is considered to be negligible since it is only about 0.02% of total full power core coolant flow. This evaluation is documented in Reference 16. MPR 1957 62 Revision 0

1 7 Materials and Fabrication ) 7.1 Material Selection / In accordance with fabrication drawings and Reference 11, a list of materials for each wedge component is provided below: Component Material Clamp ASTM A182 F316 or ASTM A240 TP316 Retainer Spring ASTM B637, UNS NO7750 TYPE 3 (X-750) Jack Bolt Washer ASTM B637, UNS NO7750 TYPE 3 (X-750) Jack Bolt Sleeve ASTM A479 or A240 TP316 Locking Pin ASTM A479,TP316 Jack Bolt ASTM B637, UNS NO7750 TYPE 3 (X-750) Spacer (Base) ASTM A182 F316 or ASTM A479 or ASTM A240 TP316 7 ) The proposed materials are highly resistant to inter-granular stress corrosion cracking (IGSCC) and are suitable for reactor environment conditions. Materials l are traceable to certified material test reports (CMTRs). 7.2 Material Procurement Specifications Material procurement is based on the fabrication specification and the detailed component drawings listed below: Fabrication specification, GE Nuclear Energy document 24A5734 l = (Reference 11). GE Nuclear Energy drawings: - 117D3264P001, Clamp, Core Plate ) ) MPR 1957 7.] Revision 0 j l j

117D3264P002, Clamp, Core Plate 117D3266P001, Retainer, Spring 117D3267P001, Washer, Jack Bolt 117D3268P001, Sleeve, Jack Bolt - 117D3269P001, Pin, Locking 117D3270P001, Bolt, Jack - 117D3271P001, Spacer (Base) These documents and drawings are not provided with this submittal, but can be provided upon request. 7.3 Material Fabrication Requirements for material fabrication are taken from Reference 11, as summarized below: Material X-750 316 Stainless Steel Specifications ASTM B-637, UNS N07750, ASTM A-182, A240, A-479, TP316 Modified Type 3 Special Cobalt s.10% Carbon s.02% Chemistry Limits IGA Testing Yes, unless: Yes, unless: Min 0.03 inch of material Min 0.03 inch of material removed from all surfaces, or removed from all surfaces Verification that no acid after solution heat treatment. pickling was used. Age Hardening / 1300 15*F (20 hrs s t s 24 2000 100*F (15 Heat Treatment hrs) minutes / inch) Air cooled Water quench (or equal) Sensitization testing per: ASTM A-262 Practice A, and/or ASTM A-262 Practice E, and/or GE Procedure E50YP20. Special/ Min Yield at 550'F 2 92 ksi Re s 92 Additional Min Ultimate at 550'F z 142.5 ksi Mechanical Procerties MPR-1957 7-2 Revision 0

7.3.1 Cutting, Forming, and Cleaning In general, fabrication processes were developed and used to avoid IGSCC susceptibility of the materials (as documented in Reference 11). Examples of these processes are as follows: ) Welding on any subassembly or part was prohibited. Cutting methods such as shearing or punching were prohibited except where the cold worked material was subsequently and completely removed by machining, sawing, or solution heat treatment. Plasma art cutting was prohibited unless a minimum of 0.25 inches of the cut surface was subsequently removed by machining. Shot penning, hammering, grinding, or power destagging of final surfaces were prohibited. Minor straightening / reforming of parts was prohibited unless: 1) hardness of the stainless steel component in the final fabricated condition did not exceed the hardness requirement noted above, and 2) cold bending strain, after solution annealing did not exceed 2%%. All parts have been cleaned to removed contaminants of concern per approved prc,cedures that comply, as a minimum, with ANSI N45.2.1 Class B. Known contaminants are chlorides, fluorides, sulfur, lead, mercury and all metals with low melting points. MPR 1957 7-3 Revision 0

1 8 l Pre-Modification and Post-Modification Inspection ? 8.1 Pre-Modification Inspection ) Visual exams will be completed prior to installation to confirm the following: that each installation site is free of obstructions and debris, and ) l that the core plate (top plate) and shroud are intact and structurally sound. i Since the wedge design imparts localized loads and stresses into the core plate and shroud, the inspections will be limited to the immediate area around each installation site. The shroud and the core plate will not be examined in their entirety. I As part of the pre-modification inspections,in-reactor measurements will be taken to determine the as-built gaps between the core plate and the shroud. These measurements will be used to determine the final wedge dimensions, specific to each installation site.' Each wedge will be labeled and designated to a specific installation location. The final installation gap between the wedges, core plate and shroud will be approximately.02 to.03 inches. ) 8.2 Post-Modification Inspection ) 8.2.1 Prior to RPV Reassembly Prior to vessel reassembly, visual inspections will be performed to verify the installation of each wedge. Specifically, inspections will confirm that: each wedge is properly located, oriented and positioned, the retainer springs are properly engaged on the jacking bolt, 'The gap at each installation site is expected to differ slightly. The wedges were procured in " rough" machined form. The as-built gap information is to be used to " final" machine each wedge to fit at its designated site. The measurement equipment will be qualified and traceable ) to NIST. MPR 1957 81 Revision 0 l

the interference fit with the shielding support angles has been properly established, and 5 all miscellaneous installation tooling and support equipment / hardware have been removed from the vessel (a foreign material exclusion / program will be used to monitor materials in the vessel). Procedures will be used to ensure that allinspection activities are properly completed. 8.2.2 During Subsequent RefueIIng Outages i The wedge design is a benign, static structure that does not experience or impose significant loads or stresses on either itself or other reactor internal components (during normal operations). In addition, this design has been used at other facilities with no reported problems.2 As such, subsequent inspection activities will be limited. During 18R (the next refueling outage), GPU Nuclear plans to perform a baseline inspection (VT-3) on accessible areas of the wedges to confirm the originalinspections described in Section 8.2.1 above. Inspections in 19R and beyond are expected to be further limited in scope, based on confirmatory inspections in 18R. Prior to 18R, GPU Nuclear will prepare a detailed inspection plan for performance of the baseline wedge inspections to be conducted in 18R. 2The design is currently in use at Nine Mile Point Unit 1. MPR 1957 8-2 Revision 0

9 v I References [ [ 1. EPRI TR-108722, " Top Guide / Core Plate Repair Design Criteria (BWRVIP-50)," May 1998. l~ 7 2. EPRI TR-107284, "BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25)," December 1996. 3. GENE Specification 24A5733, " Oyster Creek Core Plate Wedges Design Specification," Revision 1. l 4. ASME Boiler and Pressure Vessel Code, Section III, Subsection NB,1989 Edition. i l 5. GENE B11-00709-001," Oyster Creek Core Plate Wedges Design Evaluation," 1 Revision 1. 6. ASME Boiler and Pressure Vessel Code, Section III, Subsection NG, " Core j Support Structures," 1989 Edition. l .l 7. Amendment 37 to Application for Provisional Operating License for Oyster l Creek Plant. I 8. MPR Report 1566, " Oyster Creek Nuclear Generating Station, Core Shroud i Repair, Design Report," October 1994, Revision 1 (Two Volumes). 9. MPR Report 1579," Confirmatory Dynamic Analysis of Oyster Creek Reactor ~ Internals for Core Shroud Repair Evaluation," December 1994, Revision 0. 10. GENE-771-0894," Justification for Allowable Deflections of the Core Plate and l I Top Guide-Shroud Repair," Revision 2, November 16,1994. 11. GENE Specification 24A5734," Oyster Creek Core Plate Wedges Fabrication Specification," Revision 1. 12. GENE B11-00709-003, " Oyster Creek Core Support Plate Flaw Tolerance ' Evaluation," Revision 1. ~ 13. MPR Calculation No. 083-261-BRL-1, Revision 0, " Shroud Stiffness With Failed Vertical Welds and Installed Wedges." MPR-1957 9-1 . Revision 0

14. MPR Calculation No. 083-261-BRL-2, Revision 0, " Transient Dynamic Evaluation of Oyster Creek Shroud With Vertical Welds Failed."

15. MPR Calculation No. 083-261-BRL-3, Revision 0," Wedge Loads and Shroud Stresses."

16. MPR Calculation No. 083-248-CBS-01, Revision 1, " Shroud Finite Element Evaluation." 17. USNRC Safety Evaluation of the Proposed Repair for the Oys;ter Creek Core Shroud, General Public Utilities Nuclear Corporation, Docket No. 50-219. MPR 1957 9-2 Revision 0 _ _ _ _ _ _ _ _ _ _ _}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20237D569&oldid=3925288"