CLI-09-07, Staff Assessment of the Oyster Creek 3-D Finite Element Analysis of the Drywell Shell

From kanterella
(Redirected from CLI-09-07)
Jump to navigation Jump to search
Staff Assessment of the Oyster Creek 3-D Finite Element Analysis of the Drywell Shell
ML091350358
Person / Time
Site: Oyster Creek
Issue date: 05/18/2009
From:
Division of License Renewal
To:
Eastern Environmental Law Ctr
Ashar, H, NRR/DE/EGCB, 301-415-2851
Shared Package
ML091340736 List:
References
CLI-09-07
Download: ML091350358 (21)
v • d • e


Text

ENCLOSURE STAFF ASSESSMENT OF THE OYSTER CREEK 3-D FINITE ELEMENT ANALYSIS OF THE DRYWELL SHELL INTRODUCTION The NRC Staff (Staff) completed a comprehensive review using our expertise and engineering judgment to scrutinize carefully Oyster Creek Nuclear Generating Stations (Oyster Creek) three dimensional finite element analysis (3-D FEA). Based on the Staffs review, the 3-D FEA provides a conservatively-biased, realistic analysis of the available ASME code margins utilizing a more modern computer analysis and sensitivity analyses to account for uncertainties in the extent of degradation between the measured locations in the drywell shell. Based on the Staffs comprehensive review, Oyster Creeks 3-D FEA shows that the drywell shell meets the ASME code margins under all the postulated conditions, as is discussed in more detail below.

I. Background The Staff completed a comprehensive review of the Oyster Creek license renewal application (LRA) in April 2007. Oyster Creek is currently owned and operated by Exelon Generation Company, LLC (Exelon or the applicant).1 The Staff determined that Exelon demonstrated that aging effects will be adequately managed to support operation for an additional 20 years and issued a renewed license on April 8, 2009.2 As part of the LRA review, the Staff reviewed the aging management program for the drywell shell. The plant contains a drywell shell constructed of carbon steel, which is approximately 100 feet tall and is shaped like an inverted light bulb (Figure 1). It measures approximately 70 feet in diameter at the spherical base. At an elevation of 71 feet 6 inches, it transitions from a spherical shape to a cylindrical shape that is approximately 33 feet in diameter. In the late 1980s, corrosion was discovered in an area in the spherical section of the shell marked in Figure 1 as the sandbed region, which is divided into 10 bays each designated with an odd number from 1 through 19. After removing the sand from this region, it was found that corrosion was unevenly distributed within the ten bays of the drywell shell, with Bays 1, 11, 13, 17, and 19 being the most affected bays. An analysis was performed by General Electric (GE) during that period, which is the analysis of record and Oyster Creeks current licensing basis, to confirm that the drywell shell had sufficient structural integrity for continued operations. To support that analysis, the applicant evaluated the degradation of the shell, taking approximately 1,000 ultrasonic (UT) measurements. Based on the UT measurements, areas on the external side of the shell that were considered the most thinned were ground flat for UT monitoring and grids were established for UT monitoring on the inside of the shell. The Staff reviewed the GE analysis and documented its review in the Evaluation Report on the Structural Integrity of the Oyster Creek Drywell, dated April 24, 1992. The sandbed region was subsequently coated to protect the shell from further corrosion. Since discovery of the corrosion, monitoring of the drywell shell by UT measurements has continued, with the most recent measurements taken in 2008.

1 Oyster Creek was previously owned by AmerGen Energy Company (AmerGen) and the license renewal application was filed by AmerGen. For the purposes of this report, Exelon will be used for consistency.

2 Agency Wide Document Access and Management System (ADAMS) Accession No. ML080280440.

During the review of the drywell shell, the staff evaluated the information in the license renewal application including: (1) the analysis of record, which was performed by General Electric (GE)

(References 1 and 2); (2) the confirmatory analysis performed for the NRC by Sandia National Laboratories (Sandia) (Reference 3), and (3) the applicants commitments (see Commitment 27 in Appendix A of the Staffs safety evaluation report (SER)).3 Based on the analysis of record and the Sandia confirmatory analysis, commitments, along with the other provisions of the aging management program, the Staff concluded that reasonable assurance exists that the drywell shell will withstand the postulated conditions, without exceeding the ASME code margins, during the period of extended operation.

During the review of Oyster Creeks license renewal application, the Advisory Committee on Reactor Safeguards (ACRS) met on February 1, 2007. During the February meeting, Exelon committed to perform a 3-D FEA of the Oyster Creek drywell shell in the as-found condition using more modern computational methods. The analysis of record performed by GE used many conservatisms due to the limitations of computer modeling, and a more modern 3-D FEA was intended to provide a more realistic assessment of the actual factor of safety for the drywell shell. Exelons commitment to perform the 3-D FEA was incorporated as Item 18 of Commitment 27.4 The commitment was also entered as a proposed license condition in the Staffs SER and is a current condition of the renewed license. The 3-D FEA provides a conservatively-biased but realistic quantification of the available margins above the ASME code minimum for buckling. It also adequately bounds the effect of the measurement uncertainties in the size of the thinned areas on the ASME code margins based.

Exelon submitted a Summary Report (Reference 4), describing the 3-D FEA, in fulfillment of its commitment, by letter dated January 22, 2009. As part of the Staffs review of the 3-D FEA, supporting reports and calculations were reviewed as part of the inspection completed in March 2009. Reference 4 presents the results of a baseline analysis of the postulated conditions and sensitivity analysis to account for measurement uncertainties. Enclosure 1 of the summary report, referred to as the Baseline Analysis, summarizes the analysis used to define the baseline degradation of the drywell shell, the 3-D FEA model, loads and load combinations imposed on the shell, the analysis results, the buckling evaluation, and Exelons evaluation of the available margin with respect to the ASME Section III, Subsection NE acceptance criteria. of the report, referred to as the Sensitivity Analysis, utilizes the same inputs and processes as in the Baseline Analysis with the exception that the baseline thicknesses are reduced in order to capture measurement uncertainties in existing thickness measurements of the sandbed.

In Reference 4, Exelon determined that for the two sensitivity analyses, the available safety factors exceed the ASME Section III, Subsection NE acceptance criteria for buckling. The applicant indicated that the 3-D FEA incorporates data on the drywell shell thicknesses through the applicants 2008 refueling outage inspection.

3 Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station, Docket No. 50-219, March 2007.

4 Id. The ACRS discussed Exelons commitment in its letter dated February 8, 2007, to Chairman Klein.

The Staffs assessment of the 3-D FEA consists of an evaluation of Exelons Baseline Analysis, Exelons Sensitivity Analysis, an assessment of the Exelons Conclusions, and the Staffs Conclusion.

Prior Analyses of the Drywell Shell Three independent analyses of Oyster Creeks drywell shell have been completed by GE, Sandia, and Structural Integrity Associates (SIA), who performed the 3-D FEA. In addition, the 3-D FEA performed by SIA was independently reviewed by the Staff and Becht Nuclear Services (Becht), a contractor hired by the State of New Jersey.5 A. Analysis of Record In the GE analyses, internal grid measurements of the drywell shell were compared against two criteria:6 (a) a mean thicknesses greater than 736 mils in each bay, and (b) the lowest thickness in an isolated area7 greater than 490 mils. For any external ultrasonic testing (UT) measurements that were found to be less than 736 mils (also called the locally-thinned areas),

the required thickness was 536 mils, in an area not to exceed 1 square foot. The actual measured UT thicknesses were all greater than the allowable criteria (i.e., mean thickness >

736 mils; locally thinned areas > 536 mils; pressure criterion > 490 mils). GEs analysis was intended to determine the limiting thickness acceptance criteria. In the GE analysis, the locally-thinned areas were conservatively located between the centerlines of the vents, which results in early drywell shell buckling modes. The conservatively modeled GE analysis, in general, met the safety factor of 2.0 against buckling under the most limiting load combination containing the refueling water load. The GE analysis contained many conservative assumptions, such as the location of the locally-thinned areas being more conservative than the actual location of the locally-thinned areas (near the vent lines), where considerably thicker reinforcing plates would preclude certain early buckling modes.

B. Sandia Analysis The Staff-sponsored Sandia analysis (Reference 3) conservatively considered the average of the external UT measurements in each sandbed bay as the bay thickness used in the evaluation input. The locally-thinned areas, modeled as 30 inch by 18 inch rectangles in the two most degraded bays, were assigned the lowest measured thicknesses in those bays. Under the limiting load combination containing the refueling water load, the safety factor against buckling was calculated as 2.15. The Sandia analysis also confirmed that the Oyster Creek drywell shell meets the ASME code margins under all the postulated conditions.

II. Baseline Analysis and Three Dimensional Finite Element Model The 3-D FEA, performed by SIA, is characterized as a conservatively-biased realistic analysis of the current condition of the drywell shell and evaluated the effects of the measurement uncertainties. The drywell shells thickness was divided into two major sections based on 5 Bechts review is available at ADAMS Accession No. ML091040727.

6 The GE analysis developed two acceptance criteria for drywell thickness requirements. The two criteria account for overpressure and buckling failure modes.

7 The isolated area is defined to be a 2 inch diameter circle or smaller.

elevation. The degraded areas identified by the internal grid UT measurements are separated into measurements above an elevation of 11 feet 0 inches and below this elevation (see Figure 2). In general, the measured thicknesses above 11 feet 0 inches are higher than or equal to those below this elevation, except for the thicknesses in Bay 7. The Bay 7 thicknesses are close to the nominal thickness (1154 mils) of the drywell shell in the sandbed area because the bay experienced very little degradation.

The model representing the degradation also considered any locally-thinned areas identified during the 1992 inspection, prior to the application of the epoxy coating to the exterior of the sandbed region. During subsequent inspections, the licensee confirmed that these locally-thinned areas have not experienced significant additional corrosion. These areas were modeled in the 3-D FEA as locally-thinned circular areas (instead of the square areas used in GE analysis) to facilitate more modern computer modeling. These circular areas in the 3-D FEA completely circumscribed the square areas of the GE analysis. Therefore, this modeling results in larger thinned areas than analyzed by the GE analysis. The locally-thinned areas are placed at the locations where they were actually identified, resulting in more realistic failure modes than those characterized by GE.

Table 1, which reproduces selected information from Table 3-4 of Reference 4, provides the diameters and thicknesses of the identified locally-thinned areas. This data was used to determine the modeled drywell shell thicknesses based on the internal grid UT measurements, and the external UT measurements.

Table 1: Local Thinned Area Modeling Sandbed Region Diameter (inches)

Thickness (mils)

Bay 1 51 696 Bay 13 18 658 Bay 15 18 711 Bay 17 18 (1) and 51(2) 663 (1) and 850 (2)

Bay 19 51 720 Notes: (1) Inner circle of locally thinned area.

(2) Outer ring of locally thinned area.

The assigned thicknesses in all bays in the sandbed area correspond to the UT measurements taken during the 2006 inspection.8 The applicant confirmed that the thicknesses found during the 2008 inspection are consistent with those found during the 2006 inspection. The staff finds Exelons modeling of the corroded areas in the sandbed region appropriate for a realistic analysis of the available margin to the ASME code limits.

A.

Finite Element Model - Mesh Optimization SIA used the public domain computer code ANSYS, Release 11.0, for performing the 3-D FEA.

The entire geometry of the drywell shell is mapped using areas in three dimensional space at the mid-thickness location of the shell parts including the cylindrical, upper spherical, lower spherical, and the sandbed areas. Real, type, and material numbering in ANSYS were used to distinguish the different components. Real numbers are for the thicknesses, type numbers are for the assemblies (e.g. star truss assemblies), and material numbers are for the different materials associated with the drywell shell assembly.

The 3-D FEA incorporated all penetrations greater than 3 inches in diameter, the insert plates, the reinforcing plates, the star-truss assembly that stabilizes the cylindrical portion of the shell during an earthquake event, the vent lines including the vent headers and downcomers, the constraint imposed by the concrete floors, trenches, and the components that would significantly affect the analysis results. The model also incorporates the general element size of 3 inches in the spherical and cylindrical regions of the drywell, 1.5 inches in the sandbed region, and 0.75 inches in the locally-thinned areas. The 3-D FEA contains approximately 400,000 elements.

SIA performed mesh sensitivity analyses using 2-inch elements in the spherical and cylindrical regions, 0.75-inch elements in the sandbed region, and 0.5 inch elements in the locally corroded area, constituting an additional 100,000 elements. The sensitivity analysis was performed using the load combination containing the 44 psi internal pressure. The stress comparison indicated that there was no significant difference in the results with the finer mesh element size.

8 See Figure 2.

Based on the staffs review of the geometry description of the drywell shell, representation of various components in the model, and the results of the mesh size optimization process, the staff finds that the essential parameters of the drywell shell are appropriately and adequately modeled in the 3-D FEA.

B.

Model Boundary Conditions In the Baseline Analysis, SIA discussed the boundary conditions for three cases: (a) the bottom head thermal and structural boundary condition, (b) the star truss boundary condition in the drywell shell cylindrical region, and (c) the vent-header boundary condition.

The boundary condition for the bottom head of the spherical portion of the drywell shell is treated in two parts. In the first part, SIA considered the spherical drywell shell below the bottom of the sandbed region at an elevation9 of 8 feet 111/2 inches to the bottom of the sphere at an elevation of 2 feet 3 inches, which is embedded in concrete, as fixed in the radial direction.

However, this constraint allows for movement in the meridional direction, assuming no bond exists between the drywell shell steel and the concrete. In the second part, for the spherical shell above the bottom of the sandbed region to the top of the inside concrete at an elevation of 10 feet 3 inches, SIA assumed the inside concrete does not restrain drywell shell movement in any direction.

There are 8 star truss assemblies (also called stabilizers) spaced evenly around the circumference in the cylindrical portion of the drywell shell, at an elevation of 82 feet 2 inches.

Each star truss has two truss members, which consist of double extra strong 10 inch pipe. The truss members connect the biological shield wall to the drywell shell star truss reinforcing plate and are welded onto the top plate of the shield wall. On the outside of the star truss reinforcing plate, a male lug connects the slot plate embedded into the concrete wall of the reactor building.

The star truss components transmit the forces from the drywell shell internals (i.e. reactor vessel) to the reactor building without loading the drywell shell. The star truss assemblies only impose a net force in the circumferential direction of the drywell shell. This circumferential force is resisted by the concrete wall of the reactor building through the male lug on the outside of the star truss reinforcing plate.

In the 3-D FEA model, drywell shell displacement is constrained in the circumferential direction at the male lug of the star truss. This constraint is imposed to simulate the drywell shells actual movement restraint when the male lug on the outside of the star truss comes into contact with the reactor building concrete wall within the slot. No similar constraint was imposed on the welded connections between the truss members and the shield wall.

The vent headers and downcomers are supported by vertical columns located at the bottom of the stiffener attached to the vent header for each end of the vent header segment. The 3-D FEA model incorporates this constraint as a boundary condition at the vent-line. The model constraint limits displacement in the vertical direction.

The Staff finds that SIA has realistically implemented appropriate boundary conditions for the concrete in the spherical shell, the star truss assemblies in the cylindrical shell, and the supports for the vent-line headers and downcomers. The boundary constraints represent the physical geometry of the shell components, and are in accordance with good engineering practice.

9 Elevations on the drywell shell are shown in Figure 1, above.

C.

Loading Input The primary loads considered in the analysis include: dead loads (including the water load during refueling), live loads, loss of coolant accident loads (internal pressure and associated temperature), pipe reaction loads, and earthquake loads (Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE)) including anchor movements where applicable, external pressure, and post-accident flooding loads.

The original loading conditions and load combinations were defined in the Oyster Creek Final Safety Analysis Report (FSAR), upon granting the full power license to the plant. Since that time, there have been NRC approved updates and modifications to the original FSAR. The modeled loading conditions and combinations utilize the modified and updated information as reflected in the current FSAR. Table 2 shows the loads and load combinations utilized in the 3-D FEA. The loads and load combinations analyzed using the modified loads are shown in Table 7-6 of 403R, and are reproduced in Table 2 of this document.

Table 2: Load Combinations10 Load Combination ASME Level Condition Load Cases LC1 A

Design/Test Prs1 + Grvty + OBE + SAM LC2 A

Design/Test Prs1 + Grvty - OBE - SAM LC3 A/B Normal Prs2 + Grvty + Mch/Live + OBE + SAM +

EPOBE + EPThrm + Thrm2 LC4 A/B Normal Prs2 + Grvty + Mch/Live - OBE - SAM -

EPOBE + EPThrm + Thrm2 LC5 A/B Refueling Prs2 + Grvty + Mch/Live + OBE + SAM +

EPOBE + EPThrm + Thrm2 + Refuel LC6 A/B Refueling Prs2 + Grvty + Mch/Live - OBE - SAM -

EPOBE + EPThrm + Thrm2 + Refuel LC9 C

Post-Accident Prs2 + Grvty + Mch/Live + SSE +

SAM(SSE) + EPSSE + Flood LC10 C

Post-Accident Prs2 + Grvty + Mch/Live - SSE -

SAM(SSE)- EPSSE + Flood Notes:

SAM - Seismic Anchor Movement Prs1 - Internal Pressure Prs2 - External Pressure Thrm2 - Operating Temperature EP - External Piping OBE - Operating Basis Earthquake SSE - Safe Shutdown Earthquake 10 This information in this table is also available at Table 7-6 of Reference 4.

The Staff finds that the loads and load combinations modeled in the 3-D FEA are consistent with those in the current FSAR. The staff also finds that the ASME Service Levels assigned to the load combinations are in accordance with the prevailing practice, except for the use of Service Level D for LC9 and LC10. Since SIA applied a more restrictive standard, Level C for LC9 and LC10, the Staff finds this modeling to be conservative and consistent with good engineering practice.

D.

Stress Evaluation The original Code of Record for the design and construction of the Oyster Creek drywell shell was the ASME Boiler and Pressure Vessel (B&PV) Code,Section VIII, 1962 Edition, with Nuclear Case Interpretations 1270 N-5, 1274 N-5 and 1272 N-5. The original Code of Record and Nuclear Case Interpretations do not provide stress intensity limits at different operating conditions (i.e. post-accident conditions) or limits for the local membrane stress due to the thickness reductions from local to general corrosion effects.

To address the need for guidance for the above areas, the code stress intensity requirements and limits from the ASME B&PV Code,Section III, Subsection NE, Class MC Components 1989 Edition including 1991 Winter Addenda were used. This ASME B&PV Code,Section III was used in previous re-evaluations of the Oyster Creek drywell stress analyses.11 The stress intensity limits, as specified in Subarticle NE-3320, must also be satisfied.

The applicable allowable stress intensity, Smc, for the Oyster Creek drywell materials are presented in Table 3, below. If the allowable stress intensity was not available for a material, the allowable value was appropriately determined by SIA based on the rules provided in Appendix III of the ASME B&PV Code,Section III. SIA also explained the use additional Subsection NE provisions as follows:

(a)

Per NE-3213.9, the bending stress at a gross structural discontinuity is classified as secondary stress. This requirement includes bending stresses due to internal pressure and external loads or moments per Table NE-3217-1, and for stresses near the nozzle, other openings, junctions to heads or flanges, knuckles or junctions to the shell.

(b)

Per NE-3213.10, a stress region may be considered local, if the distance over which the membrane stress intensity exceeds 1.1 Smc, does not extend in the meridional direction more than (Rt), where R is the minimum mid-surface radius of curvature and t is the minimum thickness in the region considered.

(c)

Per NE-3221.3, the primary general or local membrane plus primary bend stress intensity is derived from the highest value across the thickness of a section. In addition, it is stated in Note (2) of Figures NE-3221-1 and NE3221-3 that the provisions apply to the solid rectangular section.12 11 See References 2 and 3.

12 Reference 4.

Among the load combinations presented in Table 2, the refueling cases (LC5 and LC6) are the limiting cases in Levels A and B service conditions, because of the additional water load during refueling. For Level C, the post-accident flooding cases (LC9 and LC10) are the limiting cases.

These load cases were selected as the bounding cases for stress evaluation.

The materials used in the Oyster Creek drywell shell construction and the ASME Code allowable stress intensities are provided in Table 7-2 of Enclosure 113 and are reproduced in Table 3 of this document.

Table 3: Basic Code Allowable Stress Intensities14 Material Allowable Stress Intensities, Smc (ksi) 100 °F 200 °F 300 °F SA-516, Grade 70 19.3 19.3 19.3 SA-516, Grade 60 16.5 16.5 16.5 SA-320, L7(4) 27.5 27.5 27.5 USS-T1(1) (SA-514, Grade F) 27.5 27.5 27.5 SA-333, Grade O(2) 15.1 15.1 15.1 SA-333, Grade 1 15.1 15.1 15.1 SA-312, Type 304 18.8 17.8 16.6 SA-276, Type 304(1, 3) 18.8 17.8 16.6 Notes: (1) Stress intensity allowable is based on Appendix III, Article III-3000, using tensile strength of 110 ksi at room temperature.

(2) Stress intensity allowable not available, uses the allowable of SA-333, Grade 1.

(3) Use the allowable of SA-312, Type 304.

(4) Use the allowable of SA-193 Grade B7.

For each of the ASME Service Levels, SIA has determined the allowable limits as a function of the basic allowable stress intensities shown in Table 3. SIA points out that for LC9 and LC10 (Post-Accident Flooding and SSE) the current practice permits the use of Service Level D allowable limits, which are significantly higher than the Applied Service Level C allowable limits.

However, Oyster Creeks FSAR utilizes Service Level C, for these load combinations. SIA used Service Level C allowable limits for the 3-D FEA. The stresses generated in the drywell shell components from the analysis, with respect to the ASME III, Subsection NE allowable limits demonstrated that:

1) Under load combinations LC5 and LC6, incorporating the refueling load, the stresses in the cylindrical shell, stiffeners and in the star truss assemblies are below service Level B allowable limits,
2) Under LC5, the stresses in the sandbed bays and the vent pipes are comparatively low (i.e., <35% of the service Level B allowable limits), and 13 Reference 4.

14 See also Table 7-2 of Reference 4.

3) Under load combinations LC9 and LC10 and incorporating the post-accident flooding load and SSE, the stresses in the cylindrical and sandbed region are less than 80% of the Level C allowable limits.

SIA also performed a fatigue evaluation of the drywell shell material for the six stipulated conditions in NE-3221.5(d) of the ASME Code. The evaluation was based on the following assumptions, which have been and are being monitored under the current licensing basis:

1) The average temperature of the drywell is 150º F,
2) The number of significant pressure-temperature fluctuations is 200,
3) The startup-shutdown cycles are likely to be no more than one per year,
4) The leak test cycle is one every 10 years, and
5) An average (rather than instantaneous) coefficient of thermal expansion of the materials is used.

These assumptions fairly represent the plant characteristics, are appropriate for performing the fatigue evaluation, and utilize good engineering practice. The evaluation considered assumed cycles for each drywell shell material for the six stipulated conditions and showed that a detailed fatigue analysis is not required.

The staff finds that the stresses generated from the analysis are below the allowable limits for the critical load combinations, the required ASME code fatigue evaluation, and show no adverse effects on the drywell shell components.

E.

Buckling Evaluation The basic compressive allowable stress values, referred to in ASME Code Section III NE-3222.1, correspond to a factor of safety of 2 in Code Case N-284-1. This factor is applied to the buckling stress values that are determined by classical analysis and reduced by an appropriate capacity reduction factor (CRF).15 The CRF accounts for (1) the effects of imperfections and non-linearity in geometry and boundary conditions and (2) the plasticity reduction factors which account for nonlinearity in material properties. The stability stress limits in NE-3222.2 correspond to the following factors of safety (FS):

1) FS = 2.0 for the Level A/B Service Limits would apply to LC-3 to LC-6,
2) FS = 1.67 for Service Limit C would apply to LC-9 and LC-10, and
3) FS = 1.34 for Service Limit D - Not applicable for Oyster Creek.

The buckling evaluation of the drywell shell components subject to meridional compressive stresses requires a geometric description as well as the basic material properties of the components. The buckling evaluation consists of a determination of: (1) load factors from the buckling mode shapes, (2) CRF, modified by the hoop tensile stresses (if applicable), (3) plasticity reduction factor (if applicable), and (4) calculation of safety factors as the theoretical buckling stress divided by the meridional compressive stress.

15 The selection of use of the CRF is discussed below in a separate section.

1. Capacity Reduction Factor For the load combination consisting of refueling load and external pressure, the Sandia analysis did not give credit to the hoop tensile stresses generated due to gravity load, and utilized unmodified capacity reduction factor (CRF) given in the ASME Code Case N-284. The GE analysis, which is the analysis of record, and the SIA 3-D FEA gave credit to the tensile stresses generated by the refueling load. GE and SIA modified the CRF accordingly. As such, the CRF has been the subject of considerable interest and differing views amongst the parties to the license renewal hearing process. The Staff notes that for the SIA analysis, the refueling loading case was not the limiting case for buckling, but the post-accident flooding which used the same CRF as the GE analysis and Sandia analysis.

The analysis of record for Oyster Creeks current licensing basis was performed by GE in 1992.

GE used extremely conservative inputs in accordance with the modeling capabilities that were in use at that time. As discussed during the license renewal hearings, GE made very conservative assumptions regarding the physical state of the sandbed region. GE uniformly thinned the drywell shell in the sandbed region in its analysis even though large areas of the sandbed region experienced little corrosion.

In GEs analysis, a modified CRF was applied to the drywell shell based on tests on metal spheres16 performed by Dr. Clarence Miller. Code Case N-284, which was used for this analysis, allows higher values of CRF when appropriate justification is provided.

In 10 CFR 50.55a, the regulations state that nuclear power plants must meet the requirements of the ASME Boiler and Pressure Vessel Code, which is formulated and approved through a consensus process with experts. The NRC endorses the code and its code cases through the rulemaking process, or through approving its use for certain applications. In this case, the Staff evaluated this approach in its review of the GE analysis in 1994. In conjunction with the 1994 review, the Staff had structural engineering experts at Brookhaven National Laboratory (BNL) review Oyster Creeks use of the modified CRF. Considering the input from the technical evaluation from BNL,17 which explicitly discussed the use of the modified CRF for Oyster Creeks drywell shell, the staff found the modification to the CRF acceptable.18 a) Advisory Committee on Reactor Safeguards At the February 8, 2007, meeting of the Advisory Committee on Reactor Safeguards (ACRS) on the Oyster Creek license renewal application, the technical basis for the Code Case and test results were presented regarding the modified CRF use for analyzing Oyster Creeks drywell shell. In its February 8, 2007, letter to the Staff, the ACRS concurred with the Staffs position that found the modified CRF acceptable.

16 The sandbed bay has the same geometry as a sphere.

17 The BNL evaluation of the modified capacity reduction factor is available at Evaluation Report on the Structural Integrity of the Oyster Creek Drywell, dated April 24, 1992.

18 Id.

b) Sandia Analysis During the staffs review of Oyster Creeks license renewal application, the staffs structural engineering expert asked Sandia to perform a confirmatory analysis of the drywell shell using a more modern finite element analysis approach. Sandia made significant conservative assumptions and concluded that the drywell shell met ASME code margins even without using a modified CRF. The Staff notified the ACRS in a March 8, 2007, memorandum that Sandia did not use a modified CRF because they did not have the results from Dr. Millers tests available to them.19 Therefore, Sandia was unable to justify the modified CRF without confirming Dr. Millers tests. Because Sandias analysis found that the drywell shell met ASME Code margins without modifying the CRF, the Staff determined that the application of the modified CRF would only increase the available margin already shown to exist.

c) SIA 3-D FEA Exelon committed during the February 8, 2007, ACRS meeting to perform a 3-D finite element analysis of the Oyster Creek drywell, to better quantify the existing margin above the minimum required ASME code factor of safety in the as-found drywell shell using more modern methods.

Exelon submitted a summary report of the analysis, performed by SIA, to the NRC Staff on January 22, 2009, which is the subject of this report. This analysis used the modification to the CRF supported by Dr. Millers tests.

d) Becht Review The State of New Jersey (New Jersey) engaged a contractor, Becht Nuclear Services (Becht), to perform an independent review of the SIA analysis of the drywell shell. Bechts review of the 3-D FEA concentrated mainly on the SIA analysis and the Sandia analysis. Becht concluded that [t]he Code requirements are satisfied for the drywell in its current (2006-2008) state of degradation.20 Becht also concluded that the required code buckling factor of safety (FS) is acceptable without use of Millers modified capacity reduction factor.21 Finally, Becht believes that the uncertainty associated with the wall thinning measurements has been treated adequately for the measurements provided to-date, and as evaluated in two sensitivity cases in SIAs [analysis].22 Although Becht concluded that the use of the modified CRF negatively impacted how conservative the 3-D FEA was when taken alone, SIAs conservative treatment of the locally high theoretical buckling stress would offset any negative conservatism arising from the modified CRF. Becht also states that none of the negative conservatisms are singly or in-combination significant enough to warrant further action. These conservatisms are of the types covered by margins included in the ASME Codes and Standards.

19 Memo from P. Kuo, Director of DLR/NRR to F. Gillespie, ACRS, Re:ACRS Review of Oyster Creek SER (Mar. 8, 2007) (ADAMS Accession No. ML070650376).

20 Becht Report at 1.

21 Id.

22 Id.

e) Staff Conclusion In reviewing the 3-D FEA, other analyses of the drywell shell, and independent reviews of the analyses, the Staff notes that numerous analytical approaches would accurately and adequately account for stresses present in any structure when utilized consistent with the ASME code by a well-qualified and experienced structural engineer. As such, every decision requires technical judgment and balancing on the part of the structural engineer. For example, SIA imposed conservative local stresses and a less conservative CRF, where Bechts preferred methodology would have imposed less conservative local stresses and a more conservative CRF. As Becht concluded above, the ASME code factor of safety is met regardless of the CRF.

SIAs use of the modified CRF is accepted by structural engineering experts in the ASME code community, who have formally endorsed this approach to modifying capacity reduction factors in a recently approved Code case.23 Article 1500 of Code Case N-284-1 allows for the CRF to be increased for tensile stresses resulting from internal pressures. However, the Code Case does not address the effects of tensile stresses that result from other loadings (e.g., gravity loads).

Based on the research performed by Dr. Miller, the applicant used the modified CRF for other loads resulting in tensile stresses.24 The staff found the use of the modified CRF acceptable for use.

The methodology utilized by the SIA and as described by Dr. Miller for modification of the CRF was formally published in ASME Code Case N-759 (2006), AAlternative Rules for Determining Allowable External Pressure and Compressive Stresses for Cylinders, Cones, Spheres, and Formed Heads.@ The staff is in the process of reviewing the Code case for endorsement in Revision 35 of Regulatory Guide 1.84 for, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III.

When this methodology is applied to a shell with a non-uniform wall thickness, as in the case of Oyster Creek drywell shell, where the calculated stresses in both the hoop (tensile) and meridional (compressive) directions are not uniform among different elements, certain approximations have to be made to compute the magnitude of the modified CRF. The staff recognizes the complexity involved in developing a reasonable estimate of the CRF due to the variation in the ratio of tensile hoop stresses to compressive meridional stresses among different elements from the finite element model. The Staff finds the approach used by SIA in developing the modified CRF to be realistic and reasonable for this purpose. The Staff also notes that even if SIA had used an unmodified CRF, the analysis indicates that Oyster Creeks drywell shell exceeds the minimum safety factors as prescribed by the ASME code. Thus, the Staff recognizes that the modification of the capacity reduction factor requires expert structural engineering judgment due to the complexity of the analysis, but does not agree that its use leads to a non-conservative analysis for the reasons discussed above.

2. Buckling Modes and Safety Factors The controlling load cases considered for the buckling analysis are (1) the refueling load case and (2) the post-accident flooding load case. SIA performed bifurcation buckling analysis for each load case, and generated 200 eigenvalue buckling modes with corresponding load factors.

23 ASME Code Case N-759 (2006), Alternative Rules for Determining Allowable External Pressure and Compressive Stresses for Cylinders, Cones, Spheres, and Formed Heads.

24 Reference 2 at Section 2.

After a review of the buckling modes, corresponding load factors, and displacements, the following buckling modes governed each section of the drywell shell.

1) For the refueling load case, the first mode governs the cylindrical portion of the drywell, while the postulated buckling of the upper, middle and lower spherical region of the drywell occurs in the 15 th

, 25 th

, and 32 nd modes. The postulated buckling of the sandbed region occurs at modes higher than the 45 th mode with a relatively high load factor of 11.584 (high load factor means high factor of safety).

2) For the post-accident flooding case, the postulated buckling in the cylindrical region occurs at the 56 th mode and that of the upper and middle spherical region occurs at the 200 th mode. The postulated buckling of the lower spherical region occurs at the 15 th mode with a load factor of 7.344. In the sandbed region, a very small amount of displacement (0.01 inch) occurs in the 11 th mode in Bay 1. SIA has used the 11 th mode load factor of 7.162 for calculating the buckling safety factor calculations.

A summary of the pertinent safety factors for each major component of the drywell are given in Table 4 and Table 5. The staff finds the process used in arriving at the load factors, theoretical buckling loads, and safety factors consistent with good engineering practice and ASME Section III requirements, as supplemented by Code Case N-284-1.

Table 4: Buckling Evaluation25 Loading Region Cylindrical Spherical Upper Middle Lower Refueling SF - Safety Factor Min.

Reqd - 2.00 3.39 4.27 3.60 3.60 Flooding SF - Safety Factor Min.

Reqd -1.67 3.46 7.57 4.76 2.22 25 Reference 4.

Table 5: Buckling Evaluation, Sandbed Region26 Loading Sandbed Region Bays 1

3 5

7 9

11 13 15 17 19 Refueling Safety Factor Min.

Reqd -

2.00 3.70 3.54 3.56 3.57 3.56 3.69 3.78 3.69 3.67 3.74 Flooding Safety Factor Min.

Reqd -

1.67 2.08 2.36 2.25 2.24 2.23 2.12 2.13 2.29 2.14 2.02 26 Id.

F.

Other Conservative Assumptions SIA has listed additional conservative assumptions and approaches included in its Baseline Analysis, as follows:

1) Seismic Response Spectra: The drywell is supported at two locations, at the 10 feet 3 inches base, and at the 82 feet 9 inches star truss location. The seismic spectra at the 82 feet 9 inches star truss support location are significantly higher than the spectra at the 10 feet 3 inches base. The bounding seismic response spectra are used as the seismic input for the drywell. This introduces significant conservatism in the seismic response of the structure.
2) Post Accident Flooding Case Contributing Water Mass: All of the water is included as added mass in the drywell in the seismic analysis. In a seismic event, a substantial portion of the water will be acting on the bioshield wall, which has been conservatively neglected in the Baseline Analysis. The assumption that the entire seismic energy of the water is absorbed by the drywell is conservative.
3) Welded Steel Structure Damping Ratio: As implemented by the SIA analysis, NRC Regulatory Guide 1.61 (issued in October 1973) recommends a damping value of 2% to be used with the Operating Basis Earthquake analysis for welded steel structures. Revision 1 of Regulatory Guide 1.61 (issued in March 2007) recommends a 3 percent damping value for welded structures. The use of an increased damping value will reduce the component stresses in the refueling case, in which the Operating Basis Earthquake is included as one of the loads.
4) Service Level C Limits versus Service Level D Limits: The Post Accident Flooding Case is evaluated using the Service Level C Limits, which are significantly lower than the Service Level D Limits. The Post Accident Flooding condition is a highly unlikely event and is generally categorized as a Service Level D event. Using the Service Level D Limits would have yielded additional margins on the analysis results for the Post Accident Flooding Load Case.
5) Support Provided by Star Truss/Bioshield Wall: There are a total of 8 star trusses connecting the drywell shell to the bioshield wall at an elevation of 82 feet 9 inches. The structural support provided by the Star Truss/Bioshield Wall has been conservatively neglected from the analyses. The structural support will take on some load, which is a conservatism in the analyses, especially in the buckling analysis.
6) Sizes of Locally Thinned Areas: The sizes of the locally thinned areas in the sandbed region have been conservatively mapped. An extended circular boundary has been drawn to enclose the postulated locally thinned areas.

These conservatisms result in a larger, thinner area being modeled in the analysis, which produces conservative analysis in the sandbed region.

III. Sensitivity Analysis Two Sensitivity Cases were considered as part of 3-D analysis of the Oyster Creek drywell shell. The summary analysis is provided in Reference 4. The two cases considered are:

Case 1 (S1): Reduce the wall thickness of the defined locally thinned area in Bay 1 from 696 mils (see Figure 2) to 596 mils, keeping the general area of the thickness in Bay 1 constant (826 mils - as in 403R) with all other inputs the same as in the Base Case.

Case 2 (S2): Reduce the wall thickness of the general area in Bay 19 from 826 mils (see Figure 2) to 776 mils, keeping the thickness of the locally thinned area constant (720 mils) and all other inputs the same as in the Base Case.

To capture the potential uncertainties in identifying the location and degradation of the locally thinned areas, the applicant has reduced the thickness of the 51-inch diameter circle (see Figure 2) by 100 mils in Bay 1, the bay most affected by corrosion. As the UT measurements at certain elevations are not available in certain bays, in addition to the approach taken by the applicant in the Baseline analysis, the applicant has reduced the shell thickness by 50 mils in Bay 19 in S2. This will quantify of the effects of thickness reduction in the sandbed region on the buckling safety factors. The staff finds the use of these sensitivity cases conservative and acceptable for considering the effect of the size and thickness of degraded areas on the available margin.

The stress analyses and buckling evaluation for the two sensitivity cases follow the same methodology used in the baseline analysis. The load combinations and ASME Code evaluation are performed in the same manner as those in the baseline analysis. The loading input, analysis procedures, 3-D FEA model, and the acceptance criteria for the sensitivity cases are the same as those used in the baseline analysis. The reduced wall thicknesses in a locally thinned area in Bay 1 and reduced thicknesses in Bay 19 are the only variations from the baseline analysis.

The buckling results from these two sensitivity cases are compared to the results obtained in the baseline analysis.27 As the regions above the sandbed area are not affected in S1 and S2, the stresses in these regions are essentially the same as those in the baseline analysis. A summary of the stress intensity results from the Baseline analysis, S1, and S2 is provided in Reference 4. The analysis shows less than an eight percent increase in the stress intensities in Bays 1 and 19, for S1 and S2. These nominal increases remain well below the ASME code limits.

Using the same methodology as in the Baseline analysis, SIA developed the load factors and displacements for S1 and S2 for the controlling load combinations with the refueling load and post-accident flooding load. The affected region remains the sandbed region. A review of the 27 SIAs sensitivity analyses adequately vary the size of the thinned area and the depth of the thinned areas such that reasonable conclusions can be drawn regarding these effects and are adequately analyzed for these effects on the drywell shell. The three analyses performed by GE, Sandia, and SIA modeled the drywell shell degradation in different geometries. Based on the variation in analysis techniques, reasonable conclusions can be drawn about the effect of variations similar to performing additional sensitivity analyses.

safety factors in the sandbed region indicates that the safety factor in Bay 19 is decreased by about 9 percent (from 3.74 to 3.40) in S1, under the refueling load. In the post-accident flooding load case, the safety factors for S1 and S2 show minimal variations to the baseline analysis. In the load cases considered, the safety factors always exceed the ASME allowable minimums.

IV. Overall Staff Assessment The staffs assessment indicates that the 3-D FEA of the Oyster Creek drywell shell has been performed using modern methods of analysis, using the actual geometry of the shell, with appropriate boundary conditions. The applicants consultant, SIA, utilized UT thickness measurements taken during the 2006 outage. In the cover letter accompanying the Summary Report, dated January 22, 2009, the applicant confirms that the UT measurements taken during the 2008 outage correspond to the measurements taken during the 2006 outage. In Section 3.11 of the staffs Inspection Report (Reference 6), the staff noted that the technical evaluation also compared the 2008 data values to the corresponding values recorded by the 2006 UT examinations in the same locations and concluded that there were no significant differences in measured thicknesses and no observable widespread corrosion.

The applicants analysis is performed for the two controlling cases (i.e., the refueling load case and the post-accident flooding load case). These cases have different ASME allowable safety factors (2.0 for the refueling load case and 1.67 for the post-accident flooding load case). Table 6, below, reflects the normalized safety factors.28 A normalized safety factor greater than 1 shows that excess margin exists to the ASME code limits. In the table, only Bays 1 and 19 of the sandbed region, the most affected bays, are included. In the refueling load case, the effects of the sensitivity cases are not appreciable. The lowest calculated buckling margins in the sandbed region are controlled by the post-accident flooding load.

Table 6: Comparison of Normalized Safety Factors29 LOAD REGION BASELINE S1 S2 Refueling Cylindrical Shell 1.65 1.67 1.67 Upper Spherical Shell 2.13 2.13 2.13 Middle Spherical Shell 1.80 1.80 1.80 Lower Spherical Shell 1.80 1.69 1.80 Sandbed Bay 1 1.85 1.74 1.81 Sandbed Bay 19 1.87 1.70 1.84 Flooding Cylindrical Shell 2.07 2.07 2.07 Upper Spherical Shell 4.53 4.54 4.53 Middle Spherical Shell 2.85 2.84 2.84 Lower Spherical Shell 1.33 1.31 1.31 Sandbed Bay 1 1.25 1.26 1.23 Sandbed Bay 19 1.21 1.20 1.19 28 SFn=SFactual/SFallowable 29 The normalized safety factors shown in the table are used to ease comparison to the ASME code criteria for the various components of the drywell shell. Any number in excess of 1.0 shows available margin to the ASME code criteria.

The following observations can be made from the Table:

1) For all the cases, the normalized safety factors are well above 1.0 (a margin of 1.0 would indicate that the factor of safety is at the ASME Code allowable limit).
2) For the flooding load case, the lowest normalized safety factor of 1.19 is indicated in sandbed Bay 19.
3) For the refueling load case, the lowest normalized safety factor is indicated in the cylindrical shell.
4) The refueling load case affects the normalized safety factors in S1, but not in S2 insandbed Bays 1 and 19.
5) Based on the comparison of the normalized safety factors in Table 6, it is apparent that no substantial difference exists in the normalized safety factors between the Base Case and Sensitivity Case 2, where the Bay 19 thickness was reduced by 50 mils. If the 50 mils reduction had been extended to the two adjacent Bays (i.e., Bay 19 and Bay 1), it is the staffs position that the normalized safety factors would have been, at a minimum, greater than 1.0.

Overall, the analysis demonstrated that all the components of the drywell shell show adequate margins against instability under refueling loads and post-accident flooding loads.

A review of the refueling load case stress evaluations performed for the Baseline analysis, S1, and S2, indicates that the stresses in the cylindrical shell, stiffeners/gussets and star truss stiffeners indicate stresses less than the allowable ASME code limits. It is the Staffs assessment that the sensitivity analyses and other analyses of the drywell shell adequately bound the intent of Judge Barratas separate opinion 30 and demonstrate that additional sensitivity analyses will not result in significant new information. For the refueling load, the margin against buckling is the lowest in the cylindrical shell region, an area that has experienced little corrosion.

V. Concluding Remarks The staff has performed an in depth review of Oyster Creeks 3-D FEA of the drywell shell.

Based on that review, the Staff concludes that the 3-D FEA fulfilled Oyster Creeks commitment to provide a modern analysis that quantifies the available safety margin including performing sensitivity analyses. The 3-D FEA shows that Oyster Creeks drywell shell complies with the ASME code limits for the postulated conditions. This analysis was performed utilizing good engineering practices and judgment and applied conservatively-biased realistic assumptions.

The sensitivity analyses performed by Oyster Creek also adequately bound the intent of Judge Barratas separate opinion. Although this 3-D FEA is not the analysis of record for Oyster Creeks current licensing basis, it also demonstrates that Oyster Creeks drywell shell meets the ASME code limits even when accounting for measurement uncertainties in the thicknesses of the shell.

30 See Amergen Energy Co., LLC (Oyster Creek Nuclear Generating Station) LBP-07-17, 66 NRC 327,373-76 (2007)

VI. References

1.

An ASME Section VIII Evaluation of Oyster Creek Drywell for Without Sand Case, Part I, Stress Analysis," GE Report, Index No. 9-3, Revision 0, DRF # 00664, February 1991, prepared by GE Nuclear Energy, San Jose, California.

2.

An ASME Section VIII Evaluation of Oyster Creek Drywell for Without Sand Case, Part II, Stability Analysis, GE Report, Index No. 9-4, Revision 2, DRF # 00664, November 1992, prepared by GE Nuclear Energy, San Jose, California.

3.

Structural Integrity Analysis of the Degraded Drywell Containment - Oyster Creek Nuclear Generating Station, Sandia National Laboratories, January 2007, ML070120395.

4.

Exelon Nuclear, Structural Evaluation of the Oyster Creek Drywell - Summary Report, prepared by Structural Integrity Associates, San Jose, California, January 2009, ML090270871, ML090270872, and ML090270873.

5.

Miller, C.D., Applicability of ASME Code Case N-284-1 to Buckling Analysis of Drywell Shell, June 2006.

6.

Oyster Creek Generating Station - NRC License Renewal Follow-up Inspection Report 05000219/2008007, Dated January 21, 2009, ML090210106.

Retrieved from "https://kanterella.com/w/index.php?title=CLI-09-07,_Staff_Assessment_of_the_Oyster_Creek_3-D_Finite_Element_Analysis_of_the_Drywell_Shell&oldid=5315854"