ML12277A280

BWROG Linear Elastic Fracture Mechanics TR Draft SE Letter

ML12277A280
Person / Time
Site:	Boiling Water Reactor Owners Group
Issue date:	11/02/2012
From:	Stuchell S Licensing Processes Branch (DPR)
To:	Schiffley F GE-Hitachi Nuclear Energy Americas
Golla J
References
BWROG-TP-11-023, Rev 0, TAC ME7650
Download: ML12277A280 (11)

Text

November 2, 2012 Mr. Frederick P. Schiffley, II BWROG Chairman c/o GE Hitachi Nuclear Energy P.O. Box 780 3901 Castle Hayne Road, M/C F-12 Wilmington, NC 28402

SUBJECT:

DRAFT SAFETY EVALUATION FOR THE BOILING WATER REACTOR OWNERS GROUP LICENSING TOPICAL REPORT BWROG-TP-11-023, LINEAR ELASTIC FRACTURE MECHANICS EVALUATION OF GENERAL ELECTRIC BOILING WATER REACTOR WATER LEVEL INSTRUMENT NOZZLES FOR PRESSURE-TEMPERATURE CURVE EVALUATIONS, REVISION 0, NOVEMBER 2011 (TAC NO. ME7650)

Dear Mr. Schiffley:

By letter dated November 17, 2011 (Agencywide Documents Access and Management System Accession No. ML11325A073), the Boiling Water Reactor Owners Group (BWROG) submitted BWROG-TP-11-023, Revision 0 dated November 2011, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations, to the U.S. Nuclear Regulatory Commission (NRC) staff for review. Enclosed for BWROG review and comment is a copy of the NRC staffs draft safety evaluation (SE) for the Licensing Topical report.

Twenty working days are provided for you to comment on any factual errors or clarity concerns contained in the draft SE. The final SE will be issued after making any necessary changes and will be made publicly available. The NRC staffs disposition of your comments on the draft SE will be discussed in the final SE.

To facilitate the NRC staffs review of your comments, please provide a marked-up copy of the draft SE showing proposed changes and provide a summary table of the proposed changes.

If you have any questions, please contact Joseph Golla at 301-415-1002.

Sincerely,

/RA/

Sheldon D. Stuchell, Acting Chief Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 691 cc w/ encl: See next page

ML12277A280 NRR-106 OFFICE PLPB/PM PLPB/LA EVIB/BC PLPB/BC (A)

DPR/DD PLPB/PM NAME JGolla DBaxley SRosenberg SStuchell SBahadur JGolla DATE 10/17/2012 10/16/2012 10/17/2012 10/18/2012 10/31/2012 11/2/2012

Boiling Water Reactor Owners Group Project No. 691 cc:

BWROG Chairman Frederick P. Schiffley c/o c/o GE Hitachi Nuclear Energy PO Box 780 3901 Castle Hayne Road, M/C F-12 Wilmington, NC 28402 Frederick.schiffley@exeloncorp.com Craig J. Nichols GE Hitachi Nuclear Energy PO Box 780 M/C F12 3901 Castle Hayne Road Wilmington, NC 28402 Craig.nichols@ge.com Lucas Martins GE Hitachi Nuclear Energy PO Box 780 M/C F12 3901 Castle Hayne Road Wilmington, NC 28402 Lucas.martins@ge.com

DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION LICENSING TOPICAL REPORT (LTR) BWROG-TP-11-023, REVISION 0 LINEAR ELASTIC FRACTURE MECHANICS EVALUATION OF GENERAL ELECTRIC BOILING WATER REACTOR WATER LEVEL INSTRUMENT NOZZLES FOR PRESSURE-TEMPERATURE CURVE EVALUATIONS BOILING WATER REACTORS OWNERS= GROUP PROJECT NO. 691

1.0 INTRODUCTION AND BACKGROUND

In a letter dated November 17, 2011, the Boiling Water Reactor (BWR) Owners Group (BWROG) submitted Licensing Topical Report (LTR) BWROG-TP-11-023, Revision 0, dated November 2011, Linear Elastic Fracture Mechanics [(LEFM)] Evaluation of General Electric Boiling Water Reactor Water Level Instrument [(WLI)] Nozzles for Pressure-Temperature [(P-T)]

Curve Evaluations (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML113250288), to the U.S. Nuclear Regulatory Commission (NRC) for review and acceptance for referencing in subsequent licensing actions. LTR BWROG-TP-11-023, Revision 0 (hereafter referred to as the LTR), provides a fracture mechanics solution for a partial penetration WLI nozzle for an internal pressure load case and a 100 Degree-Fahrenheit-per-hour (°F/hr) thermal load case. The solution will then be used in developing plant-specific P-T curves.

This review also includes an evaluation of the BWROGs responses to the NRC staffs requests for additional information (RAIs), which were provided to the NRC in a letter from the BWROG dated June 13, 2012 (ADAMS Accession No. ML12167A239). Please note the NRC staff asks that you revise the final -A version of the LTR to reflect the clarification made in your response to RAI-2.

The objective of the topical report (TR) process is, in part, to add value by improving the efficiency of other licensing processes, for example, the process for reviewing license amendment requests (LARs) from commercial operating reactor licensees. The purpose of the U.S. Nuclear Regulatory Commission (NRC) TR program is to minimize industry and NRC time and effort by providing for a streamlined review and approval of a safety-related subject with subsequent referencing in licensing actions, rather than repeated reviews of the same subject.

A TR is a stand-alone report containing technical information about a nuclear power plant safety topic, which meets the criteria of a TR. A TR improves the efficiency of the licensing process by allowing the NRC staff to review a proposed methodology, design, operational requirements, or other safety-related subjects that will be used by multiple licensees, following approval, by referencing the approved TR. The TR provides the technical basis for a licensing action.

ENCLOSURE

During the review of the BWROGs LTR BWROG-TP-11-023, Rev. 0, the NRC staff found that, in general, the LTR meets the objectives of an LTR and reinforces previously established NRC regulations and guidelines as noted within this SE. The NRC has evaluated this LTR against the criteria of 10 CFR Part 50, and has determined that it does not represent a backfit.

Specifically, NRC Staff technical positions outlined in this SE are consistent with the aforementioned regulations and established staff positions, while providing more detailed discussion concerning the methodology and data required to support linear elastic fracture mechanics evaluations of General Electric BWR WLI Nozzles for P-T Curve Evaluations. This SE endorses staff positions previously established through licensing actions and interactions with industry.

2.0 REGULATORY EVALUATION

The NRC has established requirements in Appendix G, Fracture Toughness Requirements, of Part 50 to Title 10 of the Code of Federal Regulations (10 CFR Part 50), in order to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The regulation at 10 CFR Part 50, Appendix G, requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G, Fracture Toughness Criteria for Protection Against Failure, to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) were used to generate the P-T limits.

The regulation at 10 CFR Part 50, Appendix G, also requires that applicable surveillance data from reactor pressure vessel (RPV) material surveillance programs of 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Table 1 to 10 CFR Part 50, Appendix G, provides the NRC staff=s criteria for meeting the P-T limit requirements of ASME Code,Section XI, Appendix G, as well as the minimum temperature requirements of the rule for bolting up the vessel during normal and pressure testing operations.

In addition, the NRC staff regulatory guidance related to P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials,@ and Standard Review Plan Chapter 5.3.2, APressure-Temperature Limits Upper-Shelf Energy and Pressurized Thermal Shock.

The regulation at 10 CFR Part 50, Appendix H, provides the NRC staff=s criteria for the design and implementation of RPV material surveillance programs for operating light-water reactors.

In March 2001, the NRC staff issued RG 1.190, ACalculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.@ Fluence calculations are acceptable if they are performed with approved methodologies or with methods which are shown to conform to the guidance in RG 1.190.

For RPV instrument nozzles located in the beltline region, such as the WLI nozzles, the increased neutron fluence may cause an embrittlement concern during the period of extended operation of plants that have received a license extension. Subparagraph G-2223(c) of the

ASME Code,Section XI, Appendix G states that, fracture toughness analysis to demonstrate protection against nonductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5 in. (63 mm) or less, provided the lowest service temperature is not lower than RTNDT plus 60 °F (33 °C). Therefore, when Subparagraph G-2223(c) is referenced, the ART value (i.e., RTNDT adjusted to account for the effects of radiation) must be determined for the instrument nozzles and the associated nozzle-to-RPV welds to determine if the lowest service temperature criterion will be met through the end of the evaluation period. If not, the instrument nozzles and the associated nozzle-to-RPV welds must be considered in the P-T limits in accordance with 10 CFR Part 50, Appendix G, and the ASME Code,Section XI, Appendix G.

On January 31, 1996, the NRC staff issued Generic Letter 96-03 to inform licensees that they may request a license amendment to relocate the P-T limit curves from the Technical Specifications (TS) into a pressure temperature limits report (PTLR) or other licensee-controlled document that would be controlled through the TS.

3.0 TECHNICAL EVALUATION

3.1 The BWROGs Evaluation The LTR provides an LEFM evaluation of the WLI nozzles to be used in P-T limit applications.

Section 1, Introduction, provides the relationship among P-T limits, PTLR, and the proposed LEFM evaluation for WLI nozzles. Section 2, Methodology, provides the proposed LEFM methodology, starting from the stress analysis based on the finite element method (FEM) to the LEFM analysis based on the boundary integral/influence function (BIE/IF) method. Section 3, Assumptions, provides assumptions adopted in each step of the proposed methodology, such as the heat transfer coefficients for the WLI nozzle and RPV external and internal surfaces, material properties for various components, and the stress free temperature for evaluating cladding stresses. Section 4, Finite Element Model, provides information on development of the 38 FEM models first introduced in Section 2, considering various types of nozzle and FEM models with and without a crack. This section also addressed FEM model validation through mesh density check for models with and without a crack. Section 5, Instrument Nozzle Load Cases, addresses the loads on the nozzle: internal pressure, thermal transient, and pipe reaction load. Section 6, Pressure, Thermal, and Piping Load Results, presents FEM results (i.e., stresses and applied stress intensity factors (KIs)) under these three types of load.

Section 7, Observations and Discussions, offers the BWROGs observation of the behavior of the applied KIs and a discussion of the modeling choices that could affect the results. Section 8, Generic Methodology for KI Estimation, provides generic KI Formulas derived from the FEM results for a variety of nozzles for licensees to use for WLI nozzles in their plant-specific P-T limit applications. Section 9, Summary, provides summary findings and conclusions.

3.2 The NRC Staffs Evaluation Since Section 1 of the LTR provides only an introduction on P-T limits, PTLR, and the proposed LEFM evaluation for WLI nozzles, the staffs evaluation starts from Section 2, which discusses the stress analyses due to pressure, thermal transient, and pipe reaction loads using quarter and full FEM models. Section 2.2 of the LTR states, Since there is no specific material specification identified for the CS [(i.e., carbon steel)] nozzle inserts from Table 2-1 of Reference [13], SA-541 Class 1 is assumed. Depending on the material properties used, this

assumption could change the calculated stresses significantly. Hence, the NRC staff issued RAI-1. RAI-1 also includes questions regarding the assumption of Alloy 600 material for the weld metal for all nozzles and the assumption of certain heat transfer coefficients for the inside and outside surface of the vessel and nozzle.

The BWROGs June 13, 2012, response to RAI-1 indicated that; in addition to Alloy 600, stainless steel, and CS as the WLI nozzle insert material as reported in the LTR; low alloy steel (LAS) is also found to be the material for the WLI nozzle insert. Therefore, a revision was made to the LTR to reflect this new finding. The BWROG performed a LEFM analysis based on FEM on a representative nozzle using CS and LAS material properties and found that their KI values under the thermal transient are within 3.7 percent. Since this magnitude of difference is within the accuracy of the inputs and overall methodology, the staff accepts the difference in KI due to material difference. Therefore, the first part of RAI-1 is resolved. For the second part of RAI-1 regarding the assumption of using Alloy 600 material for the weld metal for all nozzles, the BWROG revised the LTR to limit the application of this LTR to only WLI nozzle configuration/design using Alloy 600 material for the weld metal. Hence, the second part of RAI-1 is resolved. For the last part of RAI-1 regarding the use of certain heat transfer coefficients for the inside and outside surface of the vessel and nozzle, the BWROG revealed that the convection coefficient for the outside surface is from NEDO-21821-A, Boiling Water Reactor Feedwater Nozzle/Sparger Final Report, and the convection coefficient for the inside surface is consistent with the range of values considered in the same report. In addition, an evaluation was provided in the BWROGs response to demonstrate that further increase in the assumed convection coefficient has an insignificant effect on the maximum thermal stresses.

Based on these, the staff considers the last part of RAI-1 resolved.

For the load cases discussed in Section 2.3 of the LTR, it is not clear whether piping loads are loads from only the piping attached to the safe-end of the instrument nozzle, or also from the various large piping (not shown in Figure 4-1) attached to the RPV. This is the basis for RAI-2.

The BWROGs June 13, 2012, response to RAI-2 confirmed that piping loads are loads from only the piping attached to the safe-end of the instrument nozzle. RAI-2 is therefore resolved but the BWROG is asked to revise the LTR to reflect this clarification.

The 45° line (Figure 2-1) is chosen for extracting hoop stresses from the FEM analysis.

Although the orientation of this path is consistent with the necessary inputs for the BIE/IF solution for the nozzle corner crack, it has not been established in the LTR that this approach will produce conservative, or the most accurate results. This is the basis for RAI-3.

The BWROGs June 13, 2012, response to RAI-3 cited EPRI NP-339, Improved Evaluation of Nozzle Corner Cracking, and an Oak Ridge National Laboratory (ORNL) report ORNL/TM-2010/246, Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles, to support the conclusion that good agreement exists between the stress intensity factors calculated from detailed analysis and the simplified one-dimensional stress distribution. The staff confirmed that the ORNL report concluded that the combined KI due to both pressure and thermal loading is estimated reasonably well in the analyses using uncracked and cracked FEM models for the BWR WLI nozzle where the hoop stress extraction path is the 45° line through the nozzle thickness.

Therefore, using the same extraction path in the current application should also produce reasonable and acceptable results. RAI-3 is resolved. After resolution of RAI-1 to RAI-3, the staff has no additional concerns over the proposed methodology, pending evaluation of the remaining sections of this LTR.

Section 3 of the LTR discusses assumptions used in the analysis. The staff had a question on the BWROGs applying the weaker base metal properties instead of weld material properties in the FEM analyses. This is the basis for RAI-4.

The BWROGs June 13, 2012, response to RAI-4 clarified that base metal is intended to mean an equivalent ASME Code material for the weld metal, not to mean the use of LAS properties.

The LTR is revised to reflect this clarification. Therefore, RAI-4 is resolved. For other assumptions summarized in Section 3 of the LTR, Assumption 1, Assumption 2, and part of Assumption 4 have already been discussed in Section 2 regarding the proposed methodology.

The staffs evaluation and acceptance of these assumptions can be found above in the discussion of the BWROGs response to RAI-1. Assumption 5 regarding constant density and Poissons ratio is appropriate because it has only secondary effect on the results and is consistent with the common practice, and the remaining part of Assumption 4 regarding vessel pad material and nozzle-to-safe end weld material has negligible effect due to their distance from the location of interest. Assumption 6 regarding a stress free temperature of 550° F is commonly used in stress and fracture mechanics analyses for RPV with cladding, and is acceptable.

Section 4 of the LTR provides information for the 3-dimentional (3-D) FEM un-cracked models for the structural and thermal analyses. Moderate details for 3-D FEM cracked models for the structural and thermal analyses are also provided for validating the BIE/IF KI results.

To demonstrate that a proprietary FEM meshing algorithm used in the analyses can produce accurate results, the BWROG provides a benchmark of its FEM fracture mechanics modeling methodology in Appendix A of this LTR. The staff examined the comparison of results from various approaches and accepted the BWROGs use of this meshing algorithm in the current application because the algorithm is not very sensitive to mesh size and the results are more conservative than those provided by the closed-form solution. However, the staff is not sure that coupling the nodes on the free end of the safe end in the nozzle axial direction best simulates the real situation. This is the basis for RAI-5.

The BWROGs June 13, 2012, response to RAI-5 clarified that coupling the nodes on the free end of the safe end is used in the mesh sensitivity study to demonstrate that the spatial discretization selected for the FEM models was sufficient to resolve the parameters of interest.

The BWROG further stated that for the piping load cases analyzed as part of the evaluation in Sections 5.3 and 6.3 of the LTR, the nodes at the end of the safe end are connected to a pilot node on which the appropriate forces and moments are applied. The staff accepts this explanation because coupling the nodes on the free end of the safe end is not used in the nozzle FEM models under piping load cases discussed in Section 6.3 and, instead, a more realistic modeling of the piping end is used. RAI-5 is resolved. In addition, mesh density checks for FEM un-cracked and cracked models were also performed by the BWROG with results shown in Tables 4-1 and 4-2 and Figure 4-8. The staff examined these results based on different mesh densities and determined that reasonable stability of the stresses and KIs has been achieved by the current FEM models and the results presented later in Section 6 are credible.

Section 5 of the LTR provides information for the three types of loadings considered in the LTR:

internal pressure, thermal transient, and pipe reaction loads. Additional FEM modeling details

such as specific elements used and boundary conditions are also provided here. Except for the concern raised in RAI-2 regarding the source of piping loads which has been resolved as mentioned before, the staff found that all sources of loading have been considered and the top-level modeling schemes are consistent with industry practice.

Section 6 of the LTR presents results from applying the three types of loading discussed in Section 5 of the LTR, and thus required close staff examination because of its importance. Four RAIs were generated for this section: RAI-6 on the determination of the critical crack plane orientation for KIs, RAI-7 on adjustment of large elastic pseudo-stresses, RAI-8 on definition of a hoop stress extraction path with and without RPV cladding, and RAI-9 on the very different variation of KI values along the crack front under pressure and thermal loading.

The BWROGs June 13, 2012, response to RAI-6 states that, The postulated crack is always placed in an orientation such that it is normal to the maximum hoop stress in the RPV from the pressure load case. It further states that, the thermal stress distribution around the circumference of the nozzle blend radius does not exhibit substantial variation. Consequently, locating the postulated crack, for the thermal load case, in an orientation identical to that selected for the pressure load case results in the bounding combination of the thermal and pressure contributions to KI, for all nozzles. Since the postulated crack is located in an orientation having the bounding combination of the thermal and pressure contributions to KI for all nozzles, RAI-6 is resolved. The BWROGs June 13, 2012, response to RAI-7 states that, No adjustment to stress is made at any time.The comment included in the LTR was intended only to convey that, for the purposes of the figure, the contour scale on the plot was truncated such that the peak stress in the vicinity of the discontinuity was not shown. This is acceptable because no adjustment is made to the calculated stresses, and the adjustment on stresses is only for better contour plotting.

Regarding the definition of a hoop stress extraction path including or not including RPV cladding, the BWROGs June 13, 2012, response to RAI-8 states that, It is confirmed that in both cases [i.e., paths including and not including RPV clad] the thermal loading was applied to the cladding and the thermo-elastic stress analysis was performed with cladding in the finite element model. Two different paths were defined to extract path stress distributions: one starting at a radial location corresponding to the ID of the cladding, and one starting at a radial location corresponding to the ID of the LAS shellNeither path shown in Figure 6-7 passes through the air gap between the nozzle insert and the RPV shell. RAI-8 is resolved because the BWROG confirmed the staffs interpretation of the definition of paths for extracting stresses and confirmed that neither path passes through the air gap between the nozzle and the RPV.

Regarding the very different variation of KI values along the crack front under pressure and thermal loading, the BWROGs June 13, 2012, response to RAI-9 provided physical explanation of the phenomenon based on stresses. The staff considers the explanation reasonable and, therefore, RAI-9 is resolved. In conclusion, RAI-6, RAI-7, RAI-8, and RAI-9 are related to nozzle FEM and BIE/IF modeling, and successful resolution of them has cleared the staffs concerns over the results summarized in Figures 6-3 to 6-5 and Figure 6-9. Consequently, the staff can rely on the analytical results presented in Section 6 to determine acceptability of the LTR.

Section 7 of the LTR states that, For P-T curve analysis, a conservative 1/4 thickness flaw is assumed; a real flaw does not exist. Consequently, the inherent conservatism in assuming a 1/4 thickness flaw is considered sufficient such that requiring use of the maximum KI along the

entire crack front is considered to be excessively conservative. The staff had comments on this statement and did seek to gain additional information through RAI-10. RAI-10 also requests the BWROG assess the practicality of applying a factor of 1.1 to the applied pressure KI values based on the BIE/IF and 1.4 to the applied thermal KI values based on the BIE/IF to bound the applied KI values and find out (1) whether these factors will always make the instrument nozzle limiting for the P-T limit curves and (2) whether plant operation will be severely limited if the instrument nozzle becomes limiting.

The BWROGs June 13, 2012, response to RAI-10 provides, among background information and conservatism in the current ASME Code,Section XI, Appendix G approach, convincing arguments regarding the proposed BIE/IF approach:

1. Use of near surface KI values, rather than the deepest point is not consistent with ASME

[Code Appendix] G methods, as well as,

2. The observation that near a free surface the failure is governed more appropriately by a fracture toughness elevated above the plane strain fracture toughness, which implies greater margin to failure at this location, and,

3. The pressure term, which is shown to be 24 % conservative compared to the peak FE LEFM K, is further amplified by the Code required structural factor of 2.0 compared to the Code required structural factor of 1.0 for thermal loads.

4. The observation that a single nodal KI, near the free surface, from the FE LEFM analysis falls above the conservative BIE/IF solution does not represent a reduction in margin required by the ASME [Code Section] XI, Appendix G methods.

Based on the information provided by the BWROG and 10 CFR Part 50, Appendix G, ASME Code,Section XI, Appendix A and Appendix G, the staff believes that, unlike the ASME Code,Section XI, Appendix A method for flaw evaluation of a detected flaw where the nearly peak KI along the crack front and all load contributions to KI such as the differential thermal expansion of the clad and welding residual stresses have to be considered, the ASME Code,Section XI, Appendix G method is meant to establish fracture toughness requirements for all RPVs under the normal operation condition based on a postulated flaw, considering specific load contributions to KI. As long as the proposed LEFM evaluation for WLI nozzles follows the same assumptions made in the ASME Code Section XI, Appendix G method, the WLI nozzles will have the similar margin against brittle fracture as all RPVs when they are operated within the P-T limits based on the ASME Code,Section XI, Appendix G. Therefore, for interpretation of the BIE/IF results, the staff determined that, consistent with the ASME Code,Section XI, Appendix G method, the KI of the deepest point of the crack front, not the one close to the surface point, should be used when compared to the FEM results. As shown by Figure 6-5 for the pressure load case and Figure 6-9 for the thermal load case, the KI values based on BIE/IF at points closer to the deepest point of the crack front always bound the KI values based on FEM at these points, establishing the acceptability of the proposed KI methodology based on BIE/IF. Therefore, RAI-10 is resolved.

The BWROG summarized the BIE/IF and the FEM root-mean-square (RMS) KI results in Table 7-1 of the LTR. The table showes that the BIE/IF KI results using the proposed method are conservative by 39 percent for the pressure case and 46 % for the thermal load case when

compared with those from the FEM method. Consistent with the ASME Code,Section XI, Appendix G methodology, the staff used the FEM KI results close to the deepest point of the flaw (instead of the RMS KI results used by the LTR) and revised the conservatism estimate to 26 percent for the pressure case (see Figure 6-5 of the LTR for the FEM KI value close to the deepest point of the flaw) and 100 percent for the thermal load case (see Figure 6-9 of the LTR for the FEM KI value close to the deepest point of the flaw). These conservative estimates will be used to resolve RAI-11.

Section 8 of the LTR presents generic, empirical applied KI equations under pressure and the thermal ramp load based on curve-fitting of applied KI results for a variety of nozzles. Since the proposed equations are best estimate, instead of bounding, linear equations, the staff issued RAI-11 for additional information.

The BWROGs June 13, 2012, response to RAI-11 states that, Considering that the BIE/IF solution was shown to be conservative with respect to a plant specific FEA, by 46 percent for the thermal ramp load case and 39 percent for the internal pressure load case, in this same LTR, it is not considered necessary to add further conservatism to this methodology.

Consequently, we believe that keeping the curve fit equations as best-estimate curve fits rather than upper bound curve fit equations is justified. The staff determined that using quantified conservatism in one area to offset quantified none-conservatism in another is acceptable as long as the former is greater than the latter. In this case, the staffs revised conservatism estimates of 26 percent for the pressure case and 100 percent for the thermal load case far exceed the non-conservatism of 5 percent for the pressure and 15 percent for the thermal load case for using the best estimate equation instead of the bounding equation. Therefore, RAI-11 is resolved. The staff agrees with the BWROG that using the BIE/IF solution in the P-T limit curve analysis is appropriate.

4.0 CONDITIONS AND LIMITATIONS Based on the evaluation in Section 3.2, the staff determined that no conditions or limitations are necessary for future potential applicants to address in their application of this LTR to their plant-specific P-T limit submittals.

5.0 CONCLUSION

Based on the evaluation, the NRC staff concludes that LTR BWROG-TP-11-023, Revision 0 provides acceptable methodology for BWR licensees to obtain plant-specific stress intensity factors for an internal pressure load case and a 100 °F/hr thermal ramp load case for use in developing plant-specific P-T limit curves for RPV WLI nozzles. Since the analyses assumed that Alloy 600 material was used for the weld metal for all nozzles, the BWROG revised the LTR to limit the application of this LTR to only WLI nozzle configuration/design using Alloy 600 material for the weld metal. This proposed methodology is consistent with 10 CFR Part 50, Appendix G and the ASME Code,Section XI, Appendix G.

Principal Contributor: S. Sheng Date: October 31, 2012