ML111090402

Issuance of License Amendment Related to the Revision to the Reactor Vessel Head Drop Methodology

ML111090402
Person / Time
Site:	Point Beach
Issue date:	06/01/2011
From:	Beltz T Plant Licensing Branch III
To:	Meyer L Point Beach
beltz T, NRR/DORL/LPL3-1, 301-415-3049
References
TAC ME4006, TAC ME4007
Download: ML111090402 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 1, 2011 Mr. Larry Meyer Site Vice President NextEra Energy Point Beach, LLC Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241-9516

SUBJECT:

POINT BEACH NUCLEAR PLANT (PBNP), UNITS 1 AND 2 - ISSUANCE OF LICENSE AMENDMENT RELATED TO THE REVISION TO THE REACTOR VESSEL HEAD DROP METHODOLOGY (TAC NOS. ME4006 AND ME4007)

Dear Mr. Meyer:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment Nos. 242 and 246 to Renewed Facility Operating License Nos. DPR-24 and DPR-27, for PBNP. Units 1 and 2.

The amendments revise the licensing basis requirements for a postulated reactor vessel head drop event to conform to the NRC staff-endorsed guidance provided in Nuclear Energy Institute (NEI) 08-05, "Industry Initiative on Control of Heavy Loads," Revision 0, in response to your June 1, 2010, application, as supplemented by letters dated July 9, 2010, and November 22.

2010. The amendments support a change to the licensing basis in a revision to the Final Safety Analysis Report, Chapter 14.3.6, Reactor Vessel Head Drop Event.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

~--------

Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosures:

1. Amendment No. 242 to DPR-24

2. Amendment No. 246 to DPR-27

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 242 Renewed License No. DPR-24

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by NextEra Energy Point Beach, LLC (the licensee), dated June 1, 2010, as supplemented by letters dated July 9,2010, and November 22, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, by Amendment No. 242, the license is amended to authorize revision to the updated Final Safety Analysis Report (FSAR), as set forth in the application dated June 1, 2010, as supplemented by letters date July 9, 2010, and November 22, 2010.

The licensee shall update the FSAR to incorporate the licensing basis requirements related to a postulated drop of the reactor vessel head as described in the licensee's application and supplements, and the NRC staff's safety evaluation attached to this amendment, and shall submit the revised description authorized by this amendment with the next update of the FSAR.

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3.

The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance. The FSAR changes shall be implemented in the next periodic update of the FSAR in accordance with 10 CFR 50.71(e).

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: J"Lme 1, 2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 246 Renewed License No. DPR-27

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by NextEra Energy Point Beach, LLC (the licensee), dated June 1,2010, as supplemented by letters dated July 9,2010, and November 22,2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, by Amendment No. 246, the license is amended to authorize revision to the updated Final Safety Analysis Report (FSAR), as set forth in the application dated June 1, 2010, as supplemented by letters date July 9,2010, and November 22,2010.

The licensee shall update the FSAR to incorporate the licensing basis requirements related to a postulated drop of the reactor vessel head as described in the licensee's application and supplements, and the NRC staff's safety evaluation attached to this amendment, and shall submit the revised description authorized by this amendment with the next update of the FSAR.

- 2

3.

The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance. The FSAR changes shall be implemented in the next periodic update of the FSAR in accordance with 10 CFR 50.71(e).

FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: June 1, 2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 242 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND AMENDMENT NO. 246 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301

1.0 INTRODUCTION

By application dated June 1, 2010 (Reference 1), as supplemented by the letters dated July 9, 2010 (Reference 2), and November 22, 2010 (Reference 3), NextEra Energy Point Beach, LLC (NextEra, the licensee) requested a license amendment to revise the licensing basis requirements for a postulated reactor vessel head (RVH) drop event to conform to the U.S.

Nuclear Regulatory Commission (NRC) staff-endorsed guidance provided in Nuclear Energy Institute (NEI) OB-05, "Industry Initiative on Control of Heavy Loads," Revision 0, issued July 200B (Reference 9).

The supplements dated July 9, 2010, and November 22, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on September 21, 2010 (75 FR 57526).

2.0 REGULATORY EVALUATION

Balance of Plant Review The NRC staff developed guidelines for control of heavy load lifts in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," issued July 19BO (Reference 10). The guidelines were developed to assure safe handling of heavy loads in areas where a load drop could impact stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. In Section 5.1.1 of NUREG-0612, the guidelines to reduce the likelihood that heavy load handling would affect important-to-safety structures, systems, and components (SSCs) by addressing (1) the development of safe load paths; (2) the development of procedures for load-handling operations; (3) the training of crane Enclosure

- 2 operators; and (4) the design, testing, inspection, and maintenance of cranes and lifting devices.

The guidelines in Sections 5.1.2 through 5.1.6 of NUREG-0612 address alternatives that either further reduce the likelihood that heavy load handling would affect important-to-safety SSCs or establish bounds on the operation of the heavy load handling system to ensure the consequences of a handling system failure would be acceptable. These alternatives include performance of load drop consequence analyses.

In a letter dated December 22, 1980, later identified as NRC Generic Letter (GL)80-113 (Reference 12), as modified by GL 81-07 (Reference 13), the NRC staff requested that all licensees describe how they would fully satisfy the guidelines of NUREG-0612 and what additional modifications, if any, would be required. The NRC staff divided implementation of this request into two phases (Phase I and Phase II). The NRC staff requested that Phase I responses address Section 5.1.1 of NUREG-0612, and that Phase II responses address Sections 5.1.2 through 5.1.6 of NUREG-0612, as applicable.

The Wisconsin Electric Power Company (WEPCo), the licensee for the Point Beach Nuclear Plant (PBNP) at that time, responded to these letters through several submittals from 1981 through 1983. In a letter dated November 22, 1982, WEPCo provided a conservative evaluation of the effects of a RVH drop (Phase II information) based on completely elastic material behavior and indicated the reactor vessel support structure could buckle. The NRC staff accepted the Phase I actions proposed by WEPCo in a Safety Evaluation (SE) issued on March 27, 1984, but the NRC staff did not fully address the Phase" information provided by WEPCo.

In GL 85-11 (Reference 11), the NRC staff concluded that a detailed review of the Phase II responses received from licensees was not necessary based, in part, on acceptable improvements resulting from the Phase I review. In this manner, the NRC staff's closeout of the Phase II reviews resulted in a lack of consistency in plant licensing bases with respect to control of heavy loads.

During a review of the PBNP Unit 2 reactor head replacement in 2005, the NRC staff identified the inconsistency in the evaluation of the postulated RVH drop. The Nuclear Management Company (NMC), LLC, the licensee for PBNP in 2005, prepared and submitted a RVH drop analysis for NRC staff review to resolve NRC staff questions pertaining to a postulated RVH drop event. In a letter dated September 23,2005 (Reference 5), the NRC issued Amendment Numbers 220 and 226 to the facility operating licenses for PBNP, Units 1 and 2, respectively.

The license amendments incorporated a PBNP Unit 1 RVH drop accident analysis into Chapter 14.3.6 of the PBNP Final Safety Analysis Report (FSAR) and revised the PBNP Unit 2 RVH drop accident analysis. The analyses involved a structural analysis of the impact on the reactor vessel, reactor coolant system (RCS) piping, and the reactor vessel supporting structures. The analyses included a radiological consequence analysis predicated on an assumption that the impact would result in a fuel clad gap release 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.

The analyses also included a core cooling evaluation based on the presumptive failure of the bottom mounted instrumentation (BMI) conduits located beneath the reactor vessel and the subsequent water loss from the reactor vessel.

An industry initiative to address the lack of consistency in plant licensing bases related to control of heavy loads on a generic basis was proposed by NEI in a letter dated September 14, 2007 (Reference 14). The industry initiative included the following elements:

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• For all heavy load lifts, adequately implement commitments to safe load paths, load handling procedures, training of crane operators, use of special lifting devices, use of slings, crane design, and inspection, testing, and maintenance of the crane.

• For RVH lifts, establish a load drop analysis (generic or plant-specific) that bounds your planned lifts with respect to load weight, load height, and medium present under the load.

• If load drop analyses are used, reflect restrictions on load height, load weight, and medium present under the load in plant procedures.

• In the safety analysis report, include a summary description of the basis for conducting safe heavy load movements, including commitments to safe load paths, load handling procedures, training of crane operators, use of special lifting devices, use of slings, crane design, and inspection, testing, and maintenance of the crane. If the safety basis includes reliance on a load drop analysis, then that fact should be included in the summary description within the safety analysis report.

The industry task force established by NEI to implement the initiative met with NRC staff in several public meetings to develop guidelines for industry use in improving the consistency of licensing basis information related to control of heavy loads.

In a letter dated July 28,2008 (Reference 4), NEI transmitted NEI 08-05, "Industry Criteria for Reactor Vessel Head Load Drop and Consequence Analysis," Revision 0, for NRC review. The guidance contained in Section 2 of NEI 08-05, includes (1) a comparison of NUREG-0612 guidelines for analyses of postulated RVH drops and the industry initiative guidelines; (2) general guidelines for the analysis; (3) selection of material properties; (4) analytical modeling requirements; and (5) acceptance criteria when evaluating the effects of postulated heavy load drops. The industry limited the scope of the guidelines on RVH drop analyses to those necessary to demonstrate that, after a postulated RVH drop accident, the core would remain covered with coolant and that sufficient cooling would be available.

Through a letter to NEI dated September 5, 2008 (Reference 7), the NRC staff forwarded its Safety Evaluation Report (SER) addressing NEI 08-05, Revision O. In the SE, the NRC staff accepted the guidelines and methodology presented in Section 2 of NEI 08-05, with one exception regarding acceptance criteria for coolant pressure boundary components. The NRC staff found that limiting the scope of the evaluation of RVH drops to assuring the core would remain covered with coolant was acceptable because:

• implementation of the guidelines in Section 5.1.1 of NUREG-0612 had reduced the probability of RVH drop (GL 85-11)

• if dropped, the RVH cannot credibly contact the fuel in the vessel directly

• previous evaluations have indicated that the consequences of impacts between the upper vessel internals and the fuel were not significant with respect to public health and safety (due in part to the strength of the fuel in its longitudinal direction and damping provided by the water in the vessel)

- 4 Consequently, the NRC staff found that an evaluation of the radiological consequences of a postulated RVH drop was unnecessary to assure adequate safety. However, the NRC staff found that the acceptance criteria of Appendix F to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, (rather than the less-restrictive, industry-proposed strain-based criteria presented in NEI 08-05) would be appropriate for evaluation of coolant-retaining component performance following a postulated RVH drop.

The proposed change to the licensing basis would revise FSAR Chapter 14.3.6. The proposed revision to FSAR Chapter 14.3.6 includes removing the radiological consequence analysis associated with the postulated RVH drop; incorporating the results of a new, dynamic analysis of BMI conduit integrity; and removing elements of the core cooling evaluation associated with coolant loss through the BMI conduits.

Mechanical and Civil Engineering Branch Review The NRC staff's regulatory guidelines associated with the control of heavy loads are outlined in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," issued in July 1980.

Appendix A to NUREG-0612, "Analyses of Postulated Load Drops," includes NRC staff-provided guidelines for performance of load drop consequence analyses, such as the drop of a RVH. As part of the actions taken at PBNP, Units 1 and 2, to address the control of heavy loads at the facility, the results of the original RVH drop analysis were presented to the NRC by letter dated November 22, 1982. Facility Operating License Amendments 220 and 226, for PBNP Units 1 and 2, respectively, incorporated the RVH drop analysis into Chapter 14 of the PBNP FSAR.

The corresponding NRC staff SE related to these license amendments was issued on September 23, 2005 (Reference 5). This SE was subsequently revised by letter dated January 12, 2006 (Reference 6). The June 1, 2010, license amendment request (LAR) and supplements would revise the current licensing basis (CLB) requirements related to the RVH drop event consequence analyses described in the PBNP FSAR.

The NRC staff's acceptance criteria in the areas of mechanical and civil engineering, related to the review of the proposed change to the RVH drop accident analyses, are based on General Design Criteria (GDC)-4, which requires, in part, that SSCs important to safety be designed to accommodate the dynamic effects of missiles, such as a RVH, that may result from equipment failures. The GDCs are located in Title 10 of the Code of Federal Regulations (10 CFR),

Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants." In the PBNP FSAR, Section 1.3, "General Design Criteria," indicates that the GDCs documented in the PBNP FSAR are similar in content to the Atomic Industrial Forum versions of the proposed (1967) GDCs, instead of the GDCs documented in 10 CFR Part 50, Appendix A. The GDCs documented in 10 CFR Part 50, Appendix A, were published after the construction permits for PBNP were issued. The NRC staff's review of the PBNP FSAR, Table 1.3-1, "Point Beach General Design Criteria," indicates that PBNP GDC 40 is similar to GDC 4.

Additional guidance and acceptance criteria related to the NRC staff's review of the proposed revision to the PBNP RVH drop event analysis can be found in NEI 08-05. Section 2 of NEI 08 05 provides specific guidance on the analytical aspects of the RVH drop accident. As indicated in this section of the NEI guidance, the RVH drop accident is a "beyond design basis" accident scenario and the acceptance criteria are limited to ensuring that the core remains covered with coolant and sufficient core cooling is available following the accident. Therefore, the analysis of

-5 such an event can be performed by exploiting analytical tools which can be used to provide realistic or best estimate predictions of the structural behavior of an SSC which is subject to the effects of a RVH drop impact.

The NRC staff's review of NEI 08-05 is documented in an SE dated September 5, 2008 (Reference 7), and includes clarifications and conditions on the implementation of the guidance found in NEI 08-05. As indicated in NEI 08-05 and the NRC staff's associated SE, the issuance of this NEI report was the result of NRC staff and industry recognition that the actions requested through previous NRC generic communications had resulted in a lack of consistency in plant licensing bases with respect to the control of heavy load lifts. Accordingly, Section 2.2 of NEI 08-05 contains a detailed comparison between the guidance within NEI 08-05 and that provided in NUREG-0612 regarding the RVH drop analyses. Regulatory Issue Summary (RIS) 2008-28, "Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts," dated December 1, 2008 (Reference 15), documented the NRC staff's approval of the use of the methods found in NEI 08-05 and also cited the NRC staff's associated SE for further information on the conditions and clarifications for licensee's using the methods in the NEI report.

=

Background===

The NRC staffs mechanical and civil engineering technical evaluation focused on the structural integrity of SSCs which could be impacted by the licensee's proposed changes to the PBNP licensing basis as part of this LAR. In addition to the NRC staff's review of the June 1, 2010, LAR, a summary of the current analysis of record (AOR) related to the PBNP RVH drop accident is presented below due to the fact that the LAR cites information which was previously reviewed and approved by the NRC staff in References 5 and 6. The current AOR related to the PBNP RVH drop accident evaluated the structural integrity of many of the SSCs which could be impacted by the postulated drop of the RVH The analyses presented as part of this LAR do not affect the conclusions reached by the licensee and the NRC staff regarding the structural integrity of previously evaluated SSCs. The NRC staff's review presented below is limited to SSCs whose structural integrity was previously unevaluated.

3.1 Current Analysis of Record Related to a Reactor Vessel Head Drop Event The current AOR related to the RVH drop event is described in Section 14.3.6 of the PBNP FSAR. The primary assumptions used in the RVH drop event analysis are as follows: (1) the RVH weighs approximately 200,000 pounds, and (2) the RVH is dropped from an elevation of 26.4 feet above the reactor pressure vessel (RPV) closure flange, with a flat, concentric impact of the RVH on top of the RPV flange. The load path for this postulated accident includes the RPV, RPV supports at the four vessel nozzles and two brackets, the support girder box frame, and the six pipe columns and their supports, which rest on the concrete foundation. The results of the finite element analyses (FEA) performed for the current AOR demonstrated that the structural integrity of those SSCs evaluated during the analyses would not be compromised during a postulated RVH drop accident, including the RCS loop piping. The NRC staff concluded in its September 23, 2005, SER that the licensee provided reasonable assurance that structural integrity of the RCS loop piping would be maintained in the event that a postulated RVH drop.

- 6 With respect to the current AOR, the September 23, 2005, SER also notes that the 36 BMI conduits which are welded to the nozzles at the bottom of the PBNP RPVs were previously assumed to be severed during a postulated RVH drop accident. The displacement time histories determined during the FEA supporting the current AOR revealed that the maximum displacement of the RPV during the postulated RVH drop accident was found to be larger than the clearance between a portion of the BMI tubes and the floor beneath them. As such, the licensee assumed in the current AOR that all of the BMI conduits were severed during a postulated RVH drop event.

3.2 Removal of Radiological Consequence Analysis In the July 24, 2005, application (Reference 16) associated with Amendments 220 and 226 to the operating licenses for PBNP Units 1 and 2, respectively, NMC proposed an evaluation of the radiological consequences of a RVH drop that assumes cladding damage to 100 percent of the fuel assemblies such that a complete gap release occurs. This assumption was intended to result in a bounding assessment of radiological consequences associated with the postulated RVH drop event. However, the assumed damage was not mechanistically related to the postulated RVH drop.

As addressed in the SE associated with NEI 08-05, the NRC staff found that a radiological consequence analysis of a postulated RVH drop was not required. The bases for this finding included the following:

• Phase I actions to reduce the probability of a heavy load drop to low values, Inability of the RVH to credibly make direct contact with fuel in the reactor vessel, and

• Several previous evaluations of postulated reactor vessel internal component drops that indicated the radiological consequences of such events would be small because of the longitudinal strength of the fuel and damping provided by the vessel water.

Accordingly, the NRC staff found the proposed removal of the radiological consequence analysis for a postulated RVH drop from the PBNP FSAR acceptable.

3.3 Bottom Mounted Instrument Conduit Evaluation As indicated in Enclosure 2 to Reference 2, the 36 BMI conduits are each connected to a nozzle, which conjoins each BMI conduit to the bottom of the RPV. The BMI conduits make up a portion of the RCS boundary, beginning at the RPV junction and terminating at the conduit seal table, and a failure of any of the conduits would result in a loss of RCS inventory. The licensee's proposed revision to the CLB requirements related to the RVH drop accident is limited to the behavior of the BMI conduits during a postulated drop of the RVH onto the RPV, given that the other SSCs which could be affected by the postulated drop event have been previously structurally qualified. Acceptability, as stated in NEI 08-05 and the NRC staffs SE endorsing NEI 08-05, is based on demonstrating that the core will remain covered with coolant and core cooling capabilities will be maintained following a RVH drop accident. From a structural integrity perspective, this is ensured by maintaining the BMI conduit stress levels below applicable stress limits such that the conduits do not rupture in the event of a RVH drop.

The licensee indicated that the evaluations described in Reference 1, which were performed in support of the proposed modification to the postulated RVH drop accident consequences, are

-7 based on a dynamic FEA performed using the ANSYS computer program. The licensee stated in Reference 1 that the FEA were performed in accordance with the NRC staff-approved guidance of NEI 08-05. The stresses induced in the 8MI conduits during the RVH drop event, as determined by the FEA, were compared with the applicable stress limits of the ASME 8&PV Code, Division 1,Section III, Appendix F, "Rules for Evaluation of Service Loadings with Level D Limits." The FEA and resulting comparisons to the ASME 8&PV Code stress limits are described in detail below.

3.3.1 Finite Element Analysis Model The FEA performed in support of the proposed LAR was developed by Westinghouse Electric Company, LLC, and utilizes the same RPV displacement time-histories previously determined for the current AOR. Four different FEA cases were performed, with three of the cases utilizing a combination of the large deflection option in ANSYS and explicit floor contact between the 8MI conduits and the concrete beneath the conduits. The FEA modeled two different 8MI conduits; one representing the longest 8MI conduit (conduit 29) and the other representing the shortest 8MI conduit (conduit 32). The licensee indicated in Enclosure 2 to Reference 2 that the same model for each conduit was used for both P8NP Units 1 and 2, given the fact that their design and construction are identical. While the physical model used for both units was the same, each of the four cases was run for each unit, given that each unit has unique displacement time-histories. In response to an NRC staff RAI, which requested justification for only modeling two of the 36 conduits, the licensee stated in Reference 3 that the thick-walled conduits are not pressurized during RVH lift operations and the construction of the racks which support the conduits ensures that the conduits move in-phase, in the event that the RVH should drop onto the RPV. As such, the licensee stated that incidental conduit-to-conduit contact would not result in unacceptable stress levels, thus providing the rationale for only modeling the longest and shortest conduits.

The finite element model is described in Section 5.1.1 of Enclosure 2 to Reference 2. As described, the model was developed using dimensional and walk down data applicable to the 8MI conduits and their associated supports, in order to ensure the most accurate configuration of the model possible. Figure 5-1 of Enclosure 2 to Reference 2 illustrates 8MI conduits and the boundary conditions representing the 8MI conduit supports. As Reference 2 indicates, the supports are rigidly modeled and constrained in two translational directions while all rotational degrees of freedom are permitted. In modeling the seal table boundary condition, the node representing the seal table was fixed for all degrees of freedom, preventing any displacement at this node. For the two cases where floor contact was assumed to occur, the floor surface was modeled as a rigid target.

In the course of the NRC staff's acceptance review, the NRC staff queried the licensee regarding whether the structural behavior of the supports, including stresses and displacements seen by the supports during the RVH drop event, had been considered in the FEA. As the description above indicates, the structural behavior of the supports was not considered and the supports were modeled only as rigid boundary conditions, i.e., the supports themselves were not explicitly modeled. The licensee notes in Enclosure 1 to Reference 2 that the supports are constructed such that they are substantially stiffer than the 8MI conduits, thus negating the need to assess their structural integrity for this accident analysis. Additionally, by modeling the supports as rigid boundary conditions, the supports absorb no energy, providing a conservative

- 8 estimate for the stresses induced in the BMI conduits by maximizing the energy absorbed by the conduits.

The applied loads used in the BMI conduit structural evaluation are based on the displacement time-histories developed in the current AOR for PBNP Units 1 and 2. These displacement time histories represent the downward displacement of the RPV once it is impacted by the RVH. The licensee noted in Reference 1 that when the displacement time-histories were applied to the finite element model at the BMI conduit-to-RPV interface, incorrect applied accelerations resulted due to "noise;" this numerically anomalous input is visible in Figure 4-3 of Enclosure 2 to Reference 2. To eliminate the noise, the licensee stated that a spring-mass system was added to the finite element model to better represent the actual behavior of the displacement time-histories being applied through the RPV. This improved response is visible in Figures 5-5, 5-6, and 5-7 of Enclosure 2 to Reference 2, which illustrates the applied displacement time histories and the corresponding displacement time-histories of the top of the spring-mass system (node 10000000) and the BMI conduit-RPV interface (node 1), respectively.

Consistent with the guidance in NEI 08-05, the licensee indicated in Reference 1 that a true stress-strain curve was utilized in the FEA to model the material behavior of the BMI conduits during the RVH drop accident. Additional information regarding the construction of the true stress-strain curve was provided in response to an NRC staff RAI (Reference 3). As indicated in the licensee's RAI response, the stress-strain curve used in the FEA, described above, was developed by applying percentage increases to the ASME B&PV Code minimum material properties for the yield and ultimate stresses of Type 304 Stainless Steel (SS). The percentage increases were used to develop "typical" values for Type 304 SS material. Using the increased minimum values for the yield and ultimate stresses, a standard relationship between engineering and true stresses and strains was applied to develop the final, true stress-strain curve used in the FEA. A truncated version of the curve is illustrated in Figure 4-6 of to Reference 2, with the licensee noting that the final, ultimate stress and strain values used in the analysis were 115,500 pounds per square inch and 0.337 inch per inch, respectively.

3.3.2 Finite Element Analysis Results The results of the FEA described above are presented in Tables 1 through 4 of Reference 1.

These tables summarize the highest stresses induced in the BMI conduits by the four FEA cases performed for each unit, described above. The tables also compare these stresses with the applicable allowable stress limits of the 1998 Edition through 2000 Addenda of the ASME B&PV Code Section III, Appendix F; these stress limits were also utilized to structurally qualify the SSCs evaluated in the licensee's current AOR. In establishing the membrane and membrane plus bending allowable stress intensity limits, it was stated in Enclosure 2 to Reference 2 that these values incorporated the use of the ASME B&PV Code minimum values for the ultimate stress of the Type 304 SS conduit material.

An examination of the results indicates that the largest stresses developed in the BMI conduits, due to the application of the RVH drop displacement time-histories, were located at the BMI conduit-to-RPV nozzle location (node 1) for each of the four cases considered. Additionally, the most limiting case evaluated was the case where the large deflection option was incorporated into the FEA, without floor contact, as provided in Table 2 of the June 1, 2010, application.

Further examination of the results indicates that for all four analysis cases considered, the

- 9 membrane stress intensity and the membrane plus bending stress intensity allowable values, as established by the applicable ASME B&PV Code provisions, were satisfied. The most limiting conduit in PBNP Unit 1 was BMI conduit 32, which maintains a structural margin of 3.86 percent when compared against the ASME B&PV Code stress limits. The most limiting conduit for PBNP Unit 2 was also BMI conduit 32, which maintains a structural margin of 4.01 percent against the ASME B&PV Code stress limits.

Additionally, with respect to the results presented in Reference 1, it was noted that only the results for conduit 29 were presented for both units in the case where floor contact is explicitly considered, due to the fact that BMI conduit 32 does not deflect enough to make contact with the floor. Therefore, the results for BMI conduit 32 in Tables 1 and 2 would remain unchanged for the two cases which considered the stresses induced in the BMI conduits resulting from floor contact.

3.3.3 Summary The NRC staff considers the licensee's structural evaluation of the BMI conduits acceptable, based on (1) the adequacy of the licensee's FEA, which conforms to NRC staff-approved guidance for these analyses, and (2) the licensee's quantitative demonstration that the applicable stress limits for the BIVII conduits were not exceeded in the evaluation performed for the conduits; these bases are expanded on below. As previously indicated, the NRC staff approved guidance of NEI 08-05 describes the RVH drop event as a one-time, "beyond design basis" accident, which permits plastic deformation of SSCs associated with the RPV, provided that the aforementioned core cooling criteria are satisfied. As such, NEI 08-05 and RIS 2008-28 state that best estimate analyses may be performed to provide a more accurate prediction of the structural behavior of SSCs potentially affected by the accident. The NRC staff notes in a number of instances, described below, that the licensee utilized conservatisms in its analyses which were not required by NEI 08-05. While employing these conservatisms may not provide best estimate structural behavior predictions, implementing these conservatisms provided bounding estimates of the stresses induced in the BMI conduits during the RVH drop accident.

The NRC staff finds the licensee's FEA model and methodology acceptable, given that its development was consistent with the NRC staff-endorsed guidance of NEI 08-05. The FEA model of the 8MI conduits was based on actual dimensional data and walkdown information from PBNP, providing high confidence that the model accurately represents the true plant conditions and enables an accurate FEA. The NRC staff also finds the licensee's assumption that the supports need not be evaluated acceptable, based on the information provided in Reference 3, stating that 1) the supports are much stiffer than the conduit and 2) neglecting the ability of the supports to absorb any of the RVH drop impact energy is conservative, due to the fact that it enables the BMI conduits to absorb additional energy, resulting in higher induced stresses which are subsequently compared to the applicable stress limits. The NRC staff also finds the licensee's application of the applied loads to only two 8MI conduits in the FEA adequate, based on the information provided in Reference 3 which adequately demonstrates that the BMI conduits will move in phase during the postulated RVH drop event and any conduit conduit contact stresses are negligible, when compared to the stresses which are induced at the conduit-to-RPV junction, where no conduit-to-conduit contact can occur.

The NRC staff finds the licensee's use of a true stress-strain curve acceptable, given that it is one of the primary analytical requirements outlined in NEI 08-05, given that the use of such

- 10 material data is in accordance with the best estimate philosophy for this type of analysis. The NRC staff also finds the licensee's true stress-strain curve data acceptable. This acceptance is based on a comparison of the licensee's curve to other Type 304 SS true stress-strain curves found in the American Society of Metals International, Atlas of Stress-Strain Curves, "Stainless Steel," issued in 2002; the data is recommended for use in NEI 08-05 as a viable source for true stress-strain curves. The data comparison shows that the licensee's curve is more conservative than the Type 304 SS curves found in Reference 8. Therefore, based on the fact that the data used by the licensee is conservative compared to data previously approved by the NRC staff, the NRC staff considers the licensee's curve acceptable.

The NRC staff finds the licensee's results of the FEA acceptable based on the following: 1) the licensee provided the results of four bounding cases performed for the postulated RVH drop event and 2) the stresses induced in the BMI conduits in the FEA were shown to be within the stress limits of the applicable ASME B&PV Code provisions. Tables 1 through 4 of Reference 1 provide a quantitative conclusion that the BMI conduits satisfy the stress limits of the applicable provisions of Section III, Appendix F of the ASME B&PV Code. Additionally, by utilizing these stress limits, the NRC staff notes that the licensee has also satisfied the NRC staff's conditions regarding the use of NEI 08-05, identified in Reference 7, and RIS 2008-28, which state that the ASME B&PV Code stress criteria of Appendix F should be utilized as the structural acceptance criteria for these evaluations, in lieu of the industry-proposed strain-based criteria.

The guidance in NEI 08-05 states that the large deflection option should be utilized in the FEA performed to support evaluations of a postulated RVH drop event. The analyses outlined in to Reference 2 and the results of these analyses, provided in Tables 1 through 4 of Reference 1, demonstrate that the licensee has satisfied this criterion by employing the large deflection option in the FEA. The most limiting case evaluated for the structural evaluation of the BMI conduits assumed no floor contact with the large deflection option enabled. The NRC staff notes that while this iteration did not consider the realistic contact with the floor which BMI conduit 29 would have experienced, the licensee performed other iterations which did consider floor contact, including a case which modeled floor contact with the large deflection option enabled.

The NRC staff concludes that there is reasonable assurance that the structural integrity of the 8MI conduits will be maintained during the accident outlined above, which postulates the drop of the RVH at PBNP onto the RPV. This conclusion is based on the licensee's conformance to the NRC staff-approved guidance associated with the analytical evaluations of the RVH drop accident and the licensee's quantitative demonstration that the stresses induced in the BMI conduits during such an accident will remain less than the stress limits provided by the applicable provisions of the ASME B&PV Code. By satisfying the stress criteria used to demonstrate structural integrity of the BMI conduits during the RVH impact event, the !\\IRC staff further concludes that the licensee has adequately satisfied the acceptance criteria required for the RVH drop accident, which requires that the core remains covered with coolant and sufficient core cooling is available following the accident.

3.4 Modification to the Core Cooling Evaluation In the application for Amendments 220 and 226 associated PBNP Units 1 and 2, respectively (Reference 16), the !\\IMC concluded and the NRC staff accepted that adequate reactor core cooling and makeup capability would be maintained following the expected deflection of the

- 11 reactor vessel from a postulated RVH drop. However, the licensee assumed complete severance of the BMI conduits, which would result in a non-isolable loss of coolant from the reactor vessel. To provide adequate cooling in this assumed scenario, the licensee credited containment closure and the availability of safety injection pumps, residual heat removal pumps, refueling water storage tank inventory, and containment sump recirculation capability.

The analysis described in the LAR submitted by letter dated June 1, 2010, determined that the BMI conduits would maintain pressure boundary integrity following the postulated RVH drop.

Since the pressure boundary integrity of the reactor vessel and attached systems following a postulated RVH drop was adequately demonstrated, the requirement for two operable residual heat removal loops specified in Technical Specification 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level," ensures adequate capability to maintain the core covered with coolant and provide sufficient cooling. Therefore, the evaluation criteria for core cooling following a postulated head drop would be satisfied by required equipment, and the staff found removal of the description of containment closure and availability of safety injection pumps and containment sump recirculation capability from Section 14.3.6 of the PBNP FSAR acceptable.

3.5 Conclusion Based on its review described above, the NRC staff considers the proposed LAR, regarding the revision of the PBNP Units 1 and 2 licensing basis requirements related to the postulated drop of the RVH, acceptable. This acceptance is outlined above and is based on the licensee's conformance to the NRC staff-approved criteria for evaluating the structural integrity of SSCs which could be affected by a postulated RVH drop and demonstrating that these SSCs (the BMI conduits) will maintain their structural integrity during such an event. Furthermore, based on its review described above, the NRC staff has concluded that the regulatory requirements described in SE Section 2.0 will continue to be satisfied following the implementation of the proposed revision to the licensing basis requirements related to this postulated accident.

Therefore, there is reasonable assurance that the structural integrity of the affected SSCs will be maintained following the implementation of the proposed re-rack.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no Significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on September 21, 2010 (75 FR 57526). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

- 12 Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1)

Letter from L. Meyer, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, License Amendment Request 265, Revision to the Reactor Vessel Head Drop Methodology," dated June 1, 2010 (ADAMS Accession No. ML101520200).

2)

Letter from L. Meyer, NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, License Amendment Request 265, Revision to the Reactor Vessel Head Drop Methodology, Supplement 1," dated July 9,2010. (ADAMS Accession Nos. ML102030115 (Cover Letter and Enclosure 1) and ML102030116 (Enclosure 2>>.

3)

Letter from L. Meyer (signed by Charles Trezise), NextEra Energy Point Beach, LLC, to NRC Document Control Desk, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50 266 and 50-301, Renewed License Nos. DPR-24 and DPR-27, License Amendment Request 265, Revision to the Reactor Vessel Head Drop Methodology, Response to Request for Additional Information," dated November 22, 2010 (ADAMS Accession No. ML103270128).

4)

Letter from A. R. Pietrangelo, NEI, to E. J. Leeds, NRC, "Industry Initiative on Control of Heavy Loads," dated July 28, 2008 (ADAMS Accession No. ML082180666).

5)

Letter from H. K. Chernoff, NRC, to D. L. Koehl, Nuclear Management Company, LLC, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendment re: Incorporation of Reactor Vessel Head Drop Accident Analysis into the Final Safety Analysis Report (TAC Nos. MC7650 and MC7651)," dated September 23,2005 (ADAMS Accession No. ML052560089).

6)

Letter from C. F. Lyon, NRC, to D. L. Koehl, Nuclear Management Company, LLC, "Point Beach Nuclear Plant, Units 1 and 2 - Revision to Safety Evaluation for Amendment Nos. 220 and 226 (TAC Nos. MC7650 and MC7651)," dated January 12, 2006 (ADAMS Accession No. ML052850005).

- 13

7)

Letter from W. H Ruland, NRC, to T. C. Houghton, Nuclear Energy Institute, "Industry Initiative on Control of Heavy Loads," dated September 5,2008 (ADAMS Accession No. ML082410532).

8)

American Society of Metals International, Atlas of Stress-Strain Curves, "Stainless Steel," 2002, Materials Park, OH.

9)

Nuclear Energy Institute (NEI) 08-05, "Industry Initiative on Control of Heavy Loads,"

Revision 0, issued July 2008 (ADAMS Accession No. ML082180684).

10)

NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," issued July 1980 (ADAMS Accession No. ML070250180).

11)

Generic Letter 85-11, "Completion of Phase II of "Control of Heavy Loads at Nuclear Power Plants," NUREG-0612," dated June 28, 1985 (ADAMS Accession No. ML031150689).

12)

Generic Letter 80-113, "Control of Heavy Loads," dated December 22, 1980 (ADAMS Accession No. ML071080219).

13)

Generic Letter 81-07, "Control of Heavy Loads," dated February 3,1981 (ADAMS Accession No. ML031080524).

14)

Letter from A. R. Pietrangelo, NEI, to J. E. Dyer, NRC, "Industry Initiative on Heavy Load Lifts," dated September 14,2007 (ADAMS Accession No. ML072670127).

15)

Regulatory Issue Summary 2008-28, "Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts," dated December 1,2008 (ADAMS Accession No. ML082460291).

16)

Letter from D. L. Koehl, Nuclear Management Company, LLC, to NRC Document Control Desk, "Point Beach Nuclear Plant, Units 1 and 2, Dockets 50-266 and 50-301, License Nos. DPR-24 and DPR-27, Request for Review of Heavy Load Analysis," dated July 24, 2005 (ADAMS Accession No. ML052140556).

Principal Contributors: W. Jessup, NRR S. Jones, I\\IRR Date: June 1, 2011

ML111090402

• concurrence via memo

.. NLO with comment OFFICE NRRlLPL3*1/PM NRR/LPL3-1/PM OGC B/BC NRRISBPB/BC (A) NRR/LPL3*1/BC NAME TBeltz BTuily LSubin **

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• GPurciarelio

• RPascarelli DATE 05/11/11 05/11111 05/20/11 04/13/11 03/25/11 06/01/11