ML18149A466

Point Beach Nuclear Plant, Units 1 and 2 - Construction Truss License Amendment Request 278 Draft Updated Final Safety Analysis Report (UFSAR) Revision

ML18149A466
Person / Time
Site:	Point Beach
Issue date:	05/29/2018
From:	Craven R Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2017-LLA-0209, NRC 2018-0030
Download: ML18149A466 (13)

Text

NEXT era POINT BEACH May 29, 2018 NRC 2018-0030 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Point Beach Nuclear Plant, Units 1 and 2 Docket 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 NextEra Energy Point Beach. LLC, Construction Truss License Amendment Request 278 Draft Updated Final Safety Analysis Report (UFSAR) Revision

Reference:

1. NextEra Energy Point Beach, LLC, License Amendment Request 278, Risk-Informed Approach to Resolve Construction Truss Design Code Nonconformances , dated March 31, 2017 (ML 17090A511)

2. NRC Letter, Point Beach Nuclear Plant , Units 1 and 2 -Request for Additional Information for Point Beach Nuclear Plant , Units 1 and 2, Regarding License Amendment Request to Resolve Nonconformances Relating to Containment Dome Truss (EPID L-2017-LLA-0209), dated January 31 , 2018 (ML 18025C043)

3. NextEra Energy Point Beach, LLC, Construction Truss License Amendment Request 278 , Response to Request for Additional Information, dated April 12, 2018 (ML 18102B164)

By letter dated March 31, 2017 (Reference 1 ), NextEra Energy Point Beach (NextEra) submitted License Amendment Request (LAR) 278 to resolve legacy design code nonconformances associated with the Point Beach Units 1 and 2 containment dome construction trusses. By letter dated January 31 , 2018 (Reference 2), NRC requested additional information related to LAR 278. A response was provided on April 12 , 2018 (Reference 3). In Reference 1, NextEra provided a &aft Updated Final Safety Analysis Report (UFSAR) revision for information. The enclosure to this letter provides a revised draft UFSAR revision, also for information.

The UFSAR will be updated after LAR'approval and after LAR 278 regulatory commitments have been implemented.

The final UFSAR verbiage will be determined upon receipt of the license amendment associated with LAR 278. This letter contains no new regulatory commitments.

Ne x tEra Energy Point Beach , LLC 6610 Nuclear Road, Two Rivers, WI 54241 Document Control Desk Page 2 If you have any questions please contact Mr. Eric Schultz, Licensing Manager, at (920) 755-7854.

Sincerely , NextEra Energy Point Beach, LLC Robert Craven Site Director cc: Director, Office of Nuclear Reactor Regulation Administrator , Region Ill, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Project Manager, Point Beach Nuclear Plant, USNRC

Enclosure:

1. Draft Updated Final Safety Analysis Report Revision in Support of LAR 278 Enclosure 1 Draft Updated Final Safety Analysis Report Revision in Support of LAR 278 (1 O pages follow)

NOTE: UFSAR Update to occur after Ul modifications are complete 5.6 CONS1RUCTION

5.6.1 CONSTRUCTIO

N METHODS 5.6.Ll APPLICABLE CODES Construc tion FSAR Section 5.6 n1e following codes of practice are ll5 ed t o establish standards of construction procedure:

ACI301 ACI306 ACI318 ACI347 ACI605 ACI613 ACl614 ACI315 ASME AISC PCI 5.6.1.2 Specification for Structural Concrete for Buildings (Proposed)

Recommended Practice for Cold Weather Concreting Building Code Requirements for Reinforced Concre t e Recommended Practice for Concrete Formwork Reconunended Practice for Hot Weather Concreting Recommended Practice for Selecting Proportions for Concrete Recommended Practice for Measuring.

Mixing and Placing Concrete Manual of Standard Practice for Detailing Reinforced Concrete Structures Boiler and Pressure Vessel Code , Sections III , VIII and IX Steel Con.5truction Ivfanual Inspection Manmtl CONCREIB Cast-in-place concrete was used to construct the containment shell. The base sfab construction was performed utilizing farge block pours. After the completion of the base slab steel liner erection and testing , an additional 18 in_ thick concrete slab was p l aced to provide protection for the floor liner. The concrete placement in the walls was done in I 0 ft. high lifts with vertical joints at the radial center line of each of six buttresses. Cantile\ ered jump forms on the exterior face and the interior steel wall liner seived as the forms for the wall concrete. The dome liner pla t e, t emporarily supported by 18 radial steel trusses and purlins , served as an inner form for the initial 8 in thick pour in the dome. The weight of the subsequent pour was supported in tum by the initial 8 in. pour. Tue tmsses were lowered away from the liner plate after the initial 8 in. of concrete reached design strength , but prior to the placing of the balance of the dome concrete. TI1e horizontal and the ve

• 1 con stru ction joints were prepared by dry sandblasting follo\11 ed by deaning and wetting. Horizo 1 surface s were co v ered with approximately 1/4 in. thick mortar of the same cement-sand ratio as ed in the concrete inuuediately before concrete placing. 5.6.1.3 REINFORCING STEEL Prior t o placing , vis u al of the sho abricated reinforcing steel was perfom1ed to ascertain dimensional conform.mce with design ecifications and the drawings. This was followed b y a check " in place " performed by the 1g inspector to as s ure the dimension.11 and location conformance. UFSAR WlO Page 5.6-1 of 2 9 The truss structures have remained in the lowered position since construction and are used as a support for containment spray piping, containment air recirculation cooling system (VNCC) ductwork , post-accident containment ventilation (PACV) piping, and miscellaneous lights and associated conduits.

See Section A.5.10 for resolution of design code nonconformances associated with as-built discrepancies related to the truss structures and the ability of the truss structures to provide support to the associated components/structures.

A.5.2 SEISMIC CLASSIFICATION OF STRUCTURES AND EQUIPMENT Particular structure and equipment classifications are given below: Buildings and Structures Containment , including all penetrations and airlocks , the concrete shield , the liner , and the interior structures , and the dome truss (for suppo11 of the containment spray piping and the containment air recirculation cooling system (V1'tCC) ductwork)

Containment dome truss structures Spent fuel pool Control room Diesel generator room Pumphouse (to the extent that water is always available to the service water pumps) UFSAR2017 Page A.5-4of38 Inse11 Note I at the end of the table , on a later page: Note 1: Seismic Class I III (See Note 1) I I I I The containment dome truss structures were originally construction aids with the function to supp011 the containment dome liner and concrete during construction.

In the as-left configuration , the truss structures are classified as Seismic Class lII structures, as no qualifications or assessments exist for the truss structures to support a higher classification; however the truss structures maintain functions which include providing suppot1 to Seismic Class I systems/components and preventing Seismic WI interaction.

Refer to Section A.5.10 , for further detail.

l_I Sei s mic De si gn F SAR Appendix A.5 AS.9 SEISJ\IIIC ANALYSIS OF THE DIESEL GENERATOR BUil..D IN G (DGB) 1he mathematical model of the DGB consisted of several stick element s representing the reinforced concrete s h ear wal ls wi th nodes at each floor level. Eac h of the s e nodes was connected by rigid links , repre s en ti ng the rigid diaplu-agm action of the floor s fa b. The soil-structure interaction wa s accounted for b y usin g six soil springs (tfilee transla t ions and three rotat i on s in a Cartesian system), attached to the rigid foundation 111at. The Housner horizonta l des i gn spectra with a pe.ak ground acceleratio n of 0.06g for an opera t ing basi s earthquake a n d 0.12g for a safe sh u tdown earthquake were used as grou n d input mot i o ns. The vertical component of grou n d accelera t ion wa s 2/3 of the magnitude of t11e horizontal component.

The respo ns es (deflect i o ns , moments, s hear s , etc.) offhe buildin were obtained tlu-ough the response spectnun met h od using one set of soil spring v a l ue s. Inseit new A.5 .1 o Response spectra curve s fi equipment located in the DGB were obtained tlu-ough time histo ry analysis. TI1e analy s is a rt ed wi th t h e design earthquake tinie histories inp u t at the bottom of the matl1ematic mode th e DGB. TI1e t ime hi st ories for the three directions of motio n (n"io horizontal an ne v ertica l), at each floor were then obtained as a result of the analy s is. B y appl y ing :s e floor tinie histories to a single-degree-of-freedom o s cillator , response s pectra ere obtained for each of th e floors of the DGB. (Reference 16 and Reference

17) AS.1 0 REFERENCES L NRC Safety E v aluation dated September 30 , 1983 , Amendment No. 75 to Facility OperatingLicense N o. DPR-24. 2. WE Letter to NRC , VPNPD-91-112 , " Status U pdate Electrical Distribution System Function.al Inspection Point Beach Nuclear Plant Units 1 and 2," dated March 28 , 1991. 3. NRC Safety Evaluation Dated September 17 , 1986 , " Safety Evaluation of Topical Report (WCAP-10858)," " AMSAC Generic Design Package." 4. WE Letter to NRC , " Additioml Response To NRC Generic Letter 81-14 , " Point Beach Nuclear Plant , Units 1 and 2 , dated May 4 , 1982. 5. NRC Letter , Status Report and Technical Evaluation Report , " Seismic Qualification Of The Auxiliary Feedwater System, " Point Beach Nuclear Plant Units 1 and 2 , dated January 16 , 1985. 6. NRC Safety Evaluation, Amendment Nos. 45/50 to Facility Operating License Nos. DPR-24 and DPR-27 for tlie Point Beach Nuclear Plant , Units 1 and 2 , " Low Temperature Overpre:ssureMitigating Systems ," dated May 20 , 1980. 7. NRC Letter , " NUREG-0737 Item IlB.l Re.actor Coolant System Vents -Point Beach Nuclear Plant Units 1And2 ," dated September 22 , 1983. 8. WE Letter to NRC , " Reactor Coolant System Gas Vent System Point Beach Nuclear Plan t, Units 1 and 2 ," dated June 18 , 1982. 9. NRC Safety Evaluation, Addendum No. 5 to the Safety Evaluation in the Matter of Poin t Beach Nuclear Plant Units 1 and 2 , dated November 2, 1971. UF SAR 2014 PageA.5-21 o f3 8 S e imlic D e s i gn An.1lysfa FSAR A ppe n di x A.5 10. NRC Safety Evaluation, Amendment Nos. 35/41 to Facility Operating License No s. DPR-24 and DPR-27 for the Point Beach Nuc l ear Plant , Units 1 and 2 , " Modification of The Spent Fuel Storage Pool, " dated April 4 , 1979. 11. V.lE Lettec to NR.C , " Reactor Vessel Overpressurization., " Point Beach Nuc l ear Plant , Units land 2, dated December 20 , 1976. 12. NRC Safety Evaluation., " 1\/Iain Steam Line Break with Continued Feedwater Addition., " Point Beach Nuclear Plant , Units 1 and 2 , dated October 8 , 1982. 13. WE Letter to NR.C , " Final Lettec 81-14 Seismic Qualification of Auxiliary Feedwater System, " Point Beach Nuclear Plant , Units 1 And 2 , dated April 26 , 1985. 14. NRC Safety Evaluation, " Seismic Qualification of the Aurilia.ry Feedwater System, " Point Beach Nuclear Plant Units 1 And 2 , dated September 16 , 1986. 15. WE Lettec to NR.C , .. Seismic Qualification of the Auxiliary Feedwater System, " Point Beach Nuclear Plant Units 1 and 2 , dated December 15 , 1982. 16. VPNPD-93-171 , " Design Summary for the Installation of Two additional Emergency Diesel Generators -Point Beach Nuclear Plants , Unit 1 and 2 ," dated September 24 , 1993 and attached Report REP-0026 , " PBNP Diesel Project Design Submittal, " Revision 0 , dated September 21 , 1993. 17. NRC Safety Evaluation 94-003 , " Emergency Diesel Generator Addition Project , Point Beach Nuclear Plant ," October 24 , 1994. 18. US NRC Generic Letter USI A-46 Resolution., Seismic Evaluation Report, Revision 1, dated January 1996. 19. NRC SE , " Response to Supplement No. 1 to Generic Letter 87-02 for the Point Beach Nuclear Plant , Units. 1 and 2 ," dated July 7 , 1998. 2 0. US NRC SE, ".Amendment No. 24 0 to Renewed Facility Operating License No. D PR-24 and Amendmen t No. 244 t o Renewed Facility Operating License No. DPR-27 , NeA-1Era Energy Point Beac:h, LLC , Point Beach Nuclear Plant , Units 1 and 2 , Docket Nos. 50-266 and 50-301, " datedApril 14 2011. lnseri new references 21 -3 6 UF SAR20 1 4 Page A.5-22 of 38 INSERT A.5.10 STRUCTURAL QUALIFICATION OF THE CONTAINMENT DOME CONSTRUCTION TRUSS STRUCTURES The containment dome construction truss structures were initially erected during site construction to support the containment dome liner steel during the initial 8 in. pour of the containment dome concrete (Section 5.6.1.2).

The truss structures were subsequently lowered away from the dome liner approximately 3 in., when the initial 8 in. of concrete reached design strength, but prior to placing the balance of the dome concrete.

The truss structures have remained in the lowered position since construction within each respective containment building and are used to provide supp01i for:

• the containment spray piping ring headers (including a portion of upstream piping),

• a p01iion of the containment air recirculation cooling system (VNCC) ductwork,

• the post-accident containment ventilation (P ACV) piping,

• and miscellaneous lighting and associated conduits.

The design functions of the truss structures in the post-construction configuration include:

• Maintaining sufficient structural integrity to preclude seismic interaction with Seismic Class I SSCs located adjacent to (i.e., containment liner and building) and below the truss structures before, during, and after a design basis accident or event.

• Providing supp01i without impeding the design functions of the attached Seismic Class I systems: o Containment spray piping o Containment air recirculation cooling system (VNCC) ductwork.

• Providing support to non-seismic equipment (P ACV piping, lighting and associated conduit) to preclude seismic interaction with Seismic Class I SSCs. The original containment dome construction truss structures functioned as a construction aid, and had no seismic classification specified in the FSAR. In the as-left configuration, the truss structures were documented as having been evaluated to demonstrate that the trusses and attached piping would not collapse from applied loading due to the maximum earthquake (Reference 35). The truss structures in the original as-left configuration were implicitly classified as Seismic Class III structures (since no assessments, documentation, or evaluations existed to support higher seismic classification).

The design functions, as listed above, of the truss structures included providing supp01i to Seismic Class I SSCs. In the post construction configuration with the attached supp01ied systems, the truss structures and the attached containment spray ring headers did not meet the code of record acceptance limits as required per Section A. 5 .1, Seismic Design Classifications.

• The containment dome construction truss structures were evaluated (Reference

30) for seismic loading (without any applied pipe support loads) to determine the stiffness and seismic amplification of the containment response spectra at the El. 105 ft. for evaluation of the containment spray piping. A revision to the truss structural analysis (Reference

30) was not perf01med to incorporate all applied loading (pipe supp01i loads, ductwork, etc.) for all postulated loading scenarios (design basis accident and event loads).

To address the attached equipment loading on the truss structures, a structural analysis was pursued. Initial assessment of the truss structures to evaluate for the applied piping and seismic loads determined that the as-built configuration of the truss was not consistent with the designed and previously analyzed truss. The as-found configuration was analyzed and determined to result in stresses that were nonconforming to the original code of record (AISC 6th Ed.) (Reference 36). Subsequent walkdowns and follow-on reviews of photos of the truss structures identified that, in addition to the as-built discrepancies, the clearance between the trusses and the containment liner at certain locations around the containment circumference were postulated to result in contact between the trusses and the containment liner due to thermal expansion during a design basis accident or truss deflection from applied seismic loads during a design basis event. The potential contact load would result in code (see Section 5.1.2.2, Mechanical Design Bases) nonconformances for the containment liner/structure.

Additionally, field walkdowns identified that the anchor bolts for the truss structure bearing housings at several truss locations were positioned at or near the end of the slotted hole. The as-found configuration would limit thermal movement of the truss structures during a design basis accident, leading to additional stresses that did not conform to the design code of record. The legacy nonconformances were identified in both Units 1 and 2. Modifications (Reference

31) (Reference

32) were completed to relocate the anchor bolts centered within the slotted hole of the bearing housing to permit free thermal growth of the truss. To address the nonconformance to site design basis guidelines and codes ofrecord, a informed license amendment request (LAR) was submitted (Reference 21). The basis of the LAR was a risk-informed evaluation (Reference

33) that was performed to determine the risk associated with acceptance of the trusses in the as-built configuration for Unit 2, and an modified configuration for Unit 1 (see discussion below), considering the occurrence of a seismic or thermal event (Reference 21). To support the risk-informed evaluation, a series of engineering calculations were performed to identify the limiting truss members and the associated fragility values for the truss structures for both applied design basis thermal and seismic loading. The structural calculations served a secondary function of demonstrating that the truss structures maintained structural integrity before, during, and after applied loading from a design basis accident or event. The engineering calculations used alternate evaluation methods and acceptance criteria, as the evaluated structures/components did not meet the original design criteria.

The alternate evaluation methods and acceptance criteria, which are different than the original codes of record, formed the basis to suppmi the risk evaluation, and upon regulatory approval, became the codes and guidelines to be used for current and future evaluation of the truss structures.

The following guidance and criteria are applicable to the evaluation of the trusses (Reference 28):

• The ground seismic input is the site specific ground motion response spectra (GMRS) as documented in Reference

34.

• In-structure seismic response spectra is dete1mined through soil-structure interaction (SSI) analysis.

Ground motion time histories shall meet Section 2.4 of ASCE/SEI 43-05 with the limitations identified in NUREG/CR-6926.

• Damping for the truss structures is 7%.Damping for the containment spray piping attached to the truss structures is 4%.AISC N690-1994(R2004), American National Standard Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities, is used as the code for evaluating the truss structural components, using the increased allowable stresses for dead load and seismic load combinations and dead load and thermal load combinations.

o For truss members of the upper and/or lower chord that do not meet the limits of AISC N690-l 994(R2004), the maximum pe1missible strain is limited to 1.5% for combined axial and flexure or flexure only.

• The allowable contact load on the containment liner is based on guidance in ASME B&PV Code,Section III, Division 1, 1983, Appendix F: o The allowable load under seismic or design basis accident loads is the minimum of the load that develops a maximum primary stress intensity of 0.9Su (ultimate strength) and 2/3 of the maximum sustainable load. o Liner integrity for applied cyclic loading is assessed by comparing the accumulation in strains and the change in strains between cycles, in combination with the fatigue curve from Figure 1-9 .1 of ASME Boiler and Pressure Vessel Code,Section III, 1983. o Localized exceedance of pe1missible concrete strain truss contact points, with an allowable limit of 0.003 in/in per ACI 318-63. o The concrete compressive strength is based on the compressive strength from test data as permitted in ACI 318-63. Note: The above criteria are limited in application to the truss structures and adjacent or supported equipment near the truss structures which was used to resolve the nonconformances addressed in Reference

21. All of the equipment supported by the truss structures, such as the containment spray piping, P ACV piping, associated pipe supports, VNCC ductwork, lighting, and associated conduits shall use the design code of record for evaluation.

The above criteria was used to calculate the seismic fragility and a thermal probability of failure for the trusses and attached components for use in the probabilistic risk assessment (Reference 33), based on which the trusses were accepted.

The analyses evaluated Unit 1 assuming completion of a modification to trim the truss structures at six designated locations to increase clearance between the trusses and the containment liner, and evaluated Unit 2 in the found/post-construction condition with no modifications pending. Moving forward, future evaluations/modifications for the trusses and/or attached components shall follow the above criteria.

As long as the above criteria are met, the probabilistic risk assessment remains valid. Changes to the above criteria that reduce the seismic fragility and thermal probability of failure shall require a probabilistic risk assessment.

To meet the new acceptance criteria (Reference 21), the following modifications (Reference

23) were completed:

• The modification of one containment spray pipe support in Unit 1, SI-301R-1-H202.

• A modification to the Unit 1 truss structures to improve clearances between the construction trusses and the containment liner at six truss locations.

The clearance modification to Unit 1 results in stress reduction and a configuration bounded by the Unit 2 thermal fragility analysis.

The suppo1ting calculations demonstrated that following completion of the Unit 1 truss and containment spray pipe support modifications, structural integrity, i.e., the ability to support cairied loads and not interfere with supported equipment functions was maintained in both Units with adequate margin. The risk informed resolution of the nonconfmmances included implementation of new thermal and seismic limits to initiate assessment of the construction trusses, equipment supported by the trusses, and the containment/containment liner, as necessary, for any event exceeding the specified limits. Any event reaching or exceeding the specified limit(s) requires Unit shutdown and inspection and/or analysis to ensure the affected structures/components can withstand a subsequent design basis accident without adversely impacting the SSCs' design function(s).

THERMAL LIMIT VALUE Unit 1 maximum containment atmospheric temperature 227°F (Reference

26) Unit 2 maximum containment atmospheric temperature 236°F (Reference

26) SEISMIC LIMIT VALUE Horizontal peak ground acceleration 0.05g (Reference

27) Vertical peak ground acceleration 0.04g (Reference

27) The risk-informed resolution of the code nonconformances was approved by License Amendment Nos. _and_ dated __ (Reference 22). The truss structures continue to remain classified as Seismic Class III structures as no assessments, documentation, or evaluations have been developed to support higher seismic classification as the required input necessary for higher seismic qualification does not exist and cannot be replicated (i.e., material test repmis, weld inspection, weld procedures and qualifications, final as-built dimension validation, etc.). The truss structures' design functions remain unchanged following the resolution of the nonconformances, which includes continuing to provide support to Seismic Class I SSCs.

New References for Section A.5.11: 21. License Amendment Request 278, Risk-Informed Approach to Resolve Construction Truss Design Code Non-conformances , dated March 31 , 2017 22. US NRC Safety Evaluation , "Amendment No. XX:X to Renewed Facility Operating License No. DPR-24 and Amendment No. XX:X 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 , dated [Month] [Day], [Year] 23. EC 282198, Modification to Unit 1 Dome Truss to Increase Available Liner Gap; includes SI-301R-1-H202 modification

24. Not Used. 25. Not Used. 26. Calculation 11Q0060-C-036, The1mal Evaluation of Units 1 and 2 Containment Dome Trusses for Lesser Events 27. Calculation 11Q0060-C-037, Seismic Evaluation of Units 1and2 Containment Dome Trusses for Lesser Events 28. Calculation 11Q0060-RPT-002, Methodology and Criteria to Dete1mine the Strength Capacity of the Point Beach Nuclear Plant Containment Dome Trusses and Attached/

Adjacent Components in Support of a Risk-Informed License Amendment Request 29. Not Used. 30. Calculation 6904-15-TR , Calculation for Adequacy of Containment Dome Construction Truss 31. EC 281440, Ul C Dome Truss Bearing Box Bolted Connection Work 32. EC 281403, U2C Dome Truss Bearing Box Bolted Connection Work 33. Probabilistic Risk Assessment Evaluation PBN-BFJR-17-019 , Rev. 1, Point Beach Units 1 & 2 Construction Truss PRA Evaluation . . 34. Con-espondence NRC 2014-0024, Dated March 31 , 2014, Subj. NextEra Energy Point Beach , LLC, Seismic Hazard and Screening Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(£) Regarding Recommendation 2.1 of the Te1m Task Force Review of Insights from the Fukushima Dai-ichi Accident

35. Correspondence PBB-W-3162, Dated July 15, 1970, Subj. Containment Dome Trusses Seismic Analysis 36. AR 01750123, Unit 1 & 2 Containment Dome Truss Analysis Preliminary Results