ML031630824
| ML031630824 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/19/2003 |
| From: | Salas P Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 24370-TR-C-003-A, Rev 1 | |
| Download: ML031630824 (115) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 May 19, 2003 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D. C. 20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket No.
50-327 SEQUOYAH NUCLEAR PLANT -
ACCEPTED VERSION OF TOPICAL REPORT NO. 24370-TR-C-003-A, STEAM GENERATOR COMPARTMENT ROOF MODIFICATION"
Reference:
NRC letter to TVA dated April 18, 2003, Safety Evaluation of Topical Report No. 24370-TR-C-003, "Steam Generator Compartment Roof Modification, Revision 1" (TAC NO. MB5387)
The purpose of this submittal is to provide the accepted version of the subject topical report as requested in the reference letter.
The accepted topical report now includes a copy of the reference letter, historical review information, and any original report pages that were replaced.
The "-A" has been included in the topical report number to designate NRC acceptance.
There are no commitments contained in this letter.
This letter is being sent in accordance with NRC RIS 2001-05.
If you have any questions about this change, please telephone me at (423) 843-7170 or J. D. Smith at (423) 843-6672.
and Industry Affairs Manager Enclosure Pnted r necycILd paper
,-Do -) C)
QA Record 888 030515 SEQUOYAH UNIT 1 STEAM GENERATOR REPLACEMENT STEAM GENERATOR COMPARTMENT ROOF MODIFICATION
-- ITOPICAL REPORT
'II PROJECT Sequoyah DISCIPLINE N
CONTRACT 99N5B-253631 UNIT I
DESC. Topical Report - SG Enclosure Roof Modification DWG/DOC NO.
24370-TR-C-003-A SHEET OF REV.
01 DATE 05/15/03 ECN/DCN FILE N2N-059 1
02/07/03 Revised to Address NRC RAls SWK MRA JVS 0
02/19/02 Issued for TVA use SWK JC JVS REV.
DATE REASON FOR REVISION BY EGS PE JOB NO.: 24370 DOCUMENT NO.:
24370-TR-C-003-A RIMS, WTC A-K JI 801 APPROVED Ths pDtovl does not f64IsS. th.
Contrsator from any Part f his e-gponsIbIlIty lot the orretnles of dsign. OWSa. snd imslons.
Letter No. TVBEC-0415 Vs May 15, 2003 T112"tVALLTY AUTHOWITY SoUp N
OY P.O. Tiudel
SEQUOYAH UNIT 1 STEAM GENERATOR REPLACEMENT STEAM GENERATOR COMPARTMENT ROOF MODIFICATION TOPICAL REPORT
Table of Contents NRC Acceptance Letter and Safety Evaluation Report for Topical Report 24370-TR-C-003,
'Steam Generator Compartment Roof Modification, Revision 1"................................ 2a-2f 1.0 Abstract....................................................
3 2.0 Introduction....................................................
3 3.0 Objectives....................................................
8 4.0 Regulatory RequirementslCriteria for Ice Condenser Divider Barriers............................. 8 4.1 SRP Section 3.8.3 - Concrete and Steel Internal Structures of Steel or Concrete Containments...........................................
8 4.2 SRP Section 6.2.1.2 - Subcompartment Analysis..........................................
I 1 5.0 Description of Concrete Work to be Performed..........................................
11 6.0 Description of Existing Design Basis and Original Analyses.......................................... 12 7.0 Description of Modification to the Structure and New Analyses.25 8.0 Results of New Analyses..............................................
34 9.0 Summary and Conclusions...............................................
36 10.0 References..............................................
37 Appendices A
No Significant Hazards Consideration Determination..............................................
39 B
Historical Review Information..............................................
43a-43bl Tables 6-1 (UFSAR Table 3.8.3-1) Loading Combinations and Allowable Stresses for the Interior Concrete Structure...............................................
15 6-2 (UFSAR Table 3.8.3-2) Loading Combinations and Load Factors.................................... 16 6-3 (UFSAR Table 3.8.3-6) Original Design Stress Margin Table 3.8.3-1 Criteria Versus Table 3.8.3-2 Criteria (4).
17 7-1 (Table CC-3230-1 from ASME Section III, Division 2,1975) Load Combinations and Load Factors...................................................................
29 7-2 Loading Combinations, Load Factors and Allowable Stresses for SG Compartment Roof Modification (5)(6)...................................................................
30 9-1 Differences Between Original and New Steam Generator Compartment Analyses.......... 36 Figures 2-1 Equipment - Reactor Building (UFSAR Figure 1.2.3-11).................................................... 5 2-2 Equipment - Reactor Building (UFSAR Figure 1.2.3-12).................................................... 6 2-3 Equipmnent - Reactor Building (UFSAR Figure 1.2.3-13).................................................... 7 6-2 Concrete Steam Generator and Pressurizer Compartment - Reinforcement.................... 19 6-3 Concrete Steam Generator and Pressurizer Compartment - Reinforcement.................... 20 6-4 Concrete Crane Wall Outline...................................................................
21 6-5 Temperature Gradient (UFSAR Figure 3.8.3-2)................................................................ 22 6-6 Steam Generators 1 and 4 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-1)....................................................................
23 6-7 Steam Generators 2 and 3 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-2).................................................................... 24 7-1 Sensitivity of Peak Compression Pressure to Deck Bypass (UFSAR Figure 6.2.1-22)..... 31 7-2 Steam Generator Compartment Roof Connection Frame Layout and Connection Details 32 7-3 Finite Element Model SGE1" and SGE2' and Element Groups and Global Coordinate Systems (Reference 6)....................................................................
33 8-1 Areas of Critical Stresses...................................................................
36 Page 2 of 43
I UjJILUI npuivI L
I F-u uI-v.-vvo-r-NRC Acceptance Letter and Safety Evaluation Report for Topical Report 24370-TR-C-003, "Steam Generator Compartment Roof Modification, Revision 1" Page 2a of 43
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~ r1'.vnQ%ULH1A I Jn T %oVIVIIVIlZblUN WASHINGTON, D.C. 20555-0001 kpll 18, 2)
Mr. J. A. Scalice Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801
SUBJECT:
SAFETY EVALUATION OF TVA TOPICAL REPORT NO. 24370-TR-C-003, "STEAM GENERATOR COMPARTMENT ROOF MODIFICATION, REVISION "
Dear Mr. Scalice:
On March 28, 2002, the Tennessee Valley Authority (TVA, the licensee) requested the U. S. Nuclear Regulatory Commission (NRC) staff's approval of Topical Report No. 24370-TR-C-003, "Steam Generator Compartment Roof Modification," proposing an alternative methodology for the reconstruction of the steam generator compartment concrete roof. The staff rejected the original proposed methodology in a letter dated January 10, 2003.
Subsequently, on February 14, 2003, TVA resubmitted Topical Report No. 24370-TR-C-003, "Steam Generator Compartment Roof Modification, Revision 1" (Topical Rev. 1) for Sequoyah Nuclear Plant Unit 1. Topical Rev. 1 contains a new design and analysis of the reconstruction method for the Unit 1 steam generator compartment roof modification.
The enclosed NRC safety evaluation contains the NRC staff's review. The NRC staff has reviewed the submittal and determined that the altemative method contained in Topical Rev. 1 is acceptable.
In accordance with the guidance provided on the NRC web site, we request that TVA publish an accepted version of this topical report within 3 months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed safety evaluation between the title page and the abstract. It must be well indexed such that information is readily located. Also, it must contain in appendices historical review information, the questions and accepted responses, and original report pages that were replaced. The accepted version shall include an -A' (designated accepted) following the report identification symbol.
If the NRC's criteria or regulations change so that the conclusions in this lefter are invalidated, thus making the topical report unacceptable, TVA will be expected to revise and resubmit its
Mr. J. A. Scalice respective documentation, or submit justification for the continued applicability of the topical report without revision of the respective documentation.
If you have any questions conceming this matter, please contact Eva Brown at (301) 415-2315.
Sincerely, Michael L Marshall, Jr., Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-327
Enclosure:
Safety Evaluation cc w/encl: See next page UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 255-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR SAFETY EVALUATION OF TOPICAL REPORT NO. 24370-TR-C-003, "STEAM GENERATOR COMPARTMENT ROOF MODIFICATION, REVISION 1" TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-327
1.0 INTRODUCTION
In a letter dated March 28, 2002, Tennessee Valley Authority (TVA, the licensee) requested approval of Topical Report No. 24370-TR-C-003, Steam Generator Compartment Roof Modification," for use at Sequoyah Nuclear Plant, Unit 1 (SQN1). This submittal described an alternate methodology for the reconstruction of the steam generator (SG) compartment concrete roof. The staff rejected the original proposed methodology by the licensee in a letter dated January 10, 2003. Subsequently, in a letter dated February 14, 2003, the licensee submitted Revision 1 of the subject topical report for U. S. Nuclear Regulatory Commission (NRC) review and approval. Topical Report No. 24370-TR-C-003, Steam Generator Compartment Roof Modification, Revision 1" (the Topical Rev. 1) contained a new design and analysis of the reconstruction method for the Unit 1 SG compartment roof modification.
2.0 DESIGN STANDARDS NUREG-0800, Revision 1, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), Section 3.8.3, outlines the standards for use by the NRC staff during the review of concrete containment intemal structures. The SP compartment roof or divider barrier is designed in the event of a loss-of-coolant-accident to contain the steam released from the reactor coolant system, and to channel the steam through venting doors to the ice-condenser, temporarily serving as a pressure-retaining envelope.
SRP 3.8.3, Section 11.3.d, indicates that the loads and load combinations for the divider barrier are required to be evaluated against Article CC-3000 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Code, 1975 Edition (the Code)Section III, Division 2 with some exceptions. The design and analysis of the modification are contained in Section 111, Division 2 of the Code with the specified limits for stresses and strains requirements being contained in Subsection CC-3430.
Enclosure 3.0 EVALUATION 3.1 Descridtion of Compartment Roof Modification The four SGs of the SQN1 will be replaced during the spring of 2003. To support the replacement of the old.SGs with the replacement SGs, access openings wilJibe created in.the roof of the SG compartments. Each access opening will be sized and cut to allow the removal and replacement of the SG in the compartment.
To provide an access opening for SG replacement, a section of the compartment concrete roof over each SG will have to be cut out. Cutting of the concrete will be accomplished by first core-boring holes around the perimeter of the cut, then using wire saws to cut straight lines between the cores. The cores also serve as the bolt holes for the through-bolts used to connect the concrete section back to the existing compartment roof concrete. After removal, the edges of the concrete section will be bush-hammered to provide a gap that ranges from 3/4 inch to 1-1/4 inches between the cut-out portion of the concrete and the existing compartment roof concrete.
The cut-out portion of the concrete will be re-attached to the existing compartment roof concrete once the replacement SG and associated piping are placed inside the compartment.
A top and bottom steel connecting frame will sandwich the cut-out portion of the concrete. The steel frames will be through-bolted by four 2-inch diameter threaded rods and span over the existing compartment roof concrete. The steel frames will also be through-bolted by six 2-1/2 inch and eighteen 2-inch diameter threaded rods along the perimeter of the cut line. The threaded rods will be pretensioned to a stress level of 70 percent of its yield stress.
Approximately 30 tapered steel shim sets will be installed along the perimeter of the cut line.
Each tapered shim set will comprise a tapered shim attached to the sectional surface of the cut-out portion of the concrete with anchor bolts and a loose tapered shim that will be driven into the gap between the fixed tapered shim and the existing compartment roof concrete. The loose tapered shim will be welded to the fixed shim to prevent movement. The bolt holes and the remaining annular space will be grouted using nonshrink grout.
3.2 Evaluation of Proposed Modification The licensee analyzed the roof of the SG compartments using a finite element computer code (STRUDL). Manual calculations were performed at various locations to confirm results obtained from the computer analysis. The analysis results indicated that the maximum concrete and rebar stresses in the modified roof are within the allowable stress limits for normal and abnormaVextreme environmental conditions as specified in Secfion CC-3000 of Section 111, Division 2 of the Code. The maximum calculated bending stress in the connecting frame beams and the maximum calculated bearing stress on concrete and the tapered steel shims were determined to be below the allowable limits.
Vertical loads generated by the vertical seismic inertia of the cut-out portion of the concrete roof and the maximum design basis accident to the existing compartment roof concrete. This force transfer would occur because the cut-out portion of the concrete roof is not only sandwiched between two steel frames, but, also through-bolted to the frames that span over the existing compartment roof concrete and are connected to it by through-bolts along the perimeter of the cut line. Horizontal loads generated by the horizontal seismic inertia of the cut-out portion of the concrete roof would be transferred through steel shims to the existing compartment roof concrete. The steel frames, in conjunction with the through-bolts and the steel shims, eliminate any significant movement between the cut-out portion of the concrete roof and the existing compartment roof concrete and provide a positive connection between the two. The NRC staff finds that the proposed modification method provides.a positive connection between the cut-out portion of the concrete roof and the existing compartment roof concrete and is, therefore, reasonable and acceptable.
The licensee used STRUDL computer code to analyze the structure of the proposed roof modification and verified the adequacy of the computer results by manual calculations at several locations. The staff finds that the manual verffication adds confidence to the analysis results. The acceptance criteria were based on Secton III, Division 2 of the Code requirements, which are acceptable to the staff. The analysis results indicate that the stresses in concrete and steel of the roof modification structure, under all loading combinations prescribed by the Secton I1, Division 2 of the Code, are within the allowable specified stress limits. The NRC finds that the licensee has used appropriate analysis methods and criteria to analyze the modified roof compartment, and that the analysis results indicate conformance with the design code requirements.
4.0 CONCLUSION
The NRC staff has reviewed the proposed SG compartment roof modification method for SQN1. Based on the information provided by the licensee, the NRC staff has concluded that the load and load combinations proposed are conservative, the design and analysis were completed consistent with appropriate industry standards, and the allowable stresses and strains are reasonable and acceptable. Therefore, the proposed modification satisfies the design requirements at the SQN1.
Principal Contributor: John S. Ma, NRR Date:
Ajl 18, 23
I uPIU-a reIu L Q,3I v I-I-s, 1.0 Abstract The four steam generators of the Sequoyah Nuclear Plant Unit 1 will be replaced during the spring of 2003. To support the replacement of the old steam generators (OSGs) with the replacement steam generators (RSGs), access openings will be created in the roof of the steam generator (SG) compartments inside containment. An appropriately sized access opening will be made in each SG compartment roof by cutting out a section of concrete from the roof of the compartments using wire saws. Upon completion of installation of the RSGs, the original cut concrete section (plug) of the SG compartment roof will be reattached to the respective compartment roof by means of through-bolted connections, comprised of steel connection frames and threaded rods. The plug will be attached to the top and bottom connection frames using four 2-inch diameter threaded rods that are installed in core bore holes through the plug. The top and bottom connection frames will clamp the concrete plug to the complimentary portion of the SG compartment using six 2-1/2 inch and eighteen 2-inch diameter threaded rods. The threaded rods are installed in the core bore holes located around the perimeter of the concrete plug and will be pre-tensioned. A series of steel shims will be driven into the annular space (created at the cut line) and mechanically locked into place. The annular space will be grouted.
The original design of the SG compartment was based in part on the load combinations defined in Table 3.8.3-2 of the UFSAR. This UFSAR table is based on Table CC-3200-1 of the Proposed ASME Section 1II, Division 2, 1973, Proposed Standard Code for Concrete Reactor Vessels and Containments, Section CC-3000 which was issued in 1973 (the time of original design) by the ACI-ASME Committee on Concrete Pressure Components for Nuclear Service, for trial use and comment. The purpose of this topical report is to provide the technical basis for use of the slightly modified load combinations and allowable stresses in the adopted 1975 edition of ASME Section 1I1, Division 2, instead of those described in the UFSAR. Analyses performed using the adopted ASME load combinations have shown that the modified SG compartment roof design will not exceed allowable stresses in the concrete, rebar and structural steel when subjected to the design basis differential pressure of 24 psi combined with the other design basis loads such as seismic, pipe thrust, dead load and live load. This design differential pressure is approximately 23% higher than the maximum compartment accident pressure differential of 19.52 psi.
2.0 Introduction The steam generator compartments are designed and constructed as cast in-place reinforced concrete structures. As indicated in UFSAR Section 3.8.3.6.1, the minimum compressive strength of the containment interior concrete structures is 5000 psi.
UFSAR Section 3.8.3.1.7 describes the steam generator compartments. Two double-compartment structures house the four steam generators in pairs on opposite sides of the containment. For each pair of steam generators, divider barrier walls exist around the two steam generators and are capped with a three-foot thick concrete roof spanning over the steam generators from the crane wall. A wall between each pair of steam generators extends from the divider walls to the crane wall, completing the double compartment. The center wall does not extend up to the concrete roof. This area above l the wall, except for the portions occupied by the main steam pipe restraint beam, reduces the compartment pressure buildup in a single compartment by venting the steain to the other compartment. These features are depicted on UFSAR Figures 1.2.3-11, 1.2.3-12, and 1.2.3-13 (provided as Figures 2-1, 2-2, and 2-3, respectively).
Page 3 of 43
- -I-.-.---.
The steam generator compartments form part of the interior concrete structure that is referred to as the divider barrier. UFSAR Section 3.8.3.1.1 defines the divider barrier as that part of the interior structure that separates the upper containment from the lower containment. This barrier forces steam that is released from a LOCW DBA to pass through the ice condenser. The failure of any part of the divider barrier is considered critical since it would allow LOCA/DBA steam to bypass the ice condenser, thereby increasing the pressure within the primary containment. The original design loads for the compartment concrete were based on preliminary accident pressurization calculations. Conservative design basis loads were used in the original design to bound potential changes between the preliminary and the final pressurization analysis results.
UFSAR Section 3.8.3.2 details the codes and standards to which the intemal concrete structures were designed. The load combinations and allowable stresses for the intemal concrete structures including the divider barrier are detailed in UFSAR Tables 3.8.3-1 and 3.8.3-2 (provided as Tables 6-1 and 6-2, respectively).
There are no Technical Specifications (TSs) associated specifically with the steam generator compartments. However, there are TSs associated with other portions of the divider barrier. TSs 314.6.5.3, 314.6.5.5, and 314.6.5.9 address the ice condenser doors, divider barrier personnel access doors and equipment hatches, and divider barrier seal, respectively. The planned changes to the steam generator compartment roof will restore the leaktightness of the roof and will not affect the ice condenser doors, divider barrier personnel access doors and equipment hatches, or divider barrier seal. Therefore, the TSs will not be affected by the planned changes to the steam generator compartment roof portion of the divider barrier.
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Topical Report 24370-TR-C-003-A Figure 2 Equipment-Reactor Bullding (UFSAR Figure 1.2.3-11)
Page 5 of 43
Topical Report 24370-TR-C-003-A Figure 2 Equipment - Reactor Building (UFSAR Fgura.2.3-12)
Page 6 of 43
Topical Report 24370-TR-C-003-A Flgure 2 Equlpment-Reactor Building (UFSAR Flgure 1.2.3-13)
Page 7 of 43
Topical Report 24370-TH-G-003-A 3.0 Objectives To describe the current steam generator compartment roof design and proposed modification.
To present data that supports and justifies the reinstallation of the cut steam generator compartment roof concrete sections using frames installed on the top and bottom of the section and then through-bolted together.
To support a license amendment for using load combinations and allowables for reinforced concrete provided in "adopted" ASME Section 1I1, Division 2, 1975 instead of the load combinations provided in Proposed" ASME Section III, Division 2, 1973.
4.0.
Regulatory Requirements/Criteria for Ice Condenser Divider Barriers Detailed below are regulatory requirements/criteria that are relevant to the design of the divider barrier portion of internal structures in an ice condenser containment. Since the SG compartment roof is part of the divider barrier, the planned modification to the roof must conform to the requirements/criteria below. Following each requirement/criteria is an italicized discussion of how the requirement/criteria is met and/or where the requirement/criteria is addressed within this topical report.
4.1 SRP Section 3.8.3 - Concrete and Steel Internal Structures of Steel or Concrete Containments Standard Review Plan (SRP) 3.8.3 details the information required for NRC review of containment internal structures and the criteria for NRC acceptance of these structures.
This review is performed to assure conformance with the requirements of 1 OCFR50.55a and 10CFR50, Appendix A, General Design Criteria (GDC) 1, 2,4,5, and 50. The parts of these regulations that are relevant to the divider barrier design are:
1 ) 1 OCFR50.55a and GDC 1 as they relate to the divider barrier being designed, fabricated, executed, and tested to quality standards commensurate with the importance of the safety function to be performed.
The quality standards used in the design, fabrication, execution, and testing of the modified divider barrier are the same or equivalent to those used for the original divider barrier.
- 2) GDC 2 as it relates to the design of the divider barrier being capable to withstand the most severe earthquake and appropriate combination of all loads.
The modified SG compartment roof has been designed for the same loads and load combinations as the original design (described in Section 6.0), except as noted in Section 7.0. The results described in Section 8.0 show that it is capable of withstanding the most severe earthquake loads and the appropriate combination of other loads.
- 3) GDC 4 as it relates to the divider barrier being capable of withstanding the dynamic effects of equipment failures including missiles, pipe whips and blowdown loads associated with the loss of coolant accidents.
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As described in Sections 7.0 and 8.0, the modified SG compartment design has been evaluated for the dynamic effects of pipe whip andjet impingement loads following a pipe break inside the SG compartment.
- 4) GDC 5 as it relates to the sharing of structures important to safety.
The divider barrier is not a shared structure. Therefore, conformance to GDC 5 is not applicable for the modified SG compartment.
- 5) GDC 50 as it relates to the divider barrier being designed with sufficient margin of safety to accommodate appropriate design loads.
As described in Sections 7.0 and 8.0, the modified SG compartment design is capable of withstanding the same design pressure as the original SG compartment design without exceeding allowable stresses in the concrete, rebar and structural steel. This design pressure is 23% greater than the maximum calculatedpost-LOCA differential pressure. Since the design pressure and the maximum calculated accident pressure have not changed, there is no reduction in the margin of safety for the modified SG compartment design.
The descriptive information provided is considered acceptable if it meets the minimum requirements set forth in Section 3.8.3.1 of NRC Regulatory Guide (RG) 1.70. This RG indicates that the descriptive information relevant to the divider barrier that should be provided includes plan and section views to define the primary structural aspects and elements relied upon to perform the safety-related function of the divider barrier.
General arrangement diagrams and the principal features of the divider barrier should be described.
A description of the revised SG compartment roof design is provided in Section 7.0.
Figure 7-2 provides details for the frames to be installed on the top and the bottom of the compartment concrete section and the layout of the connection through-bolts. Other aspects of the divider barrier design will remain as described in the Sequoyah UFSAR.
An update to the UFSAR will be prepared to reflect the revised Unit 1 SG compartment roof design.
The design, materials, fabrication, erection, inspection, testing, and in-service surveillance of the divider barrier are covered by the following codes, standards, and regulatory guides:
1 ) ACI-349 As indicated in Section 1. 1 of Part 1 of ACI-349, structures covered by ASME Section I, Division 2 are specifically excluded from the requirements of this standard. As discussed in Section 7.0, the modified SG compartment roof design conforms to ASME Section I, Division 2. Therefore, this standard is not applicable to the modified SG compartment roof design.
- 2) ASME Section III, Division 2 Conformance of the original design of the SG compartment roofs to the ASME Code is discussed in Section 6.0. As detailed in Section 7.0, the reinforced concrete part of the modified SG compartment roof design is consistent with the adopted edition of Page 9 of 43
Topical Report 2437U--I K-L;-UU3-A the ASME Code. The basis andjustification for use of the later edition of the Code is also provided in Section 7.0.
- 3) ANSI N45.2.5, "Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants".
Addressed under the response to RG 1.94 below.
- 4) Regulatory Guide 1.94, Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants" RG 194 endorses ANSI N45.2.5-74, but specifies additional requirements related to use of other codes and standards, RG 1.55, concrete consolidation, and rebar splice welding. The TVA Nuclear Quality Assurance Plan (NOAP) (Reference 15) follows this regulatory guide, but also provides altematives to the regulatory guide guidance.
The installation, inspection, and testing activities associated with the through-bolted connection frame modification to the SG compartment roofs will conformn to the RG 1.94 guidance or the altematives allowed by the TVA NOAP.
- 5) Regulatory Guide 1.142, Safety-Related Concrete Structures for Nuclear Power Plants" RG 1.142 endorses ACI 349-76. As discussed in Section 7.0, the modified SG compartment roof design conforms to ASME Section III, Division 2 (1975). As such, the modified SG compartment roof design is not required to be evaluated against the requirements of RG 1.142 or ACI 349-76.
The divider barrier design is reviewed to determine if the loads and load combinations used meet the acceptance criteria. For concrete pressure-resisting portions of the divider barrier, the loads and load combinations of Article CC-3000 of ASME Section 1I1, Division 2 Code apply.
As described in Section 7.0, the load combinations of Table CC-3230-1 of Article CC-3000 of ASME Section Ill, Division 2, 1975 were used in the evaluation of the modified SG compartment roof design.
The design and analysis procedures utilized for the divider barrier are acceptable if they are in accordance with ACI 318.
As described in Section 6.0, the original SG compartment structural design is in compliance with a combination of AC) 318 and the Proposed ASME Section 111, Division 2, 1973. Section 7.0 describes how the modified SG compartment design complies with ASME Section 111, Division 2, 1975 (ACI 359-74).
The structural acceptance criteria for the divider barrier are acceptable if the specified stress and strain limits are in accordance with Subsection CC-3430 of ASME Section III, Division 2. The 33-1/3% increase in allowable stresses is only permitted for temperature loads and not for OBE seismic or wind loads.
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Topical Report 24370-TR-C-003-A As described in Section 8.0, the stresses in the reinforced concrete of the modified SG compartment roof stresses under the load combinations defined in Table CC-3230-1 of ASME Section I, Division 2, 1975 are less than or equal to the stress allowables defined in Section CC-3400 of ASME Section I1, Division 2, 1975. The 33-1/3%
increase in allowable stresses was only used for temperature loads. The structural steel through-bolted connection frames are designed in accordance with Reference 3.
The specified materials of construction and quality control programs for the divider barrier are reviewed. Information on the materials used and the extent of compliance with ANSI N45.2.5 should be provided to support this review. Information on special, new, or unique construction techniques should also be provided in order to assess their effects on the structural integrity of the completed divider barrier.
The materials used in the modified SG compartment design are detailed in Section 7.0.
Installation, inspection and testing of the modified SG compartment roof will conforn to the quality assurance requirements of ANSI N45.2.5. Other than tensioning or preloading the threaded rods, there are no special, new, or unique construction techniques that will be used during installation of the modified SG compartment roof.
4.2 SRP Section 6.2.1.2 - Subcompartment Analysis SRP 6.2.1.2 details the information required for NRC review of the design differential pressure analyses for containment subcompartments. This review is performed to assure conformance with the requirements of 1 OCFR50, Appendix A, GDC 4 and 50.
The parts of these regulations that are relevant to the divider barrier design are:
- 1) GDC 4 as it relates to the ability of the divider barrier to accommodate the dynamic effects of missiles, pipe whipping, and discharging fluids that may occur during normal operations or during an accident.
As described in Sections 7.0 and 8.0, the modified SG compartment design has been evaluated for the dynamic effects of pipe whip andjet impingement loads following a pipe break inside the SG compartment.
- 2) GDC 50 as it relates to the divider barrier being designed with sufficient margin to prevent fracture of the barrier due to pressure differential across the barrier.
As described in Sections 7.0 and 8.0, the modified SG compartment design is capable of withstanding the same design pressure as the original SG compartment design without exceeding the allowable stresses in the concrete, rebar or structural steel. This design pressure is 23% greater than the maximum calculatedpost-LOCA differential pressure.
5.0 Description of Concrete Work to be Performed The modification of the steam generator compartment roof will first entail cutting out a section of the concrete roof over each steam generator. Cutting of the concrete will be accomplished by first core-boring holes a.round the perimeter of the cut, then using wire saws to cut the straight lines between the cores. The cores also serve as the bolt holes for the through-bolts used to connect the concrete section back to the structure. After removal, the edges of the concrete section will be bush-hammered to provide an annular gap of about 1" upon reinstallation of the concrete section. Each concrete section will be Page 11 of 43
Topical Report 24370-TR-C-003-A sized to allow the removal and replacement of the stean generator in the compartment.
The concrete section will be re-installed once the RSG and associated piping are placed inside the compartment. Restoration of the SG compartments will involve re-attaching the cut out concrete sections to the existing structure using a top and bottom frame sandwiching the cut out concrete sections and connecting the frames with through-bolted threaded rods around the perimeter of the cut. Tapered steel shims will be placed in the annular gap between the concrete sections and the bolt holes and annular space will be grouted using non-shrink grout. Additional details of the through-bolted connection frame design and the capability of the non-shrink grout to limit bypass leakage through the divider barrier is provided in Section 7.0.
The steam generator compartments have been re-evaluated, with specific focus on the modified roof, for the effects on structural response and found to be acceptable. The through-bolted connection frames and the tapered steel shims have been designed to be adequate for the applicable design loadings. Details of these evaluations are provided in Section 7.0. The design of the repaired steam generator compartments is in compliance with the requirements of Reference 2.
6.0 Description of Existing Design Basis and Original Analyses The original design bases of the concrete internal structures, which includes the SG compartments, is discussed in detail in Section 3.8.3 of the UFSAR and Section 2.9 of Reference 2. UFSAR Section 3.8.3.2 states that the structural design of the interior concrete structures is in compliance with the American Concrete Institute (ACI) 318-63 Building Code Working Stress Design Requirements for load combinations shown in UFSAR Table 3.8.3-1 (provided as Table 6-1), including LOCA calculated pressures with moisture entrainment received from the NSSS contractor, or the ACI-ASME (ACI 359)
Article CC3000 document, "Proposed Standard Code for Concrete Reactor Vessels and Containments" (Proposed ASME Section I, Division 2, 1973), and ACI 318-71 for the load combinations shown in Table 3.8.3-2 (provided as Table 6-2), including LOCA calculated pressure. Section 3.8.3.2 of the UFSAR also states that the design and construction of the interio concrete structures is based on the appropriate sections of NRC Standard Review Plan 6.2.1.2, "Subcompartment Analysis".
The original design loads for the SG compartment concrete were based on preliminary accident pressurization calculations. Because of the uncertainties associated with these preliminary accident analyses, conservative design basis loads were used in the original design to bound potential changes between the preliminary and the final pressurization analysis results. The preliminary accident pressurization loads were higher than the final accident loads, which resulted in a conservative SG compartment design.
The maximum differential pressure used in the original design was 21.3 psi which is a 25% increase over the design basis accident (DBA) differential pressure of -17 psi (Reference 5) for the SG compartment provided by Westinghouse (i.e., 1.25 x 17 psi).
The original design was based on loads, load combinations and allowable stresses documented in Table 3.8.3-1 of the UFSAR (provided as Table 6-1).
As detailed in UFSAR Section 3.8.3.4.1, each component of the interior concrete structure was evaluated individually. Its boundary conditions and degrees of fixity were established by comparative stiffness; loads were applied, and moments, shears, and direct loads determined by either moment distribution or finite element methods of analysis. UFSAR Section 3.8.3.4.1 also states that reinforcing steel was proportioned Page 12 of 43
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i-for the component sections in accordance with UFSAR Tables 3.8.3-1 or 3.8.3-2 and the ultimate strength provisions of ACI 318-71 Building Code were used to check the combined effects of torsion, shear, and direct tensile loads.
At the construction permit stage, a factor of 1.4 was applied to the DBA pressure provided by Westinghouse. The structural adequacy of the steam generator compartments was checked based on the 40 percent margin and the recommendations of the ACI/ASME Joint Committee contained in Proposed Standard Code for Concrete Reactor Vessels and Containments". Accordingly, the SG compartment design was evaluated for a maximum design internal differential pressure of 24 psi (i.e., 1.4 x 17 psi) using loads, load combinations, and allowable stresses documented in UFSAR Table 3.8.3-2 (provided as Table 6-2). This is reflected in Section 3.8.3.4.1 of the UFSAR, which indicates that a factor of 1.4 was applied to the design pressures resulting from a LOCA during the construction stage. The results are tabulated in UFSAR Table 3.8.3-6 (provided as Table 6-3).
NRC Standard Review Plan 6.2.1.2, Subcompartment Analysis,Section II.B.5, addresses the application of peak differential pressure to be used in the design of the subcompartment. At the construction permit stage, a factor of 1.4 is applied to the calculated peak differential pressure to establish the differential pressure used for design of the subcompartment. At the operating permit stage, the calculated peak differential pressure should not exceed the design pressure. As noted in UFSAR Section 3.8.3.3 and consistent with SRP 6.2.1.2,Section II.B.5, the maximum calculated differential compartment pressures were increased by 40% to account for uncertainties. At the Operating License stage, the design pressures equaled or exceeded the peak calculated differential pressure. Therefore, the design conformed to the requirements of SRP 6.2.1.2.
UFSAR Section 6.2.1.3.10 indicates that the SG compartments were originally designed for two separate pressure loadings. These loadings are (1) a 24 psi maximum internal differential pressure from a break in the main steam line and (2) a uniform internal pressure of 43 psi. The SG compartments were also designed to resist the jet thrust force (910 kips on the roof per Reference 5) that would result following a main steam line break.
The largest blow-down flow results from the severance of the main steam pipe. As indicated in UFSAR Section 3.6.7.6.3, postulated main steam line break locations are shown on UFSAR Figures 3.6.7-1 and 3.6.7-2 (provided as Figures 6-6 and 6-7, respectively). Operating thermal conditions and accident thermal effects accompanying a pipe break (See UFSAR Figure 3.8.3-2, provided as Figure 6-5) were also accounted for.
The blow-down flow analysis of the main steam breaks described in Section 6.2.1.3.10 of the UFSAR resulted in a maximum pressure differential of 19.15 psi compared to the design differential pressure of 24 psi. The UFSAR analysis assumed the main steam flow restrictor is located downstream of the pipe break. Reanalysis of the main steam line break, based on the RSG design with the flow restrictor upstream of the pipe break, resulted in the maximum pressure differential increasing to 19.52 psi. Thus, the design pressure exceeds the maximum calculated differential pressure by 23%, and is therefore conservative.
Page 13 of 43
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'440 uI n--uuo-i As stated in UFSAR Section 3.8.3.4.8, the SG compartment was also originally designed to resist a 43-psi hypothetical pressure from a reactor coolant pipe break. This loading was used to provide a high degree of conservatism in the preliminary design of the SG compartment.
The center wall and the beam below the concrete roof are used as bumper points for main steam pipe whip restraints. These members restrain pipe whip in case of a pipe break and transmit forces to the roof and/or to the wall. It is noted that these whip restraints are bumpers that provide restraint against the pipe-whip in one direction only.
Additionally, they also provide lateral restraint by means of saddle/bracket devices.
The original design of the steam generator compartments, in particular, is documented in Reference 5 and summarized in UFSAR Section 3.8.3.4.8. The roof of the SG cbmpartments was analyzed using a combined member-grid and flat plate finite element STRUDL model. Manual calculations were performed at various locations to confirm computer results. The inverted T-beam, which stiffens the roof, was analyzed for the dynamic effects of a main steam pipe breaking and loading the flange of the beam. The roof was also independently analyzed as a plate using the finite element plate-bending program, GENDEK 3. The roof was analyzed both as a beam-stiffened slab and a uniform slab, neglecting the effects of the beam. The edges of the roof were considered fixed.
From Reference 16 and Figure 6-1, the design compressive strength of the SG compartment concrete at 28 days is 5000 psi. Note that the estimated in-place design compressive strength of the SG compartment roof concrete at 90 days is 5700 psi (Reference 5, Sheets 2e and 2f). The reinforcing used for the interior structures conforms to ASTM A615 Grade 60 (Reference UFSAR Section 3.8.3.2). Figures 6-2 and 6-3 provide additional details of the pre-modification design of the SG compartment roofs. This paragraph provides the historical data as to the required design strength and actual strength of the in-situ steam generator compartment concrete.
Page 14 of 43
Topical Report 24370-TR-C-003-A Table 6-1 (UFSAR Table 3.8.3-1)
Loading Combinations and Allowable Stresses for the Interior Concrete Structure COMBINATIONS LOADINGS 1
IA 2
2A 3
53A 4
5A DEAD LOAD X
X X
x X
X LIVE LOAD X
X X
x x
NORMAL TEMP.
X X
X LOCA PRESSURE X
X X
X LOCA TEMP.
X X
X HYPOTHETICAL x
PRESSURE Y2 SSE X
SSE X
X X
PIPE FORCES x
INITIAL JET PIPE FORCES SATURATED x
(REDUCED) JET OR ANCHOR W.S.D. ALLOWABLE DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER STRESSES BARRIER BARRIER BARRIER BARRIER BARRIER BARRIER fc 0.45 fc 0.45 f'c 0.45 f'c 0.45 f'c 0.60 f'c 0.75 f'c 0.60 rc 0.75 fc 0.60 fc 0.75 Fc fs 0.40 fy 0.40 fy 0.50 fy 0.50 fy 0.72 fy 0.90 fy 0.72 fy 0.90 fy 0.72 fy 0.90 fy U.S.D. LOAD FACTORS 1.25 1.0 1.0 1.25 1.0 1.25 1.0 f'c = Ultimate strength of concrete fy = Yield strength of reinforcement Page 15 of 43 I
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Loading Combinations and Load Factors
- 1. Includes all temporary construction loading during and after construction of containment.
- 2. V is lower for tension members and is essentially the same as given by (ACI 318-71).
LOADS NOMENCLATURE:
D Dead loads, or their related intemal moments and forces Feqo Operating basis earthquake Feqs Design basis earthquake L
Live load, or their related internal moments and forces Pa Accident/incident maximum pressure Ro Piping loads during operating conditions Ra Piping loads due to increased temperature resulting from the design accident Ta Thermal loads under the thermal conditions generated by the postulated break and including To.
To Operational temperature Yr Reaction load on broken pipe due to fluid discharge
- The term "design basis earthquake" has the same meaning as the term "safe shutdown earthquake."
Page 16 of 43 Category Ta D
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Pa To Fego Feqs Ro Ra Yr Allowable S tresses Service:
Const 1.0 1.0 1.0 (Flexure)
Normal 1.0 1.0 1.0 1.0 or 1.0 f= 0.45 fc Factored:
fs= 0.50 fy (Shear)
Extreme 1.0 1.0 1.0 1.0 1.0 50% of Factored Environ-mental Abnormal 1.0 1.0 1.0 1.5 1.0 an /or 1.0 (Flexure) fc= 0.75 c AbnormaV 1.0 1.0 1.0 1.25 1.25 1.0 and/or 1.0 f = 0.90 fy Severe Environ-(Shear) mental AbnormaV 1.0 1.0 1.0 1.0 1.0 1.0 a or 1.0 Extreme (2) V Environ-mental f = 0.85
Topical Report 24370-TR-C-003-A Table 6-3 (UFSAR Table 3.8.3-6)
Original Design Stress Margin Table 3.8.3-1 Criteria Versus Table 3.8.3-2 Criteria (4)
TABLE 3.8.3-1 CRITERIA TABLE 3.8.3-2 CRITERIA LOCA PRESSURE + 20%
LOCA PRESSURE + 40%
DESIGN FEATURE (2) CONTROLLING STRESS MARGIN (%)
(3) CONTROLLING STRESS MARGIN (%)
LOAD SHEAR MOMENT LOAD COMBINATION SHEAR MOMENT COMBINATION REACTOR VESSEL ANNULUS WALL @ R.C. PUMP SUPPORT SA
-(1) 18.5 ABNORMAL
-(1) 80
'REACTOR CAVITY COLUMNS 4-FLEXURE 17 18.5 ABNORMAL/SEVERE 64 22 2-SHEAR ENVIRONMENTAL
'CONTROL ROD DRIVE MISSILE SHIELD 4
9 7
ABNORMAL 70 61 CRANE WALL @ EL. 679.78 5
0 0
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0 ENVIRONMENTAL CRANE WALL COLS @ 194'-08'-24' & 2040-3V-57' 5A 7
19 ABNORMAUSEVERE 20 10 ENVIRONMENTAL
'STEAM GEN COMPTS, SIDE WALL @ CRANE WALL 1
58 17.5 ABNORMAL 87 34
- PRESSURIZER COMPT @ CRANE WALL 4
16 11 ABNORMAL
>100
>100
'FLOOR EL 733.63 @ INTERSECTION W/CRANE WALL 1
9 8.5 ABNORMAL 19 39
- FLOOR EL 721.0 @ CRANE WALL 1
62 73 ABNORMAUSEVERE 68
>100 ENVIRONMENTAL MISC COMPTS, RADIAL WALL @ CRANE WALL 1
25 61 ABNORMAL 36
>100 FILL SLAB EL. 679.78 @ CRANE WALL 5
>20 0
ABNORMAUEXTREME
>20 0
ENVIRONMENTAL
-CANAL WALL (SPAN C -VERT POS MOM) 1
-(1) 3.5 ABNORMAL
- (1) 51
'CRANE WALL (SPAN C - NEG MOM @ OPERATING FLOOR) 1 40 3.5 ABNORMAUSEVERE 28 11 ENVIRONMENTAL CRANE WALL, EL. 714.0, HORIZ, NF 1
- (1) 5.5 ABNORMAL
-(1) 36
- DENOI DIVIDERF BARRIER (1) NEGLIGIBLE SHEAR STRESSES IN THESE AREAS (2) SEE TABLE 3.8.3-1 FOR LOADS (3) SEE TABLE 3.8.3-2 FOR LOADS (4)
I Tis tabie does not reiect tme evaluatons aocumented in txnlDlt F oT report ujii tib-1 Y-C.
Tabulated stress margins are from the original calculations and do not reflect later evaluations.
Changes have been documented in calculation packages.
Page 17 of 43
Topical Report 24370-TR-C-003-A Page 8 of 43
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Page 22 of 43
Topical Report 24370-TR-C-003-A S4 - 4
- POSTULATED BREAKS s - 4 E
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Figure 6 Steam Generators 1 and 4 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-1)
Page 23 of 43 4
Topical Report 24370-TR-C-003-A EAST VALVE ROOM WALL POSTULATED BREAKS SZ-4 l
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Figure 6 Steam Generators 2 and 3 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-2)
Page 24 of 43
Topical Report 24370-TR-C-003-A 7.0 Description of Modification to the Structure and New Analyses After installation of the replacement steam generators, the removed concrete section (plug) of the steam generator compartment roof will be reattached to the complimentary portion of the existing SG compartment by means of top and bottom steel connection frames. The plug will be attached to the top and bottom connection frames using four 2-inch diameter threaded rods that are installed in core bore holes through the plug. The top and bottom connection frames will clamp the concrete plug to the complimentary portion of the SG compartment using six 2-1/2 inch and eighteen 2-inch diameter threaded rods. The threaded rods are installed in the core bore holes located around the cut line as shown on Figure 7-2. The frames consist of box beams made from 1-1/4 inch ASTM A572 Grade 50 material with a yield stress of 50 ksi. The threaded rods conform to ASTM Al 93 Grade B7 material with a yield stress (Fy) of 105 ksi. The threaded rods will be preloaded to a stress level of 0.7 (Fy) after the concrete plug is installed. This configuration will transfer all the vertical forces from the concrete plug to the complimentary portion of the existing SG compartment structure. The lateral forces will be transferred to the existing SG compartment structure by a series of steel shims (ASTM A36 material) that will be driven into the annular space around the perimeter of the plug and mechanically locked into place. The annular space between the concrete plug and the complimentary portion of the SG compartment structure will be grouted.
The width of the opening between the concrete plug and the complimentary portion of the SG compartment will vary as the wire rope used to make the cuts wears. The surface of the cutout section of concrete will be prepared to provide a gap that ranges from 4-inches to 1 -% inches. The non-shrink grout to be used to fill the annular gap and the core bore holes is Masterflow 928 or Masterflow 713 Plus as manufactured by ChemRex. This grout is produced under a Quality Assurance program and is certified to comply with the requirements of ASTM C1107. This ASTM standard requires that the grout be tested for height change and compressive strength. The non-shrink grout, like the surrounding concrete, could 'theoretically" experience the formation of micro-cracks when subjected to the design pressure load. Conservative estimates (Reference 8) of the flow path through these micro-cracks yield values that are 1.6 percent of the total design bypass leakage flow area of 5 square feet discussed in UFSAR Section 6.2.1.3.5.
The design leakage area is composed of a known leakage area of approximately 2 square feet and an undefined leakage area. Any leakage through cracks in the grout would be part of this undefined leakage area. UFSAR Figure 6.2.1-22 (provided as Figure 7-1) shows that this percentage increase in bypass area would result in a very small increase in the upper containment pressure. Therefore, micro-cracks resulting from the design pressure load will have a negligible effect on the function of the divider barrier and the analyses that depend on the divider barrier. The SG compartment roof modification described above is detailed on Figure 7-2.
The above mode of restoration results in a modified configuration to the roof of the SG compartment. The use of steel through-bolted connection frames essentially results in a more flexible boundary condition along the cut-line. In other words, this boundary condition behaves more like a hinge. This means that the reinstalled concrete section of the roof is more flexible than the original configuration, and therefore, subjected to higher deflections and bending moments towards its center. The frame structure is designed to accommodate this increased deflection. Also, the inverted concrete T-beam section under the concrete roof acts like a spacer transmitting the whip-restraint forces from the main steam pipe to the 3 feet thick roof. In the original configuration, the T-beam provided considerable strength in resisting the pipe whip loads. It is noted that since the Page 25 of 43
Topical Report 24370-TR-C-003-A reinstalled concrete section in the modified configuration is more flexible than the original design, the forces are redistributed within the reinstalled concrete section. The effects on the walls surrounding the SG compartment (3 feet thick crane wall, 2 feet thick compartment wall and the center wall) were also evaluated. Therefore, as described below, the evaluation of the modified configuration included the T-beam, roof, crane wall, SG compartment walls, and center wall.
The modified SG compartment roof was evaluated to load combinations, load factors, and allowable stresses tabulated in Table 7-2. Table 7-2 is based on Sections CC-3200 and CC-3400 of ASME Section II, Division 2, 1975, which are generally consistent with UFSAR Table 3.8.3-2. Exceptions to UFSAR Table 3.8.3-2 are the load factors associated with the Yr load and the allowable stresses when thermal effects are included with other loads. The Yr load factors used to evaluate the modified SG compartment roof are consistent with ASME Section 1II, Division 2, 1975. The allowable stresses due to thermal effects are consistent with both the Proposed ASME Section 1II, Division 2, 1973 and ASME Section III, Division 2, 1975. The structural steel through-bolted connection frames are designed in accordance with Reference 3.
As noted in Section 6.0, the load combinations in Table 3.8.3-2 of the UFSAR are based on Table CC-3200-1 of the Proposed ASME Section 1I, Division 2, 1973, Proposed Standard Code for Concrete Reactor Vessels and Containments, Section CC-3000 which was issued in 1973 (the time of original design) by ACI-ASME Committee on Concrete Pressure Components for Nuclear Service for trial use and comment. The purpose of this topical report is to support taking an exception for the load factors associated with the Yr load (reaction load due to fluid discharge on broken pipe, which in the present case is the pipe thrust load) for the Abnormal and AbnormaVSevere Environmental Load Categories as described below. Use of this exception is consistent with the adopted 1975 and later editions of ASME Section III, Division 2 (Reference 12).
In the original design analyses the Yr load was combined with load factors of 1.5 and 1.25 that are associated with the DBA design pressures for the Abnormal and AbnormaVSevere Environmental Load Categories, respectively. The jet impingement I pipe-whip / pipe break loading (Yr) will rapidly increase, peaking shortly after pipe break and then rapidly decrease in amplitude. The associated DBA pressure loadings will take considerable time following pipe break to reach their design basis peak amplitude values. It is, therefore, overly conservative to combine the DBA pressures with design basis pipe-whip load. The adopted 1975 and later editions of ASME Section III Division 2 (Reference 12) do not include this load combination. The load combinations and allowables used in this analysis for the Abnormal and Abnormal/Severe Environmental Load Categories were based on Table CC-3230-1 (included in this report as Table 7-1) of the adopted 1975 Edition of ASME Section IlIl Division 2 (Reference 12), which superseded the Proposed Code (Reference 11). Note that the load denoted as Rr in Reference 12 corresponds to the Yr load in Reference 11. Also, as allowed by Section CC-3400 of both the proposed 1973 and adopted 1975 versions of ASME Section IlIl, Division 2, credit is taken for the allowable stresses in concrete and rebar to be increased by 33-1/3% for service loads, and the tensile strain in rebar to exceed yield for factored loads when thermal gradient effects are included in the load combinations.
It is also noted that it is acceptable to use a later edition of the ASME Section III code for repairs and replacement per ASME Section Xl (Reference 13). Further, it is noted that the design DBA differential pressure of 24 psi being used in the SG compartment roof evaluation is conservative since it is higher than the maximum calculated differential Page 26 of 43
Topical Report 24370-TR-C-003-A pressure of 19.52 psi by 23%. These conservatisms further justify the use of load factors for the Abnormal and AbnorrnaVSevere Environmental Load Categories based on the adopted 1975 Edition of ASME Section IlIl, Division 2 (Reference 12) without compromising the integrity of the modified SG compartment roof.
The modified configuration of the SG compartment was analyzed for design loads using a 3D finite element ANSYS (Version 5.6) model (Reference 6). Although the roof remains the focus of the evaluation, the model (provided as Figure 7-3) included five components - the 3 feet thick roof, entire SG compartment wall, center wall, 180' sector of the crane wall, and the whip restraint beam; to obtain an accurate representation of the system. The finite elements used were SHELL43 elements for the roof and walls, BEAM44 elements for the whip restraint beam, and BEAM4 elements for the portions of the crane wall where it has openings to the ice condenser. The top of the SG compartment roof is at elevation 778.69'. The compartment wall was modeled as fixed at elevation 733.63 at the top of the containment operating floor; and the crane wall (Figure 6-4) is modeled as fixed at elevation 721' where the ice condenser floor is located. The nodes at the cut-line along which the connection frames and tapered steel shims are located were realistically modeled to transmit vertical forces and in-plane compression only. The material properties used in the model for the concrete were consistent with those used in the original analysis in Reference 5.
The loads, load combinations and allowable stresses to which the modified SG compartment was evaluated are documented in Reference 7 and summarized in Table 7-2. The modified configuration of the SG compartment roof was analyzed for the following design loads: dead load, live load, design pressure differential of 24 psi from a DBA (main steam pipe break), operating and accident temperature effects, seismic effects (OBE and SSE), and pipe thrust load on the whip-restraint beam from a broken main steam pipe. Design pressure, seismic, and pipe thrust effects were modeled as equivalent static loads. The pipe thrust load applied was 926.25 kips, which is based on the blowdown load documented in Reference 14 and conservatively includes a factor of 1.5 to account for the gap between the MS piping and the restraint (as used in the original analysis).
As noted in Section 6.0, the SG compartments were originally designed for a hypothetical pressure of 43 psi resulting from the rupture of a reactor coolant pipe. This pressure was used to provide a high degree of conservatism in the original design, which allowed the structure to accommodate a range of possible equipment configurations and final analysis results. The concrete strength used in the roof evaluation is the in-place compressive strength of the SG compartment roof concrete at 90 days, which is 5700 psi (Reference 5, Sheets 2e and 2f).
The steel through-bolted connection frames and tapered steel shims were designed and evaluated for the load combinations as described in the previous discussion based on criteria in Section 5.1 of Appendix A to Reference 3.
The vertical design loads on the concrete plug will be transferred into the SG compartment structure around the perimeter of the plug by the clamping forces induced by the through-bolts connecting the top and bottom steel connection frames. For example, a vertical load in the upward direction, acting on the concrete plug, would be transferred to the compartment structure as follows:
Page 27 of 43
Topical Report 24370-TR-C-003-A The vertical load from the plug will be transferred by bearing between the concrete plug and the steel bearing plates (located between the concrete and the steel frame), to shear in the steel frame, to tension in the through-bolt, back to shear in the lower frame, to bearing between the steel bearing plates and the concrete of the SG compartment.
The horizontal design loads on the concrete section will be transferred into the SG compartment structure via tapered steel shim sets. Each tapered shim set will be comprised of a tapered shim attached to the face of the concrete section and a loose tapered shim that will be driven into the gap between the fixed tapered shim and the existing compartment concrete. When installed snugly, the loose tapered shim will be welded to the tapered fixed shim to prevent movement. Approximately 30 tapered shim sets (15 top and -15 bottom) will be installed around the perimeter of the compartment concrete section. Conservatively, only four (4) tapered shim sets will be considered to transfer all the horizontal design loads between the concrete section (with frame attached) and the compartment structure. The grout between the concrete section and compartment structure will not be considered to transfer any design basis loads.
The Divider Barrier will be restored by covering the annular space around the perimeter of the plug on the bottom side of the 3-foot thick SG compartment roof and filling the space with non-shrink grout.
Page 28 of 43
Topical Report 24370-TR-C-003-A Table 7-1 (Table CC-3230-1 from ASME Section III, Division 2,1975)
Load Combinations and Load Factors Category D
L1 F
Pt Pa Tr TO Ts Eo Ess W
Wt Ro Re R..
Pv Hq Service:
Sever evrn ntl1.0 1.0
.1.0
- 1.
.0.
.0.
.0.
.0.
Fan rct io:.
1 Sevrena rnena 1.0 1.3 1.0 1.0
.5.
1.0 1.0 Setre environmental 1.0 1.0
'1.0 1.0 1.0 1.0 Abnormal
~~~~1.0 1.0 1.0
- 1....
1.0..
1.0 1.0..
AbmlSevere environmental 1.0 1.0 1.0 12...
1.0..
1.25 1.0..
1.0 1.0 1.0 12...
1.0..
1.25 0
1.0..
AbolExtreme environmental 1.0 1.0 1.0 1.0 1.0 1.0 NOTE:
~
~
~
~ ~
10 10 10 10 (1)bncluealltmoaycnruin loain durin and afte costucio of. cotim..
1 Page 29 of 43
Topical Report 24370-TR-C-003-A Table 7-2 Loading Combinations, Load Factors and Allowable Stresses for SG Compartment Roof Modification (5)(6)
NOTES:
- 1. Includes all temporary construction loading during and after construction of containment.
- 2.
vc is lower for tension members and is given by v: = 2f.
(1 + 0.002NJAg), with Nu negative for tension.
- 3. The allowable stress is increased by 33-1/3% when temperature effects are combined with other loads.
- 4. The tensile strain may exceed yield when the effects of thermal gradients are included in the load combination, i.e., f, can be <= fy, and Es can be > cy when thermal effects are included.
- 5. The load combinations, load factors and allowable stresses in this table are based on the ASME Section III Division 2, 1975, which are, in general, consistent with the proposed ACI 359 - ASME Section I Division 2, 1973 with the exception of load factors associated with the Yr load.
- 6. Structural steel components of the through-bolted connection frames and tapered steel shims were designed in accordance with TVA Design Criteria SQN-DC-V-1.3.2, Miscellaneous Steel Components for Class I Structures.
LOADS NOMENCLATURE:
D F.qo F.qs L
Pa Ro Ra Ta To Yr Dead loads, or their related intemal moments and forces Operating basis earthquake Design basis earthquake Uve load, or their related intemal moments and forces Accidentincident maximum pressure Piping loads during operating conditions Piping loads due to increased temperature resulting from the design accident Thermal loads under the thermal conditions generated by the postulated break and including To Operational temperature Reaction load on broken pipe due to fluid discharge (corresponds to R, in ASME Section III, Division 2,1975)
'The term design basis earthquake" has the same meaning as the term safe shutdown earthquake."
Page 30 of 43 Allowable Category Ta D
L.i Pa T.
FeO Fas R.
Ra Yr Stresses Service:
(Flexure)
=0.45 fc Const 1.0 1.0 1.0 f, = 0.50 fy (3)
Normal 1.0 1.0 1.0 1.0 1.0 (Shear) 50% of Factored (3)
Factored:
Extreme 1.0 1.0 1.0 1.0 1.0 (Flexure)
Environmental fc = 0.75 f'c
= 0.90 fy (4)
Abnormal 1.0 1.0 1.0 1.5 1.0 (Shear)
(2)v.= 2j7 AbnormaV 1.0 1.0 1.0 1.25 1.25 1.0 0.85 Severe Environmental AbnormaV 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Extreme Environmental I
I
Topical Report 24370-TR-C-003-A U-
= 0.6 _
Z2 a-r_
Li. 0.2 -
ci,
@2 N
0-00
@2
@2
,M.
-0.2 0-
.4-a
-100
-7;
-SO
-25 0
25 50 75 100 OUXNGE IN DECK BYASS AM (PERCNT)
Figure 7 Sensitivity of Peak Compression Pressure to Deck Bypass (UFSAR Figure 6.2.1-22)
Page 31 of 43
w~~~~tS vfr4-lo E mear ane roof
-M i§t1!{i EM ;! B PLS*!e£X.stPIW
/a,-,
I Sia
\\ATt tEO
=E, K I3 aa.__
7 A ewe" 4-Flgure I 2. - Steam Generator Compartmen R=EM-4t Cnc.
Fa Lao!
an Connect.on Detalls Fiur 72 Sea GnraorCopatmntRofCon-t-n-rae-a-u-ad-onecio--ta-Topical Report 2437O-TR-C-OO3 A Page 32 of 43
Topical Report 24370-TR-002-A "ROOF" - Enclosure Roof "CRANEW" - Crane Wall LU
\\
'EL 733.63' \\
"L "COLUMN - part of crane "EWALL - SG Enclosure Wall wall with openings to the ice condensor
'ROOF"- Enclosure Roof EL 778.69' "BEAM" - Whip Restraint Beam.
y 4J "CENTERWV-Center Wall separates two SGs Figure 7 Finite Element Model "SGE1" and "SGE2" and Element Groups and Global Coordinate Systems (Reference 6)
Page 33 of 43
Topical Report 24370-TR-002-A 8.0 Results of New Analyses The modified configuration of the steam generator compartment roofs has been evaluated for the design loads and load combinations documented in Reference 7 as described in Section 7.0. Except as noted in Section 7.0, these design loads and load combinations are consistent with those used in the original analyses for the SG compartments. The structural adequacy of the modified SG compartment roof configuration under these design loads and load combinations was evaluated in Reference 8. The design of the steel through-bolted connection frames and tapered steel shims is documented in Reference 9. The results are briefly summarized below.
Normal service load combinations used to evaluate the modified SG compartment roof configuration were the same as those used for the original configuration. Under normal service load conditions, the maximum concrete and rebar stresses in the modified roof are within the allowable normal service concrete and rebar stress limits as specified in Section CC-3430 of ASME Section III, Division 2, 1975 (summarized in Table 7-2). The critical areas where these stresses occur are near the middle surface of the cut section at the junction of the roof and the end of the whip restraint beam (Reference Area 1 on Figure 8-1). The stress levels in other areas are generally much lower. Therefore, the modified SG compartment roof configuration is acceptable under normal service conditions.
The load combinations evaluated for the modified roof were based on Table CC-3230-1 (included in this report as Table 7-1) of the adopted 1975 Edition of ASME Section il Division 2 (Reference 12), which replaced the Proposed Code (Reference 11) as discussed in Sections 6.0 and 7.0. These load combinations are similar to those used for the original SG compartment roof design except for the Abnormal and Abnormal /
Severe Environmental load categories for which the Yr load is now not considered in the load combination. For factored load combinations on the modified roof configuration, the most critical load combinations are the Abnormal and Abnormal I Extreme Environmental load categories. The critical areas of high stresses for the Abnormal load combination are the approximately triangular corner areas of the existing roof bounded by the cut-line near each end of the center wall (Reference Areas 2 and 3 on Figure 8-1). For the Abnormal / Extreme Environmental load combination the critical area included the area near the middle of the cut section at the junction of the roof and the end of the whip restraint beam (Reference Area 1 on Figure 8-1) in addition to the corner areas identified for the Abnormal load combination. It is noted that the maximum stresses/forces occurred only in the localized areas mentioned above. The stresses in other areas are lower. The maximum stresses, in these critical areas, for the factored load combinations were found to be within the allowable concrete and rebar stresses based on limits specified in Section CC-3400 of ASME Section 1I1, Division 2,1975. The maximum vertical deflection occurred for the Abnormal / Extreme Environmental load combination at the middle of the roof near the end of the whip restraint beam.
It is noted that the design DBA.differential pressure of 24 psi was used in the modified SG compartment roof stress evaluation. Even though the calculated stresses under accident conditions equaled the allowable stresses in some locations, this analysis is conservative since it used a differential pressure that is 23% higher than the maximum calculated differential pressure of 19.52 psi.
Page 34 of 43
Topical Report 24370-TR-002-A The influence of the modified roof configuration on stresses in the SG compartment wall sections adjacent to the roof has been determined to be insignificant and the wall and roof stresses remain within design allowables.
The design of the steel through-bolted connection frames and tapered steel shims documented in Reference 9 is described in Section 7.0 and shown on Figure 7-2. The through-bolts will be installed with a pre-tension load based on 0.7Fy. Using conservative design checks, the maximum calculated bending stress in the connection frame beams and the maximum calculated bearing stress on concrete and the tapered steel shims were determined to be below allowables. The connection frame beams will be used in conjunction with the through-bolts to provide the clamping action that will transfer the vertical design basis loads from the concrete section to the compartment.
The connection frame beams span over all of the connection through-bolts. Since all the connection frame beams are connected together, rigid body rotations of the beams about the bolt axes are prevented at all concrete section/compartment connections.
The connection frame beams have been designed to transfer all vertical design loads, at the concrete section/compartment interface, via bending and shear stresses. The beams have been designed such that the maximum stresses in the beam plates and connecting welds are less than the allowable stresses.
The connection frame beams are sized such that the concrete bearing stresses under the beams are below allowables due to both the connection through-bolt pre-tension loads and due to all design basis loads.
The connection frame beams are connected by web angles or connection plates. The welded angles/plates are designed to be flexible in order to transfer all vertical design loads between beam members of the frame, as pinned connections. Vertical loads are due to the vertical seismic inertia from the concrete and the maximum DBA pressure (seismic inertia loading from the steel frame is negligible). As the concrete section deflects, it lifts the individual frame members, hence, inducing vertical loads at the beam-to-beam connections and vertical prying loads at the through-bolt connections. The beam connection angles/plates are also designed to transfer all horizontal seismic loads due to the maximum accelerations of the frame.
Based on the evaluations in the calculations noted above, the modified SG compartment roofs have been found to be structurally adequate for the loads associated with the design loading conditions/combinations which are in general consistent with the original design except as noted above and in Section 7.0.
The modifications to the steam generator compartment roofs do not affect the structural capability of the steam generator compartments to contain the intemal pressure associated with the design bases main steam line breaks. The modifications do not affect temperature differentials through the compartment roof or the radiation shielding capacity of the structures.
As discussed in Section 6.5.6.3 of the UFSAR, there is a maximum calculated leakage of 250 cfm between the upper and lower containment through the divider barrier, of which the steam generator compartments are part. The amount of leakage between the two sections of the containment will not be affected by the restoration of the steam generator compartment roofs. The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries Page 35 of 43 I
Topical Report 24370-TR-002-A between upper and lower containment. It is noted that any leakage due to possible cracks in the grout, particularly under design DBA loading, will be extremely small and therefore insignificant (Reference 8).
Area 2 Cutline T-Beam Centerline Area3 CenterWall Centerline Figure 8-1 Areas of Critical Stresses 9.0 Summary and Conclusions Restoration of the SG compartment will be accomplished by reattaching the removed section of concrete using through-bolted structural steel connection frames and tapered steel shims in the annular gap. The SG compartments have been reanalyzed to determine that the modified configuration is acceptable. This analysis follows the same basic approach as documented in the existing SG compartment design calculations, the Sequoyah design criteria, and/or the Sequoyah UFSAR. Areas where the two analyses differ are summarized in Table 9-1.
Table 9-1 Differences Between Original and New Steam Generator Compartment Analyses Original Analyses New Analyses Analyzed compartment structure as Analyzed compartment structure several individual components (roof, using a three dimensional ANSYS enclosure wall, center wall, and crane finite element model comprised of wall) using two-dimensional model.
system components.
Evaluated compartment structure for a Did not evaluate compartment 43-psi hypothetical pressure.
structure for a 43-psi hypothetical Page 36 of 43 I
I
Topical Report 24370-TR-002-A Use of the methodologies, loads and load combinations discussed in this topical report are either consistent with the original design basis or based on accepted industry design standards. The proposed modifications to the SG compartment design are therefore justified.
10.0 References
- 1.
Sequoyah Nuclear Plant Updated Final Safety Analysis Report, Amendment 16.
- 2.
TVA Design Criteria SQN-DC-V-1.1, Design of Reinforced Concrete Structures.
Revision 16.
- 3.
TVA Design Criteria SQN-DC-V-1.3.2, Miscellaneous Steel Components for Class I Structures, Revision 10.
- 4.
TVA Design Criteria SQN-DC-V-1.3.3.1, Additions After November 14, 1979 -
Reinforced Concrete, Structural, and Miscellaneous Steel, Revision 6.
- 5.
TVA Calculation SCG-1-40, Steam Generator Compartment, Final Design, Revision 4.
- 6.
TVA Calculation SCG-1S-607, Evaluation of Steam Generator Compartment Modification - 3D Finite Element Model, Revision 0.
- 7.
TVA Calculation SCG-1S-608, Evaluation of Unit 1 Steam Generator Compartment Modification - Load Conditions and Allowable Stresses, Revision 0.
- 8.
TVA Calculation SCG-1S-609, Evaluation of Steam Generator Compartment Modification - Finite Element Analysis Results, Revision 0.
- 9.
WVA Calculation SCG-1S-610, Evaluation of Unit 1 Steam Generator Compartment Modification - Design of Roof Support Frames, Revision 1.
- 10.
Bechtel Calculation 24370-C-013, Rev. 0, ANSYS 5.6 Verification.
Page 37 of 43 Original Analyses New Analyses pressure.
Analyzed compartment structure Analyzed compartment structure for a initially for a maximum differential maximum design internal differential pressure of 21.3 psi which is a 25%
pressure of 24 psi as specified in the increase over the DBA pressure UFSAR using loads, load differential of -17 psi for the SG combinations and allowable stresses compartment provided by documented in Table 7-2.
Westinghouse (i.e., 1.25 x 17 psi). Per NRC request, a 40% increase in DBA differential pressure (i.e., 1.4 x 17 psi) was investigated later.
Evaluated compartment roof globally Evaluated the modified roof globally for an equivalent static jet thrust force for an equivalent static pipe thrust
(-910 kips on the roof) that would load of 926.25 kips which is based on result following a main steam pipe the shock spectrum from the MS Blow break inside a single compartment.
Down Analysis.
Analyzed the compartment structure Analyzed the modified compartment using the load combinations, load structure using load combinations and factors, and allowable stresses shown allowable stresses in Table 7-2. Load in UFSAR Tables 3.8.3-1 or 3.8.3-2.
factors for the load combinations and allowable stresses were based on Table CC-3230-1 and Section CC-3400, respectively, of the 1975 Edition of ASME Section III, Division 2.
I
Topical Report 24370-TR-002-A
- 11.
Proposed ASME Section III Division 2, 1973, Proposed Standard Code for Concrete Reactor Vessels and Containments, Section CC-3000 (This draft code was issued in 1973 by ACI-ASME Committee on Concrete Pressure Components for Nuclear Service for Trial Use and Comment).
- 12.
ASME Section III Division 2, 1975 Edition, Concrete Reactor Vessels and Containments, Section CC-3000.
- 13.
ASME Section Xl, Rules for Inservice Inspection of Nuclear Power Plant Components.
- 14.
TVA Calculation 0600117-S002, RO, Blow Down Analysis - Main Steam System.
- 15.
TVA-NQA-PLN89-A, Nuclear Quality Assurance Plan, Revision 10.
- 16.
General Engineering Specification G-2, Plain and Reinforced Concrete, Revision 7.
Page 38 of 43
Topical Report 24370-TR-002-A Appendix A No Significant Hazards Consideration Determination DESCRIPTION OF THE PROPOSED CHANGE The four steam generators of the Sequoyah Nuclear Plant Unit 1 will be replaced during the spring of 2003. To support the replacement of the old steam generators (OSGs) with the replacement steam generators (RSGs), access openings will be created in the roof of the steam generator (SG) compartments inside containment. An appropriately sized access opening will be made in each SG compartment roof by cutting out a section of concrete from the roof of the compartments.
Upon completion of installation of the RSGs, the original cut section (plug) of the SG compartment roof will be reinstalled using a modified configuration from the original.
The concrete plug removed from each of the SG compartment roofs will be reattached to the complimentary portion of the SG compartment roof by means of top and bottom steel connection frames. The plug will be attached to the top and bottom connection frames using four 2-inch diameter threaded rods that are installed in core bore holes through the plug. The top and bottom connection frames will clamp the concrete plug to the complimentary portion of the SG compartment using six 2-1/2 inch and eighteen 2-inch diameter threaded rods. These threaded rods are installed in the core bore holes located around the plug cutline and will be pre-loaded. The frames consist of box beams made from 1-1/4 inch steel. A series of steel shims will be driven into the annular space around the perimeter of the plug and mechanically locked into place.
The core bores and the annular space between the concrete plug and the complimentary portion of the SG compartment roof will be grouted using non-shrink grout that conforms to ASTM C 1107, thereby sealing the roof.
II.
REASON FOR THE PROPOSED CHANGE The process for restoration of the steam generator compartment roof using the through-bolted connection frames results in less construction debris in containment since the concrete cuts will not require chipping for rebar splicing. The process is also simpler and faster than splicing new rebar and pouring new concrete.
Ill.
SAFETY ANALYSIS Normal service load combinations used to evaluate the modified SG compartment roof configuration were the same as those used for the original configuration. Under normal service load conditions, the maximum concrete and rebar stresses in the modified roof are within the allowable normal service concrete and rebar stress limits as specified in Section CC-3430 of ASME Section 1I1, Division 2, 1975. The critical areas where these stresses occur are near the middle surface of the cut section at the junction of the roof and the end of the whip restraint beam. The stress levels in other areas are generally much lower. Therefore, the modified SG compartment roof configuration is acceptable under normal service conditions.
The load combinations evaluated for the modified roof were based on Table CC-3230-1 of the adopted 1975 Edition of ASME Section Ill Division 2, which replaced the proposed 1973 ASME Section 1I1, Division 2. These load combinations are similar to those used Page 39 of 43
Topical Report 24370-TR-002-A for the original SG compartment roof design except for the Abnormal and Abnormal/Severe Environmental load categories for which the Yr load is now not considered in the load combination. For factored load combinations on the modified roof configuration, the most critical load combinations are the Abnormal and AbnormaVExtreme Environmental load categories. The critical areas of high stresses for the Abnormal load combination are the approximately triangular corner areas of the existing roof bounded by the cut-line near each end of the center wall. For the AbnormaVExtreme Environmental load combination the critical area included the area near the middle of the cut section at the junction of the roof and the end of the whip restraint beam in addition to the corner areas identified for the Abnormal load combination. It is noted that the maximum stresses/forces occurred only in the localized areas mentioned above. The stresses in other areas are lower. The maximum stresses for the factored load combinations were found to be within the allowable concrete and rebar stresses based on limits specified in Section CC-3400 of ASME Section III, Division 2, 1975. The maximum vertical deflection occurred for the AbnormaVExtreme Environmental load combination at the middle of the roof near the end of the whip restraint beam.
It is noted that the design DBA differential pressure of 24 psi was used in the modified SG compartment roof stress evaluation. Even though the calculated stresses under accident conditions equaled the allowable stresses in some locations, this analysis is conservative since it used a differential pressure that is 23% higher than the maximum calculated differential pressure of 19.52 psi.
The influence of the modified roof configuration on stresses in the SG compartment wall sections adjacent to the roof have been determined to be insignificant and the wall and roof stresses remain within design allowables.
The bolts used in the steel through-bolted connection will be preloaded to a stress level of 0.7 Fy. By conservative analysis, the maximum calculated bending stress in the connection frame beams and the maximum calculated bearing stress on concrete and the tapered steel shims were determined to be below allowables. The connection frame beams will be used in conjunction with the through-bolts to provide the clamping action that will transfer the vertical design basis loads from the concrete section to the compartment. The connection frame beams span over all of the connection through-bolts. Since all the connection frame beams are connected together, beam rigid body rotation about the bolt axes are prevented at all concrete section/compartment connections.
The connection frame beams have been designed to transfer all vertical design loads, at the concrete section/compartment interface, via bending and shear stresses. The beams have been designed such that the maximum stresses in the beam plates and connecting welds are less than the allowable stresses.
The connection frame beams are sized such that the concrete bearing stresses under the beams are below allowables due to both the connection through-bolt pre-tension loads and due to all design basis loads.
The connection frame beams are connected by web angles or connection plates. The welded angles/plates are designed to transfer all vertical design loads between beam members of the frame, as pinned connections. Vertical loads are due to the vertical seismic inertia from the concrete and the maximum DBA pressure (seismic inertia Page 40 of 43
Topical Report 24370-TR-002-A loading from the steel frame is negligible). As the concrete section deflects, it lifts the individual frame members, hence, inducing vertical loads at the beam-to-beam connections and vertical prying loads at the through-bolt connections. The beam connection angles/plates are also designed to transfer all horizontal seismic loads due to the maximum accelerations of the frame.
The modified SG compartment roofs have been found to be structurally adequate for the loads associated with the design loading conditions/combinations which are in general consistent with the original design except as noted above.
The modifications to the steam generator compartment roofs do not affect the structural capability of the steam generator compartments to contain the intemal pressure associated with the design bases main steam line breaks. The modifications do not affect temperature differentials through the compartment roof or the radiation shielding capacity of the structures.
As discussed in Section 6.5.6.3 of the UFSAR, there is a maximum calculated leakage of 250 cfm between the upper and lower containment through the divider barrier, of which the steam generator compartments are part. The amount of leakage between the two sections of the containment will not be significantly affected by the restoration of the steam generator compartment roofs. The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries between upper and lower containment. It is noted that.any leakage due to possible cracks in the grout, particularly under design DBA loading, will be extremely small and therefore insignificant.
IV.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA has concluded that operation of SQN Unit 1, in accordance with the proposed modification to the steam generator compartment roof, does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92(c).
A.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The probability of occurrence or the consequences of an accident are not increased as presently analyzed in the safety analyses since the objective of the event mitigation is not changed. No changes in event classification as discussed in UFSAR Chapter 15 will occur due to the modification of the Unit 1 steam generator compartment roof design.
The grout used to fill the gap between the replaced concrete and the surrounding concrete, like the surrounding concrete, could "theoretically" experience the formation of micro-cracks when subjected to the design pressure load.
Conservative estimates of the flow path through these micro-cracks yield values that are numerically insignificant when compared to the allowable divider barrier bypass leakage. Micro-cracks resulting from the design pressure load will have a negligible effect on the function of the divider barrier and the analyses that depend on the divider barrier. Therefore, the containment design pressure is not challenged, thereby ensuring that the potential for increasing offsite dose limits Page 41 of 43
Topical Report 24370-TR-002-A above those presently analyzed at the containrment design pressure of 12.0 pounds per square inch is not a concern.
Therefore, the proposed modification to the Unit 1 steam generator compartment roof design will not significantly increase the probability or consequences of an accident previously evaluated.
B.
The proposed amendment does not create the possibility of a new or different kind of accident from anv accident previouslV evaluated.
The possibility of a new or different accident situation occurring as a result of this condition is not created. The steam generator compartment roof forms part of the divider barrier. This barrier is not an initiator of any accident and only serves to force steam that is released from a LOCA/ DBA to pass through the ice condenser. The failure of any part of the divider barrier is considered critical since it would allow LOCA/DBA steam to bypass the ice condenser, thereby increasing the pressure within the primary containment.
As discussed in Section 6.5.6.3 of the UFSAR, there is a maximum calculated leakage of 250 cfm between the upper and lower containment through the divider barrier. The amount of leakage between the two sections of the containment will not be significantly affected by the restoration of the steam generator compartment roofs. The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries between upper and lower containment. It is noted that any leakage due to possible cracks in the grout, particularly under design DBA loading, will be extremely small and therefore insignificant.
Therefore, the potential for creating a new or unanalyzed condition is not created.
C.
The proposed amendment does not involve a significant reduction in a margin of safety.
A design DBA differential pressure of 24 psi was assumed in the original design of the steam generator compartment roof. This differential pressure is 23%
higher than the maximum calculated differential pressure of 19.52 psi. Since the same design differential pressure was also used in the modified SG compartment roof stress evaluation, the margin of safety was not reduced.
As discussed previously, the amount of leakage that bypasses the divider barrier will not be affected by the restoration of the steam generator compartment roofs.
The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries between upper and lower containment. Hence, the worse-case accident conditions for the containment will not be affected by the proposed modifications.
Therefore, a significant reduction in the margin to safety is not created by this modification.
Page 42 of 43
Topical Report 24370-TR-002-A V.
ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.
Page 43 of 43
Topical Report 24370-TR-O2-A Appendix B Historical Review Information Appendix B Table of Contents NRC Letter to TVA dated January 8, 2003, Summary of December 23, 2002 Meeting with TVA to Discuss the Sequoyah Unit I Steam Generator Compartment Roof Modifications...............................................................
43b-43h NRC Letter to TVA dated January 10, 2003, Sequoyah Nuclear Plant Unit 1 - Steam Generator Compartment Roof Modification Topical Report Technical Assessment... 43i-431 NRC Letter to TVA dated January 22, 2003, Summary of January 16, 2003 Meeting with TVA to Discuss the Sequoyah Unit 1 Steam Generator Compartment Roof Modifications...............................................................
43m-43p 1VA Letter to NRC dated February 14, 2003, Steam Generator Replacement Project -
Topical Report No. 24370-TR-C-003, Steam Generator Compartment Roof Modification, Revision 43q43bl Page 43a
Topical Report 24370-TR-002-A NRC Letter to TVA dated January 8, 2003, Summary of December 23, 2002 Meeting with TVA to Discuss the Sequoyah Unit I Steam Generator Compartment Roof Modifications Page 43b
JVU1. LEMFl 1=ULLA I I..Urk'IIVII;iiN 0.,,),
3 WASHINGTON. D.C. 20555-0001 Jaiuary 8, 2003 LICENSEE:
Tennessee Valley Authority FACILITY:
Sequoyah Nuclear Plant, Unit 1
SUBJECT:
SUMMARY
OF DECEMBER 23,2002, MEETING WITH TVA TO CD1SCU11IE SEQUOYAH UNIT 1 STEAM GENERATOR COMPARTMENT RtYF MODIFICATIONS (MB5387)
On December 23, 2002, the U.S. Nuclear Regulatory Commission (NRC) staff met with representatives of Tennessee Valley Authority (TVA) and Bechtel at the NRC Headquarters in Rockville, Maryland. TVA requested this meeting to discuss altematives to the through-bolted splice-plated methodology proposed in Bechtel's Topical Report No. 24370-TR-C-003 [ADAMS Accession No. ML020910792]. The NRC staff has reviewed the topical report and has concluded that the through-bolted splice-plate connection methodology to repair the steam generator (SG) compartment roof during the replacement of the SGs for Sequoyah Nuclear Plant Unit 1, was inadequate [ADAMS Accession No. ML023530628].
This meeting was classified as a Category 1 meeting, which provided an opportunity for members of the public to communicate with the NRC staff after the business portion, but before the meeting was adjoumed. However, there were no members of the public in attendance. is a list of attendees, and Enclosure 2 is a copy of TVA's handout distributed during the meeting.
The licensee discussed four altematives for the repair. The alternatives proposed used various combinations of steel reinforcement and poured concrete to reestablish the SG compartment roofs. The licensee indicated their intention to finalize two of the four alternatives. As a result, TVA requested another meeting to present their detailed design for the two altematives. The participants agreed to meet on January 16, 2003.
The staff noted that the above meeting was beneficial in gaining an understanding of TVA's plans and activities regarding the pending SG replacement for Sequoyah Unit 1.
9cNcw-A Raj K. Anand, Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-327
Enclosures:
As stated
-cc w/encls: See next page 43
ATTENDANCE LIST NUCLEAR REGULATORY COMMISSION (NRC) MEETING WITH TENNESSEE VALLEY AUTHORITY (TVA) AND BECHTEL Sr AT NRC HEADQUARTERS DECEMBER 23. 2002 NRC Ronald Hernan Raj Anand Eva Brown John Ma TVA Pedro Salas Dennis Lundy Paul Trudel David Ryder Bechtel J. V. Smith Eugene Thomas Myron Anderson ENCLOSURE 1
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Topical Report 24370-TR-002-A NRC Letter to TVA dated January 10, 2003, Sequoyah Nuclear Plant Unit I -
Steam Generator Compartment Roof Modification Topical Report Technical Assessment Page 43i
As NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 10, 2003 Mr. J. A. Scalice Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
SEQUOYAH NUCLEAR PLANT UNIT 1-STEAM GENERATOR COMPARTMENT ROOF MODIFICATION TOPICAL REPORT TECHNICAL ASSESSMENT (TAC NO. MB5387)
Dear Mr. Scalice:
On March 28, 2002, the Tennessee Valley Authority submitted a topical report to the U.S.
Nuclear Regulatory Commission (NRC) for review and approval. The topical report provided an alternate methodology for the reconstruction of the steam generator compartment roof during the Sequoyah Nuclear Plant (SQN) Unit 1 steam generator (SG) replacement project. The proposed method involved cutting four pieces of concrete roof slabs and reattaching them to the remaining uncut concrete roof slabs by using "through-bolted splice-plate connections,"
located along the concrete cut line. Based on the information provided by the licensee, the NRC staff determined that the assumptions made for the analysis performed did not reflect the actual boundary condition at or near the concrete cut line. The NRC staff has, therefore, concluded that the proposed repair of the SG compartment roof is inadequate in that it degrades the capability of the roof to withstand its design loads. The NRC staff's assessment is enclosed.
As discussed with Mr. Pedro Salas on December 23, 2002, the NRC has scheduled a public meeting on January 16, 2003, to discuss in greater detail SQN's proposed alternatives.
If you have any questions, please feel free to call Ms. Eva Brown at (301) 415-2315 or Mr. Allen Howe at (301) 415-2024.
Sincerely, Raj Anand, Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-327
Enclosure:
See next page cc wlenclosure: See next page 4k1
NUCLEAR REGULA10HY UlMMISSIUN 4
i WASHINGTON, D.C. 20555-0001 NRC STAFFS ASSESSMENT OF TVA TOPICAL REPORT NO. 24370-TR-C-003. "STEAM GENERATOR COMPARTMENT ROOF MODIFICATION" TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNIT 1 DOCKET NO. 50-327
1.0 INTRODUCTION
By a letter dated March 28, 2002, Tennessee Valley Authority (the licensee) of Sequoyah Nuclear Plant (SON) submitted a topical report for an alternate methodology for the reconstruction of the steam generator (SG) compartment concrete roof. The U.S. Nuclear Regulatory Commission (NRC) staff and the licensee held meetings on October 24 and December 23, 2002, where the licensee provided handout information conceming this submittal.
2.0
SUMMARY
OF SUBMITTALS The topical report proposed an alternate methodology for the reconstruction of the SG compartment concrete roof that would be cut to enable the removal of the old SGs and installation of new SGs. The proposed method involved reattaching the original cut section of the concrete roof slab to the remaining uncut concrete roof slab by using 23 pieces of through-bolted splice-plate connections," located along the concrete cut line.
The thickness of the roof slab varied from 2 feet (ft) 3 inches to 3 ft. The proposed detail of the through-bolted splice-plate connection involved (1) the use of two 3-inch thick steel plates, 2 ft long in the, direction perpendicular to the cut line with varied widths along the cut line; one at the top side of the roof slab and the other at the bottom side, (2) maintaining a 1-inch gap space filled with grout between the cut section of the concrete roof slab and the uncut section of the concrete roof slab, and (3) placing a steel bolt vertically near the center of the plates through the 1-inch gap space and tying the two plates together by a nut and a washer at the bolt ends.
3.0 -TECHNICAL ASSESSMENT The licensee assumed that the proposed details of the through-bolted splice-plate connection could be modeled as a positive connection in one analysis and a cantilevered support in a second analysis. However, the NRC staff found the licensee's assumptions incorrect and the proposed details of the through-bolted splice-plate connection unacceptable. The staff's reasons are presented below.
Enclosure 431l A splice-plate is commonly used to join two separate structural members together. A splice is utilized when a single piece of steel is bolted to two separate pieces of steel or concrete roof slabs in this case. The splice transfers force from one structural member to the other through bolts and the splice plate. A splice is considered as a positrve connection if it can reliably or positively transfer force from one structural member to the other. The licensee assumed the through-bolted splice-plate connection along the cut line would act as a 'hinge' boundary condition in one analysis. The assumption in a hinge boundary condition is that the two structural members joined by the hinge may rotate around the hinge, but remain connected by the hinge so that in-plane forces will be positively transferred through the hinge. However, the bolt in the proposed detail of the through-bolted splice-plate connection is not installed through either piece of the structural concrete roof slabs, but, in the 1-inch gap space. Under such a condition, any horizontal force transfer between the two structural concrete roof slabs is through the frictional force between the steel plates and the concrete. This frictional force is unreliable and small, thus the hinge boundary condition assumption for the through-bolted splice-plate connections along the cut line is unrealistic because horizontal forces cannot reliably and positively transfer between the two structural concrete roof slabs.
The licensee performed another analysis by assuming that the proposed through-bolted splice-plate connection would function as a cantilevered support for the cut section of the concrete roof slab. A cantilevered support requires the support itself to be firmly anchored on one end so that its deformation or deflection under load can be reliably predicated. However, the bolt in the proposed detail is not through-bolted in the uncut portion of the structural concrete roof slab, but in the 1-inch grouted gap space. The grout acts as space filler only and cannot be counted as structural material because there is no reinforcing bars to bond the grout to the structural concrete roof slab. The grout is likely to shrink as the grout ages and would likely crack when subjected to thermal loads. Therefore, the grout cannot offer a solid bearing condition for the bolt, and the bolt may move within the 1-inch gap space in the radial direction and in the direction along the concrete cut line. Furthermore, the splice-plate can rotate freely in any direction once the small and unreliable frictional force between the steel and concrete is overcome by force generated due to seismic loads. The potential movement of the bolt and plates invalidates the analysis assumptions and disqualifies the proposed detail of the through-bolted splice-plate as being a cantilevered support.
4.0 CONCLUSION
Based on the above discussion, the NRC staff finds that the proposed location of the bolt in the 1-inch gap space to be unacceptable for reconnecting the cut portion of the roof slab since it does not provide positive connection to the other portion of the roof. The assumptions made for the analysis performed by the licensee do not reflect the actual boundary condition at or near the concrete cut line, because the actual condition is neither a positive connection nor a canblevered support The staff, therefore, concludes that the proposed repair of the SG compartment roof is inadequate in that it degrades the capability of the roof to withstand its design loads.
- 43D~
Topical Report 24370-TR-002-A NRC Letter to TVA dated January 22, 2003, Summary of January 16, 2003 Meeting with TVA to Discuss the Sequoyah Unit 1 Steam Generator Compartment Roof Modifications Page 43m
C WASHINGTON, D.C. 20555-0001 Januar 22, 2003 LICENSEE.
Tennessee Valley Authority FACILITY:
Sequoyah Nuclear Plant, Unit-1
SUBJECT:
SUMMARY
OF JANUARY 16, 2003, MEETING WITH TVA TO bISCUSS THE SEQUOYAH UNIT 1 STEAM GENERATOR COMPARTMENT ROOF MODIFICATIONS (MB5387)
On January 16, 2003, the U.S. Nuclear Regulatory Commission (NRC) staff met with representatives of Tenriessee7Valley Authority (TVA) and Bechtel at the NRC Headquarters in Rockville, Maryland. TVA requested this meeting to discuss an altemative to the through-bolted splice-plated methodology proposed in Bechtel's Topical Report No. 24370-TR-C-003 [ADAMS Accession No. ML020910792]. By letter dated January 10, 2003 the NRC staff informed TVA that the through-bolted splice-plate c6nnection methodology to repair the steam generator (SG) compartrhent roof during the replacernent of the SGs for Sequoyah Nuclear Plant Unit 1, was inadequate [ADAMS Accession No. ML023530628].
This meeting was classified as a Category I meeting, which provided an opportunity for members of the public to communicate with the NRC staff after the businest portion, but before the meeting was adjouned. However, there Were no members of the public in attendance. is a list of attendees, and Enclosure 2 is a copy of TVA's handout distributed during the meeting.
The licensee discussed the selected altemative for the repair. The altemative proposed uses steel reinforced beams to reestablish the SG compartment roofs. The licensee indicated their intention to finalize the design for this alternative. As a result, TVA requested another meeting.
to present the final design. The participants agreed to meet again on February 12, 2003.
The staff noted that the above meeting was beneficial in gaining an understanding of TVA's plans and activities regarding the pending SG replacement for' Sequoyah Unit 1.
Raj K. Anand, Project Manager, Section 2 Project Directorate II Division of Ucensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-327
Enclosures:
As stated cc wlencis: See next page
ATTENDANCE LIST NUCLEAR REGULATORY COMMISSION (NRC) MEETING WITH TENNESSEE VALLEY AUTHORITY (TVA) AND BECHTEL AT NRC HEADQUARTERS JANUARY 16.2002 NRC Allen Howe Raj Anand Eva Brown John Ma TVA Pedro Salas Dennis Lundy Paul Trudel David Ryder Bechtel J. V. Smith Eugene Thomas Myron Anderson ENCLOSURE 1
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Topical Report 2437U-K-U-UUS-A TVA Letter to NRC dated February 14, 2003, Steam Generator Replacement Project -
Topical Report No. 24370-TR-C-003, "Steam Generator Compartment Roof Modification, Revision 1" Page 43q
Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 February 14, 2003 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of
)
Docket No.
50-327 Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT -
STEAM GENERATOR REPLACEMENT PROJECT -
TOPICAL REPORT NO. 24370-TR-C-003, "STEAM GENERATOR COMPARTMENT ROOF MODIFICATION, REVISION 1"
Reference:
NRC letter to TVA dated January 10, 2003, Sequoyah Nuclear Plant Unit 1 -
Steam Generator Compartment Roof Modification Tpical Report Technical Assessment (TAC NO. MB5387)
In response to the reference letter, TVA is submitting for NRC review and approval Revision 1 of the subject topical report.
Revision 1 of the topical report includes a new design and analysis of the reconstruction method for the Unit 1 steam generator compartment roof modification.
The reconstruction method provides for additional combinations of steel reinforcement and poured concrete that ensure the original design boundary conditions at the concrete cut line are met. The enclosure to this letter contains Topical Report No. 24370-TR-C-003, Revision 1.
NRC and TVA met on February 12, 2003, to discuss the topical report.
At the conclusion of the meeting, TVA was informed there were no outstanding regulatpry issues, and that NRC expected to issue its approval by mid-March 2003.
TVA needs the NRC approval before it can proceed with the modification work activities once the unit is removed from service on March 16, 2003.
4-3f
U.S. Nuclear Regulatory Commission Page 2 February 14, 2003 This letter is being sent in accordance with NRC RIS 2001-05.
There are no commitments contained in this letter.
If you have any questions about this change, please telephone me at (423) 843-7170 or J. D. Smith at (423) 843-6672.
~E>ro gs Lice ing and Industry Affairs Manager Enclosures cc (Enclosures):
Mr. Raj K. Anand, Senior Project Manager U.S. Nuclear Regulatory Commission MS 0-8G9 One White. Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 4-3 s
ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)
UNIT 1 DOCKET NO.
327 TOPICAL REPORT NO. 24370-TR-003, REVISION 1 STEAM GENERATOR COMPARTMENT ROOF MODIFICATION 4-4 I1
SEQUOYAN UNIT 1 STEAM GENERATOR REPLACEMENT STEAM GENERATOR COMPARTMENT ROOF MOIDIFICATION TOPICAL REPORT l
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SEQUOYAH UNIT I STEAM GENERATOR REPLACEMENT STEAM GENERATOR COMPARTMENT ROOF MODIFICATION TOPICAL REPORT 4,7V
Topical Report 24370-TR-C-003 Table of Contents 1.0 Abstract....................................................................
3 2.0 Introduction...................................................................3 3.0 Objectives................................................................... 8 4.0 Regulatory Requirements/Criteria for Ice Condenser Divider Barriers.............
................... 8 4.1 SRP Section 3.8.3 - Concrete and Steel Internal Structures of Steel or Concrete Containments....................................................................
8 4.2 SRP Section 6.2.1.2 - Subcompartment Analysis............................................................. 11 5.0 Description of Concrete Work to be Performed.....................................
11 6.0 Description of Existing Design Basis and Original Analyses.............................................. 12 7.0 Description of Modification to the Structure and New Analyses......................................... 25 8.0 Results of New Analyses....................................................................
34 9.0 Summary and Conclusions....................................................................
36 10.0 References................................................................... 37 Appendices A
No Significant Hazards Consideration Determination........................................................ 39 Tables 6-1 (UFSAR Table 3.8.3-1) Loading Combinations and Allowable Stresses for the Interior Concrete Structure...................................................................
15 6-2 (UFSAR Table 3.8.3-2) Loading Combinations and Load Factors.16 6-3 (UFSAR Table 3.8.3-6) Original Design Stress Margin Table 3.8.3-1 Criteria Versus Table 3.8.3-2 Criteria (4).17 7-1 (Table CC-3230-1 from ASME Section III, Division 2, 1975) Load Combinations and Load Factors.................
29 7-2 Loading Combinations, Load Factors and Allowable Stresses for SG Compartment Roof Modification (5)(6)...............
30 9-1 Differences Between Original and New Steam Generator Compartment Analyses.......... 36 Figures 2-1 Equipment - Reactor Building (UFSAR Figure 1.2.3-11)..................................................... 5 2-2 Equipment - Reactor Building (UFSAR Figure 1.2.3-12)..................................................... 6 2-3 Equipment - Reactor Building (UFSAR Figure 1.2.3-13)..................................................... 7 6-2 Concrete Steam Generator and Pressurizer Compartment - Reinforcement.................... 19 6-3 Concrete Steam Generator and Pressurizer Compartment - Reiriforcement.................... 20 6-4 Concrete Crane Wall Outline................................................................
21 6-5 Temperature Gradient (UFSAR Figure 3.8.3-2)................................................................ 22 6-6 Steam Generators 1 and 4 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-1)...............................
........ 23 6-7 Steam Generators 2 and 3 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-2)................................................................ 24 7-1 Sensitivity of Peak Compression Pressure to Deck Bypass (UFSAR Figure 6.2.1-22)................................................................
31 7-2 Steam Generator Compartment Roof Splice Plate Layout and Connection Details...........
2 7-3 Finite Element Model SGEI' and tSGE2" and Element Groups and Global Coordinate Systems (Reference 6)................................................................
33 8-1 Areas of Critical Stresses................................................................
36 Page2of43
Topical Report 24370-TR-C-003 1.0 Abstract The four steam generators of the Sequoyah Nuclear Plant Unit I will be replaced during the spring of 2003. To support the replacement of the old steam generators (OSGs) with the replacement steam generators (RSGs), access openings will be created in the roof of the steam generator (SG) compartments inside containment. An appropriately sized access opening will be made in each SG compartment roof by cutting out a secbon of concrete from the roof of the compartments using wire saws. Upon completon of installabon of the RSGs, the original cut concrete section (plug) of the SG compartment roof will be reattached to the respective comparbrnent roof by means of through-bolted connections, compnsed of steel connection franmes and threaded rods. The p l'g will be attached to the top and botton connection frames using four 2-inch diameter threaded ro-ds tat are installed in core bore holes through the p g. T he top and bottom connection framres.wi clan-ap the concrete plug to then complimentary portion of the SG comoararint using six 2-1t2 inch and eighiteen 2-inch diameter threaded rods. The threar:ed rods are instaUled in te core bore holes located around th.e perimeter of the concrete plug and will be pre-tensioned. A ser_s of steel sh`ms w-L be driven into the annular space (created at he cut ine' and mechanIcalIy locked into place. The annular space wiill be grouted.
The original design of the SG compartment was based in part on the load combinations defined in Table 3.8.3-2 of the UFSAR. This UFSAR table is based on Table CC-3200-1 of the Proposed ASME Section 111, Division 2, 1973, Proposed Standard Code for Concrete Reactor Vessels and Containments, Section CC-3000 which was issued in 1973 (the time of original design) by the ACI-ASME Committee on Concrete Pressure Components for Nuclear Service, for trial use and comment The purpose of this topical report is to provide the technical basis for use of the slightly modified load combinations and allowable stresses in the adopted 1975 edition of ASME Section liI, Division 2, instead of those described in the UFSAR. Analyses performed using the adopted ASME load combinatons have shown that the modified SG compartment roof design will not exceed allowable stesses in the concrete, rebar and structural steel when subjected to the design basis differential pressure of 24 psi combined with the other design basis loads such as seismic, pipe thrust, dead load and live load. This design differential pressure is approximately 23% higher than the maximum compartment accident pressure differential of 19.52 psi.
2.0 Introduction The steam generator compartments are designed and constructed as cast in-place reinforced concrete structures. As indicated in UFSAR Section 3.8.3.6.1, the minimum compressive strength of the containment interior concrete structures is 5000 psi.
UFSAR Section 3.8.3.1.7 describes the steam generator compartments. Two double-compartment structures house the four steam generators in pairs on opposite sides of the containment. For each pair of steam generators, divider barrier walls exist around the two steam generators and are capped with a three-foot thick con-rete roof spanning over the steam generators from the crane wall. A wall between each pair of steam generators extends from the divider walls to the crane wall, completing the double compartment The center wall does not extend up to the concrete roof. This area above the wall, except for the portions occupied by the main steam pipe restraint beam, reduces the compartment pressure buildup in a single compartment by venting the steam to the other compartment. These features are depicted on UFSAR Figures 1.2.3-11, 1.2.3-12, and 12.3-13 (provided as Figures 2-1, 2-2, and 2-3, respectively).
Page3of4-3
Topical Report 24370-TR-C-003 The steam generator compartments form part of the interior concrete structure that is referred to as the divider barrier. UFSAR Section 3.8.3.1.1 defines the divider barrier as that part of the interior structure that separates the upper containment from the lower containment. This barrier forces steam that is released from a LOCA/ DBA to pass through the ice condenser. The failure of any part of the divider barrier is considered critical since it would allow LOCA/DBA steam to bypass the ice condenser, thereby increasing the pressure within the primary containment. The original design loads for the compartment concrete were based on preliminary accident pressurization calculations. Conservative design basis loads were used in the original design to bound potential changes between the preliminary and the final pressurizaffon analysis results.
UFSAR Section 3.8.3.2 detaifs the codes and standards to which the intemal concrete structures were designed. The load combinations and allowable stresses for the intemal concrete structures including the divider barrier are detailed in UFSAR Tables 3.8.3-1 and 3.8.3-2 (provided as Tables 6-1 and 6-2, respectively).
There are no Technical Specifications (TSs) associated specifically with the steam generator compartments. However, tere are TSs associated with other nortiors f th divider barrier. TSs 34.-.5.3, 314.6.5.5. and 3!4.6.5.9 address the ice condenser doors.
divider barmer personn_l access doors an equipment hatches, and divider barrier seat, respectively. The planned changes to te steam generator compartment roof, wil restore the leakia htness of the rm and will not affect tfe ice condenser doors, divider b-.rier personnel access doors and equ.msnt hatches, or divider barrier sea.. Therefore, the TSs will not be affected by the planned changes to the steam generator compartment roof portion of the divider barrier.
Page 4 of 43
Topical Report 24370-TR-VC003 Building e
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Topical Report 24370.TR-C-003
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Topical Report 24370.TR-C.003 Page 7 of 43
Topical Report 24370-TR-C-003 3.0 Objectives To describe the current steam generator compartment roof design and proposed modification.
To present data that supports and justifies the reinstallation of the cut steam generator compartment roof concrete sections using frames installed an tte top and bottom of the sectiDn and then trc.gg-olited toger.
To support a license amendment for using load combinations and allowables for reinf-orced concrete provided in uadoptedu ASME Section III, DMsion 2, 1975 instead of the load combinations provided in Proposed' ASME Section III, Division 2, 1973.
4.0 Regulatory RequirementslCriteria for Ice Condenser Divider Barriers Detailed below are regulatory requirements/criteria that are relevant to the design of the divider barrier portion of internal structures in an ice condenser containment. Since the SG compartment roof is part of the divider barrier, the planned modification to the roof must conform to the requirements/criteria below. Following each requirement/criteria is an italicized discussion of how the requirement/criteria is met and/or where the requirement/criteria is addressed within this topical report.
4.1 SRP Section 3.8.3 - Concrete and Steel Internal Structures of Steel or Concrete Containments Standard Review Plan (SRP) 3.8.3 details the information required for NRC review of containment intemal structures and the criteria for NRC acceptance of these structures.
This review is performed to assure conformance with the requirements of IOCFR50.55a and 1 OCFR5O, Appendix A, General Design Criteria (GDC) 1, 2, 4, 5, and 50. The parts of these regulations that are relevant to the divider barrier design are:
1 ) 10CFR50.55a and GDC 1 as they relate to the divider barrier being designed, fabricated, executed, and tested to quality standards commensurate with the importance of the safety function to be performed.
The quality standards used in the design, fabrication, execution, and testing of the modified divider barier are the same or equivalent to those used for the original divider bamer.
- 2) GDC 2 as it relates to the design of the divider barrier being capable to withstand the most severe earthquake and appropriate combination of all loads.
The modified SG compartment roof has been designed for the same loads and load combinations as the original design (described in Section 6.0), except as noted in Section 7.0. The results described in Section 8.0 show that it is capable of withstanding the most severe earthquake loads and the appropriate combination of other loads.
- 3) GDC 4 as it relates to the divider barrier being capable of withstanding the dynamic effects of equipment failures including missiles, pipe whips and blowdown loads associated with the loss of coolant accidents.
Page 8 of 43 0 3 c
Topical Report 24370-TR-C-003 As described in Sections 7.0 and 8.0, the modified SG compartment design has been evaluated for the dynamic effects of pipe whip andjet impingement loads following a pipe break inside the SG compartment.
- 4) GDC 5 as it relates to the sharing of structures important to safety.
The dividerbarrieris not a shared structure. Therefore, conformance to GDC 5is not applicable for the modified SG compartment
- 5) GDC 50 as it relates to the divider barrier being designed with sufficient margin of safety to accommodate appropriate design loads.
As described in Sections 7.0 and 8.0, the modified SG compartment designis capable of withstanding the same design pressure as the original SG compartment design without exceeding allowable stresses in te concrete, rebar and stictural sfee!. This design pressure is 23% greater than the maximum calculated post-LOCA differentialpressure. Since the design pressure and the maximum calculated accident pressure have not changed, there is no reduction in the margin of safety for the modified SG compartment design.
The descriptive information provided is considered acceptable if it meets the minimum requirements set forth in Section 3.8.3.1 of NRC Regulatory Guide (RG) 1.70. This RG indicates that the descriptive information relevant to the divider barrier that should be provided includes plan and section views to define the primary structural aspects and elements relied upon to perform the safety-related function of the divider barrier.
General arrangement diagrams and the principal features of the divider barrier should be described.
A description of the revised SG compartment roof design is provided in Section 7.0.
Figure 7-2 provides details for the frames to be ;ns.a.'d o.n the
'tp and the botr of the compartment conrete section and the layou of the 6oncflc!ion through-bolts. Other aspects of the divider barrier design will remain as described in the Sequoyah UFSAR.
An update to the UFSAR will be prepared to reflect the revised Unit 1 SG compartment roof design.
The design, materials, fabrication, erection, inspection, testing, and in-service surveillance of the divider barrier are covered by the following codes, standards, and regulatory guides:
- 1) ACI-349 As indicated in Section 1.1 of Part I of ACI-349, structures covered byASME Section i, Division 2 are specifically excluded from the requirements of this standard. As discussed in Section 7.0, the modified SG compartment roof design conforms to ASME Section 11, Division 2. Therefore, this standard is not applicable to the modified SG compartment roof design.
- 2) ASME Section III, Division 2 Conformance of the original design of the SG compartment roofs to the ASME Code is discussed in Section 6.0. As detailed in Section 7.0, the relnf orced conc.ree part of the modified SG compartment roof design is consistent with the adopted edition of Page 9 of 43 4;-fit I0
Topical Report 24370-TR-C-003 the ASME Code. The basis andjustffication for use of the later edition of the Code is also provided in Section 7.0.
- 3) ANSI N45.2.5, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants'.
Addressed under the response to RG 1.94 below.
- 4) Regulatory Guide 1.94, Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants' RG 1.94 endorses ANSI N45.2.5-74, but specifies additional requirements related to use of other codes and standards, RG 1.55, concrete consolidation, and rebar splice welding. The TVA Nuclear Quality Assurance Plan (NQAP) (Reference 15) follows this regulatory guide, but also provides altematives to the regulatory guide guidance.
7he installation, inspection, and testing activities associated with the through-bol-ted connection fame modification to the SG compartment roofs will conform to the RG 1.94 guidance or the alternatives allowed by the TVA NQAP.
- 5) Regulatory Guide 1.142, Safety-Related Concrete Structures for Nuclear Power Plants' RG 1.142 endorses ACI 349-76. As discussed in Section 7.0, the modified SG compartment roof design conforms to ASME Section lll, Division 2 (1975). As such, the modified SG compartment roof design is not required to be evaluated against the requirements of RG 1.142 orACI 349-76.
The divider barrier design is reviewed to determine if the loads and load combinations used meet the acceptance criteria. For concrete pressure-resisting portions of the divider barrier, the loads and load combinations of Article CC-3000 of ASME Section III, Division 2 Code apply.
As described in Section 7.0, the load combinations of Table CC-3230-1 of Article CC-3000 of ASME Section 111, Division 2, 1975 were usedin the evaluation of the modified SG compartment roof design.
The design and analysis procedures utilized for the divider barrier are acceptable if they are in accordance with ACI 318.
As described in Section 6.0, the original SG compartment structural design is in compliance with a combination of ACI 318 and the Proposed ASME Section ll, Division 2, 1973. Section 7.0 describes how the modified SG compartment design complies with ASME Section ll, Division 2, 1975 (ACI 359-74).
The structural acceptance criteria for the divider barrier are acceptable if the specified stress and strain limits are in accordance with Subsection CC-3430 of ASME Section III, Division 2. The 33-1/3% increase in allowable stresses is only permitted for temperature loads and not for OBE seismic or wind loads.
Page lO of 43 4:5oLp-
Topical Report 24370-TR-C-003 As described in Section 8.0, he stresses in he rainforceda oncrete of the modified SG compartment roof stresses under the load combinations defined in Table CC-3230-1 of ASME Section III, Division 2, 1975 are less than or equal to the stress allowables defined in Section CC-3400 of ASME Section lil, Division 2, 1975. The 33-1/3%
increase in allowable stresses was only used for temperature loads. he strctu.rafst.eel rnugh-olSted coannection, frmaes are desioned in accordance,trh Reforance 3.
The specified materials of construction and quality control programs for the divider barrier are reviewed. Information on the materials used and the extent of compliance with ANSI N45.2.5 should be provided to support this review. Information on special, new, or unique construction techniques should also be pfovided in order to assess their effects on the structural integrity of the completed divider barrier.
The materials used in the modified SG compartment design are detailed in Section 7.0.
Installation, inspection and testing of the modified SG compartment roof will conform to the quality assurance requirements of ANSI N45.2.5. Other than tensioning or preloading the threaded rods, there are no special, new, or unique construction techniques that will be used during installation of the modified SG compartment roof.
4.2 SRP Section 6.2.1.2 - Subcompartment Analysis SRP 6.2.1.2 details the information required for NRC review of the design differential pressure analyses for containment subcompartments. This review is performed to assure conformance with the requirements of 1 OCFR50, Appendix A, GDC 4 and 50.
The parts of these regulations that are relevant to the divider barrier design are:
- 1) GDC 4 as it relates to the ability of the divider barrier to accommodate the dynamic effects of missiles, pipe whipping, and discharging fluids that may occur during normal operations or during an accident.
As described in Sections 7.0 and 8.0, the modified SG compartment design has been evaluated for the dynamic effects of pipe whip andjet impingement loads following a pipe break inside the SG compartment.
- 2) GDC 50 as it relates tD the divider barrier being designed with sufficient margin to prevent fracture of the barrier due to pressure differential across the barrier.
As described in Sections 7.0 and 8.0, the modified SG compartment design is capable of withstanding the same design pressure as the original SG compartment design without exceeding te allowable sztroses in the concrete, rebar or structurai steel. This design pressure is 2% greater than the maximum calculated post-LOCA differential pressure.
5.0 Description of Concrete Work to be Performed The modification of the steam generator compartment roof will first entail cutting out a secton of the concrete roof over each steam generator. Cutting of the concrete will be accomplished by first core-boring holes around the perimeter of the cut, then using wire saws to cut the straight lines between the cores. The cores also serve as the bolt holes for the through-bofts used to connect the concrete section back to the structure. After removal, the edges of the concrete section will be bush-hammered to provide an annular gap of about 1 upon reinstallation of the concrete section. Each concrete section will be Page11of43 I
Topical Report 24370-TR-C-003 sized to allow the removal and replacement of the steam generator in the compartment The concrete section will be re-installed once the RSG and associated piping are placed inside the compartment. Restoraton of the SG compartments will involve re-attaching the cut out concrete sectons to the existing structure using a op and bottom frame santwiching the cut outl concrete sections antd connecting the frames with through boted threaded rods around the perime+ter of the cut Tapered-steel shims wfll be placed in tie annular gap between the oncrete sections and the bolt holes and annular space will be grouted using non-shrink grout. Additional details of the through-bolted connection framne design and the capability of the non-shrink grout to limit bypass leakage through the divider barrier is provided in Section 7.0.
The steam generator compartments have been re-evaluated, with specific focus on the modified roof, for the effects on structural response and found to be acceptable. The throuch-bolted connection frames and the tapered steel shims have been designed to be adequate for the applicable design loadings. Details of these evaluations are provided in Section 7.0. The design of the repaired steam generator compartments is in compliance with the requirements of Reference 2.
6.0 Description of Existing Design Basis and Original Analyses The original design bases of the concrete internal structures, which includes the SG compartments, is discussed in detail in Section 3.8.3 of the UFSAR and Section 2.9 of Reference 2. UFSAR Section 3.8.3.2 states that the structural design of the interior conrete structures is in compliance with the American Concrete Institute (ACI) 318-63 Building Code Working Stress Design Requirements for load combinations shown in UFSAR Table 3.8.3-1 (provided as Table 6-1), including LOCA calculated pressures with moisture entrainment received from the NSSS contractor, or the ACI-ASME (ACI 359)
Artcle CC3000 document, Proposed Standard Code for Concrete Reactor Vessels and Containments" (Proposed ASME Section IlIl, Division 2, 1973), and ACI 318-71 for the load combinations shown in Table 3.8.3-2 (provided as Table 6-2), including LOCA calculated pressure. Section 3.8.3.2 of the UFSAR also states that the design and construction of the interior concrete structures is based on the appropriate sections of NRC Standard Review Plan 6.2.1.2, "Subcompartment Analysis'.
The original design loads for the SG compartment concrete were based on preliminary accident pressurization calculations. Because of the uncertainties associated with these preliminary accident analyses, conservative design basis loads were used in the original design to bound potential changes between the preliminary and the final pressurization analysis results. The preliminary accident pressurization loads were higher than the final accident loads, which resulted in a conservative SG compartment design.
The maximum differential pressure used in the onginal design was 21.3 psi which is a 25% increase over the design basis accident (DBA) differential pressure of -17 psi (Reference 5) for the SG compartment provided by Westinghouse (i.e., 1.25 x 17 psi).
The original design was based on loads, load combinations and allowable stresses documented in Table 3.8.3-1 of the UFSAR (provided as Table 6-1).
As detailed in UFSAR Section 3.8.3.4.1, each component of the interior concrete structure was evaluated individually. Its boundary conditions and degrees of fixity were established by comparative stiffness; loads were applied, and moments, shears, and direct loads determined by either moment distribution or finite element methods of analysis. UFSAR Section 3.8.3.4.1 also states that reinforcing steel was proportioned Page 12 of 43
Topical Report 24370-TR-C-003 for the component sections in accordance with UFSAR Tables 3.8.3-1 or 3.8.3-2 and the ultimate strength provisions of ACI 318-71 Building Code were used to check the combined effects of torsion, shear, and direct tensile loads.
At the construction permit stage, a factor of 1.4 was applied to the DBA pressure provided by Westinghouse. The structural adequacy of the steam generator compartments was checked based on the 40 percent margin and the recommendations of the ACI/ASME Joint Committee contained in Proposed Standard Code for Concrete Reactor Vessels and Containments". Accordingly, the SG compartment design was evaluated for a maximum design intemaldifferential pressure of 24 psi (i.e., 1.4 x 17 psi) using loads, load combinations, and allowable stresses documented in UFSAR Table 3.8.3-2 (provided as Table 6-2). This is reflected in Section 3.8.3.4.1 of the UFSAR, which indicates that a factor of 1.4 was applied to the design pressures resulting fror a LOCA during the construction stage. The results are tabulated in UFSAR Table 3.8.3-6 (provided as Table 6-3).
NRC Standard Review Plan 6.2.1.2, Subcompartment Analysis,Section II.B.5, addresses the application of peak differential pressure to be used in the design of the subcompartment. At the construction permit stage, a factor of 1.4 is applied to the calculated peak differential pressure to establish the differential pressure used for design of the subcompartment. At the operating permit stage, the calculated peak differential pressure should not exceed the design pressure. As noted in UFSAR Section 3.8.3.3 and consistent with SRP 6.2.1.2.Section II.B.5, the maximum calculated differental compartment pressures were increased by 40% to account for uncertainties. At the Operating License stage, the design pressures equaled or exceeded the peak calculated differential pressure. Therefore, the design conformed to the requirements of SRP 6.2.1.2.
UFSAR Section 6.2.1.3.10 indicates that the SG compartments were originally designed for two separate pressure loadings. These loadings are (1) a 24 psi maximum internal differential pressure from a break in the main steam line and (2) a uniform intemal pressure of 43 psi. The SG compartments were also designed to resist the jet thrust force (910 kips on the roof per Reference 5) that would result following a main steam line break.
The argest-blow-down flow results from the severance of the main steam pipe. As indicated in UFSAR Section 3.6.7.6.3, postulated main steam line break locations are shown on UFSAR Figures 3.6.7-1 and 3.6.7-2 (provided as Figures 6-6 and 6-7.
respectively). Operating thermal conditions and accident thermal effects accompanying a pipe break (See UFSAR Figure 3.8.3-2, provided as Figure 6-5) were also accounted for.
The blow-down flow analysis of the main steam breaks described in Section 6.2.1.3.10 of the UFSAR resulted in a maximum pressure differential of 19.15 psi compared to the design differential pressure of 24 psi. The UFSAP analysis assumed the main steam flow restrictor is located downstream of the pipe break. Reanalysis of the mair stear lne break, based on the RSG design with the flow restrictor upstream of the pipe break, resulted in the maximum pressure differentia' increasing to 19.52 psi. Thus, the design pressure exceeds the rnaximum calculated differential pressure by 2'%, and is therefore conservative.
Page13of43 43 ath
Topical Report 24370-TR-C-003 As stated in UFSAR Section 3.8.3.4.8, the SG compartment was also originally designed to resist a 43-psi hypothetical pressure from a reactor Coolant pipe break. This loading was used to provide a high degree of conservatism in the preliminary design of the SG compartment The center wall and the beam below the conc-ete roof are used as bumper points for main steam pipe whip restraints. These members restrain pipe whip in case of a pipe break and transmit forces to the roof and/or to the wall. It is noted that these whip restraints are bumpers that provide restraint against the pipe-whip in one direction only.
Additionally, they also provide lateral restraint by means of saddle/bracket devices.
The original design of the steam generator compartments, in particular, is documented in Reference 5 and summarized in UFSAR Section 3.8.3.4.8. The roof of the SG compartments was analyzed using a combined member-grid and flat plate finite element STRUDL model. Manual calculations were performed at various locations to confirm computer results. The inverted T-beam, which stiffens the rof, was analyzed for the dynamic effects of a main steam pipe breaking and loading the flange of the beam. The roof was also independently analyzed as a plate using the finite element plate-bending program, GENDEK 3. The roof was analyzed both as a beam-stiffened slab and a uniform slab, neglecting the effects of the beam. The edges of the roof were considered fixed.
From Reference 16 and Figure 6-1, the design compressive strength of the SG compartment concrete at 28 days is 5000 psi. Note that the estmated in-place design compressive strength of the SG compartment roof concrete at 90 days is 5700 psi (Reference 5, Sheets 2e and 20. The reinforcing used for the interior structures conforms to ASTM A615 Grade 60 (Reference UFSAR Section 3.8.3.2). Figures 6-2 and 6-3 provide additional details of the pre-modification design of the SG compartment roofs. This paragraph provides the historical data as to the required design strength and actual strength of the in-situ steam generator compartment concrete.
Page14of43 4Av
Topical Report 24370-TR-C-003 Loading Combinations Table 6-1 (UFSAR Table 3.8.3-1) and Allowable Stresses for the Interlor Concrete Structure f'c = Ultimate strength of concrete fy = Yield strength of reinforcement Page 15 of 43
.F IA COMBINATIONS LOADINGS 2
2A 3
3A 4
5 5A DEAD LOAD X
X X
X X
X LIVE LOAD X
X X
X X
NORMAL TEMP.
X X
X LOCA PRESSURE X
X X
X LOCA TEMP.
X X
X HYPOTHETICAL x
PRESSURE
% SSE X
SSE X
X X
PIPE FORCES X
INITIAL JET PIPE FORCES SATURATED x
(REDUCED) JET OR ANCHOR W.S.D. ALLOWABLE DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER DIVIDER OTHER STRESSES BARRIER BARRIER BARRIER BARRIER BARRIER BARRIER fe 0.45 rc 0.45 c 0.45 fc 0.45 rc 0.60 fec 0.75 rc 0.60 fe 0.75 c 0.60 fc 0.75 f'c fs 0.40 y 0.40 fy 0.50 fy 0.50 fy 0,72 fy 0.90 fy 0.72 fy 0.90 fy 0.72 fy 0.90 fy U.S.D. LOAD FACTORS 1.25 1.0 1.0 1.25 1.0 1.25 1.0 I
I
Topical Report 24370-TR-C-003 Table 6-2 (UFSAR Table 3.8.3-2)
Loading Combinations and Load Factors
- 1. Includes all temporary construction loading during and after construction of containment.
- 2. V is lower for tension members and is essentially the same as given by (ACI 318-71).
LOADS NOMENCLATURE:
D Dead loads, or their related internal moments and forces Feqo Operating basis earthquake Feqs Design basis earthquake L
Live load, or their related intemal moments and forces PI Accidentlincident maximum pressure R,
Piping loads during operating conditions Ra Piping loads due to increased temperature resulting from the design accident Ta Thermal loads under the thermal conditions generated by the postulated break and including T.
T.
Operational temperature Y,
Reaction load on broken pipe due to fluid discharge The term design basis earthquake' has the same meaning as the term 'safe shutdown earthquake.!
Page 16 of 43 4: c,k Category Ta D
Ll)
Pa To Fego Feqs Ro Ra Yr Allowable
~~~~~Stre-sses Service:
Const 1.0 1.0 1.0 (Fiexure)
Normal 1.0 1.0 1.0 1.0 or 1.0 f=0.45 fc Factored:
f= 0.50 fy (Shear)
Extreme 1.0 1.0 1.0 1.0 1.0 50% of Factored Environ-mental Abnormal 1.0 1.0 1.0 1.5 1.0 andlor 1.0 (Flexure) 0
=D.75 ft AbnorrnaV 1.0 1.0 1.0 1.25 1.25 1.0 andlor 1.0
- f. = 0.90 fy Severe Environ-(Shear) mental Abnormal/
1.0 1.0 1.0 1.0 1.0 1.0 ador 1.0 Extreme (2) V.
2j Environ-mental f = 0.85
Topical Report 24370-TR-C-003 Table 6-3 (UFSAR Table 3.8.3-6)
Original Design Stress Margin Table 3.8.3-1 Criteria Versus Table 3.8.3-2 Criterla (4)
TABLE 3.8.3-1 CRITERIA TABLE 3.8.3-2 CRITERIA LOCA PRESSURE + 20%
LOCA PRESSURE + 40%
DESIGN FEATURE (2) CONTROLLING STRESS MARGIN (
(3) CONTROLUNG STRESS MARGIN (%)
LOAD SHEAR MOMENT LOAD COMBINATION SHEAR MOMENT COMBINATION REACTOR VESSEL ANNULUS WALL @ R.C, PUMP SUPPORT SA
-(1) 15.5 ABNORMAL (1) 80
- REACTOR CAVITY COLUMNS 4-FLEXURE 17 18.5 ABNORMALSEVERE 64 22 2-SHEAR ENVIRONMENTAL "CONTROL ROD DRIVE MISSILE SHIELD 4
9 7
ABNORMAL 70 B1 CRANE WALL @ EL. 679.78 5
0 0
ABNORMAUEXTREME 0
0 ENVIRONMENTAL
'CRANE WALL COLS c} 194'-08' & 204'-31'-57 5A 7
19 ABNORMAL/SEVERE 20 10 ENVIRONMENTAL
'STEAM GEN COMPTS, SIDE WALL Q CRANE WALL 1
58 17.5 ABNORMAL 87 34
- PRESSURIZER COMPT CRANE WALL 4
16 11 ABNORMAL
>100
>100
'FLOOR EL 733.63 @ INTERSECTION WICRANE WALL 1
9 0.5 ABNORMAL 19 39
'FLOOR EL. 721.0 @ CRANE WALL 1
62 73 ABNORMAUSEVERE B8
>100 ENVIRONMENTAL MISC COMPTS. RADIAL WALL § CRANE WALL 1
25 01 ABNORMAL 36
>100 FILL SLAB EL. 679.78 § CRANE WALL 5
>20 0
ABNORMAL/EXTREME
>20 0
ENVIRONMENTAL
'CANALWALL(SPANC.VERTPOSMOM) 1
-(1) 3.5 ABNORMAL
.1) 51 CRANE WALL (SPAN C - NEG MOM OPERATING FLOOR) 1 40 3.5 ABNORMAL/SEVERE 28 11 ENVIRONMENTAL CRANE WALL, EL 714.0, HORIZ, NF 1
-(1) 5.5 ABNORMAL
-(1) 36
- flKIfl;:>
flivml ln nrn L..umcn
~t (1) NEGLIGIBLE SHEAR STRESSES IN THESE AREAS (2) SEE TABLE 3.8.3-1 FOR LOADS (3) SEE TABLE 3.8.3-2 FOR LOADS I is taule uues not reiec te evaluatons cocumenieo in Exhlolt I-oi report 't.u CO-14-k.
Tabulated stress margins are from the original calculations and do not reflect later evatuatlons.
Changes have been documented In calculation packages.
(-i Page 17 of 43 14 )
Topical Report 24370-TR-C-003 4h Page 1 of 43 I jJj
Topical Report 24370.TR-C.003 Page 9 of 43 I
Topical Report 24370.TR-C.003 Figure 3 - Conerete Steam Generator and Pressurizer Compartment Reinforcement Pege20or I
Topical Report 24370TR-C-003 Page 21 of 43
Topical Report 24370-TR-C-003 Figure 6 Temperature Gradient (UFSAR Figure 3.8.3-2)
Page22of43 l 4-3 o-7
Topical Report 24370-TR-C-003 Figure 6 Steam Generators I and 4 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-1)
Page 23 of 43 4-o-r
Topical Report 24370-TR-C003 Figure 6 Steam Generators 2 and 3 Postulated Break Locations and Fixes (UFSAR Figure 3.6.7-2)
Page 24 of 43
Topical Report 24370-TR-C-03 7.D Description of Modification to the Structure and New Analyses After installaton of the replacement steam generators, the removed concrete section (plug) of the steam generator compartment roof will be reattached to the complimentary portion of the existng SG compartment by neans of top a-d botorn st^el connection frames. The plug will t-attached to the top and bottom connection fraens using fur 2-inch diameter treaded rods that are instal-=ed in core bore holes through the plug. The top and boltom connection frames wi darp the concrete plug to the complimentary portion of the SG comparntment us.ng six 2-112 inch and eighteen 2-inch diameter threaded rods. The threaded rods are nstailed in the core bore holes localed around the cut line as shown on Figure 7-2. The franmes consist of ox beams made fromrn 1-1/44 inrh ASTM A572 Grade 50 material.it a ayield stress of 50 ksi. The threaded rods conform to ASTM Al 93 Grade B7 material with a yield stress (Fr) f 1 I05 ksi. The threaded rods will be preloaded to a stress level of 0.7 (t) after the concrete lug is installed. This configuration will transfer al the vertical forces from the concrete plug to the complimentary portr of the exisling SG comarrment st.ucture. Tie ateral forces Nill be transferred to the exlsting SG compartment striture by. a series of steel shins (ASTM: A-36 material) that wil" be driven into tfe annAuar space around tWe perimeter of the plug and mechanically locked nto place. Te annular. space between the corcete plug and the complimentary portion of ihe SG compartnent structure will be grouted.
The widCh of tho openinc between the concrete plug and the cmplimentar, portion of the SG compartment vs,il vary, as the wire rope used to make the cuts w,zars. The surface of the cutout section of concrete wff be preoared to provide a gap that ranges fr-m 3-inches to 1-1/4/
4nies. The non-shrnk grourt to be used to fil the annular gap an; the core bore holes is Maserfiow 928 or M;asterf:ow 713 Plus as manufactured by ChemRex. This grout is produced under a Quality Assurance program and is certified to comply with the requiremerts of AS1,TM Clii 07.
his ASTM standard requires that the
- rout be tested For height cnange and compressive strengh. The non-shrnk grout, like the surrounding concrete, could theoretically" experience the formation of micro-cracks when subjected to the design pressure load. Conservative estimates (Reference 8) of the flow path through these micro-cracks yield values that are 1.6 percent of the total design bypass leakage flow area of 5 square feet discussed in UFSAR SectiDn 6.2.1.3.5.
The design leakage area is composed of a known leakage area of approximately 2 square feet and an undefined leakage area. Any leakage through cracks in the grout would be part of this undefined leakage area. UFSAR Figure 6.2.1-22 (provided as Figure 7-1) shows that this percentage increase in bypass area would result in a very small increase in the upper containment pressure. Therefore, micro-cracks resulting from the design pressure load will have a negligible effect on the function of the divider barrier and the analyses that depend on the divider barrier. The SG compartment roof modification described above is detailed on Figure 7-2.
The above mode of restoration results in a modified configuration to the roof of the SG compartment The use of steel through-bolted connection frame-s essentially results in a more flexible boundary condition along the cut-line. In other words, this boundary condition behaves more like a hinge. This means that the reinstalled concrete section of the roof is more flexible than the original configuration, and therefore, subjected to higher deflections and bending moments towards its center. The frame structure is desigrned to accommodate this increased deflection. Also, the inverted concrete T-beam secton under the concrete roof acts like a spacer transmitting the whip-restraint forces from the main steam pipe to the 3 feet thick roof. In the original configuration, the T-beam provided considerable strength in resisting the pipe whip loads. It is noted that since the Page 25 of 43 4 -to
Topical Report 24370-TR--03 reinstalled concrete section in the modified configuration is more flexible than the original design, the forces are redistributed within the reinstalled concrete section. The effects on the walls surrounding the SG compartment (3 feet thick crane wall, 2 feet thick compartment wall and the center wall) were also evaluated. Therefore, as described below, the evaluation of the modified configuration included the T-beam, roof, crane wall, SG compartment walls, and center wall.
The modified SG compartment roof was evaluated to load combinations, load factors, and allowable stresses tabulated in Table 7-2. Table 7-2 is based on Sections CC-3200 and CC-3400 of ASME Secfion 1II, Division 2, 1975, which are generally consistent with UFSAR Table 3.8.3-2. Exceptions to UFSAR Table 3.8.3-2 are the load factors associated with the Yr load and the allowable stresses when thermal effects are included with other loads. The Yr load factors used to evaluate the modified SG compartment roof are consistent with ASME Section II, Division 2, 1975. The allowable stresses due to thermal effects are consistent with both the Proposed ASME Section III, Division 2, 1973 and ASME Section III, Division 2, 1975. The structural steel through-bolted connection frames are designed in accordance with Reference 3.
As noted in Section 6.0, the load combinations in Table 3.8.3-2 of the UFSAR are based on Table CC-3200-1 of the Proposed ASME Section III, Division 2, 1973, Proposed Standard Code for Concrete Reactor Vessels and Containments, Section CC-3000 which was issued in 1973 (the time of original design) by ACI-ASME Committee on Concrete Pressure Components for Nuclear Service for trial use and comment. The purpose of this topical report is to support taking an exception for the load factors associated with the Yr load (reaction load due to fluid discharge on broken pipe, which in the present case is the pipe thrust load) for the Abnormal and AbnormalSevere Environmental Load Categories as described below. Use of this excepton is consistent with the adopted 1975 and later editions of ASME Section III, Division 2 (Reference 12).
In the original design analyses the Yr load was combined with load factors of 1.5 and 1.25 that are associated wih the DBA design pressures for the Abnormal and Abnormal/Severe Environmental Load Categories, respectively. The jet impingement I pipe-whip I pipe break loading (Yr) will rapidly increase, peaking shortly after pipe break and then rapidly decrease in amplitude. The associated DBA pressure loadings will take considerable time following pipe break to reach their design basis peak amplitude values. It is, therefore, overly conservative to combine the DBA pressures with design basis pipe-whip load. The adopted 1975 and later editions of ASME Section lit Division 2 (Reference 12) do not include this load combinaton. The load combinations and allowables used in this analysis for the Abnormal and Abnormal/Severe Environmental Load Categories were based on Table CC-3230-1 (included in this report as Table 7-1) of the adopted 1975 Edition of ASME Section III Division 2 (Reference 12), which superseded the Proposed Code (Reference 11). Note that the load denoted as Rr in Reference 12 corresponds to the Yr load in Reference 11. Also, as allowed by Section CC-3400 of both the proposed 1973 and adopted 1975 versions of ASME Section 111, Division 2, credit is taken for the allowable stresses in concrete and rebar to be increased by 33-113% for service loads, and the tensile strain in rebar to exceed yield for factored loads when thermal gradient effects are included in the load combinabons.
It is also noted that it is acceptable to use a later edition of the ASME Section III code for repairs and replacement per ASME Section Xl (Reference 13). Further, it is noted that the design DBA differential pressure of 24 psi being used in the SG compartment roof evaluation is conservative since it is higher than the maximum calculated differential Page 26 of 43
Topical Report 24370-TR-C-003 pressure of 19.52 psi by 23%. These conservatisms further justify the use of load factors for the Abnormal and Abnormal/Severe Environmental Load Categories based on the adopted 1975 Edition of ASME Section III, Division 2 (Reference 12) without compromising the integrity of the modified SG compartment roof.
The modified configuration of the SG compartment was analyzed for design loads using a 3D finite element ANSYS (Version 5.6) model (Reference 6). Although the roof remains the focus of the evaluation, the model (provided as Figure 7-3) included five components - the 3 feet thick roof, entire SG compartment wall, center wall, 180 sector of the crane wall, and the whip restraint beam; to obtain an accurate representation of the system. The finite elements used were SHELL43 elements for the roof and walls, BEAM44 elements for the whip restraint beam, and BEAM4 elements for the portions of the crane wall where it has openings to the ice condenser. The top of the SG compartment roof is at elevation 778.69'. The compartment wall was modeled as fixed at elevation 733.63 at the top of the containment operating floor; and the crane wall (Figure 6-4) is modeled as fixed at elevafion 721' where the ice condenser floor is located. The nodes at the cut-line along which the connection frames and tapered steell shims are located were realistically modeled to transmit vertical forces and in-plane compression only. The material properties used in the model for the concrete were consistent with those used in the original analysis in Reference 5.
The loads, load combinations and allowable stresses to which the modified SG compartment was evaluated are documented in Reference 7 and summarized in Table 7-2. The modified configuration of the SG compartment roof was analyzed for the following design loads: dead load, live load, design pressure differential of 24 psi from a DBA (main steam pipe break), operating and accident temperature effects, seismic effects (OBE and SSE), and pipe thrust load on the whip-restraint beam from a broken main steam pipe. Design pressure, seismic, and pipe thrust effects were modeled as equivalent static loads. The pipe thrust load applied was 926.25 kips, which is based on the blowI-own load documented in Reference 14 and conservaYLvehv Indudes a factor of 1,5 to account for the -ap between the MS piping and te restraint (as used n the original analysis).
As noted in Section 6.0, the SG compartments were originally designed for a hypothefical pressure of 43 psi resulting from the rupture of a reactor coolant pipe. This pressure was used to provide a high degree of conservatism in the original design, which allowed the structure to accommodate a range of possible equipment configurations and final analysis results. The concrete strength used in the roof evaluation is the in-place compressive strength of the SG compartment roof concrete at 90 days, which is 5700 psi (Reference 5, Sheets 2e and 2f).
The steel through-bored connection frames and tapered steel shins were designed and l
evaluated for the load combinations as described in the previous discussion based on criteria in Section 5.1 of Appendix A to Reference 3.
Thie vertcal design loads on the concrete plug oili be tansferred into the SG comnpartment structure around the perimeter of the plug by the clamp.ng forces induced by the through-b olis cone:ing the top and bottom steel connection frarnes. ror exa.nple, a vertical oad i the upw,ard direction, acting on te concrete plug, w.ould be transferred to the compartment structure as followl.s:
Page 27 of 43 l 43c -V
Topical Report 24370-TR-C-003 The vertical oad from the.lug wil be transe-bred by bearing between the concrete
-hc and the steel bearing plates located between the concrete and the steel frare), to shear Ir te steel Frame, to tens:on in the through-bolt, back to shear in the ower 'rame, to bearing beheen the steel bearing plates and the concrete of the SG cmpartment.
The horizontal design loads on the conrete section wvfl be transferred nt te SG campartment structure via tapered steel shim sets. Each tapered shm set will be c&rmnmsed of a tapered shim atiached toU the face of the concrete section and a loose tapered shim tha-. il-be driven into the gap bet'Neen te Fixed tapered shi'M and the ex!sting c:ompartrent concrete. When nstalled snugly, the loose tapered shim wi be wveldedto te tapered fixed shim to prevent mrvemerit Approximately 3n1 tapered shim se.s (-'1 5top and -15 bo3ttom) ill be installed around the perimteterof the rconpartment concrete section.
ronsevativ-ly, only four 4) tapered sh ir sets wllI be considered to transirer all the horizontal design jads betwer the concrete section wth frame attachac} and the compartment structure. The grout between the concrete sectiol and orroart-rnent st'ructure will not be considered to vansfer any designi bass oads.
The Divider Barrder wi-be restored by covering the annu ar space around the perimeter of the plug on the bottom side of the 3-foot Lhick SG compaertt roof ard fiffing th.e space with norshrink orout.
Page 28 of 43 43dv1AJ
Topical Report 24370-TR-C-003 Table 7-1 (Table CC-3230-1 from ASME Section III, Division 2,1975)
Load Combinations and Load Factors Category D
L.1 F
W(
RL, Re Rr P.,
Hq Service:
Tostruto 1.0 1.0 1.0 1.0 Severeuenionmetl1.0 1.0 1.0 1.0 1..
.0.
Faor ed:.
10 Severe environmental 1.0 1.3 1.0 1.0 1.5 1.0 1.0 Abnormal
~~~~1.0 1.0 1.0 1.5 10
- 1.
10 1.0 AbmlSevere environmental1. 1.0 1.03,
- 1.
25...
1.0 1.25 1..
1.0..
1.0 1.0 1.0 1.25
- 1.
- 1.
1.25
.0 1.0..
AbolExtreme environmental 1.0 1.0 1.0
.0.
0...
1.0 1.0 NOTE:
(1) Includes all temporary construction loading during and after construction of containment.
%jj
'X
~~~~~~~~~~~~~~~~~
29 of 43 1
Topical Report 24370-TR-C-003 Table 7-2 Loading Combinations, Load Factors and Allowable Stresses for SG Compartment Roof Modification (5)(6)
NOTES:
- 1.
Includes all temporary construction loading during and after construction of containment 2,
v, is lower for tension members and is given by v. = 2f7 (I + 0.002NJAg). with Nu negative for tension.
- 3. The aflowable stress is increased by 33113>% when temperature effects are combined with other loads.
- 4.
The tensile strain may exceed yield,when the effects of thermal gradients are included in the load combination.
i.e., f, can be <= f1. and v. can be > cy when thermal effects are included.
- 5. The load combinations, load factors and allowable stresses in this table are based on the ASME Section III Division 2, 1975, which are, in general, consistent with the proposed ACI 359 -ASME Section III Division 2.
1973 with the exception of load factors associated with the Yr load.
- 6.
Structural steel components of the tough-bolted connection rarnas and tapered stee; shirr,s were designed in accordance with TVA Design Criteria SQN-DC-V-1.3.2, Miscellaneous Steel Components for Class I Structures.
LOADS NOMENCLATURE:
D Feqo Feqs L
Ps R,
Ts Ye Dead loads, or their related intemal moments and forces Operating basis earthquake Design basis earthquake Live load, or their related intemal moments and forces Accidentincident maximum pressure Piping loads during operating conditions
.Piping loads due to increased temperature resulting from the design accident Thermal loads under the thermal conditions generated by the postulated break and including To Operational temperature Reaction load on broken pipe due to fluid discharge (corresponds to Rr in ASME Section 1II, Division 2, 1975)
' The term design basis earthquake has the same meaning as the term safe shutdown earthquake.'
Page 30 of 43 4.2 oY Allowable Category T.
D L1)
P.
T.
Fe.
F,.
R.
Ra Yr Stresses Service:
(Flexure) f, = 0.45 fc Const 1.0 1.0 1.0 f = 0.50 f (3)
Normal 1.0 1.0 1.0 1.0 1.0 (Shear) 50% of Factored (3)
Factored:
Extreme 1.0 1.0 1.0 1.0 1.0 (Flexure)
Environmental f, = 0.75 fc f,=0.90fy (4)
Abnormal 1.0 1.0 1.0 1.5 1.0 (Shear)
(2)v.= 2f/7 AbnormaV 1.0 1.0 1.0 1.25 1.25 1.0
=0.85 Severe Environmental AbnormaV 1.0 1.0 1.0 1.0 1.0 1.0 1.0 Extreme Environmental I
Topical Report 24370-TR-C-003 Figure 7 Sensitivity of Peak Compression Pressure to Deck Bypass (UFSAR Figure 6.2.1-22)
Page 31 of 43 l
Topical Report 24370.TR-C-003
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Figure 7*2 - Steam Generator Compartment Roof Cunnoction P rama Layout and Connection Details page 32of 43 l t:,jjY.,
i:
Topical Report 24370-TR-C-003 Figure 7 Finite Element Model "SGE1" and "SGE2" and Element Groups and Global Coordinate Systems (Reference 6)
Page 33 of 43 43S6
Topical Report 24370-TR-C-003 8.0 Results of New Analyses The modified configuration of the steam generator compartment roofs has been evaluated for the design loads and load combinations documented in Reference 7 as described in Section 7.0. Except as noted in Section 7.0, these design loads and load combinations are consistent with those used in the original analyses for the SG compartments. The structural adequacy of the modified SG compartment roof configuration under these design loads and load combinab'ons was evaluated in Reference B. The design of the steel thrcugh-blte3-d connection franes and tapered steel shims is documented in Reference 9. The results are briefly summarized below.
Normal service load combinations used to evaluate the modified SG compartment roof configuration were the same as those used for the original configuration. Under normal service load conditions, the maximum concrete and rebar stresses in the modified roof are within the allowable normal service concrete and rebar stress limits as specified in Section CC-3430 of ASME Section III, Division 2, 1975 (summarized in Table 7-2). The critical areas where these stresses occur are near the middle surfaoe of the cut section at the junction of the roof and the end of the whip restraint beam (Reference Area 1 on Figure 8-1). The stress levels in other areas are generally much lower. Therefore, the modified SG compartment roof configuration is acceptable under normal service conditions.
The load combinations evaluated for the modified roof were based on Table CC-3230-1 (included in this report as Table 7-1) of the adopted 1975 Edition of ASME Section III Division 2 (Reference 12), which replaced the Proposed Code (Reference 11) as discussed in Sections 6.0 and 7.0. These load combinations are similar to those used for the original SG compartment roof design except for the Abnormal and Abnormal I Severe Environmental load categories for which the Yr load is now not considered in the load combination. For factored load combinations on the modified roof configuration, the most critical load combinations are the Abnormal and Abnormal / Extreme Environmental load categories. The critical areas of high stresses for the Abnormal load combination are the approximately triangular comer areas of the existing roof bounded by the cut-line near each end of the center wall (Reference Areas 2 and 3 on Figure 8-1). For the Abnormal / Extreme Environmental load combination the critical area included the area near the middle of the cut section at the junction of the roof and the end of the whip restraint beam (Reference Area 1 on Figure 8-1) in addition to the comer areas identified for the Abnormal load combinaton. It is noted that the maximum stresses/forces oocurred only in the localized areas mentioned above. The stresses in other areas are lower. The maximum stresses, in these critical areas, for the factored load combinaUons were found to be within the allowable concrete and rebar stresses based on limits specified in Section CC-3400 of ASME Section 111, Division 2, 1975. The maximum vertical deflection occurred for the Abnormal I Extreme Environmental load combination at the middle of the roof near the end of the whip restraint beam.
It is noted that the design DBA differential pressure of 24 psi was used in the modified SG compartment roof stress evaluation. Even though the calculated stresses under accident conditions equaled the allowable stresses in some locations, this analysis is conservative since it used a differential pressure that is 23% higher than the maximum calculated differential pressure of 19.52 psi.
Page 34 of 43 4s bc
Topical Report 24370-TR-C-003 The influence of the modified roof configuration on stresses in the SG compartment wall sections adjacent to the roof has been determined to be insignificant and the wall and roof stresses remain within design allowables.
I The design of the steel through-bolted connection framres and tapered steel shims documented in Reference 9 is described in Section 7.0 and shown on Figure 7-2. The throuchlols will be nstalled w^;ith a ore-tension load. base on Q.7F 1. Usingl conservative design checks, the maximum calculated bending stress in the connection frarme beams and the maximum calculated bearing stress on concrete and the tapered steel shims were determined to be below allowables. The connecor frame beams w:ill be us--d in conjunc:on with the through-bolts to provide 'he clanping action that vil transfer the verical design basis loads from ths ooncrete section to te compartment.
The conec&ton frame beams scan over all of the connection truch-o-lts. Since all tl-e connection frame beans are connected t-gether, riid body rotations cf the beams about -he bolt axes are prevented at all concrete section,'compartment Connections.
The connection frame beams have been designed to transfer al vertical design loads, at the concrete seciorcomartmem.t interface, via bending and shear stresses. The beams have been designed such that the maxlnum stresses in the beam dates ar connectin welds are less har the alowab e stresses.
The coinnection frarre beams are sized such thSt the concrete bearing stresses under the beams are below alI'owables due to both the connection tirough-bolt pr-tension loads and due to all design basis loads.
The connection frame beams are connected by web angles or connection plates. The welded angles!plates are designsd to be flexible in order to transfer all vedcal rein loads between b-eam mernbers of the frame, as pinned connect-iors. Vertical aalds are due to the vertical seismic inernia from t-e concrete and the maxinum DBA pressure tseismic inertia loading wfom tLe steel frame is negligible). As the concrete seCtion deflects, it lifts the individual framre mermber-s, hence, inducing vertical oads at the beam-to*b*am connections and ve_rucal prying loads at the throuzgh-bolt connections. The beam connection angleplates are also designed to trans5er all horizotal seismnc oads due to the ax rurn accelerations of the frame.
Based on the evaluations in the calculations noted above, the modified SG compartment roofs have been found to be structurally adequate for the loads associated with the design loading conditions/combinations which are in general consistent with the original design except as noted above and in Section 7.0.
The modifications to the steam generator compartment roofs do not affect the structural capability of the steam generator compartments to contain the intemal pressure associated with the design bases main steam line breaks. The modifications do not affect temperature differentials through the compartment roof or the radiation shielding capacity of the structures.
As discussed in Section 6.5.6.3 of the UFSAR, there is a maximum calculated leakage of 250 cfm between the upper and lower containment through the divider barrier, of which the steam generator compartments are part. The amount of leakage between the two sections of the containment will not be affected by the restoration of the steam generator compartment roofs. The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries Page 35of43 l 4S
Topical Report 24370-TR-C-003 between upper and lower containment. It is noted that any leakage due to possible cracks in the grout, particularly under design DBA loading, will be extremely small and therefore insignificant (Reference 8).
Figure 8-1 Areas of Critical Stresses 9.0 Summary and Conclusions Restoration of the SG compartment will be accomplished by reattaching the removed section of concrete using trough-bolted structural steel connection frames and Tapered steel shims in the annular gap. The SG compartments have been reanalyzed to determine that the modified configuration is acceptable. This analysis follows the same basic approach as documented in the existing SG compartment design calculations, the Sequoyah design criteria, and/or the Sequoyah UFSAR. Areas where the two analyses differ are summarized in Table 9-1.
Table 9-1 Differences Between Original and New Steam Generator Compartment Analyses Original Analyses New Analyses Analyzed compartment structure as Analyzed compartment structure several individual components (roof, using a three dimensional ANSYS enclosure wall, center wall, and crane finite element model comprised of wall) using two-dimensional model.
system components.
Evaluated compartment structure for a Did not evaluate compartment 43-psi hypothetical pressure.
structure for a 43-psi hypothetical Page 36 of 43 4-3;be2 I
Topical Report.24370-TR-C-003 Use of the methodologies, loads and load combinations discussed in this topical report are either consistent with the 6riginal design basis or based on accepted industry design standards. The proposed modificafions to the SG compartment design are therefore justified.
10.0 References
- 1.
Sequoyah Nuclear Plant Updated Final Safety Analysis Report, Amendment 16.
- 2.
TVA Design Criteria SQN-DC-V-1.1, Design of Reinforced Concrete Structures.
Revision 16.
- 3.
WVA Design Criteria SQN-DC-V-1.3.2, Miscellaneous Steel Components for Class I Structures, Revision 10.
- 4.
TVA Design Criteria SQN-DC-V-1.3.3.1, Additions After November 14, 1979 -
Reinforced Concrete, Structural, and Miscellaneous Steel, Revision 6.
- 5.
TVA Calculation SCG-1-40, Steam Generator Compartment, Final Design, Revision 4.
- 6.
TVA Calculation SCG-1 S-607, Evaluation of Steam Generator Compartment Modification - 3D Finite Element Model, Revision 0.
- 7.
TVA Calculation SCG-1S-608, Evaluabon of Unit I Steam Generator Compartment Modification - Load Conditions and Allowable Stresses, Revision 0.
- 8.
TVA Calculation SCG-lS-609, Evaluation of Steam Generator Compartment Modification - Finite Element Analysis Results, Revision 0.
- 9.
WVA Calculation SCG-1S-610, Evaluation of Uni I Steam Generator Compartment Modification - Design of Rooff Support Frames, Revision 1.
- 10.
Bechtel Calculation 24370-C-013, Rev. 0, ANSYS 5.6 Verification.
Page 37 of 43 4;W Original Analyses New Analyses pressure.
4 Analyzed compartment structure Analyzed compartment structure for a initially for a maximum differential maximum design internal differential pressure of 21.3 psi which is a 25%
pressure of 24 psi as specified in the increase over the DBA pressure UFSAR using loads, load differental of -17 psi for the SG combinations and allowable stresses compartment provided by documented in Table 7-2.
Westinghouse (i.e., 1.25 x 17 psi). Per NRC request, a 40% increase in DBA differential pressure (i.e., 1.4 x 17 psi) was investigated later.
Evaluated compartment roof globally Evaluated the modified roof globally for an equivalent static jet thrust force for an equivalent static pipe thrust
(-910 kips on the roof) that would load of 926.25 kips which is based on result following a main steam pipe the shoc-spectrum from he MS Blow break inside a single compartment Do.,rn Analysis, Analyzed the compartment structure Analyzed the modified.compartment using the load combinations, load structure using load combinations and factors, and allowable stresses shown allowable stresses in Table 7-2. Load in UFSAR Tables 3.8.3-1 or 3.8.3-2.
factors for the load combinations and allowable stresses were based on Table CC-3230-1 and Section CC-3400, respectively, of the 1975 Edition of ASME Section III, Division 2.
I
Topical Report 24370-TR-C-003
- 11.
Proposed ASME Section III Division 2, 1973, Proposed Standard Code for Concrete Reactor Vessels and Containments, Section CC-3000 (This draft code was issued in 1973 by ACI-ASME Committee on Concrete Pressure Components for Nuclear Service for Trial Use and Comment).
- 12.
ASME Section III Division 2, 1975 Edition, Concrete Reactor Vessels and Containments, Section CC-3000.
- 13.
ASME Section XI, Rules for Inservice Inspectlon of Nuclear Power Plant Components.
- 14.
TVA Calculation 0600117-S002, RO, Blo-w Dowm Anlysis - Main Steam System.
- 15.
TVA-NQA-PLNB9-A, Nuclear Quality Assurance Plan, Revision 10.
- 16.
General Engineering Specffication G-2, Plain and Reinforced Concrete, Revision 7.
Page 38 of 43
Topical Report 24370-TR-C-003 Appendix A No Significant Hazards Consideration Determination DESCRIPTION OF THE PROPOSED CHANGE The four steam generators of the Sequoyah Nuclear Plant Unit 1 will be replaced during the spring of 2003. To support the replacement of the old steam generators (OSGs) with the replacement steam generators (RSGs), access openings will be created in the roof of the steam generator (SG) compartments inside containment. An appropriately sized access opening will be made in each SG compartment roof by cutting out a section of concrete from the roof of the compartments.
Upon completion of installation of the RSGs, the original cut section plug) of the SG compartment roof will be reinstalled using a modified configuraton from the original.
The concrete plug removed from each of the SG compartment roofs will be reattached to the complimentary portJon of the SG compartment roof by means of top and bottom steel connection frames. The plug will be attached to the top and bottom connection frames usnr our 2-inch dianeter threadev' r--s hat are installed in corn bore holes through the plug. T-m top and bottom connection, frames wil c-lamp the concrete plug to he cnnplimentery portion of th.e SG compartment usinrg six 2-112 ich and eichteen 2-inIc dia.meter threaded rds. T hese threaded rods are nstalled in the core bore holes located around the plig cutline and %.li be preloaded. The frames consist. of b-ox beams made fnom 1-114 nc-h steel. A series of steed shims wiAll be driven into th.e annular space around the peri-meter of the piug and me-hanica.ly locked into ae.
The coe bores and the annular space betwveen the concrete plug and te corplirnentary poiron of the SG compartment roofw til be grouted using non-shrink grout that conforms to ASTM C 1107, thereby sealing the roof.
REASON FOR THE PROPOSED CHANGE The process for restoration of the steam generator compartment roof using the through-l boked connection frarres results in less construction debris in containment since te concrete cuts will not require chipping for rebar splicing. The process is also simpler and faster than splicing new rebar and pouring new concrete.
Ill.
SAFETY ANALYSIS Normal service load combinations used to evaluate the modified SG compartment roof configuration were the same as those used for the original configuration. Under normal service load conditions, the maximum concrete and rebar stresses in the modified roof are within the allowable normal service concrete and rebar stress limits as specified in Section CC-3430 of ASME Section 1I1, Division 2,1975. The critical areas where these stresses occur are near the middle surface of the cut section at the junction of the roof and the end of the whip restraint beam. The stress levels in other areas are generally much lower. Therefore, the modified SG compartment roof configuration is acceptable under normal service conditions.
The load combinations evaluated for the modified roof were based on Table CC-3230-1 of the adopted 1975 Edition of ASME Section Ill Division 2, which replaced the proposed 1973 ASME Section III, Division 2. These load combinations are similar to those used Page 39 of 43 4:75
Topical Report 24370-TR-C-003 for the original SG compartment roof design except for the Abnormal and Abnormal/Severe Environmental load categories for which the Yr load is now not considered in the load combination. For factored load combinations on the modified roof configuration, the most critical load combinations are the Abnormal and AbnormalExtreme Environmental load categories. The crifical areas of high stresses for the Abnormal load combination are the approximately triangular corner areas of the existing roof bounded by the cut-line near each end of the center wall. For the AbnormaVExtreme Environmental load combinaton the critical area included the area near the middle of the cut secton at the junction of the roof and the end of the whip restraint beam in addition to the comer areas identified for the Abnormal load combination. It is noted that the maximum stresses/forces occurred only in the localized areas mentioned above. The stresses in other areas are lower. The maximum stresses for the factored load combinations were found to be within the allowable concrete and rebar stresses based on limits specified in Section CC-3400 of ASME Section 111, Division 2, 1975. The maximum vertical deflection occurred for the Abnormal/Extreme Environmental load combination at the middle of the roof near the end of the whip restraint beam.
It is noted that the design DBA differential pressure of 24 psi was used in the modified SG compartment roof stress evaluation. Even though the calculated stresses under accident conditions equated the allowable stresses in some locations, this analysis is conservative since it used a differential pressure that is 23% higher than the maximum calculated differential pressure of 19.52 psi.
The influence of the modified roof configuration on stresses in the SG compartment wall sections adjacent to the roof have been determined to be insignificant and the wall and l
roof stresses remain within design allowables.
The bolts used in the stee ttrouo h-bolted connection will be preloaded to a stress level of 0.7 Fy. By conservative analysis, the maximum calculated bending stress in the connection framne beams and the maximum calculated bearing stress on concrete and the tainered ste-ei shims were determined to be below allowables. The connection frame bcams will be used in conjunction with the through-bolts to provide 'the clamping ac tion that will transfer the ver.ca des,,n basis loads from the conrete section to se compatrtent The connection frcme beams spoan over all of the connection trough-bolts. Sirce a'l the conn ection frame beams are connected together, beam nigid body rotation about te bolt axes are prevented at all concrete secton/cmpartment connectiors.
The connecion frame beams have been designed to transfer a.l vertical dezian oads, at the concrete setorJcor rtment interface, via bending and shear stresses. The beams have been designed such that the maximum stresses in the beam plates and connectno welds are less than the allowable stresses.
Tne connection frane beams are sized such that the concrete bearlno stresses under the beans are below aliowables due to both [he connection through-boit pre-iension loads and due to all design basis Loads.
The conrnecron frane beams are corineicted by web angles or connection plates. The welded angles/Jplates are designed t-transfer all vertica: dsign io2ds b-_eben beam mrnembe;r of tte frame, as pirnec canneciors. Vertical loads are dua to the vertical se smc Inertia frn the concrete and te riaximtum DBA pressure seismic inertia Page 40 of 43 4-17
Topical Report 24370-TR-C-003 loading from te steei frame is negligibe). As the concrete se-tion doflects. it ifts the indiVidual frame rn-embers, hence, ir-ducing vertical loads at the beam-to-beam conanctions and vertical prying oads at the through-bolt connections. T he beam onnection anglesiplates are also designed to transfer all horizontal seismic loads due to te maximnum acelerantios of the fra.me.
The modified SG compartment roofs have been found to be structurally adequate for the loads associated with the design loading conditionslcombinations which are in general consistent with the original design except as noted above.
The modifications to the steam generator compartment roofs do not affect the structural capability of the steam generator compartments to contain the internal pressure associated with the design bases main steam line breaks. The modifications do not affect temperature dfferentials through the compartment roof or the radiation shielding capacity of the structures.
As discussed in Section 6.5.6.3 of the UFSAR, there is a maximum calculated leakage of 250 cfm between the upper and lower containment through the divider barrier, of which the steam generator compartments are part. The amount of leakage between the two sections of the containment will not be significantly affected by the restoration of the steam generator compartment roofs. The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries between upper and lower containment. t is noted that any leakage due to possible cracks in the grout, particularly under design DBA loading, will be extremely small and therefore insignificant.
IV.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA has concluded that operation of SQN Unit 1, in accordance with the proposed modification to the steam generator compartment roof, does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50.91 (a)(1), of the three standards set forth in 10 CFR 50.92(c).
A.
The proposed amendment does not involve a significant increase in the probabilitv or consequences of an accident Previously evaluated.
The probability of occurrence or the consequences of an accident are not increased as presently analyzed in the safety analyses since the objective of the event mitigation is not changed. No changes in event classification as discussed in UFSAR Chapter 15 will occur due to the modification of the Unit 1 steam generator compartment roof design.
The grout used to fill the gap between the replaced concrete and the surrounding concrete, like the surrounding concrete, could 'theoreically" experience the formation of micro-cracks when subjected to the design pressure load.
Conservative estimates of the flow path through these micro-cracks yield values that are numerically insignificant when compared to the allowable divider barrier bypass leakage. Micro-cracks resulting from the design pressure load will have a negligible effect on the function of the divider barrier and the analyses that depend on the divider barrier. Therefore, the containment design pressure is not challenged, thereby ensuring that the potential for increasing offsite dose limits Page 41 of 43 4;5bu
Topical Report 24370-TR-C-003 above those presently analyzed at the containment design pressure of 12.0 pounds per square inch is not a concem.
Therefore, the proposed modification to the Unit 1 steam generator compartment roof design will not significantly increase the probability or consequences of an accident previously evaluated.
B.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The possibility of a new or different accident situation occurring as a result of this condition is not created. The steam generator compartment roof forms part of the divider barrier. This barrier is not an iniiator of any accident and only serves to force steam that is released from a LOCAI DBA to pass through the ice condenser. The failure of any part of the divider barrier is considered critical since it would allow LOCAIDBA steam to bypass the ice condenser, thereby increasing the pressure within the primary containment.
As discussed in Section 6.5.6.3 of the UFSAR, there is a maximum calculated leakage of 250 cfm between the upper and lower containment through the divider barrier. The amount of leakage between the two sections of the containment will not be significantly affected by the restoration of the steam generator compartment roofs. The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries between upper and lower containment. It is noted that any leakage due to possible cracks in the grout, particularly under design DBA loading, will be extremely small and therefore insignificant.
Therefore, the potential for creating a new or unanalyzed condition is not created.
C.
The proposed amendment does not involve a significant reduction in a margin of safety.
A design DBA differential pressure of 24 psi was assumed in the original design of the steam generator compartment roof. This differential pressure is 23%
higher than the maximum calculated differential pressure of 19.52 psi. Since the same design differential pressure was also used in the modified SG compartment roof stress evaluation, the margin of safety was not reduced.
As discussed previously, the amount of leakage that bypasses the divider barrier will not be affected by the restoration of the steam generator compartment roofs.
The use of non-shrink grout to seal the joint created between the concrete sections and the remaining structure will maintain the boundaries between upper and lower containment. Hence, the worse-case accident conditions for the containment will not be affected by the proposed modifications.
Therefore, a significant reduction in the margin to safety is not created by this modification.
Page 42 of 43 4,36
Topical Report 24370-TR-C-003 V.
ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.
Page 43 of 43
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