ML12277A176

Steam Generator Replacement Presentation

ML12277A176
Person / Time
Site:	Prairie Island
Issue date:	10/04/2012
From:	NRC/RGN-III
To:
Atif Shaikh
References
Download: ML12277A176 (37)

Text

P i i Island Prairie I l d Nuclear N l Generating G ti Plant Pl t (PINGP)

Unit 2 Steam Generator Replacement (SGR)

Project NRC Region g 3 Presentation October 4, 2012

Purpose The purpose of this meeting is to discuss the Prairie Island Nuclear Generating Plant (PINGP) Unit 2 steam generator replacement activities including design, licensing, fabrication, installation and oversight.

2

Att d Attendees

• Karen Fili, Vice President Nuclear Capital Projects

• Scott Marty, Director Unit 2 SGR Project

• Richard Pearson Pearson, Unit 2 SGR Project Engineering Supervisor

• Gene Eckholt, Project Licensing Manager

• Glenn Adams, Project Licensing Lead Engineer 3

M ti A Meeting Agenda d

• Overview

• Design and Planning

• Replacement R l t St Steam Generator G t (RSG) Design D i

• Engineering Design and Analysis

• Radiation Protection

• Steam Generator (SG) Removal and Installation

• Post Installation Verification and Testing

• Xcel X l Energy E Oversight O i ht

• Regulatory Interface 4

O Overview i - Project P j t Scope S

Replace two Unit 2 original Westinghouse Model 51 steam generators with new AREVA Model 56/19 RSGs Two piece replacement through existing equipment hatch similar to Unit 1 5

O Overview i - Project P j t Scope S

• Run Animation 6

O Overview i - Project P j t Scope S Project includes off-site disposal of the two original steam generators Immediate preparation and packaging for off site disposal similar to Unit 1 U1 OSG Lower Part being U1 OSG Upper Part leaving prep for rail shipment at PINGP in late 2004 PINGP in late 2004 7

O Overviewi - Project P j t Initiation I iti ti

• Project initiated for economic issues including:

Increasing repair costs Potential need for planned mid-cycle inspections, forced outages Increasingg outage g time Reduction in plant output Replacement energy costs 8

O Overview i - Participants P ti i t Project Management - Xcel Energy Xcel Energy Project Management & oversight personnel RSG Engineering & Fabrication - AREVA AREVA NP SAS Engineering & Licensing support AREVA NP Major forgings - Japan Steel Works Tubing - SANDVIK Steel Installation Contractor - SNC - Lavalin SNC Lavalin (SLN) - Babcock & Wilcox Nuclear Energy Engineering and construction services required for Steam Generator removal and installation 9

O Overviewi - Schedule S h d l

• Fabrication and Delivery First lower part hydrostatic test completed and ASME Pre-Service Inspection in process On O schedule h d l for f on-site it d delivery li S Spring i 2013 using i on-site barge landing and haul routes from Unit 1 project 10

Current Timeline RSG Components Pour 1st RSG Forging g g Receipt p of Forgings g g RSGs Delivered Forgings F i Award A d - June J 2006 January 2010 and Tubing at Manufacturer to PI Full Contract - August 2006 OSG Disposal Contract Award SGR Outage 4th Qtr 2011 Fall 2013 Establish other support contracts U2 SGR Installation Services Contract and Temporary Facilities Award Dec 2009 PHASE 1 PHASE 2 Detailed Engineering (ECs)

N-4 N-3 Outage Outage N-2 Outage N-1 Outage 2006 2007 2008 2009 2010 2011 2012 2013 Laser Conceptual Work Packages Scanning Verification & Complete Precision Walkdowns Measurements Final Design Verification 11

D i and Design d Pl Planning i - RSG Design D i RSG Design Designed and fabricated by AREVA NP SAS (Paris / Chalon Chalon, France)

Two (2) piece ASME Section III Class 1 Nuclear Pressure Vessel - to be N-stamped in field Alloy 690 (TT) tubing material Final Girth Closure Weld located at top of transition cone Basic design for Unit 2 is same as Unit 1 12

D i and Design d Pl Planningi - RSG Design D i Steam outlet nozzle flow restrictor integral to RSG elliptical head Level sensing line pressure stabilizers Anti-stratification device in feedwater ring to reduce thermal stratification effects Low Delta-P deck plate Stainless Steel (SS) Feedwater Ring with JJ-tube tube outlet nozzles and welded inlet thermal sleeve Loose Parts Trapping Screen to protect tube bundle region 13

D i and Design d Pl Planningi - RSG Design D i Tube Bundle Recirculation Connection for Cold Shutdown Chemistry Control 5 sets of SS Anti-Vibration Bars (AVBs) with 9 contact points SS Quatrefoil Tube Support Plates (TSPs)

Additional handholes at tubesheet and at two elevated TSPs 14

D i and Design d PlPlanning i - RSG Design D i RSG Design (continued)

OSG RSG Thermal Output (MWt) 825 825 Thot/Tcold/Tave (nom.) (deg F) 590/530/560 590/530/560 Number of Tubes 3388 4868 Tubing OD/WT (in.) 0.875/0.050 0.75/0.043 Heat Transfer Area (sq. ft.) 51,500 61,281 Performance Coefficient UA (10E6 Btu/hr-F) 67.52 81.93 RCS Fl Flow - Best B tE Estimate ti t (10E6 lbm/hr) 38.39 38.51 RCS Pressure (psia) 2250 2250 Steam Flow (10E6 lbm/hr) 3.54 3.57 Full Load Steam Outlet Pressure (design normal operating) (psia) 750 806.8 Max. Moisture Carryover (%) 0.25 0.1 Operating Weight (lbf) 801,073 802,055 15

D i and Design d PlPlanning i - RSG Design D i Lesson L Learned L d from f Unit U it 1 RSG A small number of tubes exhibited wear scars at some TSP and AVB locations Comprehensive investigation was completed in numerous fields such as:

Design, fabrication and installation Evaluation of Non-Destructive Examination (NDE) techniques Review of operating experience Wear simulations to attempt to reproduce wear at TSPs A single root cause could not be determined Increase in depths of wear scars from 1R24 to 1R25 was limited limited, suggesting a trend to stabilization. The Operational Assessment of the growth provided justification for no inspection until 1R28 Outage 16

D i and Design d Pl Planning i - RSG Design D i RSG Improvements related to wear scars TSP Procurement - Reinforcement of flatness controls TSP Manufacturing - Recording of reworked holes and additional dimensional data TSP Assembly - Record and control parallelism, require slight contact at locking wedges and check peripheral gaps Tie Rod Assembly - Group to have same length between TSP and tie rods and tightened from the center to the exterior AVB PProcurementt - Additional Additi l recorded d d measurements 17

D i and Design d Pl Planning i - RSG Design D i Other RSG Improvement compared to Unit 1 Secondary Manway Davit Arm Design - Improved design to address binding Feedwater Ring Inspection Port Design -

Changed gasket, torque and locking washers to address loose screws and damaged gaskets AVB Anti Lift Off Device Design - Lockwashers are bigger to address binding on screw threads Tubesheet drilling is now performed after the tubesheet assembly to the lower shell section An extended lip is integrated into the tubesheet, eliminating a cylindrical ring section and one circumferential weld in the channel head 18

Design and Planning -

Fabrication Test Examples Materials and NDE per ASME Section III RSG Vessel Primary Side Hydrostatic Test Tubing Examinations At Tubing Mill Per ASME Section III Section XI per EPRI NDE Guidelines, bobbin full length, Rotating Pancake Coil (RPC) low row u-bends At AREVA - Chalon, Chalon France Helium Leak Test of tube-to-tubesheet weld Tube Expansion Profilometry Pre Pre-service service Eddy Current Testing (ECT) including RPC of expansion transition 19

D i and Design d Pl Planning i - RSG Design D i Industry I d t Lesson L Learned L d San Onofre Tube Leak A combination of unpredicted, adverse thermal hydraulic conditions diti andd iinsufficient ffi i t contact t t fforces iin th the upper ttube b bbundle dl caused fluid-elastic instability which was a significant contributor to the tube-to-tube wear, resulting in the tube leak Due to modeling errors, the San Onofre replacement generators were nott designed d i d with ith adequate d t thermal th l hydraulic h d li margin i Applicability to PINGP Fluid Elastic Instability risk evaluated with AREVA CAFCA4 (thermal-hydraulic data) and GERBOISE (vibration) codes AREVA CAFCA4 showed thermal-hydraulic results up to 4 times higher velocity than Mitsubishi FIT III Code (Reference SONGS AIT Report)

Even with conservative Connors and damping coefficients, the absence of fluid-y of the RSG tube bundle is demonstrated elastic instability Unit 1 ECT during first cycles showed no tube-to-tube contact or wear 20

D i and Design d Pl Planning i - RSG Design D i Industry Lesson Learned (continued)

San Onofre Tube Leak - PINGP versus SONGS AREVA design and fabrication, not Mitsubishi Tubes made by Sandvik, Sandvik not Sumitomo PINGP AVB design is different and does not include the SONGS retainer bar design that caused wear PINGP RSG tube bundle ~ half the size of SONGS PINGP square pitch value larger than SONGS triangular PINGP has u-bend region increasing pitch incrementation No divider plate repairs for PINGP RSGs PINGP U1 RSG Operating History

< 10 tubes preventively plugged since 2004 No challenges to tube structural integrity 21

D i and Design d Pl Planning i - EC P Packages k

Engineering Change (EC) Packages One (1) EC Package applicable to the design and analysis of the RSGs prepared by Xcel Energy Sixteen (16) EC Packages applicable to removal and installation activities prepared by SNC-Lavalin divided by system (RCS, (RCS FW, FW MS, MS etc.)

etc ) or activities (Rigging, RSG Assembly, etc.)

EC Packages are reviewed in accordance with EC process and d 10 CFR 50 50.59 59 22

D i and Design d Pl Planning i - EC P Packages k

23

B k Break 24

SG Removal R l and d IInstallation t ll ti Heavy Loads Outside Lift System (OLS) utilizes g gantry y arrangement g

with hydraulic jacking mechanism OLS designed seismic II over I to reduce risk to non-safety related Condensate Storage Tanks and Containment Structure Based on design and distance OLS can not striking D5/D6 building 25

SG Removal R l and d IInstallation t ll ti Heavy Loads Inside lifting utilizes the existing g 230-ton capacity p y Polar Crane and an Engineered Lift evaluation per ANSI B30.2 Unit 1 and Unit 2 common systems in containment evaluated for impacts and actions taken to address (photo from Unit 1 SGR) 26

SG Removal R l and d IInstallation t ll ti Heavy Loads On-site transport routes are load tested prior to SG component p transport p

SG load drop analysis is performed for the radiological consequences related to a drop of the SG 27

SG Removal R l andd IInstallation t ll ti Radiation R di ti Protection P t ti RCS Pipe End Decon Shielding Plan under development ALARA Plan under development 28

SG Removal R l andd IInstallation t ll ti Security Opening of the equipment hatch requires establishing a temporary security barrier, a common outage evolution Components will enter Protected Area (PA) through existing PA boundary gates 29

P t Installation Post I t ll ti Verification V ifi ti and dT Testingti Pre-SGR Pre SGR Outage Data Collection Reactor Coolant System (RCS) baseline flow Reactor Coolant Pump (RCP) baseline vibration Baseline of Plant Process Monitoring points Construction Tests Construction NDE - RT, MT, PT, VT Pre-service In-Service Inspections Instrument Calibration Steam Generator Hydrostatic Test - Secondary Pi Primary/Secondary

/S d System S t In-Service I S i Leak L k Test T t Requires NDE be performed in accordance with ASME Section III, 92 or later 30

P t Installation Post I t ll ti Verification V ifi ti and dT Testing ti Startup Testing Measure RCS Flow RCP Vibration / Motor Amperage Chemistry Adjustments Hot Shutdown Testing Verify RCS Flow Measure RCP Vibration Measure RCP Flow Coastdown Monitor Instruments & Controls Operation O

Operatet SG Blowdown Bl d System S t Measure / Adjust RSG Support Hot Gap Conditions Measure Rod Drop Times 31

P t Installation Post I t ll ti Verification V ifi ti and dT Testing ti Expected E t d Power P Testing T ti Evolutions E l ti Verify expected design parameters Monitor Primaryy and Secondary y Chemistryy Parameters First Stage Turbine Pressure Downstream Instrument st u e t Ca Calibrations b at o s Main Steam Flow Calibration Moisture Carryover Testing M it Steam Monitor St Line Li Drain D i Traps T

Monitor SG Blowdown 32

O Oversighti ht - AREVA NUPIC joint utility audit - 2009: Xcel Lead NUPIC joint utility audit - 2011: Xcel participant Xcel Nuclear Oversight g (NOS)

( ) visit to Japan p Steel Works to witness tubesheet Ultrasonic Test (UT)

Full time Xcel NOS representative at St. Marcel y

fabrication facility Full time Xcel NOS representative at Sandvik Project and Subject Matter Expert visits for project meetings Primary Side Hydrostatic Test meetings, Test, Eddy Current Test (ECT) and UT witness and Radiography Test (RT) radiography review 33

O Oversight i ht - SNC-Lavalin SNC L li / B&W NE Xcel X l NOS 2009 A Audit dit Xcel NOS 2012 Audit - SNC-Lavalin Engineering Xcel SGRP Engineering Staff oversight Modification development by External Design Organization Owners Acceptance and EC approval by Xcel Energy Project Engineering Staff 34

R Regulatory l t Interface I t f - LARs LAR License Amendment Request (LAR): Unit 2 exception from Containment Integrated Leak R t T Rate Testing ti (ILRT)

Revise Technical Specification (TS) 5.5.14, Containment Leak Rate Testing Basis: ASME testing better than ILRT Precedent: PINGP Unit 1 for 2004 SGRP Implementation: Prior to SGR Outage 35

R Regulatory l t Interface I t f - LARs LAR LAR: U1/U2 SG Program Changes Revise TS 3.4.19, SG Tube Integrity Revise TS 5.5.8, SG Program Revise TS 5.6.7, SG Tube Inspection Report Revise TS to recognize U2 with Alloy 690TT tubes Eliminate SG repair criteria (associated with OSGs)

Adopt TSTF-510 to SG Program Same SG Program g to both units Implementation: In conjunction with SGR Outage 36

SGR Project P j t Q&A Closing Remarks 37