IR 05000302/2011009

IR 05000302-11-009, on 01/01-04/22/2011; Crystal River Unit 3; Steam Generator Replacement Inspection

ML111330350
Person / Time
Site:	Crystal River
Issue date:	05/12/2011
From:	Mark Franke NRC/RGN-II/DRS/EB2
To:	Franke J Florida Power Corp, Progress Energy Florida
References
IR-11-009
Download: ML111330350 (328)

Text

UNITED STATES May 12, 2011

SUBJECT:

CRYSTAL RIVER NUCLEAR PLANT - STEAM GENERATOR REPLACEMENT INSPECTION PROGRESS REPORT 05000302/2011009

Dear Mr. Franke:

On April 22, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed inspections at your Crystal River Unit 3 Nuclear Plant in accordance with NRC Inspection Procedure (IP) 50001, Steam Generator Replacement Inspection. The enclosed inspection report documents inspection results, which were discussed on April 28, 2011, with you and members of your staff.

The inspections examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, interviewed personnel, and conducted plant walk downs, including visual examination of accessible portions of the containment building structure.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark E. Franke, Chief Operations Branch 2 Division of Reactor Safety Docket No. 50-302 License No. DPR-72

Enclosures:

1: Inspection Report 05000302/2011009 w/Attachment: Supplemental Information 2: Licensee Question Tracking Database

REGION II==

Docket No.: 50-302 License No.: DPR-72 Report No.: 05000302/2011009 Licensee: Progress Energy (Florida Power Corporation)

Facility: Crystal River Unit 3 Location: Crystal River, FL Dates: January 1, 2011 through April 22, 2011 Inspectors: R. Carrion, Senior Reactor Inspector L. Lake, Senior Reactor Inspector Approved by: Mark E. Franke, Chief, Operations Branch 2 Division of Reactor Safety Enclosure 1

SUMMARY OF FINDINGS

IR 05000302/2011009; 01/01-04/22/2011; Crystal River Unit 3; Steam Generator Replacement

Inspection This report covered an infrequently performed Steam Generator Replacement Project (SGRP)inspection performed by regional reactor inspectors from January 1, 2011, through April 22, 2011. This report also includes a list of issued inspection reports and a summary of the SGRP inspections performed prior to December 31, 2010. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

NRC-Identified

& Self-Revealing Findings No findings were identified.

Licensee-Identified Violations

None.

REPORT DETAILS

Background The licensee scheduled the replacement of its steam generators during Crystal River Unit 3 (CR3) Refueling Outage 16, which began in September 2009. The steam generators are housed in the reactor containment building. In preparation for the steam generator replacement, the licensee evaluated options for moving the existing steam generators out of the containment building and moving the new steam generators into the containment building. The licensee decided to make a construction opening in the containment building wall, approximately forty feet directly above the equipment hatch, to facilitate this evolution.

Fabricating the opening in the containment building wall included preparing the containment building by detensioning its pre-stressed tendons and then removing the concrete with high pressure water (a process known as hydro-demolition), removing the rebar and tendons from the area of the construction opening, and cutting the containment liner plate. A concrete delamination in Bay 3-4 of the containment structure was discovered while creating the SGR opening in October 2009. The licensee determined that the root cause of the delamination was related to the scope and sequence of the tendon detensioning process.

The licensees containment building repair plan included:

(1) additional detensioning of containment;

(2) removal of delaminated concrete;

(3) installation of reinforcement, including radial reinforcement through the delamination plane;

(4) placing of new concrete; (5)retensioning containment; and

(6) post-repair confirmatory system pressure testing. In early 2011, the licensee had completed repair steps 1 through 4 and was in the process of retensioning the containment. On March 14, 2011, during the final stages of the re-tensioning process, the licensee had indications that a new delamination occurred in Bay 5-6 of the containment structure.

Since September 2009, and through the end of this inspection period, Crystal River Unit 3 has remained shutdown, a mode of operation where containment building operability is not required.

The purpose of this inspection report is to document all inspection activities performed related to the steam generator replacement project for Unit 3 containment restoration activities including repairs due to the concrete delamination identified in October

OTHER ACTIVITIES

[OA]

4OA5 Steam Generator Replacement Inspection (IP50001)

.1 SGRP Inspection Activities through December 31, 2010

Inspection of the licensees SGRP began in September 2009 when the inspectors reviewed the preparations for heavy load movement and lifting and started the review of the design modifications associated with the project. Inspection of the licensees SGRP continued in the fourth quarter of 2009 and throughout 2010. Results of the SGRP inspections are documented in the quarterly integrated resident inspector reports. The following is a summary of the activities inspected. Details of the inspections can be found in the following inspection reports:

Inspection Report ADAMS ML # Report Section 05000302/2009004 ML093030165 4OA5.2 05000302/2009005 ML100250014 4OA5.2 05000302/2010002 ML101170619 4OA5.3 05000302/2010003 ML102090239 4OA5.2 05000302/2010004 ML103020127 4OA5.3 05000302/2010005 ML110270190 4OA5.2 Design and Planning The inspectors reviewed and examined the SGRP activities and compared them to the requirements of the American Society of Mechanical Engineers (ASME) Code. The inspectors reviewed Engineering Change (EC) 63038, Replacement Once Through Steam Generators (ROTSGs or RSGs), which included the design changes, analyses, evaluations, safety analyses, 10 CFR Part 50.59 change evaluation, configuration, materials, implementation, and post-modification testing acceptance. The inspectors reviewed EC 62500, RCS Hot Leg Cutting and Welding, EC 63016, Containment Opening; EC 63025, Main Feedwater Flow Accelerated Corrosion (FAC) Pipe Replacement, EC 63026, RCS Cold Leg Cutting and Welding, EC 63027, Secondary Side Large Bore Pipe Cutting and Welding, EC 63034, Structural Interferences, and EC 63039, Replacement Steam Generator Anchorage. The inspectors also reviewed selected work order (WO) packages prepared for the construction and implementation of the ECs to determine whether appropriate work processes and quality control hold points were implemented.

Steam Generator Removal and Replacement During the hydro-demolition process to create the construction opening in the containment wall, the licensee identified concrete cracks/separations. The concrete separations were located within the entire perimeter of the opening. An NRC Special Inspection Team was chartered to inspect the separation issues. Results of the Special Inspection, including the licensees root cause analysis, are documented in NRC Inspection Report 05000302/2009007 (ML1028610261). The licensee evaluated the containment wall cracks to modify the horizontal transfer system (HTS) supporting structures. Prior to the removal of the original steam generators (OSGs), the inspectors reviewed, observed, and evaluated the associated temporary and permanent modifications of the cutting, disconnecting, and the providing of temporary supports for the OSGs and cutoff piping. The inspectors observed lifting, rigging, downending and upending, and transporting of the OSGs, RSGs, and associated equipment; machining and preparations of the existing piping for the connections to the RSGs; welding and non-destructive examination (NDE) activities; and the radiological safety plan for the temporary storage and disposal building of the retired steam generators. The inspectors reviewed and observed the major structural modifications. The inspectors observed the licensee performance inspection of the steam generator hold-down bolts to verify that the bolts were acceptable to hold down the RSGs after the OSGs were moved from their cubicles. During the steam generator (SG) removal and replacement, the inspectors observed licensee activities associated with controls for excluding foreign material, including the primary and secondary side of the steam generators and in the related RCS openings, and the establishment of operating conditions including defueling, RCS draindown and system isolation. The inspectors also reviewed procedures, examination results, modification packages, and WO packages related to

the modifications, including the construction opening steel containment vessel (SCV)reinstallation, to ensure compliance with the requirements of the ASME Code.

RSG Fabrication, Preservice Inspection, and Baseline Inspection The inspectors reviewed records associated with the materials, fabrication, examination, and testing for the RSGs, and replacement hot leg piping subassemblies (Candy Canes), to verify compliance with the ASME Code. The inspectors also reviewed documentation and interviewed plant personnel regarding the pre-service and baseline testing of RSG tubing. The inspectors also reviewed documentation regarding the manufacture of the RSG tubing, including heat treatment records and nonconformance reports.

Welding The inspectors reviewed a sample of welding activities associated with the installation of the RSGs to evaluate compliance with licensee/contractor procedures and the applicable ASME Code. The inspectors reviewed joint configuration drawings, welding procedures, welding specifications, welding procedure qualifications, welder qualification records, weld data records, nuclear condition reports (NCRs), and post-weld heat treatment procedures.

Non-Destructive Examination The inspectors reviewed the NDE procedures, calibration and examination reports, and NCRs, and observed in-process NDEs, including liquid penetrant examinations (PTs),magnetic particle examinations (MTs), radiographic examinations (RTs), and ultrasonic examinations (UTs), and compared them to the requirements of the procedures and the ASME Code for the construction, pre-service, and baseline inspections.

Containment Construction Opening and Closure - Steel and Concrete Containment The inspectors reviewed the licensees activities associated with the concrete removal and the removal and restoration of the steel containment liner plate (SCLP) for the containment construction hatch opening, as detailed in the EC 63016, Containment Opening. The inspectors reviewed the plans for the cutting and restoration of the SCLP for the construction opening and compared post-testing requirements to the applicable ASME Code. The inspectors observed the hydro-demolition of concrete for the containment construction opening and reviewed the WO packages for the cutting of the liner plate to verify that the steps had been completed and documented. The inspectors also reviewed the welding procedures, procedure qualification records, and welder qualification records to confirm that the Code-required essential and supplemental essential welding variables were met. The inspectors reviewed the WO packages, including welding electrode receipt inspection, vacuum box leak testing, MT records, material certification records, and qualification and certification records for NDE personnel, equipment, and consumables.

Heavy Load, Rigging, Lifting, and Transporting Activities The inspectors reviewed the SG lifting preparation activities and lifting equipment load test data to ensure that they were prepared in accordance with regulatory requirements, appropriate industrial codes and standards, and to verify that the maximum anticipated

loads to be lifted would not exceed the capacity of the lifting equipment and supporting structures. The inspectors reviewed procedures, calculations, drawings, work packages, crane and equipment operator training and certificates, and load and function test records to verify that they were in accordance with regulatory requirements and appropriate industrial codes and standards. The inspectors also examined SGRP lifting, rigging, and transporting equipment, including the polar crane, mobile crane, the Temporary Lifting Device (TLD), the Horizontal Transfer System (HTS) (including its skid system), the down/upender device, the Outside Lift System (OLS), and the self-propelled modular transporter (SPMT). The inspectors observed a selective sampling of rigging, lifting, transportation, and positioning of the original and replacement SGs.

Quality Assurance (QA) Program and Corrective Actions The inspectors conducted a review of the quality assurance program and its implementation for the SG replacement to assess compliance with the requirements of 10 CFR Part 50, Appendix B. The inspectors also reviewed the surveillance reports and nonconformance reports issued for the root cause analyses, evaluations, repairs, or disposition during the manufacturing of the RSGs.

SG Post-Installation Verification and Testing The inspectors reviewed the SG post-installation verification and testing program to verify that the required post-installation verification and testing, procedural changes, and the adjustment of the instruments were properly identified.

Containment Detensioning The inspectors conducted a review of the licensees detensioning activities for the repair of the delaminated containment wall and restoration of the containment wall to it pre-construction opening condition, including associated ECs and Work Packages (WPs). The inspectors also observed vertical and horizontal tendon detensioning. The inspectors observed and reviewed the records of the acoustic monitors and strain gauges used to detect sound volumes and concrete strain changes potentially due to new cracks or compressive or tensile stress changes in the concrete during the detensioning process. The inspectors reviewed the procedures, drawings, calibrations, equipment and personnel qualifications, and the tendon detensioning communication plan associated with detensioning to verify that the licensee performed the activities in accordance with approved procedures.

Concrete Removal, Surface Preparation, and Concrete Placement Activities The inspectors conducted a review of the licensees activities associated with the removal of the damaged concrete and restoration of the containment wall. The inspectors reviewed associated documents, including ECs, WPs, specifications, drawings, test reports, and NCRs. The inspectors observed the process of the hydro-demolition of damaged concrete, the surface preparation of concrete after the hydro-demolition, and pull-out testing to assure that the concrete surface would have enough tensile strength to bond the new and original concrete. The inspectors reviewed radial rebar drilling; grouting; and identified void problems, and their respective resolutions. The inspectors observed rebar and formwork installation and tendon sleeve condition in preparation of the concrete pour. The inspectors also reviewed the associated engineering packages, WPs, and drawings to verify that licensee activities

were performed in accordance with approved documents. The inspectors observed concrete placement activities to verify that activities pertaining to concrete delivery time, flow distance, layer thickness, etc. conformed to industry standards established by the American Concrete Institute (ACI). The inspectors also observed that concrete placement activities were monitored by the licensee and contractors quality control personnel and engineers. The inspectors observed in-process concrete testing and reviewed the results for slump, air content, temperature, and unit weight, to verify that this was done in accordance with applicable American Society for Testing and Materials (ASTM) requirements. The inspectors checked the batch plant for its certification and reviewed its preparation for the concrete pour. During the concrete placement activities, the inspectors identified a finding of very low safety significance and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to establish measures to assure that testing of rebar splices would adhere to the requirements of ASME Boiler and Pressure Vessel Code. (Refer to Section 4OA5.3 of Inspection Report 05000302/2010004 (ML103020127) for additional details about NCV 05000302/2010004-03, Failure to Submit Production Splices of Swaged Mechanical Splices for Testing.)

Containment Dome Cracks On April 14, 1976, a delamination was identified in the containment dome during the final stages of containment construction and before initial plant startup. The area of the delaminated concrete was approximately circular in shape with a 105-foot diameter. The dome repair process included removal of the delaminated dome cap; removal of meridional, hoop, and radial reinforcement; and placement of a new dome cap.

Instrumentation was installed to monitor the dome during tendon detensioning, retensioning and the initial structural integrity. As part of the Special Inspection to assess the circumstances associated with the delamination discovered in 2009, NRC inspectors conducted a review of the licensees conclusion that the delamination identified in 1976 is not related to the current delamination. Additional background information and the results of these inspections are presented in NRC Inspection Report 05000302/2009007 (ML1028610261).

In 2009, NRC Special Inspection Team inspectors conducted concrete surface inspections on the containment dome and identified a rough and uneven surface condition of the dome surface. The inspectors reviewed the evaluation and corrective action that determined the rough and uneven surface condition of the dome has existed since the 1976 repair. The licensee had periodically completed numerous surface patches in an attempt to address the surface spalls. Following additional reviews of the dome tendon stresses and monitoring, the uneven surface also appeared to be a result of concrete installation and finishing from the 1976 repair and not related to settlement of the dome or the 2009 containment concrete wall delamination issue.

In 2010, the inspectors reviewed a condition assessment of the containment dome documented in the licensees engineering change package EC 74801. The licensee had included this assessment in the activities associated with its 2009 containment extent-of-condition investigation. Included was Impulse Response (IR) testing and core bores made in support of evaluating the IR data. Anomalies were identified and, to evaluate the anomalies, additional examinations were performed. A total of about 10,000 points were tested and a total of 30 core samples were removed. Visual inspections were performed of each core sample and a video scope inspection of each core hole was performed after the core sample was removed. This evaluation revealed

cracking in the plane of the dome (laminar cracking). The licensee determined that these anomalies were remnants of the repairs performed in 1976.

The inspectors reviewed the CTLGroup Project No. 059176 - Dome Report, which included the results of the examinations identified above. The information contained in this report was subsequently utilized in an engineering evaluation documented in Containment Dome Evaluation, Report No. CR-3-LI-537934-52-SE-0059. The engineering evaluation determined that the repairs made to the dome structure in 1976 are intact; that there are no significant anomalies, discrepancies, or structural issues which would affect the overall structural integrity of the dome structure; and that the structure is capable of performing its design basis functions as described in the Updated Final Safety Analysis Report (UFSAR).

Additional information is presented in NRC Inspection Report 05000302/2009007 (ML1028610261).

Containment Tendon Retensioning Plan The inspectors reviewed the containment tendon retensioning plan, testing plan, and schedule. The inspectors also interviewed licensee personnel and reviewed documents related to the retensioning and testing plans. The licensee conducted a detailed analysis to develop a tendon retensioning sequence that would minimize the possibility of causing new cracks or delaminations in the containment during the retensioning process. The retensioning process began in January 2011. The inspectors also reviewed licensee plans for Containment Building testing after completion of tendon retensioning and post-maintenance testing after restart.

.2 SGRP Inspection Activities January 1, 2011 through April 22, 2011

Discussion of Technical Issues The following issues were discussed with licensee personnel during this inspection period:

Bulges of Liner Plate The inspectors completed a review of the licensees actions related to containment liner bulges. The licensee developed a calculation to evaluate bulges in the CR3 containment liner plate. It was directed at determining an apparent cause for the bulges and establishing an analytically-based acceptance criterion for the bulges within the CR3 design basis. The analyses included finite element modeling of the liner and the associated anchorage to the concrete containment structure. The apparent cause for the bulges was determined to be a combination of elements, including geometrical imperfections in the original liner plate during construction. The calculations considered worst case configurations and a threshold for bulge size was established considering the effects that occur due to normal operation and accident conditions. The primary variables in the bulge evaluation were determined to be bulge size and thermal loading.

The calculation found that the bulges have an insignificant effect on the response of the structure due to various load combinations. The current bulges are bounded by the acceptance criteria in the analysis. To ensure that conditions are acceptable in the future, the licensee planned to include the bulges in the IWE program. The licensee added a summary evaluation to the EC, which includes steps to validate the effect of

retensioning on bulge size by measurement and evaluation of a representative sample before initiating Structural Integrity Test (SIT) pressurization as well as requirements to perform a complete baseline scan after completion of the SIT.

50.59 Evaluation The inspectors reviewed the licensees evaluation of the containment building modification resulting from the introduction of the construction opening and its subsequent restoration with respect to requirements of 10 CFR, § 50.59, Changes, Tests and Experiments, to verify that the design bases, licensing bases, and performance capability of the containment had not been degraded through the modification and to verify that the design and license basis documentation used to support changes reflect the design and license basis of the facility after the change had been made.

The inspectors review remained ongoing at the end of the inspection period. Remaining activities necessary to complete the 50.59 review included: verification that tendon retensioning activities and containment testing validated licensee design assumptions; verifying that post-modification testing adequately confirmed containment functionality via the scheduled Structural Integrity Test (SIT) and Integrated Leak Rate Test (ILRT)prior to unit startup; verifying that design basis documentation used to support changes and design basis documentation affected by changes had been adequately updated and reflected the modified design and license basis of the facility consistent with the restoration; and verifying that the licenses UFSAR had been updated accordingly.

Vertical Cracks of Containment Building One of the licensees design assumptions for containment repair was that vertical cracks discovered on the exterior wall of the containment building would close as the buildings tendons were retensioned. The inspectors walked down selected vertical cracks being monitored by the licensee to evaluate their condition. The licensee had measured the cracks periodically and determined that they were closing as the tendon retensioning process continued. The inspectors also visited the tendon control center where the retensioning process was controlled, and which housed the acoustic monitoring and strain gage instrumentation, and interviewed personnel in the center to better understand the operation of the systems being used and how the information obtained was interpreted.

The inspectors review of vertical cracks remained ongoing pending inspection of the containment building after all repairs are completed and tendons are fully retensioned, and the completion of the SIT and ILRT.

Tendon Re-tensioning Activities The inspectors reviewed the licensees re-tensioning plans, procedures, and drawings.

In addition the inspectors observed some of the re-tensioning work being performed on selected hoop tendons to verify that the work was being conducted per approved procedures.

SIT/ILRT Preparations The inspectors interviewed licensee personnel responsible for the planned SIT/ILRT to determine the status of the test preparations; walked down the containment building to verify the locations of the extensometers to be used to measure the containment movements during the SIT/ILRT; and discussed the licensees procedures to ensure that they conformed to industry standards and ASME Code requirements.

Events of March 14, 2011 On the afternoon of March 14, 2011, the licensee had completed the first retensioning sequence (Sequence #100, Hoop Tendons 42H41, 62H41, and 64H41) of the final pass (Pass #11). Per procedure, the licensee was waiting for the containment building to stabilize before beginning the next sequence and monitoring the structural behavior of the containment building via acoustical emissions monitors and strain gauges, specifically placed at various points of the structure to detect any abnormal/unexpected response to tendon retensioning. During this monitoring period, the strain gauges indicated an increase in strain and then failed high, and the acoustic monitors indicated a high level of acoustic activity in the bay bordered by Butresses #5 and #6 (Bay 5-6).

The phenomenon reportedly lasted for about twenty minutes. The licensee conducted impulse response (IR) non-destructive examination NDE techniques to determine the condition of the wall in Bay 5-6. The IR scans of the bay determined that there were numerous indications consistent with a delamination. By the end of the inspection period, the licensee had determined that the delamination was extensive in Bay 5-6 and was continuing to evaluate the condition of the entire containment structure. Future inspection activities by the NRC relating to the March 14, 2011, event are to be determined.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On April 28, 2011, the inspectors presented the inspection results to Mr. J. Franke, Site Vice President, and other members of licensee management via a telephone call. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

S. Cahill, Manager, Engineering

P. Dixon, Progress Energy

P. Fagan, RNP Technical Services Superintendent

G. Flavors, Nuclear Upgrades

J. Franke, Site Vice-President

T. Howard, Engineering

J. Holt, Site General Manager

J. Huegel, Nuclear Oversight

R. Knott, NPC Lead Engineer

M. Rigsby, Manager - Support Services

NRC personnel

D. Rich, Chief, Branch 3, Division of Reactor Projects

T. Morrissey, Senior Resident Inspector

R. Reyes, Resident Inspector

LIST OF ITEMS

Opened

None.

Closed

None.

Discussed

05000302/2010004-03 NCV Failure to Submit Production Splices of Swaged Mechanical Splices for Testing