IR 05000306/2013011
| ML14059A113 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/27/2014 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Davison K Northern States Power Co |
| Atif Shaikh | |
| References | |
| IR-13-011 | |
| Download: ML14059A113 (27) | |
Text
UNITED STATES ary 27, 2014
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2, STEAM GENERATOR REPLACEMENT INSPECTION REPORT 05000306/2013011
Dear Mr. Davison:
On January 16, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed a Steam Generator Replacement Inspection at your Prairie Island Nuclear Generating Plant, Unit 2. The enclosed inspection report documents the inspection results which were discussed on January 16, 2014, with you and other members of your staff.
The inspection 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, and interviewed personnel.
Based on the results of this inspection, two NRC-identified findings of very low safety significance were identified. One of the findings involved a violation of NRC requirements.
However, because of its very low safety significance and because the issue was entered into your Corrective Action Program, the NRC is treating this issue as a Non-Cited Violation (NCV)
in accordance with Section 2.3.2 of the NRC Enforcement Policy.
If you contest the subject or severity of any NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Prairie Island Nuclear Generating Plant.
If you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Prairie Island Nuclear Generating Plant. In accordance with Title 10, Code of Federal Regulations (CFR), Section 2.390 of the NRC's
"Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
David E. Hills, Branch Chief Engineering Branch 1 Division of Reactor Safety Docket No. 50-306 License No. DPR-60
Enclosure:
Inspection Report 05000306/2013011 w/Attachment: Supplemental Information
REGION III==
Docket No.: 50-306 License No.: DPR-60 Report No: 05000306/2013011(DRS)
Licensee: Northern States Power Company, Minnesota Facility: Prairie Island Nuclear Generating Plant, Unit 2 Location: Welch, MN Dates: August, 7, 2013, through January 16, 2014 Inspectors: A. Shaikh, Senior Reactor Inspector (Lead)
J. Bozga, Reactor Inspector N. Egan, Reactor Inspector K. Stoedter, Senior Resident Inspector E. Sanchez, Acting Resident Inspector M. Phalen, Senior Reactor Inspector Approved by: David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
Inspection Report 05000306/2013011(DRS); 08/7/2013 - 01/16/2014; Prairie Island Nuclear
Generating Plant, Unit 2; Steam Generator Replacement Inspection.
This report covers a five month announced infrequently performed inspection on steam generator replacements. The inspection was conducted by Region III based engineering, radiological, and security inspectors and the site resident inspectors. Two findings were identified by the inspectors.
One of the findings was considered a Non-Cited Violation (NCV) of NRC regulations and the other finding did not have an associated violation of NRC regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Components Within the Cross-Cutting Areas. Findings for which the SDP does not apply may be (Green) or be assigned a severity level after NRC management review. 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
and Self-Revealed Findings
Cornerstone: Initiating Events
- Green.
The inspectors identified a finding of very low safety significance (Green) involving the licensees failure to meet the requirements of the American Institute of Steel Construction (AISC) specification. Specifically, the licensee did not use the specified minimum yield strength of the outside lift system (OLS) girder material to establish an appropriate factor of safety to qualify the allowable loads that can be safely handled by the OLS girder. The AISC factor of safety to failure ensured the OLS girder would maintain structural integrity (no permanent deformation or structural failure) when subjected to the applied loads (lifted load, wind load, design basis earthquake load). This issue was entered into the licensees Corrective Action Program (CAP) as CAP 1404203, OLS calculation used actual material strength rather than ASTM. The licensee performed a functionality assessment to demonstrate that there was reasonable assurance the OLS girder remained capable of performing its intended design functions.
The inspectors determined the finding to be more than minor because the finding was associated with the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown.
Specifically, the load handling reliability of the OLS girder inherently decreased when the AISC requirements were not met. The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609,
Significance Determination Process, Attachment 0609.04, Phase 1 -- Initial Screening and Characterization of Findings, Table 3. Since the finding was associated with shutdown (defueled) conditions, the inspectors used IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined that none of the conditions constituting a loss of control were met as described in Appendix G,
Attachment 1, Phase I Operational Checklists for Both PWRS and BWRS, for this finding and no Phase II or Phase III analysis was required. Therefore, the inspectors determined that this finding was of very low safety significance. No violation of regulatory requirements is associated with this finding. The inspectors identified that there was a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding for the licensee failure to ensure the OLS girder reflected the intended design margins (Section 4OA5.2).
Cornerstone: Barrier Integrity
- Green.
The inspectors identified a finding of very low safety significance and associated NCV of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III,
Design Control, for the failure to provide adequate design control measures for the steam generator blowdown (SGBD) pipe supports 8D-2SGB-1A, 2-RBDH-5294, 2-RBDH-606, 2-RBDH-363, 2-RBDH-350, 2-RBDH-349, 2-RBDH-339, and 2-RBDH-358. Specifically the SGBD pipe supports design was non-conservative with respect to Class I requirements as defined in Updated Safety Analysis Report (USAR) Section 12, Plant Structures and Shielding, and referenced specifications. The licensee documented the violation in its CAP as CAPs 1405404 and 1412225 and performed an evaluation to demonstrate that there was reasonable assurance that the SGBD pipe supports remained capable of performing their safety functions.
The inspectors determined the finding was more than minor because the finding adversely affected the barrier integrity cornerstone and the associated cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensees calculations were not sufficient to demonstrate that the pipe supports were capable of properly supporting SGBD piping and isolation valves during design basis events, and hence ensure containment integrity. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, The Significance Determination Process (SDP) for Findings At-Power, Appendix A, Exhibit 3 (Section B). The inspectors determined that this finding was very low safety significance (Green) because each of the screening questions was answered no. Specifically, the SGBD pipe supports were subsequently determined to be capable of performing their safety function. The inspectors identified a Human Performance, Documentation (H.7) cross-cutting aspect associated with this finding for the licensees failure to ensure complete, accurate, and, up-to-date design documentation. Specifically, the licensee failed to provide adequate oversight of design calculations and documentation of as-built conditions during the SGBD pipe support re-analysis conducted to support the steam generators replacement (Section 4OA5.1).
Licensee-Identified Violations
No violations of significance were identified.
REPORT DETAILS
OTHER ACTIVITIES
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity
4OA5 Other Activities
.1 Design Changes and Modifications to Systems, Structures, and Components
a. Inspection Scope
The inspectors reviewed engineering changes associated with the replacement steam generators, secondary side piping, and steam generator support systems. During these reviews, the inspectors focused on key design aspects and modifications of the replacement steam generators and verified that changes to the facility as described in the Updated Safety Analysis Report (USAR) were reviewed and documented in accordance with 10 CFR 50.59. The inspectors used IP 71111.17, Evaluation of Changes, Tests and Experiments, and Permanent Plant Modifications as guidance, as suggested in IP 50001, to complete these reviews.
The inspectors reviewed design calculations associated with the design changes to the steam generator blowdown piping, external recirculation piping and main steam piping.
The inspectors also reviewed the design specification and design stress report for the replacement steam generators. The inspectors reviewed design calculations associated with the changes to the steam generator vertical column supports. The inspectors performed walkdowns of select main steam pipe supports and steam generator vertical column supports.
b. Findings
Steam Generator Blowdown Pipe Support Anchorages Failure to Meet Design Requirements
Introduction:
The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure the adequacy of the design of several safety-related Steam Generator Blowdown (SGBD) pipe supports. Specifically the design was non-conservative with respect to Class I requirements as defined in Updated Safety Analysis Report (USAR) Section 12, Plant Structures and Shielding, and referenced specifications.
Description:
The SGBD pipe supports which are upstream of the SGBD isolation valves are required to be Class I per USAR, Table 12.2-1, Structures, Systems, and Components Classification. The USAR Section 12.2.1.1 defines Class 1 as Those structures and components including instruments and controls whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of substantial amounts of radioactivity, and those structures and components vital to safe shutdown and isolation of the reactor. The USAR Sections 5.2.2.1.1 and 9.2.2 delineated that the safety-related function of the SGBD system was to isolate during a steam generator tube rupture event or main steam line break event and to maintain the containment boundary.
The inspectors identified the following two representative examples in which the licensee failed to demonstrate that the SGBD pipe supports met Class 1 requirements:
1) The inspectors reviewed Calculation Book No. 225, Unit 2 Reactor Building Steam Generator Blowdown System Pipe Rupture Restraints, Revision 0. This calculation was to demonstrate Class 1 compliance for SGBD Pipe Support 8D-2SGB-1A. The inspectors identified that the calculation did not validate the load path and structural integrity of the anchor bolts and anchor plate, which support and are part of Pipe Support 8D-2SGB-1A. The licensee documented these deficiencies in CAP 1405404, "Legacy calculation has missing pages", dated November 7, 2013.
2) The inspectors reviewed Calculation No. 020781-02-PISGR-45DK-0031, "Design Calculation - Blowdown Support Analysis-SGR U2," Revision 4. This calculation was to demonstrate Class 1 compliance of the SGBD pipe supports affected by the replacement of the steam generators. Pipe Supports 2-RBDH-5294, 2-RBDH-606, 2-RBDH-363, 2-RBDH-350, 2-RBDH-349, 2-RBDH-339, and 2-RBDH-358 used anchor bolt allowables from Engineering Specification 3.2.1.8, Specification for Concrete Expansion Anchors, Revision 3. The allowable loads that the anchor bolts can withstand during design bases events are specified based on a required minimum spacing and minimum edge distance. The minimum spacing is defined as centerline-to-centerline distance between adjacent anchors. The minimum edge distance is defined as the distance between the centerline of an anchor and a free edge where no concrete exists. The inspectors asked the licensee whether the minimum spacing and edge distances were met for the anchor bolts that comprised the anchorage of the aforementioned supports. The licensee subsequently performed walkdowns and discovered that they had not met the minimum spacing and edge distance for the anchorage for each of the aforementioned supports and therefore, the licensee did not appropriately reduce the allowable load that the anchor bolts can withstand during a design basis event. The licensee documented these deficiencies in CAP 1412225 Deficiency in Analysis of U2 Blowdown Supports, dated December 22, 2013.
Although these were existing, older calculations, the licensee did not identify the deficiencies in the calculations during the use of these calculations while performing a re-analysis of the SGBD pipe supports for the steam generator replacement project. As a result of the inspectors concerns, the licensee performed an evaluation of the aforementioned SGBD pipe supports to address the various deficiencies and demonstrate that there was reasonable assurance that the pipe supports remained capable of performing their intended safety functions. The inspectors reviewed the licensees evaluation and concluded there was sufficient rationale to support the final conclusion.
Analysis:
The inspectors determined the licensees failure to perform adequate evaluations to demonstrate Class I compliance of the SGBD pipe supports was contrary to the design control measures requirements per 10 CFR Part 50, Appendix B, and was a performance deficiency. Specifically, compliance with Class I requirements for the SGBD pipe supports ensures structural integrity of these pipe supports and associated piping during Class 1 design basis events. In accordance with IMC 0612, Issue Screening, Appendix B, the inspectors determined the performance deficiency affected the Barrier Integrity Cornerstone. The performance deficiency was determined to be more than minor, and a finding, because it adversely affected the associated cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensees calculations were not sufficient to demonstrate that the pipe supports were capable of properly supporting SGBD piping and isolation valves during design basis events, and hence ensure containment integrity.
The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, The Significance Determination Process (SDP) for Findings At-Power, Appendix A, Exhibit 3 (Section B).
The inspectors determined that this finding was very low safety significance (Green)because each of the screening questions was answered no. Specifically, the SGBD pipe supports were subsequently determined to be capable of performing their safety function.
The inspectors identified a Human Performance, Documentation (H.7) cross-cutting aspect associated with this finding for the licensees failure to ensure complete, accurate, and, up-to-date design documentation. Specifically, the licensee failed to provide adequate oversight of design calculations and documentation of as-built conditions during the SGBD pipe support re-analysis conducted to support the steam generators replacement.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control states, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.
Contrary to the above, 1) As of November 7, 2013, in Calculation Book No. 225, Unit 2 Reactor Building Steam Generator Blowdown System Pipe Rupture Restraints, Revision 0, the licensees design control measures failed to verify adequacy of the SGBD anchor plate and anchor bolts for SGBD Pipe Support 8D-2SGB-1A. Specifically, no calculation was performed to verify the anchor bolts and anchor plate could withstand the applied loads.
2) As of December 22, 2013, in Calculation No. 020781-02-PISGR-45DK-0031, "Design Calculation - Blowdown Support Analysis-SGR U2", Revision 4, the licensees design control measures failed to ensure adequacy of the Pipe Supports 2-RBDH-5294, 2-RBDH-606, 2-RBDH-363, 2-RBDH-350, 2-RBDH-349, 2-RBDH-339, and 2-RBDH-358 design. Specifically, the anchor bolts allowables were not based on the spacing and edge distances that existed in the plant.
The licensee entered this violation into their Corrective Action Program as CAPs 1405404 and 1412225, and performed a corrective action of evaluating the SGBD pipe supports operability. The licensee determined that the SGBD were operable but nonconforming and the inspectors reviewed the licensees evaluation and did not have any further questions.
Because this violation was of very low safety significance (Green), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000306/2013011-01).
.2 Engineering Design, Modification, Testing, and Analysis Associated with Steam Generator
Lifting and Rigging
a. Inspection Scope
The inspectors reviewed the adequacy of the lifting program for lifts of the steam generators inside and outside containment assuring that it was prepared in accordance with regulatory requirements, appropriate industrial codes, and standards; and verified that the lifting and rigging equipment, Outside Lift System (OLS), laydown areas, equipment hatch transfer system, containment polar crane and containment polar crane support structure were adequate to withstand the maximum anticipated loads to be lifted.
The inspectors reviewed engineering changes (EC) 16803 and EC 16804 that were associated with the rigging and lifting of the old and replacement steam generators inside and outside containment. The inspectors selected and reviewed samples of design specifications, corrective actions, change requests, and design calculations to confirm that the engineering changes were in compliance with applicable codes and standards. In addition, the inspectors reviewed inspection records for the containment polar crane and OLS that were done before and after each lift of the steam generator. The inspectors reviewed test records of the OLS for the maximum lifted load.
The inspectors reviewed the adequacy of the haul route evaluation and the documentation of haul route for load testing and transport of the steam generators. The inspectors verified that they had been prepared in accordance with regulatory requirements and appropriate industrial codes and standards. The inspectors also discussed the transport path load testing with the licensees steam generator replacement project (SGRP) engineering personnel and performed a walkdown of the haul route.
b. Findings
Outside Lift System Girder Failure to Meet American Institute of Steel Construction Requirements
Introduction:
The inspectors identified a finding of very low safety significance (Green)involving the licensees failure to meet the requirements of the American Institute of Steel Construction (AISC) specification. Specifically, the licensee did not use the specified minimum yield strength of the outside lift system OLS girder material to qualify the allowable loads that can be safely handled by the OLS girder.
Description:
The OLS was a non-safety-related structure that was erected during the Unit 2 Prairie Island Steam Generator Replacement to move parts of the old and replacements SGs inside and outside of containment. The OLS was comprised of a main horizontal steel girder supported at each end by vertical steel members.
Design Bases Calculation No. C-3218-12, Revision 2, Outside Lift System Structural Analysis and Interface Loads evaluated the OLS for the applied loads due to lifting the old and replacement steam generators concurrent with other design loads such as impact, dead load of OLS itself, and wind loads.
Design Basis Calculation No. PI-996-111-007-S01, Analysis of Outside Lift System for DBE Loads, Revision 1 evaluated the OLS for the applied loads due to lifting the old and replacement steam generators concurrent with other design loads such as dead lead of OLS itself, and design basis earthquake loads.
In both calculations, the licensee used the AISC specification to establish the structural adequacy of the OLS steel girder. The AISC specification required the allowable stress to be based on the specified minimum yield strength of the material. However, the licensee used certified material test report strength, which is reflective of actual material yield strength of the girder material for the evaluation of OLS girder. The use of actual material yield strength did not ensure that the appropriate AISC factor of safety was established for the OLS girder. The AISC factor of safety ensures the OLS girder would maintain structural integrity (no permanent deformation or structural failure) when subjected to the applied loads (lifted load, wind load, design basis earthquake load). The load handling reliability of the OLS girder inherently decreased when the AISC requirements were not met.
The failure to comply with the AISC specification on safety factor and demonstrate structural integrity of the OLS reduced the confidence that a failure of the OLS would not occur. A failure of the OLS could result in a heavy load drop accident that challenges plant stability. This issue was entered into the licensees Corrective Action Program as CAP 1404203, OLS calculation used actual material strength rather than ASTM, dated October 30, 2013. As part of the licensees corrective actions, the licensee performed a functionality assessment of the OLS girder for the noncompliance with AISC requirements.
The licensee functionality assessment concluded that there was reasonable assurance that the OLS girder remained capable of performing its intended design functions under the loads applied during the steam generator lifts. The inspectors reviewed the licensees functionality assessment and concluded that there was sufficient rationale to support the final conclusion.
Analysis:
The inspectors determined the licensees failure to meet AISC requirements for the OLS main support girder was a performance deficiency. In accordance with IMC 0612, Issue Screening, Appendix B, the inspectors determined the performance deficiency was more than minor, and a finding, because the performance deficiency was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown, as well as power operations.
The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase I -- Initial Screening and Characterization of Findings, Table 3. Since the finding was associated with shutdown (defueled) conditions, the inspectors used IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors determined that none of the conditions constituting a loss of control were met as described in Appendix G, Attachment 1, Phase I Operational Checklists for Both PWRS and BWRS, for this finding and no Phase II or Phase III analysis was required. Therefore, the inspectors determined that this finding was of very low safety significance. Specifically, the OLS girder remained capable of performing its intended design functions under the loads applied during the steam generator lifts. In addition, the OLS was load tested before its use during the steam generator replacement.
The inspectors identified a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding. Specifically, the licensee failed to ensure the OLS girder reflected the intended design margins.
Enforcement:
No violation of regulatory requirements is associated with this finding (FIN 05000306/2013011-02).
.3. Radiation Protection Program Controls, Planning, and Preparation
a. Inspection Scope
The inspectors reviewed radiation protection program controls, planning, and preparation in the following areas utilizing applicable portions of Baseline Inspection Procedures (IP)71124.01, 71124.02, 71124.03, 71124.04, and 71124.06 as guidance:
- As-Low-As-Is-Reasonably- Achievable (ALARA) planning;
- Dose estimates and dose tracking;
- Exposure controls including temporary shielding;
- Contamination controls;
- Radioactive material management;
- Radiological work plans and controls;
- Emergency contingencies;
- Project staffing and training plans; and
- Airborne radioactivity effluent controls
b. Findings
No findings were identified.
.4 Security Considerations Associated with Vital and Protected Area Barriers
a. Inspection Scope
The inspectors walked down areas associated with vital and protected area barriers that could be affected during replacement activities and concluded that they were operational and that the licensee was in compliance with security requirements.
b. Findings
No findings were identified.
.5 Controls and Plans to Minimize Adverse Impacts on Operating Unit and Common Systems
a. Inspection Scope
The inspectors also reviewed the licensees plan to minimize impacts to the operating unit and any common systems during SGR activities. The inspectors walked down plant areas that had the potential to be impacted by steam generator replacement (SGR) activities.
These walk downs included the area surrounding the condensate storage tanks since these tanks were at increased risk of being punctured if a heavy load was dropped during lifting/rigging activities. The inspectors verified that the licensee had appropriately managed the configuration of the operating unit, including maintaining defense-in-depth strategies for safety-related equipment. The inspectors also verified that the licensee implemented the compensatory measures discussed in their plan for the specific areas reviewed. Lastly, the inspectors ensured that any increased risk to the operating unit or common systems was appropriately reflected in the daily operating unit risk-assessments as required by 10 CFR 50.65(a)(4).
.6 Welding and Non-Destructive Examination (NDE) Activities
a. Inspection scope
The inspectors reviewed the following welding and NDE activities associated with the steam generator replacement project to evaluate compliance with the American Society of Mechanical Engineers (ASME) Code Section XI and Section V requirements. These activities were inspected in accordance with IP 71111.08, Inservice Inspection Activities.
- Special Procedures for welding and NDE on the replacement steam generators (RSGs) and connecting reactor coolant system (RCS) piping;
- Training and qualifications for personnel performing welding and NDE on RSGs and connecting RCS piping;
- NDE including radiography results and work packages for welds fabricated on RSGs and connecting RCS piping;
- Completion of pre-service NDE requirements for welds fabricated on RSGs and connecting RCS piping; and
b. Findings
No findings were identified.
.7 Activities Associated with Lifting and Rigging
a. Inspection Scope
The inspectors examined the SGRP lifting equipment necessary to perform steam generator rigging and transport; design evaluation and use of the OLS; and load drop protection. The inspectors performed direct observation of some of the heavy lifts performed both inside and outside containment to remove the old steam generators and install the new steam generators. The inspectors also verified that these activities were bound by the analyses and evaluations the licensee performed to support these activities.
In addition, the inspectors reviewed crane and OLS personnel training certifications.
b. Findings
No findings were identified.
.8 Old and New Steam Generator Cutting, Movement and Reconnection
a. Inspection Scope
The inspectors observed various portions of the process of the old steam generators being cut and lifted from the steam generator vaults through the penetrations in the steel containment vessel and shield building to the hydraulic trailer transporter. The inspectors also observed various portions of the sequence of the replacement steam generators being transferred from the hydraulic trailer transporter, upended, lifted, and positioned into their respective steam generator vaults, and reconnected to the reactor coolant system piping.
During these observations, the inspectors performed visual inspections of the OLS and the hydraulic trailer transporter.
b. Findings
No findings were identified.
.9 Steam Generator Hold-Down Bolts and Major Structural Modifications
a. Inspection Scope
There were no major structural modifications to the plant to facilitate the replacement of the steam generators for the inspectors to inspect or review. In addition, the old steam generator hold-down bolts were replaced with new hold-down bolts. The inspectors reviewed the evaluation associated with the replacement of the steam generator hold-down bolts.
b. Findings
No findings were identified.
.10 Operating Conditions throughout the Steam Generator Replacement Process
a. Inspection Scope
The inspectors routinely inspected the following activities as they occurred throughout this inspection period:
- Establishment of operating conditions including defueling, reactor coolant system drain-down, and system isolation and safety tagging/blocking;
- Implementation of radiation protection controls including: implementation of ALARA, radiological exposure, contamination, and airborne contamination controls planned for cutting, welding, and other activities including contaminated interference removal. Also, implementation of any special controls for contaminated tools and waste were reviewed;
- Implementation of controls for excluding foreign materials in the primary and secondary side of the SGs and in the reactor coolant system openings; and
- Installation, use and removal of temporary services directly related to steam generator replacement activities.
b. Findings
No findings were identified.
.11 Radiological Safety Plans for Disposal of Old Steam Generators
a. Inspection Scope
The inspectors reviewed the licensees plans for disposal of old steam generators offsite and evaluated whether the shipping documents indicated the proper shipper name; emergency response information and a 24-hour contact telephone number; accurate curie content and volume of material; and appropriate waste classification, transport index, and UN number for the following radioactive shipments:
- Shipping Record 13-047; Radioactive Waste Shipment Old Steam Generators Upper Section; November 16, 2013;
- Shipping Record 13-048; Radioactive Waste Shipment Old Steam Generators Upper Section; November 16, 2013;
- Shipping Record 13-052; Radioactive Waste Shipment Old Steam Generators Lower Section; December 26, 2013; and
- Shipping Record 13-053; Radioactive Waste Shipment Old Steam Generators Lower Section; December 26, 2013.
Additionally, the inspectors assessed whether the shipment placarding was consistent with the information in the shipping documentation.
b. Findings
No findings were identified.
.12. Steam Generators Post-Installation Verification and Testing
a. Inspection scope
The inspectors performed selective reviews and inspections, consistent with the safety significance, of the following areas: containment leak testing; post-installation inspections and verification; reactor coolant system leakage testing; calibration and testing of instrumentation for both the primary and secondary side systems affected by the SG replacement; and procedures for equipment performance testing required to confirm the design and to establish baseline measurements, during post-installation and power ascension. Specific activities observed included:
- SP 2070 - Reactor Coolant System Integrity Test;
- SP 2072.EH - Local Leak Rate Test of Equipment Hatch Containment System;
- SP 2132 - Unit 2 Personnel and Maintenance Airlock Door Leakage Test;
- SP 2750 - Unit 2 Containment Closeout Verification;
- ST 2RSG-HYDRO21 - 21 Replacement Steam Generator Hydrostatic Test; and
- ST 2RSG-HYDRO22 - 22 Replacement Steam Generator Hydrostatic Test.
b. Findings
No findings were identified.
4OA6 Meetings
.1 Exit Meeting Summary
On January 16, 2014, the inspectors presented the inspection results to Mr. Kevin Davison and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was either returned to the licensee staff or will be properly disposed of when no longer needed.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- K. Davison, Site Vice President
- J. Hallenbeck, Site Engineering Director
- S. Sharp, Plant Manager
- M. Murphy, Regulatory Affairs Director
- B. Boyer, Radiation Protection Manager
- S. Martin, Nuclear Oversight Manager
- T. Downing, Site ISI Program
- J. Hamilton, Security Manager
- J. Ruttar, Operations Manager
- D. Vincent, Regulatory Assurance
Nuclear Regulatory Commission
- K. Stoedter, Senior Resident Inspector
- E. Sanchez, Acting Resident Inspector
- T. Wengert, Senior Project Manager, Office of Nuclear Reactor Regulation
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened
- 05000306/2013011-01 NCV Steam Generator Blowdown (SGBD) Pipe Support Anchorages Failure to Meet Design Requirements (Section 4OA5.1)
- 05000306/2013011-02 FIN Outside Lift System (OlS) Girder Failure to Meet American Institute of Steel Construction (AISC) Requirements (Section 4OA5.2)
Closed
- 05000306/2013011-01 NCV Steam Generator Blowdown (SGBD) Pipe Support Anchorages Failure to Meet Design Requirements (Section 4OA5.1)
- 05000306/2013011-02 FIN Outside Lift System (OLS) Girder Failure to Meet American Institute of Steel Construction (AISC) Requirements (Section 4OA5.2)
Discussed
None Attachment