ML14204A317
| ML14204A317 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/22/2014 |
| From: | Dave Hills NRC/RGN-II/DRS/EB1 |
| To: | Lieb R FirstEnergy Nuclear Operating Co |
| Atif Shaikh | |
| References | |
| IR-13-010 | |
| Download: ML14204A317 (33) | |
See also: IR 05000346/2013010
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, IL 60532-4352
July 22, 2014
Mr. Raymond Lieb
Site Vice President
FirstEnergy Nuclear Operating Co.
Davis-Besse Nuclear Power Station
5501 N. State Rte. 2, Mail Stop A-DB-3080
Oak Harbor, OH 43449-9760
SUBJECT: DAVIS-BESSE NUCLEAR POWER STATION, STEAM GENERATOR
REPLACEMENT INSPECTION REPORT 05000346/2013010
Dear Mr. Lieb:
On July 2, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed a Steam Generator
Replacement Inspection at your Davis-Besse Nuclear Power Station. The enclosed inspection
report documents the inspection results, which were discussed on July 2, 2014, with 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, one unresolved item (URI) and three NRC-identified
findings of very low safety significance were identified. All three findings involved a violation of
NRC requirements. However, because of their very low safety significance and because the
issues were entered into your Corrective Action Program, the NRC is treating these issues as
Non-Cited Violations (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
Davis-Besse Nuclear Power Station. In addition, 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 Davis-Besse Nuclear Power Station.
R. Lieb -2-
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 James E. Neurauter Acting For/
David E. Hills, Branch Chief
Engineering Branch 1
Division of Reactor Safety
Docket No. 50-346
License No. NPF-3
Enclosure:
Inspection Report 05000346/2013010
w/Attachment: Supplemental Information
cc w/encl: Distribution via LISTSERV
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No.: 50-346
License No.: NPF-3
Report No: 05000346/2013010
Licensee: FirstEnergy Nuclear Operating Company (FENOC)
Facility: Davis-Besse Nuclear Power Station
Location: Oak Harbor, OH
Dates: December 2, 2013, through July 2, 2014
Inspectors: A. Shaikh, Senior Reactor Inspector (Lead)
J. Neurauter, Senior Reactor Inspector
V. Meghani, Reactor Inspector
N. Egan, Reactor Inspector
D. Kimble, Senior Resident Inspector
T. Briley, Resident Inspector
M. Mitchell, Senior Reactor Inspector
Approved by: David E. Hills, Chief
Engineering Branch 1
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 05000346/2013010; 12/2/2013 - 7/2/2014; Davis-Besse Nuclear Power Station; Steam
Generator Replacement Inspection.
This report covers a seven 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. One URI and three findings
were identified by the inspectors. All three findings were considered Non-Cited Violations (NCV)
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.
A. NRC-Identified and Self-Revealed Findings
Cornerstone: Barrier Integrity
Green. The inspectors identified a finding of very low safety significance (Green) and an
associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of
Special Processes, when the licensee failed to use a qualified procedure for ultrasonic
(UT) examination of the Shield Building reinforcing bars (rebar). Specifically, the licensee
used a site approved UT examination procedure that had not been qualified to examine the
total length of approximately twenty four inches of rebar as specified in the procedure due
to near field scanning limitation.
The inspectors determined that the performance deficiency of using an unqualified
procedure was more than minor and; therefore, a finding because if left uncorrected, the
performance deficiency would have the potential to lead to a more significant safety
concern. Specifically, absent NRC identification, the licensee would have continued use of
the unqualified UT examination procedure to examine potential degradation in potentially
damaged rebar in the safety-related shield building. Therefore, the licensee could
potentially have returned the shield building back to service with unacceptable flaws
existing in the rebar. The inspectors determined the finding could be evaluated using the
SDP in accordance with IMC 0609, Significance Determination Process,
Attachment 0609.04, Initial Characterization of Findings. The inspectors answered 'Yes
to the questions in Section A of Table 3; and; therefore, the finding was evaluated using the
SDP in accordance with IMC 0609, The Significance Determination Process for Shutdown
Operations, Appendix G, Attachment 1, Exhibit 4, Barrier Integrity Screening Questions.
The inspectors answered all the questions in Exhibit 4 and determined that this finding did
not result in degraded physical integrity of the containment during shutdown operations nor
did it affect any shutdown safety functions. Therefore, the finding was determined to have
very low safety significance (Green). The inspectors determined that this finding had a
cross-cutting aspect in the area of Problem Identification and Resolution, and Evaluation for
the licensees failure to thoroughly evaluate issues to ensure that resolutions address
causes and extent of conditions commensurate with their safety significance. Specifically,
the licensee failed to initially consider the entire length of rebar for potential evaluation and
hence, did not consider the appropriate extent of condition. [P.2] (Section 4OA5.9).
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Cornerstone: Barrier Integrity
Green. A finding of very low safety significance and associated Non-Cited Violation of
10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the
inspectors for the licensees failure in 2011 to properly repair concrete voiding in the shield
building that had been identified during that construction opening restoration.
The inspectors determined the performance deficiency of failure to completely repair the
void during the 2011 shield building restoration was more than minor and; therefore, a
finding because the performance deficiency was associated with the Barrier Integrity
cornerstone attribute of Design Control and adversely impacted the cornerstone objective
to provide 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 failure to completely repair the concrete
voiding in 2011 resulted in the operation of the plant with the shield building in a condition
non-conforming to its design basis. The inspectors reviewed the finding using Attachment
0609.04, Initial Characterization of Findings, Table 3 - SDP Appendix Router. The
inspectors answered No to all the questions in Section A of Table 3 and; therefore, the
finding was evaluated using the SDP in accordance with IMC 0609, The Significance
Determination Process (SDP) for At-Power Operations, Appendix A, Exhibit 3, Barrier
Integrity Screening Questions. The inspectors answered all the questions in Exhibit 3 and
determined that this finding did not represent an actual open pathway in the physical
integrity of reactor containment. Therefore, the finding was determined to have very low
safety significance (Green). The inspectors determined that this finding had a cross-cutting
aspect in the area of Human Performance, Conservative Bias, for the licensees failure to
use decision making practices that emphasize prudent choices over those that are simply
allowable. Specifically, the licensee failed to implement a conservative decision to inspect
the shield building inside surface void area after repairs had been made during the opening
restoration in 2011. Therefore, the licensee missed the opportunity to identify that they had
not adequately repaired the void. [H.14] (Section 4OA5.9).
Cornerstone: Security
Green. Details of this finding and associated NCV are documented in Official Use Only
inspection report (IR) 05000346/2014404.
B. Licensee-Identified Violations
No violations of significance were identified.
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REPORT DETAILS
4. 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 performed a review of modifications and activities related to replacement of
the steam generators. This review was performed to determine whether the replacement
steam generators were designed in accordance with Section III of the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The inspectors also
reviewed the shield building and containment vessel temporary construction openings to
determine if they were in accordance with the design assumptions and that the restoration
maintained required strength and integrity. In addition, the inspectors reviewed calculations
of structures and plant equipment and lay-down areas to determine whether there was
sufficient structural capacity to handle the heavy loads.
The inspectors reviewed certified design specifications, certified design reports, ASME
Code reconciliation reports, fabrication deviation notices, non-conformance reports (NCRs),
and design calculations to confirm that the replacement steam generators were in
compliance with the requirements of ASME Boiler and Pressure Vessel Code,Section III,
Subsection NB. The inspectors confirmed that the design specifications and design reports
were certified by registered professional engineers competent in ASME Code
requirements. The inspectors confirmed that adequate documentation existed to
demonstrate that the certifying registered professional engineers were qualified in
accordance with the requirements of the ASME Code Section III. The inspectors also
confirmed that the replacement steam generators were procured as Code NPT stamped
components.
The inspectors also reviewed a sample of piping system design calculations affected by the
steam generator replacement including large bore piping associated with the reactor
coolant system (RCS) 1/2-Loop Model, as well as small bore piping for the reactor vessel
continuous vent, the hot leg vent, and the hot leg instruments.
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b. Findings
Title 10 CFR 50.59 Safety Evaluation Not Performed for Calculated Component Damping
Introduction: The inspectors identified an unresolved item (URI) regarding the licensees
use of a calculated critical damping factor using Lubrite plate friction in the DB-1 RCS 1/2 -
1/2-Loop Model coupled system seismic evaluation. Specifically, the inspectors questioned
whether the licensee could increase its licensing basis damping factors by calculation
without prior NRC-approval.
Description: The inspectors reviewed the effect of the steam generator replacement on
RCS 1/2-Loop Model to verify the change was consistent with the design and licensing
basis reviewed by the NRC. Specifically, the inspectors reviewed Engineering Change
Package () 12-0474-000, Revision 1, and the licensees associated 10 CFR 50.59 Screen
12-03404 for the RCS 1/2-Loop coupled system model. The inspectors utilized NRC
endorsed NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, for review
of 10 CFR 50.59 Screen 12-03404.
During review of licensee Calculation 32-9191681-001, DB-1 RCS 1/2-Loop Model and
Loading Analysis, the inspectors identified that the analytical model for seismic analysis
utilized a damping value higher than the value described in the USAR. Specifically, the
licensee utilized friction properties associated with the steam generator slider support
(Lubrite plate) to calculate an effective damping factor and modify the seismic response
spectra for operating basis earthquake (OBE) and safe shutdown earthquake (SSE).
The inspectors reviewed the USAR for appropriate seismic damping. Damping for the RCS
1/2-Loop Model was described in Section 3.7.2.14, Coupled System Damping, which
states:
Although various components within a model possess different values of critical
damping (example: shield wall - 2 percent, reactor, steam generators, and
pressurizer - 1 percent and piping - 1/2 percent), the lowest damping value was
used for all components in the model by the NSSS Supplier. Re-analysis of this
model utilizes the damping factors from Subsection 3.7.1.3.
As indicated in USAR Section 3.7.1.3 and Table 3.7-1, the plant was originally licensed with
piping system damping of 1/2 percent of critical damping. In addition, the licensee
requested and the NRC approved use of ASME Code Case N-411, Alternative Damping
Values for Response Spectra Analysis of Class 1, 2, and 3 Piping,Section III, Division 1.
The inspectors noted that USAR Section 3.7.1.3 had been revised to include, Higher
damping values than those listed in Table 3.7-1 are allowed, provided proper justification
(i.e., test results, analysis etc.) is available for specific components or equipment. The
inspectors considered proper justification to include a 10 CFR 50.59 evaluation to
determine whether the increase in damping change required prior NRC-approval. The
inspectors further verified that 10 CFR 50.59 Screen 12-03404 did not evaluate the
changes to the damping value used in the RCS 1/2-Loop Model.
The inspectors further determined the licensee initially modified RCS 1/2-Loop Model
damping by calculation in 1990 in Calculation 32-1177088-00 as part of the modification to
remove the snubber seismic restraints at the base of the original steam generators. The
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inspectors verified that the supporting 10 CFR 50.50 evaluation, SE 91-0046, did not
address the calculated increase in damping factors.
This issue is considered a URI pending further review and evaluation by the NRC
staff to establish a position on whether the licensees USAR changes and supporting
10 CFR 50.59 evaluations provided appropriate rationale to support a conclusion that prior
NRC-approval is not required to increase seismic licensing basis damping by calculation
(URI 05000346/2013010-01, 10 CFR 50.59 Safety Evaluation Not Performed for
Calculated Component Damping).
.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, to determine if the program was prepared in
accordance with regulatory requirements and appropriate industry codes and standards.
In particular, the inspectors 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, temporary lifting device mounted on the
crane girders, and the temporary crane girder center post support were adequate to
withstand the maximum anticipated loads to be lifted.
The inspectors reviewed ECP 12-0067 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 records for the
containment polar crane inspections performed before and after the steam generator lift.
The inspectors reviewed the adequacy of the haul route evaluation for load testing and
transport of the steam generators. The inspectors verified that they had been prepared in
accordance with regulatory requirements and appropriate industry codes and standards.
The inspectors also discussed the transport path load testing with the licensees Steam
Generator Replacement Project (SGRP) engineering personnel, performed a walkdown of
the haul route, and also reviewed documentation of the load test.
b. Findings
No findings were identified.
.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, and 71124.08 as guidance:
- As-Low-As-Is-Reasonably- Achievable (ALARA) planning;
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- 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;
- Airborne radioactivity effluent controls; and
- Radioactive Material Storage
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
Inspectors identified a finding of very low safety significance and associated NCV of NRC
regulations. Details of this finding and associated NCV are documented in Official Use
Only inspection report (IR) 05000346/2014404.
.5 Welding and Non-Destructive Examination Activities
a. Inspection scope
The inspectors reviewed the following welding and Non-Destructive Examination (NDE)
activities associated with the Steam Generator Replacement Project to evaluate
compliance with the 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;
connecting RCS piping;
RSGs and connecting RCS piping;
connecting RCS piping; and
- Completion of baseline eddy current test (ET) examination of new SG tubes.
b. Findings
No findings were identified.
8
.6 Activities Associated with Lifting and Rigging
a. Inspection Scope
The inspectors observed and reviewed activities associated with transportation of the
steam generators into and outside containment. The inspectors reviews included the lifting
equipment necessary to perform steam generator rigging and transport, design evaluation
and use of the outside lift system (OLS), temporary lifting device (TLD), polar crane center
post support, and load drop protection. The inspectors performed direct observation of a
sample 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 personnel training
certifications.
b. Findings
No findings were identified.
.7 Old and New Steam Generator Cutting, Movement and Reconnection
a. Inspection Scope
The inspectors observed various portions of the removal of the old steam generators from
the containment vessel. These activities included but were not limited to:
- Cutting cold leg connections and the cutting and removal of existing hot leg piping;
- Installation and testing of the temporary lifting device on the containment vessel
polar crane;
- The initial jacking of each steam generator from their original locations within each
containment vessel 'D' ring;
- Lifting of the old steam generators clear of each 'D' ring and the transition to a
horizontal orientation for removal from the containment vessel;
- Removal of the old steam generators from the containment vessel and placement
on the heavy haul transporter; and
- Transport of the old steam generators to the onsite storage facility and final
placement for long-term storage.
Additionally, the inspectors observed various portions of the installation of the replacement
steam generators. These activities included but were not limited to:
- Receipt and offloading of the replacement steam generators at the site, and
transport to the replacement steam generator interim storage facility for inspection
and fit-up of various appurtenances;
- Transport of the replacement steam generators from the interim storage facility to
the shield building and transition to the outside lifting system at the shield building
construction opening;
- Movement of the replacement steam generators into the containment vessel and
transition to a vertical orientation;
- Lifting of each replacement steam generator and placement into its 'D' ring using
the temporary lifting device; and
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- Welding of cold leg connections, steam and feedwater connections, and the
installation and welding of replacement hot leg piping.
b. Findings
No findings were identified.
.8 Steam Generator Hold-Down Bolts and Major Structural Modifications
a. Inspection Scope
Inspectors reviewed the major structural modifications to the plant to facilitate the
replacement of the steam generators. Specifically, the inspectors reviewed the use of
the polar crane center post support structure for facilitating the steam generator lifts,
the openings constructed in the containment vessel and shield building, the erection of the
self-propelled modular transporter, the removal of steam generator auxiliary feedwater whip
restraints, and the evaluations documenting structural adequacy of the upper and lower
supports including the new replacement hold down bolts.
b. Findings
No findings were identified.
.9 Temporary Containment Opening, Restoration of Temporary Containment Opening,
and Containment Leakage Testing
a. Inspection Scope
The Davis-Besse containment lacked an access opening of sufficient size to permit
removal of the old steam generators and installation of the replacement steam generators.
Therefore, the licensee cut a temporary access opening in the shield building and
containment vessel of sufficient size to support the steam generator replacement.
During construction of the shield building temporary opening, the licensee identified the
presence of a void along the top wall of the opening. The licensee measured the void to be
approximately twenty-four inches at its maximum depth and had an approximately twelve
inch mouth opening on the inside (annulus) surface of the shield building. The licensee
entered the concern into its Corrective Action (CAP) Program as condition report (CR)-
2014-02896, Void in Concrete Discovered at the Top of the Shield Building Construction
Opening.
As a part of CR-2014-02896, the licensee performed a past operability evaluation of the
shield building since the void had been present from the time of the 2011 shield building
opening restoration to the 2014 construction opening. The design of the shield building
provides for: shielding from radiation sources within the shield building, controlled release
of annulus atmosphere under an accident condition, and environmental protection of the
containment vessel. In its evaluation, the licensee concluded that the shield building was
capable to perform its intended design functions during the time it was in operation with the
void present. The inspectors reviewed the licensees operability evaluation and concluded
that there was sufficient margin in the analysis to determine the shield building would have
been capable of performing its design basis functions with the as-found concrete voiding.
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Specifically, the inspectors reviewed the licensees calculations provided in the operability
evaluation of the as-found voiding and concluded the licensee had demonstrated that the
shield building remained capable of performing its design basis functions under accident
conditions and external events including earthquake, tornado wind, and tornado generated
missiles.
The licensee also performed an apparent cause evaluation (ACE) to determine the
cause(s) of this void in the shield building. In that ACE, the licensee concluded that the
direct cause of the void was lack of flowable concrete. That is, the concrete mixture did not
exhibit sufficient flow characteristics to allow it to readily penetrate the depth of the shield
building in all directions and fill the empty spaces resulting in the formation of a void. The
licensee also concluded that the contributing cause of this void was a lack of questioning
attitude amongst the licensee staff and management. Specifically, the licensee staff and
management did not question or challenge the decision to not inspect the inside wall of the
shield building after initial void repairs had been made during the 2011 shield building
restoration efforts. The licensee also identified corrective actions to ensure that the repair
of this void and the subsequent shield building restoration were performed correctly. In
particular, the licensees corrective actions emphasized concrete pour quality and pour
technique to ensure that the pour has sufficient flow characteristics and that the techniques
employed preclude the entrapment of air pockets during the pour.
The inspectors reviewed the licensees ACE and concluded that the evaluation had
adequately considered and identified the potential cause(s) and that the evaluation had
been conducted by knowledgeable, independent, and qualified licensee staff. However,
the inspectors identified an NCV associated with the licensees failure to correct the voiding
that was initially identified during the restoration of the 2011 shield building opening. The
details of this NCV are documented below in Section 4OA5.9.b.
The inspectors witnessed the repairs to the void during the 2014 outage to ensure repairs
were adequately implemented. In addition, the inspectors performed a visual inspection of
the shield building outside and inside walls after the void repairs had been completed by
the licensee to verify that the void was no longer present on the inside and outside surfaces
of the shield building.
During the construction of the shield building opening for the 2014 steam generator
replacement, the licensee identified that shield building reinforcing bars (rebar) that had
been damaged during the hydrodemolition process. Some rebar locations had cracked and
sheared off and a few other locations were observed to have cracking or bending but the
rebar had not broken off. In total, fifty-seven rebar locations were identified by the licensee
to have been damaged.
The licensee performed an ACE to determine the cause(s) surrounding this unexpected
rebar damage. In that ACE, the licensee concluded that the direct cause of the rebar
damage was alternating bending fatigue cracking due to the cyclic loading imposed on the
shield building from using a hydrodemolition process to cut the shield building opening.
The licensee also identified contributing causal factors of extremely low temperatures
during the cutting of the shield building opening, the use of mechanical couplers to join
rebar lengths during the previous 2011 shield building opening restoration, and the length
of each rebar section extending out into the shield building opening.
The inspectors reviewed the licensees ACE and concluded that the evaluation had
adequately considered and identified the potential cause(s) and that the evaluation had
11
been conducted by knowledgeable, independent, and qualified licensee staff. In addition,
the inspectors observed a sample of the laboratory testing that the licensee performed on
failed rebar sections to determine the cause(s) of failure. The inspectors concluded that
this unexpected failure of some rebar sections during construction of the temporary shield
building opening was not a performance deficiency because the licensee could not have
reasonably foreseen and prevented this particular outcome during the construction of the
temporary opening. Specifically, the hydrodemolition process used to remove the concrete
surrounding the rebar is not expected to damage rebar according to industry guidance and
operating experience on the use of this process. The complex mechanism of cracking in
the rebar induced by a combination of extremely cold temperatures, high stress cyclic
loading, long rebar lengths, and increased rebar restraint at mechanical couples presented
a unique set of circumstances that could not have been reasonably predicted and
prevented.
The inspectors noted that the rebar had been broken during the 2014 outage and repaired
prior to the shield building being placed in service at the conclusion of the outage.
Therefore, in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor
Inspection Reports, Appendix E, Examples of Minor Issues, Section 5, Work In Progress
Findings, even if the inspectors had identified a performance deficiency associated with
the issue involving rebar damage, the issue would be of minor safety significance.
Furthermore, in accordance with IMC 0612, Appendix B, Issue Screening, the issue would
not have resulted in an inspection finding/enforcement action given its minor nature.
The licensee implemented corrective actions associated with restoring the shield building
rebar by qualifying the use of existing rebar through NDE and replacing those sections of
rebar that did not pass the NDE acceptance criteria. The inspectors witnessed the NDE
performed by the licensee on the remaining rebar sections surrounding the construction
opening to qualify their return to service. Specifically, inspectors observed UT examination
performed on the shield building opening rebar sections. The inspectors identified an NCV
associated with the licensees use of an unqualified UT examination procedure used to
perform the examinations on the shield building rebar. Details of this NCV are documented
below in Section 4OA5.9.b.
In restoring the temporary construction opening in the containment vessel, the licensee
reused and reinstalled by welding, the original plate section cut from the containment
vessel. The licensee also installed new rebar to the existing original shield building rebar
via Zap Screwlok couplers and poured new concrete fabricated at an onsite batch plant.
The inspectors reviewed the licensee activities associated with the restoration of the
containment vessel and shield building access openings. Specifically, the inspectors
observed activities and reviewed records as discussed below:
- Inspectors observed the cutting of the containment vessel opening using a track-
mounted welding torch to determine if the cutting activity followed the work order (WO);
- Inspectors observed installation of the replaced containment vessel plate to determine if
the gap tolerances had been maintained in accordance with the WO;
- Inspectors reviewed the full penetration butt welds fabricated during reinstallation of
the 1.5 inch thick containment vessel access plate to determine if the welding process
followed the qualified welding procedures and to determine if weld filler materials were
traceable to certified material test reports;
12
- Inspectors reviewed the welding procedures and welder qualification records for
containment vessel closure welding activities to determine if the welding was qualified
in accordance with the ASME Code Section IX;
- Inspectors reviewed samples of the radiographic test (RT) records and magnetic
particle (MT) exam records of the containment vessel welds to determine if weld
acceptance criteria met the construction code requirements (ASME Code 1968 Edition,
1969 Summer Addenda of Section III);
- Inspectors observed repair of the concrete void identified during the 2014 shield
building construction opening including the chipping away of shield building concrete
around the void area;
- Inspectors observed the new concrete pour in the void area that was identified during
the 2014 shield building construction opening to determine if the licensee had followed
its procedures and implemented corrective actions identified in its apparent cause
evaluation of this void area;
- Inspectors performed walk downs of the void area from both the outside surface and
inside surface of the shield building to determine if the licensee had adequately repaired
the void;
- Inspectors observed NDE of the potentially damaged rebar in the shield building
construction opening and reviewed relevant procedures to determine if the
examinations were being performed by qualified personnel using qualified procedures
and that the area of examination was adequate to qualify the existing rebar in the shield
building construction opening;
- Inspectors observed laboratory testing of failed rebar sections to determine if the
licensee had performed testing in accordance with approved testing procedures on the
failed samples to determine the cause(s) and that failed rebar samples were properly
stamped and stored;
- Inspectors observed installation of couplers (reattachment by Zap Screwlok System) to
join rebar sections during restoration of the shield building opening to determine if the
licensee process conformed to the qualified procedure and design requirements;
- Inspectors reviewed the results of concrete field tests (e.g., slump and air content)
during installation to determine if the concrete had the expected properties specified for
the mix design;
- Inspectors observed the onsite and off-site storage and curing conditions for concrete
test cylinders to determine if they met the American Society for Testing and Materials
(ASTM) C31 Making and Curing Concrete Test Specimens in the Field, and ASTM
C192 Making and Curing Concrete Test Specimens in the Laboratory, prior to
acceptance testing;
- Inspectors reviewed the licensees vendor records for the source materials (e.g.,
aggregate, cement, water, and admixtures) for concrete batches used in restoration of
the shield building to determine if it these materials conformed to the design
specifications;
- Inspectors reviewed concrete cylinder compressive test report to determine if testing
was conducted in accordance with ASTM C39 Compressive Strength of Cylindrical
Concrete Specimens, and to determine if the test results demonstrated that the
concrete used for restoration of the shield building opening had adequate shear
strength to meet the USAR Section 3.8.2.3.7 minimum design compressive strength
(e.g., in excess of 4000 psi); and
- Inspectors observed the containment leakage testing to determine if the licensee had
adequately demonstrated the containment vessel opening leakage integrity following
restoration.
13
The records reviewed by the inspectors are identified in the Attachment to this report.
b. Findings
Use of Unqualified Procedure for Ultrasonic Examination of Shield Building Rebar
Introduction: The inspectors identified a finding of very low safety significance (Green) and
an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX,
Control of Special Processes, when the licensee failed to use a qualified procedure for UT
examination of the Shield Building reinforcing bars (rebar). Specifically, the licensee used
a site approved UT examination procedure that had not been qualified to examine the total
length of approximately twenty four inches of rebar as specified in the procedure due to
near field scanning limitation.
Description: During construction of the 2014 steam generator replacement outage
temporary opening in the concrete shield building, the licensee identified that some of the
rebar embedded in the concrete shield building had been damaged. This identified rebar
damage was documented in NCR 25539-200-GCE-00036 dated February 9, 2014. As part
of corrective actions to address the shield building rebar non-conformance, the licensee
performed NDE including UT examination on the rebar in the shield building temporary
opening. The purpose of this NDE, specifically, the UT examination was to examine
potential degradation in each shield building opening rebar up to twenty four inches in
length by scanning from the end of the rebar. The results of this UT examination
determined which sections of rebar were acceptable for return to service and which
sections of rebar were replaced.
The inspectors reviewed UT Examination Procedure 100-UT-104, Revision 0, Ultrasonic
Examination of Rebar with a Cross-Sectional Thickness of 0.750-in to 2-in, and observed a
sample of UT examinations performed on the shield building rebar. In Section 1.0 of this
procedure states, in part, that This technique establishes a standardized method for the
contact, longitudinal wave, straight beam, ultrasonic examination of rebar exhibiting a cross
sectional thickness (or diameter) of 3/4-in to 2-in and with a length of approximately 24-in.
The licensee qualified the use of this procedure by performing a qualification examination
on representative rebar samples using Procedure 100-UT-104. The inspectors reviewed
the qualification record and identified that the qualification examination only identified
flaws/indications at a depth of four inches and greater from the UT exam scanning surface
(ends of each rebar). In addition, the inspectors calculated the near-field distance for this
particular examination using the instrument settings stated in Procedure 100-UT-104 to be
approximately three inches. This near-field distance, which corresponds to the distance
from the UT examination scanning surface in which UT examination detection capability is
not considered reliable, was not accounted for in Procedure 100-UT-104 or the procedure
qualification record. Therefore, the inspectors questioned the quality of data obtained
during the UT examination of each rebar between the UT exam scanning surfaces to
approximately three inches in depth. Specifically, the licensee had qualified Procedure
100-UT-104 to examine approximately twenty-four inches of rebar from the UT examination
scanning surface but had not adequately demonstrated that the procedure was capable of
detecting flaws/indications between a depth of zero to three inches from the UT
examination scanning surface.
The licensee entered the inspectors identified concern into CAP under CR 2014-04754.
The licensees immediate corrective actions described in CR 2014-04754 included NDE of
14
each shield building opening rebar using the magnetic particle technique examining from
up to five inches from the end of each rebar section. The inspectors reviewed the magnetic
particle examination results and did not identify any additional concerns.
Analysis: The inspectors determined that the failure to use a qualified UT examination
procedure for examination of the shield building temporary opening rebar was a
performance deficiency.
The performance deficiency was more than minor, and a finding because if left
uncorrected, the performance deficiency would have the potential to lead to a more
significant safety concern. Specifically, absent NRC identification, the licensee would have
continued use of the unqualified UT examination procedure to examine potential
degradation in potentially damaged rebar in the safety-related shield building. Therefore,
the licensee could potentially have returned the shield building back to service with
unacceptable flaws existing in the rebar.
The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of
Findings, Table 3 - SDP Appendix Router. The inspectors answered Yes to the
questions in Section A of Table 3 and; therefore, the finding was evaluated using the SDP
in accordance with IMC 0609, The Significance Determination Process (SDP) for
Shutdown Operations, Appendix G, Attachment 1, Exhibit 4, Barrier Integrity Screening
Questions. The inspectors answered all the questions in Exhibit 4 and determined that
this finding did not result in degraded physical integrity of the containment during shutdown
operations nor did it affect any shutdown safety functions. Therefore, the finding was
determined to have very low safety significance (Green).
This finding had a cross-cutting aspect in the area of Problem Identification and Resolution,
Evaluation, for the licensees failure to thoroughly evaluate issues to ensure that resolutions
address causes and extent of conditions commensurate with their safety significance.
Specifically, the licensee failed to initially consider the entire length of rebar for potential
evaluation and hence, did not consider the appropriate extent of condition. [P.2].
Enforcement: Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B,
Criterion IX, Control of Special Processes, requires that Measures shall be established to
assure that special processes, including welding, heat treating, and nondestructive testing,
are controlled and accomplished by qualified personnel using qualified procedures in
accordance with applicable codes, standards, specifications, criteria, and other special
requirements.
Section 1.0 of UT Examination Procedure 100-UT-104, Revision 0 states, in part, that This
technique establishes a standardized method for the contact, longitudinal wave, straight
beam, ultrasonic examination of rebar exhibiting a cross sectional thickness (or diameter)
of 3/4-in to 2-in and with a length of approximately 24-in.
Contrary to the above, on February 25, 2014, the licensee failed to properly qualify
Procedure 100-UT-104. Specifically, the licensee performed the qualification
demonstration UT examination without consideration of near-field distance limitations for
the length of rebar to be examined. Therefore, the entire length of rebar up to
approximately twenty-four inches was not qualified to be examined using Procedure 100-
UT-104.
15
Because this violation was of very low safety significance and it was entered into the
licensees CAP as CR-2014-04754, this violation is being treated as a NCV, consistent with
Section 2.3.2 of the NRC Enforcement Policy. As part of its immediate corrective actions,
the licensee performed additional NDE of each shield building opening rebar using the
magnetic particle technique examining from zero to five inches from the rebar end surface.
The inspectors reviewed the magnetic particle examination results and did not have any
additional concerns. (NCV 05000346/2013010-02, Use of Unqualified Procedure for
Ultrasonic Examination of Shield Building Rebar)
Failure to Properly Repair Shield Building Concrete Voiding
Introduction: A finding of very low safety significance and associated Non-Cited Violation
(NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by
the inspectors for the licensees failure in 2011 to properly repair concrete voiding in the
shield building that had been identified during that construction opening restoration.
Description: On February 13, 2014, licensee contractors, performing shield building
hydrodemolition activities to create a temporary construction opening to support steam
generator replacement, identified concrete voiding in the shield building vertical wall on the
annulus side (inside surface). The voiding in the concrete was identified following removal
of the annulus side concrete forms that were left in-place following the 2011 shield building
construction opening restoration. The licensee had left these forms in place to serve as
blast shielding during the anticipated 2014 hydrodemolition activities. The licensees
contractors initiated NCR-25539-G61-GCE-00042 to document and repair the condition.
The inspectors also reviewed the licensees corrective action documents associated with
prior restorations of the shield building construction openings and discussed with the
licensee staff to determine why the shield building voiding was left in-place and not repaired
in 2011. As part of this review the inspectors determined:
The identified shield building concrete void was located within the 2011 temporary
construction opening that supported reactor vessel head replacement. Following
placement of the new reactor vessel head into the containment vessel, the shield building
temporary construction opening was restored to its design basis condition; replacement
steel reinforcement bars (rebar) and concrete were in accordance with the shield building
original design requirements. Temporary concrete forms were installed on the vertical wall,
exterior side and interior annulus side, to support concrete pouring. Following appropriate
concrete curing, the concrete forms on the exterior side were removed. On November 21,
2011, the licensees contractors identified concrete voiding on the exterior side at the top of
the construction opening, and initiated NCR-25539-200-G61-GCE-00020 to place concrete
to completely fill the construction opening. On November 21, 2011, the licensee entered
the condition into its CAP as CR-2011-05777.
As part of the 2011 void repair, the licensees contractors used a chipping process to
ensure only solid concrete was left in-place prior to repair. As documented in NCR-25539-
200-G61-GCE-00020, the void volume was filled with structural concrete from the exterior
side, and the seven-day compression strength tests met the original design requirements.
However, the 2011 repair did not completely fill voided concrete on the annulus side of the
shield building vertical wall. As noted in NCR-25539-200-G61-GCE-00020, the void
identified in 2011 extended from the exterior wall to interior blast shield. However, the
interior side forms were left in-place to be utilized as blast shielding during 2014
16
hydrodemolition activities. As a result, the extent of concrete voiding on the annulus side
was not identified and not fully repaired.
In addition, the inspectors reviewed licensee corrective action documents associated with
the shield building 2002 restoration for the construction opening supporting an earlier
reactor vessel head replacement. The inspectors noted that for the 2002 shield building
restoration, the licensees contractor wrote NCR-020 related to concrete voiding after
removal of concrete forms on the exterior side at the top of the construction opening and
NCR-021 related to concrete voiding at the top of the construction opening after removal of
concrete forms on the annulus side. The licensee entered the conditions into its CAP as
CR-G201-2002-07080 (exterior side) and CR-G201-2002-07472 (annulus side). The
voided concrete was repaired prior to plant start-up.
On February 14, 2014, the licensee entered the current void condition into its CAP as
CR-2014-02896. The inspectors reviewed the licensees corrective actions in CR-2014-
02896 which included rework and restoration of the concrete void. The inspectors visually
examined the degraded area to verify extent of voiding, examined exposed rebar to verify
its structural integrity, reviewed the repair procedure, interviewed licensee contractors
performing repair activities, witnessed installation of the repair concrete, reviewed concrete
compression tests, and visually examined the repaired area to verify the shield building was
restored and in conformance with its design basis.
Analysis: The inspectors determined the performance deficiency of failure to properly
repair the void during the 2011 shield building restoration was more than minor and;
therefore, a finding because the performance deficiency was associated with the Barrier
Integrity cornerstone attribute of Design Control and adversely impacted the cornerstone
objective to provide 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 failure to repair the concrete
voiding in 2011 resulted in the operation of the plant with the shield building in a condition
non-conforming to its design basis.
The inspectors reviewed the finding using Attachment 0609.04, Initial Characterization of
Findings, Table 3 - SDP Appendix Router. The inspectors answered No to all the
questions in Section A of Table 3 and; therefore, the finding was evaluated using the SDP
in accordance with IMC 0609, The Significance Determination Process (SDP) for At-Power
Operations, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors
answered all the questions in Exhibit 3 and determined that this finding did not represent an
actual open pathway in the physical integrity of reactor containment. Therefore, the finding
was determined to have very low safety significance (Green). The inspectors determined
that this finding had a cross-cutting aspect in the area of Human Performance,
Conservative Bias, for the licensees failure to use decision making practices that
emphasize prudent choices over those that are simply allowable. Specifically, the licensee
failed to implement a conservative decision to inspect the shield building inside surface void
area after repairs had been made during the opening restoration in 2011. Therefore, the
licensee missed the opportunity to identify that they had not adequately repaired the void.
[H.14] (Section 4OA5.9).
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action states,
in part, that measures shall be established to assure that conditions adverse to quality,
such as deficiencies and non-conformances are promptly identified and corrected.
17
Contrary to the above, the licensee failed to completely repair the concrete voiding
identified in the shield building during restoration of the 2011 construction opening.
Because this violation was of very low safety significance and was entered into the
licensees Corrective Action Program as CR-2014-02896, this violation is being treated as
an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000346/2013010-03, Failure to Completely Repair Shield Building Concrete
Voiding).
.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
draining, system isolation, and safety tagging/blocking;
- Implementation of radiation protection controls. These included ALARA activity
controls, radiological exposure, contamination, and airborne contamination controls
planned for cutting and welding, as well as activities associated with contaminated
interference removal. In addition, the implementation of special controls for
contaminated tools and waste were reviewed, as well as the licensee's control of air
flow across the containment building construction opening radiological boundary;
- Implementation of controls for excluding the introduction of foreign materials into the
replacement steam generators, reactor coolant system, and steam and feedwater
systems; and
- The 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 long term storage of the Old Once-
Through Steam Generators (OOTSG) onsite and evaluated whether contamination surveys
of the OOTSG were adequate for safe transport to the Old Steam Generator Storage
Building. The inspectors reviewed dose rate surveys of the storage facility with one of the
generators in place to verify that they were consistent with the licensees plans for onsite
storage. The inspectors assessed whether the radioactive material storage area was
controlled and posted in accordance with the requirements of 10 CFR Part 20, Standards
for Protection against Radiation. For materials stored or used in the controlled or
unrestricted areas, the inspectors evaluated whether they were secured against
unauthorized removal and controlled in accordance with 10 CFR 20.1801, Security of
Stored Material, and 10 CFR 20.1802, Control of Material Not in Storage, as appropriate.
The inspectors evaluated whether the licensee established a process for monitoring the
impact of long term storage (e.g., buildup of any gases produced by waste decomposition,
18
chemical reactions, container deformation, loss of container integrity, or re-release of free-
flowing water) that was sufficient to identify potential unmonitored, unplanned releases or
nonconformance with waste disposal requirements.
Additional inspection activity related to the onsite storage of the OOTSGs will be conducted
under the Reactor Oversight Program Baseline Inspection Program in 2014.
b. Findings
No findings were identified.
.12. Steam Generators Post-Installation Verification and Testing
a. Inspection scope
The inspectors performed reviews and observations of various steam generator
replacement project verification and testing activities, consistent with their overall safety
significance. Specific activities observed and/or reviewed included, but were not limited to:
- Shield building construction opening concrete restoration and concrete compressive
strength testing;
- Shield building emergency ventilation system testing following construction opening
concrete restoration;
- Containment vessel leakage testing;
- Reactor coolant system pressure testing;
- Steam and feedwater systems pressure testing;
- Various instrumentation calibration and tests associated with the replacement
- Various nondestructive surface examinations utilizing dye penetrant and/or
magnetic particle techniques;
- Various nondestructive volumetric examinations utilizing radiography and/or
ultrasonic techniques; and
- Verification of replacement steam generator thermal and hydraulic performance
during ascension to full power following plant restart.
b. Findings
No findings were identified.
4OA6 Meetings
.1 Exit Meeting Summary
On July 2, 2014, the inspectors presented the inspection results to Mr. Tom Lentz 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
19
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
T. Lentz, Manager Fleet Licensing
R. Lieb, Site Vice President
J. Hook, Manager Site Engineering
B. OMalley, Manager Fleet Oversight
P. McCloskey, Manager Regulatory Compliance
L. Rushing, SG Replacement Project Director
T. Summers, Director Site Operations
S. Osting, SG Replacement Project Engineering Lead
D. Gerren, SG Replacement Project Component Engineering Lead
D. Munson, Site NDE Level III
G. Wolf, Supervisor Regulatory Compliance
T. Chowdhary, Regulatory Compliance
Nuclear Regulatory Commission
D. Kimble, Senior Resident Inspector
T. Briley, Resident Inspector
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000346/2013010-01 URI 10 CFR 50.59 Safety Evaluation Not Performed for
Calculated Component Damping (Section 4OA5.1)05000346/2013010-02 NCV Use of Unqualified Procedure for Ultrasonic Examination
of Shield Building Rebar (Section 4OA5.9)05000346/2013010-03 NCV Failure to Completely Repair Shield Building Concrete
Voiding (Section 4OA5.9)
Closed
05000346/2013010-02 NCV Use of Unqualified Procedure for Ultrasonic Examination
of Shield Building Rebar (Section 4OA5.9)05000346/2013010-03 NCV Failure to Completely Repair Shield Building Concrete
Voiding (Section 4OA5.9)
Discussed
None
1 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
4OA5 Other Activities
- ECP-12-0067; Engineering Change Design Report, (for Rigging and Transport); Revision 4
- 12-00917; 10 CFR 50.59 Screening for ECP 12-0067 (for Rigging and Transport); Revision 4
- Procedure 25539-000-V14-MJKP-00080-001; Steam Generator Replacement Project Polar
Crane Inspection Report
- Procedure DB-MN-00006; Control of Lifting and Handling of Heavy Loads; Revision 14
- Calculation 25539-000-V14-MJKP-00006-006; Polar Crane Analysis for Lifting Steam
Generator; January 21, 2014
- Calculation 25539-000-V24-UA30-00187-005; Center Post Polar Crane; January 15, 2013
- Calculation 25539-000-SYC-0000-00001; Weight and Center of Gravity (CG) of OOTSGs-ECP
12-0067; Revision 1
- Calculation 25539-000-V24-UA30-00189-004; Lift Links; May 8, 2013
- Calculation 25539-000-V14-MJKP-00074-002; Stiffness Calculation for Center Post; June 20,
2012
- Calculation C-CSS-100.00-182; Evaluation of Underground Utilities for 18RFO Steam
Generator Replacement Haul rout; Revision 0, 0A01, 0A05
- Calculation C-CSS-100.00-184; Evaluation of Underground Utilities and Structures for 18RFO
Rigging Equipment at Construction Opening; Revision 0A2
- 25539-000-SYC-0000-00003; Haul Rout Load Test for 18RFO; Revision 1
- Calculation 25539-000-SSC-0004-00004; Lifting Trunnion Analysis; Revision 2
- Calculation 25539-000-V24-UA30-00186-003; TLD Calculation; May 3, 2013
- Calculation C-CSS-099.11-026; Polar Crane Ring Girder Evaluation for Loads Associated with
18R Steam Generator Replacement; Revision 0
- CBI Letter L-SHW-DB-00017 Bruce Ebbeson to Richard Bair; Steam Generator Rigging
Review - Final Report; December 19, 2013
- Drawing 25539-000-V24-UA30-00177-005 Page 1 of 1; General Arrangement Center Post
Polar Crane, December 18, 2012
- Drawing 25539-000-V24-UA30-00205-005 Page 1 of 1; Center Post Top Arrangement Center
Post Polar Crane; December 18, 2012
- Drawing 25539-000-V24-UA30-00206-005 Page 1 of 1; Center Post Arrangement Center Post
Polar Crane, December 18, 2012
- Drawing 25539-000-V24-UA30-00207-002 Page 1 of 1; Center Post Skid Base Center Post
Polar Crane, December 18, 2012
- Drawing 25539-000-V24-UA30-00208-005 Page 1 of 1; Center Post Base Center Post Polar
Crane, December 18, 2012
- Drawing 25539-000-V24-UA30-00384-004; General Lift Overview and Trunnion details
ROTSG; March 4, 2014
- Drawing 25539-000-V24-UA30-00385-004; General Lift Overview and Trunnion details
ROTSG; March 4, 2014
- Drawing 25539-000-V24-UA30-00151-006; General Arrangement Inside Hatch Transfer
system (IHTS); December 5, 2012
2
- Drawing 25539-000-V24-UA30-00152-006; General Arrangement Inside Hatch Transfer
system (IHTS); November 30, 2012
- Drawing 25539-000-V24-UA30-00294-005; General Arrangement Inside Hatch Transfer
system (IHTS); December 5, 2012
- Drawing 25539-000-V24-UA30-00277-005; Handling of OOTSG Inside Containment
- Drawing 25539-000-V24-UA30-00278-005; Handling of OOTSG Inside Containment
- Drawing 25539-000-V24-UA30-00279-005; Handling of OOTSG Inside Containment
- Drawing 25539-000-V24-UA30-00280-005; Handling of OOTSG Inside Containment
- Drawing 25539-000-V24-UA30-00281-005; Handling of OOTSG Inside Containment
- Drawing 25539-000-V24-UA30-00282-005; Handling of OOTSG Inside Containment
- Drawing 25539-000-V24-UA30-00283-005; Handling of ROTSG Inside Containment
- Drawing 25539-000-V24-UA30-00284-005; Handling of ROTSG Inside Containment
- Drawing 25539-000-V24-UA30-00285-005; Handling of ROTSG Inside Containment
- Drawing 25539-000-V24-UA30-00286-005; Handling of ROTSG Inside Containment
- Drawing 25539-000-V24-UA30-00287-005; Handling of ROTSG Inside Containment
- Drawing 25539-000-V24-UA30-00288-005; Handling of ROTSG Inside Containment
- Drawing 25539-000-V24-UA30-00364-004; Handling of OOTSG Outside Containment
- Drawing 25539-000-V24-UA30-00365-004; Handling of OOTSG Outside Containment
- Drawing 25539-000-V24-UA30-00366-004; Handling of OOTSG Outside Containment
- Drawing 25539-000-V24-UA30-00367-002; Handling of ROTSG Outside Containment
- Drawing 25539-000-V24-UA30-00368-002; Handling of ROTSG Outside Containment
- Drawing 25539-000-V24-UA30-00369-002; Handling of ROTSG Outside Containment
- Drawing 25539-000-CY-0000-00005; Haul Rout and Protection 18RFO; Revision 2
- Drawing 25539-000-CY-0000-00006; Haul Rout and Protection 18RFO; Revision 4
- Drawing 25539-000-CY-0000-00007; Haul Rout and Protection; Revision 0
- Drawing 25539-000-CY-0000-00008; Haul Rout and Protection; Revision 3
- Drawing 25539-000-CY-0000-00009; Haul Rout Load Test Configuration for 18RFO; Revision
2
- Drawing 25539-000-CY-0000-00013; Exposed Duct Bank Missile Protection Barrier; Revision
1
- Drawing 25539-000-CY-0000-00016; Temporary SSC Protection; Revision 1
- Drawing 25539-000-CY-0000-00018; Haul Rout Load Test for18RFO; Revision 0
- Drawing 25539-000-G4-0004-00001; Containment Building Steam Generator Rigging Path at
El. 653-0; Revision 0
- WO 200511907; Perform Haul route Load Test in Support of 1R18; January 30, 2014
- Report 25539-000-V14-MJKP-00095-001; Pre SG-1-1 Inspection Report; February 27, 2014
- Report 25539-000-V14-MJKP-00096-001; Post SG-1-1 Inspection Report; March 2, 2014
- Procedure DB-MM-06002; Polar Crane Operation; Revision 18
- CR-2014-04149; Polar Crane Rail Issues During SG#1 Removal; March 1, 2014
- CR-2014-04761; Auxiliary Feedwater Riser flange Damaged during Installation; March 11,
2014
- CR-2014-04740; Mammoet Temporary Lift Device Lift Link Interference with ROTSG Auxiliary
Feedwater Riser Flange During Upending Operations - Four Hour Delay in Critical Path;
March 11, 2014
- CR-2014-05763; Uncompleted Activities associated With Post SG 1-1 Lift Inspection Report
25539-000-V14-MJKP-00096-001; March 27, 2014
- 12-00943; 10CFR50.59 Screening for ECP 10-0463, Small Bore Piping; Revision 1
- ECP 10-463-005; Small bore and Instrumentation - SG-1 Interference Piping (Hot Leg Vent
Piping); Revision 4
- ECP 10-463-002; Small bore and Instrumentation - SG-1 Hot Leg Instrumentation Piping;
Revision 4
3
- Calculation 99A-1; Pipe Stress Analysis Steam Generator E24-1 Hot Leg Vent; Revision
D3A01
- CR-2012-11913; Calculation Open Assumptions; July 16, 2012
- 12-00911; 10CFR50.59 Screening for ECP 10-0466, SGR-18RFO-Installation of New OTSGs;
Revision 5
- ECP 0466-000; Installation of New OTSGs; Revision 5
- Drawing C-178; Steam Generator Upper Support Plan and Sections, Sheet 1; Revision 10
- DIT-SGRP-00163-00; DB SGRP Design Information Transmittal; November 1, 2013
- MEE 10-0466-05; Material Engineering Evaluation - Equivalency Evaluation for RTDs;
December 10, 2013
- MEE 10-0466-06; Material Engineering Evaluation - Equivalency Evaluation for
Thermocouples; November 7, 2013
- Calculation VC05/B001-014; Steam Generator Foundations; Revision 2A01
- Calculation VC12/B001-001; OTSG Lower Lateral Restraints; Revision 2A01
- Calculation VC11/B001-013; Steam Generator Upper Supports; Revision 5A03
- AREVA Document No. 51-9222163-000; Davis-Besse Cold Leg Elbow Boat Sample
Associated with Field Weld 1-2 and SG 1-1 Destructive Examination; Revision 0
- Condition Report CR-2014-02482; Reinforcing Steel Broken During Hydrodemolition at the
Shield Building Construction Opening; February 9, 2014
- NDE Report No. BOP-VT-14-068; System Leakage Test (VT-2) of Containment Vessel
Opening Restoration Weld; April 27, 2014
- NDE Report No. 18-MT-088; Magnetic Particle Examination of SG-2 36 inch Hot Leg Pipe to
SG Inlet Nozzle Weld; April 18, 2014
- NDE Report No. 18-UT-131; Ultrasonic Examination of SG-2 36 inch Hot Leg Pipe to SG Inlet
Nozzle Weld; April 18, 2014
- NDE Report No. 18-MT-086; Magnetic Particle Examination of SG-2 36 inch Hot Leg Pipe to
Pipe Weld; April 9, 2014
- NDE Report No. 18-UT-115; Ultrasonic Examination of SG-2 36 inch Hot Leg Pipe to Pipe
Weld; April 10, 2014
- NDE Report No. 18-MT-089; Magnetic Particle Examination of SG-1 36 inch Hot Leg Pipe to
SG Inlet Nozzle Weld; April 17, 2014
- NDE Report No. 18-UT-130; Ultrasonic Examination of SG-1 36 inch Hot Leg Pipe to SG Inlet
Nozzle Weld; April 17, 2014
- NDE Report No. 18-MT-079; Magnetic Particle Examination of SG-1 36 inch Hot Leg Pipe to
Pipe Weld; April 18, 2014
- NDE Report No. 18-UT-114; Ultrasonic Examination of SG-1 36 inch Hot Leg Pipe to Pipe
Weld; April 18, 2014
- NDE Report No. 18-MT-083; Magnetic Particle Examination of SG-2 28 inch Cold Leg Outlet
Safe End to Elbow Weld; April 9, 2104
- NDE Report No. 18-UT-117; Ultrasonic Examination of SG-2 28 inch Cold Leg Outlet Safe
End to Elbow Weld; April 10, 2104
- NDE Report No. 18-MT-084; Magnetic Particle Examination of SG-2 28 inch Cold Leg Outlet
Safe End to Elbow Weld; April 9, 2104
- NDE Report No. 18-UT-118; Ultrasonic Examination of SG-2 28 inch Cold Leg Outlet Safe
End to Elbow Weld; April 10, 2104
- NDE Report No. 18-VENDOR-273; Magnetic Particle Examination of SG-1 28 inch Cold Leg
Outlet Safe End to Elbow Weld; April 14, 2104
- NDE Report No. 18-UT-119; Ultrasonic Examination of SG-1 28 inch Cold Leg Outlet Safe
End to Elbow Weld; April 10, 2104
- NDE Report No. 18-MT-087; Magnetic Particle Examination of SG-1 28 inch Cold Leg Outlet
Safe End to Elbow Weld; April 10, 2104
4
- NDE Report No. 18-UT-116; Ultrasonic Examination of SG-1 28 inch Cold Leg Outlet Safe
End to Elbow Weld; April 10, 2104
- NDE Report No. RT-REPORT-307; Radiographic Examination of SG-1 36 inch Hot Leg Pipe
to Pipe Weld; April 4, 2014
- NDE Report No. RT-REPORT-329; Radiographic Examination of SG-1 36 inch Hot Leg Pipe
to SG Inlet Nozzle Weld; April 4, 2014
- NDE Report No. RT-REPORT-317; Radiographic Examination of Containment Vessel
Opening Weld; April 10, 2014
- NDE Report No. M44442-572938; MISTRAS Ultrasonic Examination of Shield Building Rebar;
February 26, 2014
- Procedure No. 100-UT-104; Ultrasonic Examination of Rebar With a Cross Sectional
Thickness of 0.750 inches to 2.0 inches; Revision 0
- AREVA Document No. 51-9206453-001; Davis-Besse Replacement SG and HL LBB Loads
Reconciliation; September 6, 2013
- Bechtel NCR No. 58; Aux Feedwater Flange Gouged During Upending of SG-1 into the
Containment Building; March 12, 2012
- NDE Report No. PT-REPORT-172; Liquid Penetrant Examination of SG-1 Aux FW Flange
Surface; March 20, 2014
- NDE Report No. BOP-MT-14-003; Magnetic Particle Examination of five SG-1 Aux FW Flange
Studs; March 21, 2014
- Condition Report No. 2014-04754; NRC Question on Rebar Examination; March 11, 2014
- NDE Report No. RB-MT-CO-001; Magnetic Particle Examination of Shield Building Rebar;
March 11, 2014
- Bechtel NCR No. 36; Broken Shield Building Rebar Identified During Hydrodemolition;
February 9, 2014
- Document No. 12-03404; 50.59 Screenings for Replacement SGs under ECP 12-0474;
January 31, 2014
- Condition Report No. 2014-05559; Preheat Was Not Reestablished on the SG-1 RCS 2-1 Cold
Leg Prior to Resuming Welding; March 24, 2014
- Document No. M14036 (BETA LAB Project No.); FirstEnergy Laboratory Evaluation of Shield
Building Concrete Rebar; February 28, 2014
- Document No. PP-SGRP-014-02; Steam Generator Replacement Project Missile Evaluation
Position Paper; December 12, 2013
- Document No. B&W-TR-2013-0010; Preservice Eddy Current Tubing Inspection Report for
Davis-Besse Replacement Once Through Steam Generators; September 9, 2013
- NDE Report No. 18-VT-316 & 309; Primary Reactor Coolant System Class 1 Piping System
Leak Test (VT-2); May 2, 204
- NDE Report No. 18-VT-313 & 311; Primary Reactor Coolant System Class 1 Piping System
Leak Test (VT-2), May 2, 2014
- Condition Report CR-2011-05777 (Bechtel NCR 20); Void in Concrete at the Top of the Shield
Building Pour Back; November 21, 2011
- Technical Specification No. 25539-000-3PS-DB02-Q0001; Concrete Work for Safety-related
Applications; Revision 001
- Procedure No. DB-MS-09010; Concrete Placement Procedure; June 3, 1999
- Calculation No. C-CSS-099-20-048; Seismic II/I Evaluation of Shield Building Formwork in
Annulus; November 12, 2011
- AREVA Documents No. 32-9213349-001; Davis-Besse Cladding Thickness Influence on Pipe
Stress; April 14, 2014
- AREVA Document No. 32-9206848-002; Davis-Besse Hot Leg and Cold Leg LBB Evaluation
Validation; April 17, 2014
5
- Document No. 25539-000-GMX-GCE-00001; Bechtel Special Processes Manual for
Davis-Besse Steam Generator Replacement Project; Revision 3
- CR 2014-02619; NRC Identified Safety Violation of Scaffold Access Requirements;
February 11, 2014
- CR 2014-02671; Radiation Protection Briefing for Containment Did Not Meet Standards;
February 11, 2014
- CR 2014-04099; Concern with High Radiation Area Briefs; February 28, 2014
- NOP-OP-4104; Job Coverage; Revision 6
- NOP-OP-4107; Radiation Work Permit; Revision 12
- Radiological Survey 14-0166; Down Post East D Ring from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-0167; Down Post Containment from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-0168; Down Post Containment from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-0169; Down Post Containment from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-0172; Down Post Containment from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-0175; Down Post Containment from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-0176; Down Post West D Ring from Locked High Radiation Area;
February 1, 2014
- Radiological Survey 14-7233; Welding Blank to Once Through Steam Generator Cold Leg;
February 17, 2014
- Radiological Survey 14-7234; Prep to Cut Main Steam Line and Feed Water Piping;
February 17, 2014
- Radiological Survey 14-7239; Steam Generator Hot Leg; February 18, 2014
- Radiological Survey 14-7268; Main Steam Three Foot Section Removal and Weld Blank
Flange Plate; February 20, 2014
- RWP 2014-5202; Steam Generator Replacement Scaffolding Program; Revision 0
- RWP 2014-5207; Insulation/Encapsulation Work Activities in Containment; Revision 0
- RWP 2014-5212; Reactor Coolant System Pipe Prepping, Cutting and Welding Activities;
Revision 0
- RWP 2014-5214; Steam Generator Removal/Replacement to Include Wagon Wheel and
Lower Lateral Supports; Revision 0
- RWP 2014-5215; Transport Old Once Through Steam Generator and Associated Work
Activities; Revision 0
- RWP 2014-5216-1; Main Steam, Feed Water, Auxiliary Feed Water Piping Project; Revision 0
- RWP 2014-5217; Structural Interferences; Revision 0
- Shutdown Chemistry and Source Term Reduction Plan-Davis Besse 18 Refueling Outage;
January 22, 2014
- 2RS2 Occupational ALARA Planning and Controls (71124.02)
- ALARA Plan 2014-5202; Steam Generator Replacement Scaffolding Program; Revision 0
- ALARA Plan 2014-5207; Insulation/Encapsulation Work Activities in Containment; Revision 0
- ALARA Plan 2014-5212; Reactor Coolant System Pipe Prepping, Cutting and Welding
Activities; Revision 04
- ALARA Plan 2014-5215; Transport Old Once Through Steam Generator and Associated Work
Activities; Revision 0
- ALARA Plan 2014-5216-1; Main Steam, Feed Water, Auxiliary Feed Water Piping Project;
Revision 0
6
- ALARA Plan 2014-5217; Structural Interferences; Revision 0
- ALARA Plan 2014-5241; Remove/Replace Once Through Steam Generators; Revision 1
- CR 2014-03267; PCE010-Bechtel Custom Arc Welder Working On Tack Weld Seal Plate 1-1
Cold Leg; February 19, 2014
- CR2014-03629; PCE Number 12-Personnel Contamination Event by Bechtel; February 23,
2014
- CR 2014-04189; Tracking and Trending of Radiological Events; March 2, 2014
- NOP-OP-4005; ALARA Program; Revision 04
- SAC 14-004; Site ALARA Committee Minutes; March 5, 2014
- 2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and
Transportation (71124.08)
- Radiological Survey 14-08200; Release Survey Steam Generator 1-1 to Old Steam Generator
Storage Facility; March 2, 2014
- 2011-05777; Void in Concrete at the Top of the Shield Building Pour Back (Bechtel NCR 20)
- 2011-05795; Concrete Void at Top of Shield Building Restoration Larger than Previously
Reported Per CR 2011-05777
- CR 2014-01995; Decay Heat Pump 2 Suction Gauge Isolated during RCS Drain Evolution
- CR 2011-05804; Conditional Release of Concrete for Filling Shield Building Void
- CR 2014-02896; Void in Concrete Discovered at the Top of the Shield Building Construction
Opening
- CR 2014-04019; Ladder Extension on Polar Crane Center Post Top Arrangement (PCCPTA)
Strikes Polar Crane Center Post (PCCP) Resulting in Damage to PCCPTA and Dropped
Beam Clamp
- CR 2014-04149; Polar Crane Rail issues during SG No.1 Removal
- CR 2014-04255; Rebar Contacted by the Hatch Transfer System (HTS) in the Shield Building
During Generator Transport
- CR 2014-04761; Auxiliary Feedwater Riser Flange Damaged During Installation
- CR 2014-05027; Polar Crane Contacts Temporary Mechanical Trolley Stop
- CR 2014-05110; Roll-up of recent crane issues
- CR 2014-03063; Small Smoldering Fire at Containment Opening Liner Plate
- CR 2014-03305; Hot Work Performed in West D-Ring Not Well Controlled
- CR 2014-03387; Water Mister Used to Cool a Hot Steel and Slag Nugget
- CR 2014-03594; Water Fire Extinguisher Discharged in Containment
- CR 2014-03632; Fire Extinguishers Blocked in Containment
- CR 2014-04128; Items Noted During an NRC Observation of No. 1 OTSG Lift
- CR 2014-03472; Shield Building Construction Opening Unfilled Area - Inspection Findings
- CR 2014-06832; SG 1-1 North Cold Leg Post Weld Heat Treat Performed Without Fire Watch
- CR 2014-07670; PIRC2A6 impact from CTMT pressure test
- CR 2014-09044; Containment Pressure Test Post Instrument Calibrations and
Non-Conformance Report
- PFP-CB-216; Steam Generator West D Ring Area, Room 216, Fire Area D; Revision 5
- PFP-CB-218; Steam Generator East D Ring Area, Room 218, Fire Area D; Revision 5
- WO 200511847; Construction Opening Restoration; March 24, 2014
- WO 200512648; PF3010-003 DH14 RCS Class 1; May 2, 2014
- WO 200512649; PF3010-005 RC1 RCS Class 1; May 2, 2014
- WO 200512658; PF3010-001 CF9 RCS Class 1; May 2, 2014
- WO 200512659; PF3010-002 CF10 RCS Class 1; May 2, 2014
- WO 200512660; PF3010-004 DH16 RCS Class 1; May 2, 2014
- WO 200512836; PF3010-011 DH15 RCS Class 1; May 2, 2014
- WO 200562962; Perform VT-2 and Leak Checks; May 2, 2014
7
- WO 200534740; Primary Containment Vessel Post Modification Pressure/New Weld Leakage
Inspection Test; April 27, 2014
- Procedure DB-OP-03013; Containment Daily Inspection & Containment Closeout Inspection;
Revision 9
- Procedure DB-OP-06000; Filling and Venting the Reactor Coolant System; Revision 28
- Procedure DB-OP-06002; RCS Draining and Nitrogen Blanketing; Revision 20
- Procedure DB-OP-06005; RC Pump Operation; Revision 30
- Procedure DB-OP-06012; Decay Heat and Low Pressure Injection System Operating
Procedure; Revision 59
- Procedure DB-OP-06021; Spent Fuel Pool Operating Procedure; Revision 24
- Procedure DB-OP-06900; Plant Heatup; Revision 61
- Procedure DB-OP-06901; Plant Startup; Revision 35
- Procedure DB-OP-06902; Power Operations; Revision 45
- Procedure DB-OP-06904; Shutdown Operations; Revision 45
- Procedure DB-OP-06911; Pre-Startup Checklist; Revision 24
- Procedure DB-OP-06912; Approach to Criticality; Revision 17
- Procedure DB-NE-06202; Reactivity Balance Calculations; Revision 8
- Procedure DB-FP-00007; Control of Transient Combustibles; Revision 13
- Procedure DB-FP-00009; Fire Protection Impairment and Fire Watch; Revision 19-20
- Procedure DB-FP-00018; Control of Ignition Sources; Revision 12
- Procedure NOP-WM-5003; Rigging, Lifting and Load Handling; Revision 4
- Procedure DB-MM-09193; Assembly and Disassembly of the Reactor Vessel Head and
Internals Handling Fixture (Pin Connected); Revision 0
- Procedure DB-MN-00006; Control of Lifting and Handling of Heavy Loads; Revision 17
- Procedure DB-SS-03255; Emergency Ventilation System Train 2 Refueling Interval SFAS
Drawdown Test; Revision 13
- Procedure DB-PF-03010; RCS Leakage Test; Revision 13
- Procedure DB-PF-03008; Containment Local Leakage Rate Tests; Revision 18
- Procedure DB-PF-03009; Containment Vessel and Shield Building Visual Inspection;
Revision 8
- Procedure DB-PF-10311; Primary Containment Vessel Post Modification Pressure / New Weld
Leakage Inspection Test; Revision 02
- Procedure NOBP-OP-0007; Conduct of Infrequently Performed Tests or Evolutions;
Revision 5
- Procedure OTLC-JIT-DB-1402; 18 RFO Startup JITT; Revision 0
- Procedure OTL-IPO-S104; Preparation for Plant Heatup; Revision 0
- Procedure OTL-IPO-S105; Plant Heatup Mode 5 to Mode 3; Revision 0
- Procedure OTL-IPO-S106; Reactor Startup to 18 percent Power; Revision 0
- Procedure OTL-IPO-S107; Power Operations; Revision 0
- Drawing M-007B; Piping & Instrument Diagram Steam Generator Secondary System;
Revision 59
- Drawing M-035; Piping & Instrument Diagram, Spent Fuel Cooling System; Revision 53
- Drawing OS-004; Sheet 1; Decay Heat Removal / Low Pressure Injection System; Revision 52
- Drawing OS-007; Spent Fuel Pool Cooling System; Revision 30
- Drawing M-030A; P&ID Reactor Coolant System; Revision 69
- Drawing M-033B; P&ID Reactor Coolant System Instrumentation; Revision 26
- Drawing ISID2-0030A; Reactor Coolant System; Revision 11
- Drawing OS-0001A; Reactor Coolant System; Revision 46
- Calculation C-CSS-100.05-001; Service Level 1 Non-DBA Qualified Protective Coating
Application Inventory; Revision 5 and Addendum A01 and Addendum A05
8
- Calculation C-NSA-059.01-023; Davis Besse Containment Building LOCA Debris Generation;
Revision 0
- Job 6472-11-0327; Mix ID: Davis-Besse 645-31; Set 20; Shield Building Opening
Compression Test Results; April 14, 2014
- Job 6472-11-0327; Mix ID: Davis-Besse 645-31; Set 21; Shield Building Opening
Compression Test Results; April 14, 2014
- ALARA Plan 2014-5150; Reactor Head Removal and Replacement; Revision 0
- Reactor Vessel Closure Head (RVCH) Removal Briefing Package; February 5, 2014
- Control Room Logs February 2-4, 2014
- 18RFO Shutdown Defense in Depth Report; Revisions 0-1
- Davis-Besse Unit 1, Cycle 19 Core Map
- Davis-Besse Cycle 19 Core Fuel Assembly ID Verification DVD
- Davis-Besse Cycle 19 Core Operating Limits Report; Revision 0
- Certificate of Conformance for All Work under Bechtel Subcontract No 25539-000-HC4-NEE0-
00016 (Change Order 5); April 24, 2014
- CR-2014-02896; Void in Concrete Discovered at the top of the Shield Building Construction
Opening; dated February 14, 2014
- Apparent Cause Evaluation for CR-2014-02896; Void in Concrete Discovered at the top of the
Shield Building Construction Opening; April 14, 2014
- Operability Evaluation for CR-2014-02896; Void in Concrete Discovered at the top of the
Shield Building Construction Opening; Revision 1
- CR-2014-03472; Shield Building Construction Opening Unfilled Area-Inspection Findings;
February 21, 2014
- CR-G201-2002-07080; Shield Building Concrete Voids; October, 1, 2002
- CR-G201-2002-07472; Voids in the Containment Shield Building - Annulus Side; October, 4,
2002
- CR-2011-05777; Void in Concrete at the Top of the Shield Building Pour Back (Bechtel NCR
20); November 21, 2011
- CR-2011-05795; Concrete Void at Top of Shield Building Restoration Larger than Previously
Reported Per CR-2011-05777; November 22, 2011
- CR-2014-07979; TE 1/2 Loop Model Reanalysis - AREVA Calculation 32-1177088-00; April 30,
2014
- NCR-25539-200-G61-GCE-00042; Concrete Void Discovered at Top of Construction Opening
on Annulus Side; February 13, 2014
- NCR-25539-200-G61-GCE-00020; Concrete Void Discovered at Top of Construction Opening
on Exterior Side; November 21, 2011
- NCR-020; Concrete Void at Top of Shield Building Construction Opening on Exterior Side;
October 1, 2002
- NCR-021 Concrete Void at Top of Shield Building Construction Opening on Annulus Side;
October 3, 2002
- Specification 12501-C-401Q; Technical Specification for Operational Phase for Forming,
Placing, Finishing, and Curing of Concrete; Revision 0
- DB-MS-09010; Maintenance Services Procedure: Concrete Placement Procedure; June 3,
1999
- Specification 25539-000-3PS-DB01-Q0001; Engineering Specification for Purchase of
Ready-Mix Concrete for Safety-related Applications; Revision 005
- ECP 12-0066-000; Engineering Change Package: Create and Restore Temporary Opening in
Containment Vessel Wall; Revision 2
- ECP 12-0066-001; Engineering Change Package: Create Temporary Opening in Containment
Vessel Wall; Revision 2
9
- ECP 12-0066-000; Engineering Change Package: Restore Temporary Opening in
Containment Vessel Wall; Revision 2
- Addendum A02 to Calculation C-CSS-059.01-003; Maximum External Pressure Capacity of
Containment using ASME Subsection NE; August 28, 2012
- Addendum A01 to Calculation C-CSS-059.01-004; Maximum External Pressure Capacity of
Containment using ASME Code Case N-284; May 20, 2011
- Addendum A02 to Calculation C-CSS-059.01-004; Maximum External Pressure Capacity of
Containment using ASME Code Case N-284; August 28, 2012
- Addendum A02 to Calculation C-CSS-059.01-005; Containment Vessel - External Pressure;
May 20, 2011
- Addendum A01 to Calculation C-CSS-059.01-005; Containment Vessel - External Pressure;
August 28, 2012
- Addendum A04 to Calculation C-CSS-059.01-017; Monorail System for Rigging and Storing
the Containment Vessel Opening Door Sheet; August November 26, 2013
- Addendum A04 to Calculation C-CSS-059.01-018; Evaluation of Containment Vessel for
Construction Opening; July 10, 2013
Addendum A05 to Calculation C-CSS-059.01-018; Evaluation of Containment Vessel for
Construction Opening; November 27, 2013
- Calculation C-CSS-059.01-022; ASME Evaluation of the Containment Vessel for Bracing the
Shield Building Formwork Against the Containment Vessel; Revision 0
- ECP 12-0065-000; Create and Restore Shield Building Construction Opening; Revision 3
- ECP 12-0065-002; SGR-18RFO - Shield Building Construction Opening, Cut Wall Opening;
Revision 0
- Addendum A03 to Calculation C-CSS-099.20-045; Evaluation of Shield Building for the
Construction Opening - SGR-RVCH Replacement; September 12, 2013
- ECP 12-0067-000; Steam Generator and Reactor Coolant Pump Replacement; Revision 3
- ECP 12-0474-000; Replacement Once Through Steam Generator and Replacement Hot Leg
Piping Qualification; Revision 1
- Calculation DB-08Q-340; Weld Overlay Sizing for Hot Leg Surge Nozzle; Revision 1
- Calculation 32-1177088; TE 1/2 Loop Model Reanalysis; Revision 0
- Calculation 32-9191681; DB-1 RCS 1/2-Loop Model and Loading Analysis; Revision 2
- Calculation 32-9206186; RCS Pipe Stress and Fatigue; Revision 0
- Calculation 32-9204864; DB Loop Piping - Evaluations of Nozzles and Supports; Revision 0
- Calculation 86-9205786; DB Nozzle and Spring Loads from RCS Model; Revision 2
- Certified Design Specification TS-3985; Technical Specifications; Revision 2
- Report 205S-LR-04; ASME Code Reconciliation Report; Revision 1
- Report 205S-SR-01; Replacement Once Through Steam Generators, Base Design Condition
Report; Revision 1
- Report 205S-SR-02; Replacement Once Through Steam Generators, Transient Analysis
Stress Report; Revision 0
- Report 205S-SR-08.2; Replacement Once Through Steam Generator 1-2 (NG2); Revision 0
- 10 CFR 50.59 Evaluation 91-0046; Modification 90-0079; Revision 1
- 10 CFR 50.59 Screen 12-00901; SGR-18RFO - Shield Building Construction Opening;
Revision 2
- 10 CFR 50.59 Screen 12-00901; SGR-18RFO - Shield Building Construction Opening;
Revision 2
- 10 CFR 50.59 Screen 12-01004; Large Bore Piping Changes Associated with Steam
Generator Replacement; Revision 3
- 10 CFR 50.59 Screen 12-03404; Replacement Once Through Steam Generator and
Replacement Hot Leg Piping Qualification; Revision 0
- 10 CFR 50.59 Screen 13-04350; Calculation C-NSA-000.02-011; Revision 0
10
- 10 CFR 50.59 Screen 13-04680; Calculation C-NSA-060.05-010; Revision 0
- 10 CFR 50.59 Evaluation 13-04680; Calculation C-NSA-060.05-010; Revision 0
- 10 CFR 50.59 Screen 14-00606; Calculation C-NSA-064.02-036; Revision 0
- Drawing 205SE001; Davis Besse ROTSG General Arrangement; Revision 6
- Drawing 205SE001; Davis Besse ROTSG General Arrangement Cross Sections; Revision 4
- Drawing 205SE010; AS-Built Dimensions ROTSG (1-1); Revision 1
- Drawing 205SE011; AS-Built Dimensions ROTSG (1-2); Revision 2
11
LIST OF ACRONYMS
ACE Apparent Cause Evaluation
ADAMS Agencywide Document Access Management System
ALARA As-Low-As-Is-Reasonably-Achievable
ASME American Society of Mechanical Engineers
ASTM American Society for Testing and Materials
CAP Corrective Action Program
CFR Code of Federal Regulations
CR Condition Report
ECP Engineering Change Package
IMC Inspection Manual Chapter
IP Inspection Procedure
IR Inspection Report
ISI Inservice Inspection
MT Magnetic Particle Testing
NCR Non-Conformance Report
NCV Non-Cited Violation
NDE Non-destructive Examinations
NEI Nuclear Energy Institute
NRC U.S. Nuclear Regulatory Commission
OBE Operating Basis Earthquake
OLS Outside Lift System
OOTSG Original Once Through Steam Generator
PARS Publicly Available Records System
RSG Replacement Steam Generator
RT Radiographic Test
SDP Significance Determination Process
TLD Temporary Lifting Device
USAR Updated Safety Analysis Report
UT Ultrasonic Test
R. Lieb -2-
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 James E. Neurauter Acting For/
David E. Hills, Chief
Engineering Branch 1
Division of Reactor Safety
Docket No. 50-346
License No. NPF-3
Enclosure:
Inspection Report 05000346/2013010
w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ'
DISTRIBUTION:
RidsNrrDorlLpl3-2 Resource
RidsNrrPMDavisBesse Resource
RidsNrrDirsIrib Resource
Cynthia Pederson
DRPIII
DRSIII
Patricia Buckley
ROPreports.Resource@nrc.gov
DOCUMENT NAME: G:\DRSIII\DRS\Work in Progress\DAV 2013 010 SG Replacement AAS.docx
Publicly Available Non-Publicly Available Sensitive Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII RIII
NAME AShaikh:ls DHills JCameron
DATE 07/22/14 07/22/14 07/22/14
OFFICIAL RECORD COPY