IR 05000409/2007001
| ML071730309 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/22/2007 |
| From: | Louden P NRC/RGN-III/DNMS/DB |
| To: | Berg W Dairyland Power Cooperative |
| References | |
| IR-07-001 | |
| Download: ML071730309 (19) | |
Text
une 22, 2007
SUBJECT:
NRC INSPECTION REPORT 050-00409/07-001(DNMS) -
LA CROSSE BOILING WATER REACTOR (LACBWR)
Dear Mr. Berg:
On May 30, 2007, the NRC completed an inspection at the La Crosse Boiling Water Reactor (LACBWR) facility. The purpose of the inspection was to determine whether decommissioning activities were conducted safely and in accordance with NRC requirements in the areas of facility management and control, decommissioning support, radiological safety, and spent fuel safety. At the conclusion of the inspection on May 30, 2007, the NRC inspectors discussed findings with members of your staff.
The inspection was an examination of 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. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations of activities, and interviews with personnel.
Based on the results of these inspections, the NRC has determined that one Severity Level IV violation of NRC requirements occurred. This violation is being treated as a Non-Cited Violation (NCV), consistent with Section VI.A of the Enforcement Policy. The NCV is described in the subject inspection report. If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to: (1) the Regional Administrator, Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; and (2) the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). The NRCs document system is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. We will gladly discuss any questions you have concerning this inspection.
Sincerely,
/RA by W. Snell Acting for/
Patrick L. Louden, Chief Decommissioning Branch Docket No. 050-00409 License No. DPR-45
Enclosure:
Inspection Report 050-00409/07-001(DNMS)
REGION III==
Docket No.: 050-00409 License No.: DPR-45 Report No.: 050-00409/07-001(DNMS)
Licensee: Dairyland Power Cooperative 3200 East Avenue South La Crosse, WI 54602 Facility: La Crosse Boiling Water Reactor Location: La Crosse Site Genoa, Wisconsin Dates: February 26 through May 30, 2007 Inspector: William G. Snell, Senior Health Physicist James E. Neurauter, Senior Reactor Inspector Peter J. Lee, Ph.D., CHP, Health Physicist Approved by: Patrick L. Louden, Chief Decommissioning Branch Enclosure
EXECUTIVE SUMMARY La Crosse Boiling Water Reactor (LACBWR)
NRC Inspection Report 050-00409/07-001(DNMS)
This decommissioning inspection covered aspects of facility management and control, decommissioning support activities, radiological safety, spent fuel safety, and additional security measures for reactor pressure vessel (RPV) shipment of radioactive material quantities of concern.
Facility Management and Control
- The inspectors identified one violation of 10 CFR Part 50, Appendix B, Criterion III for the failure to adequately evaluate the seismic response of the Temporary Lifting Device structure (TLD). This finding is being treated a Non-Cited Violation (NCV) consistent with Section VI.A.8 of the NRC Enforcement Policy based on the licensees immediate actions taken to correct the violation, identify the root cause, and prevent recurrence.
(Section 1.1)
- The inspectors determined that the licensee was performing audits and surveillances that were broad in scope, and actively following up to ensure identified items were being addressed in an adequate and timely manner. (Section 1.2)
- The inspectors determined that the licensee was adequately controlling decommissioning activities and radiological work areas. The performance, content, and quality of the licensees meetings to plan and coordinate decommissioning activities was adequate. (Section 1.3)
Radiological Safety
- The inspectors determined that the licensee continued to be effective in controlling workers personal exposure. (Section 2.1)
- The inspectors determined that the licensee adequately implemented its effluent monitoring program. (Section 2.2)
- The inspectors determined that the licensee complied with regulatory requirements for shipping radioactive materials. (Section 2.3)
Spent Fuel Safety
- The inspectors determined that the licensee properly maintained the fuel element storage well water level, temperature, chemistry, and cleanliness to ensure the safe wet storage of the spent fuel. (Section 3.1)
Enclosure
Other Activities
- The inspectors determined that the licensee effectively implemented the additional security measures imposed by U.S. NRC Order EA-07-014, Imposing Additional Security Measures on the Transportation of RAMQC, issued February 12, 2007.
(Section 4.1)
Enclosure
Report Details1 Summary of Plant Activities The licensees current activities were focused on routine operations regarding the safe storage of spent fuel in the fuel pool, and preparations for disposal of the reactor pressure vessel (RPV).
1.0 Facility Management and Control 1.1 Safety Reviews, Design Changes and Modifications (37801)
a. Inspection Scope The inspectors reviewed the safety evaluation performed by the licensee for the RPV support disconnection, lift, and transit to determine if the licensees evaluation was performed pursuant to Title 10, CFR Part 50, Licensing of Production and Utilization Facilities, Section 50.59, Changes, Tests, and Experiments. The inspector used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, to determine acceptability of the completed evaluation. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.
The inspectors reviewed licensee design documentation including structural and rigging calculations, drawings, and procedures that demonstrated that the RPV would be safely lifted and removed from the reactor building using the proposed Temporary Lifting Device (TLD) and existing structures to support TLD structural members. The inspectors verified that the TLD and rigging calculations were in accordance with recognized codes and standards. The inspectors also verified that the TLD was installed and supported in accordance with the design calculations and design drawings.
The inspectors reviewed the lifting lug fabrication report including non-conformance reports (NCRs) to verify the indication identified by non-destructive examination (NDE)
was evaluated and removed, and the lifting lug was reinspected and dispositioned in accordance with applicable codes and standards.
The inspectors reviewed the licensees RPV removal load path to verify that safety systems used to remove decay heat from the Fuel Element Storage Well (FESW)
would not be adversely affected if the RPV were to be dropped inside the reactor building.
The inspectors witnessed the load test of the TLD outside girders and the functional test of the strand jacks and trolley to verify that these tests were performed in
NOTE: A list of acronyms used in the report is included at the end of the report.
Enclosure
accordance with approved procedures. The inspectors witnessed the removal of the RPV from the reactor building, to verify that the RPV weight used in the design calculations bounded the actual RPV weight, and that the RPV removal was performed in accordance with approved procedures.
b. Observations and Findings b.1 Failure to Adequately Evaluate Seismic Response of TLD Structure The inspectors identified that licensee calculation 0526301.11-S008, Seismic Analysis of LACBWRs Temporary Lifting Device (TLD) Structure for Removing RPV from the Reactor Building, Revision 0, used a truncated structural model of the TLD structure and supporting system. Specifically, the truncated TLD model did not include the TLD vertical frame members outside of the reactor building. In order to determine the seismic reactions for the TLD supports at the reactor building outer wall and the FESW, the TLD model included the mass of the vertical frame members. The inspectors determined that the truncated model may produce non-conservative seismic results in the lateral horizontal direction due to rigid lateral TLD restraints used for the truncated model.
The licensee revised calculation 0526301.11-S008 to model the entire TLD structure.
As a result, the seismic lateral forces on the vertical TLD frame members increased, and additional weight was added at the frame brace restraint to ensure TLD frame stability during a DBE seismic event. This modification to the TLD supporting structure was completed prior to TLD load testing and disconnecting and lifting the RPV.
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures be established to assure that applicable regulatory requirements and the design basis, for those systems, structures and components for which this appendix applies, are correctly translated into specifications, drawings, procedures and instructions. It further requires that the design control measures provide for verifying or checking the adequacy of the design.
Contrary to the above, on May 4, 2007, the seismic design adequacy of the TLD structure was not verified. The truncated model of the TLD vertical members outside the reactor building resulted in non-conservative seismic reactions requiring additional weight to be added at the TLD brace to ensure TLD frame stability during a DBE seismic event. In accordance with the Enforcement Policy, this violation of the requirements of 10 CFR Part 50, Appendix B, Criterion III is classified as a Severity Level IV Violation because the licensee revised calculation 0526301.11-S008 and added additional weight at the frame brace restraint prior to TLD load testing and disconnecting and lifting the RPV. Although this violation was NRC identified, it is considered a Non-Cited Violation (NCV) consistent with Section VI.A.8 of the NRC Enforcement Policy based on the licensees immediate actions taken to correct the violation, identify the root cause, and prevent recurrence.
The inspectors reviewed the licensees corrective actions, Surveillance Corrective Action Report No. 37-02, Open Item A: the licensee revised calculation 0526301.11-S008 and related TLD calculations affected by changes to the results of Enclosure
calculation 0526301.11-S008, added additional weight at the frame brace restraint prior to TLD load testing and disconnecting and lifting the RPV, identified the root cause of the violation, and identified actions to prevent recurrence. The inspectors determined that the corrective actions were appropriate to address all the immediate and potential generic aspects of the violation.
b.2 10 CFR 50.59 Evaluation Observation The inspectors identified that in 10 CFR 50.59 evaluation, Document No. FC 30-06-08 (un-approved Revision 1), the licensee justified omitting seismic design requirements based on the probability that a seismic event is unlikely to occur concurrent with the RPV suspended under the TLD inside the reactor building. However, the LACBWR Decommissioning Plan SAFSTOR accident analysis for a seismic event postulates that a design basis earthquake (DBE) occurs and indicates that the reactor building and FESW structures can withstand the postulated loads.
Using NEI 96-07 Section 4.3.8.1 for guidance related to changing a method of evaluation, the inspectors determined that not evaluating the effect of the DBE load case could result in gaining margin with respect to the design loading conditions.
Therefore, removal of the seismic design requirements was a nonconservative change and a weakness in the licensees implementation of the 10 CFR 50.59 review process.
The licensee revised the evaluation to incorporate the design basis seismic requirements prior to final approval of the 50.59 evaluation.
The inspectors reviewed the licensees revised 50.59 evaluation which was addressed in Surveillance Corrective Action Report No. 37-01, Open Item B, in which the licensees approved 10 CFR 50.59 evaluation incorporated the design basis seismic requirements into the facility change prior to TLD load testing and disconnecting and lifting the RPV. The inspectors determined that the corrective actions were appropriate to address all the immediate and potential generic aspects of the issue.
b.3 Effect of TLD Loading on FESW Structure Observation The inspectors identified that licensee calculation 0526301.11-S007, Structural Integrity Evaluation of Fuel Element Storage Well and Fuel Racks Due to Reactor Pressure Vessel Removal, Revision 0, did not explicitly evaluate the FESW for the effect of supporting the TLD during RPV disconnection and lift to remove the RPV from the reactor building. The calculation relied on engineering judgment that the TLD loading on FESW structure would not be significant compared to the available design margin without providing justification in the calculation.
The inspectors reviewed the licensees corrective actions to address this issue, as documented in Surveillance Corrective Action Report No. 37-01, Open Item C. The licensees approved calculation 0526301.11-S007 explicitly evaluated the effect of supporting the TLD using the existing FESW structure prior to disconnecting and lifting the RPV. The licensee was able to demonstrate that the FESW structure had adequate design margin to support the TLD during RPV removal. However, the calculated local effect of the TLD support loading on the FESW was not insignificant Enclosure
with respect to the calculated FESW design capacity, but the corrective actions were appropriate to address all the immediate and potential generic aspects of the issue.
c. Conclusions The inspectors identified one Severity Level IV violation of 10 CFR Part 50, Appendix B, Criterion III for the failure to adequately evaluate the seismic response of the TLD structure. Based on the licensees immediate actions taken to correct the concern, identify the root cause, and prevent recurrence, this finding is being treated as a Non-Cited Violation (NCV) consistent with Section VI.A.8 of the NRC Enforcement Policy.
1.2 Self Assessment, Auditing, and Corrective Actions (40801)
a. Inspection Scope The inspectors reviewed the licensees activities associated with incident reporting, quality assurance (QA) audits, quality control (QC) surveillances, and the tracking and followup of identified deficiencies and issues. The inspectors reviewed the licensees Operational Review Committee/Safety Review Committee (ORC/SRC) Commitment Tracking list of current open items, and Administrative Control Procedure (ACP) 17.1, Incident Reports, Including Notification to the NRC, Issue 31, dated 4/2/07. The inspectors met with licensee personnel and reviewed and discussed the results of several recent audits and surveillances in which a number of items were identified.
b. Observations and Findings The licensees ORC/SRC Commitment Tracking list contained primarily open audit items and open items from QC surveillances. The Tracking list provided a summary of each open item, a response due date, who the item was assigned to, and the current status of actions taken to address the item. Approximately 30 items were on the list covering a broad area of plant issues. All the items appeared to be on track to be completed by the assigned due dates. Administrative Control Procedure 17.1 described the process for identifying and responding to incidents. Although the procedure specifically addressed incidents of some significance, such as a disabling or fatal injury, failure of a system important to safety, or significant fabrication error, the licensee also used it to track lower level items identified by the staff. Separate logs were kept to track minor injuries and maintenance requests.
c. Conclusions The inspectors determined that the licensee was performing audits and surveillances that were broad in scope, and actively following up to ensure identified items were being addressed in an adequate and timely manner.
Enclosure
1.3 Decommissioning Performance and Status Review at Permanently Shut Down Reactors (71801)
a. Inspection Scope The inspectors conducted plant tours to assess field conditions and decommissioning activities and ensure that radioactively contaminated areas were being controlled. The inspectors observed several licensee pre-job briefings and an ORC meeting to assess the performance and quality of the licensees planning and coordination of decommissioning activities.
b. Observations and Findings During site tours the inspectors noted that the material condition of facilities and equipment were commensurate with current decommissioning activities. The individuals conducting the tours with the inspectors were cognizant of the work activities in process, the dose levels throughout the facilities, and the locations of contaminated areas. Work areas were observed to be adequately controlled, postings and boundaries were appropriate, and workers were wearing personal protective clothing that was suitable for the work they were performing. The pre-job briefings were well attended, addressed all areas of the work to be performed, including safety issues, radiological controls, work sequence, hold points, and emergency contingencies. The ORC meeting was conducted to review and approve the sequence of actions regarding cutting the bolts holding the feet of the reactor vessel in place, the initial six inch lift, and the cutting of a portion of the feet off. The meeting was attended by appropriate personnel, including the plant manager, health and safety supervisor, reactor engineer, vessel lift project manager, and QA personnel. The ORC thoroughly discussed the issue and ensured all questions and concerns were adequately addressed before approving the activity.
c. Conclusions The licensee was adequately controlling decommissioning activities and radiological work areas. The performance, content, and quality of the licensees meetings to plan and coordinate decommissioning activities was adequate.
2.0 Radiological Safety 2.1 Occupational Radiation Exposure (83750)
a. Inspection Scope The inspectors reviewed the ALARA reviews and radiation work permits of the cutting of suction and discharge nozzles, perimeter nozzles, control rod extension tubes, bolts and feet from the RPV footing, and removal of the insulation of the RPV, and transferring RPV from the Reactor Building to the shipping canister, and grouting of the canister containing the RPV, to verify that the licensee had established procedures, and engineering and work controls that were based on sound radiation protection principles in order to achieve occupational exposures that were ALARA. The inspectors attended health physics staff pre job meeting to assess the radiological Enclosure
planning of select portions of RPV removal activities. The inspectors reviewed external exposure records, and breathing zone air sampling results to assess the internal doses.
b. Observations and Findings The radiological planning of the health physics staff pre-job meeting provided adequate radiation protection coverage, and the ALARA reviews were effective in minimizing unnecessary doses. The licensee provided sufficient shielding of radiation sources in a practical way. Also, to reduce potential intakes of radioactive materials, the licensee installed high efficiency particulate air filter-equipped (HEPA) filtration systems in the vicinity of the work areas, and workers also wore half-face mask respirators in the reactor low cavity area.
Based on the review of external exposure records, workers received doses less than the dose limits specified in radiation work permits. Based on the breathing zone air sampling results, workers received no significant internal doses.
The significant exposure for the RPV removal project was from January 2006 to May 2007. The total personnel external exposures were 31314 mrem and the highest dose received by an individual was 2625 mrem. Air sampling data indicated that most of the internal dose was due to americium-241. Based on the DAC-hours dose assessments, the total personnel committed effective dose equivalents (CEDE) were 341 mrem and the highest CEDE received by the individual was 60 mrem.
c. Conclusions The inspectors determined that the licensee continued to be effective in controlling workers personal exposure.
2.2 Radioactive Waste Treatment, and Effluent and Environmental Monitoring (84750)
a. Inspection Scope The inspectors evaluated the licensees activities to effectively control, monitor, and quantify releases of radioactive materials in liquid, gaseous, and particulate forms to the environment. The inspector reviewed the licensees 2006 Effluent and Environmental Monitoring Reports, and the Offsite Dose Calculation Manual (ODCM).
The inspector evaluated the dose to the general public due to potential release of radioactive materials in air during the transfer of the RPV out of the Reactor Building through the bi-parting door.
b. Observations and Findings The licensees gaseous effluent monitors and waste water effluent monitor were calibrated and checked for proper operation in accordance with station procedures.
The licensee participates in a cross check program with an off-site laboratory to confirm the quality of its analytical data. Results of a cross check of licensee laboratory results completed in calendar year 2006 indicated agreement in all analytical data.
Enclosure
The ODCM was comprehensive and contained the requirements listed in the licensees technical specifications. The effluent monitoring data indicated that release concentrations were consistent with limits specified in10 CFR Part 20, Appendix B, Table 2, and that doses to the general public were in conformance with Appendix I of 10 CFR Part 50. Further, environmental sampling results indicated only background radiation levels with no distinct contribution from the shutdown reactor.
Prior to removing the RPV from the reactor building, the insulation of the RPV was removed. The removal of the insulation of the RPV generated significant amount of contaminated dust. A grab air sample was taken to determine when to open the bi-parting door of the reactor building to minimize the potential for exposure to the general public from an air effluent release through the bi-parting door. Based on the grab sampling result, the air concentration of Am-241 was less than 4.0 x 10 -11 µCi/ml.
During transfer of the RPV, the bi-parting door was open for 45 minutes. Therefore, the potential dose received by a member of the public near the Reactor Building would be less than 10 mrem.
c. Conclusions The inspectors determined that the licensee adequately implemented its effluent monitoring program.
2.3 Transportation of Radioactive Materials (86750)
a. Inspection Scope The inspectors reviewed the radioactive materials shipping program and applicable shipping documents. The inspector evaluated whether the licensee was in compliance with NRC and Department of Transportation (DOT) shipping requirements.
b. Observations and Findings The licensee had processed four LSA-II and two SCO-II waste shipments since September 2006. The licensee shipped the radiological waste to Barnwell Waste Management Facility in Barnwell, South Carolina. The licensee verified that the results of radiation and removable contamination levels were within applicable limits.
During the inspection the RPV transportation package was being processed as a rail shipment. The entire RPV was classified as Class C waste. The transportation package is the RPV enclosed within a cylinder steel canister and concrete grouting.
This is an exclusive use Type B package to be used for a one-time shipment and disposal of the RPV at Barnwell Waste Management Facility in Barnwell, South Carolina. The inspectors verified that licensee had established a transportation security plan, and a transportation emergency response plan to effectively control the shipment. The inspectors observed the licensee conduct required surface contamination surveys and radiation level surveys of the package. No detectable surface contamination was identified. The radiation levels measured by the licensee were within regulatory limits of external radiation standards for Type B packages.
Enclosure
The licensees shipping manifests showed that personnel packaged, labeled, and marked each shipping container according to the DOT and 10 CFR Part 71 transportation requirements.
c. Conclusions The inspectors determined that the licensee complied with regulatory requirements for shipping radioactive materials.
3.0 Spent Fuel Safety 3.1 Spent Fuel Pool Safety at Permanently Shutdown Reactors (60801)
a. Inspection Scope The inspectors reviewed the licensees activities to ensure the safe wet storage of spent fuel in the Fuel Element Storage Well (FESW). The review included the verification of water temperature, and water level requirements of Technical Specification (TS) 4.1.2, the surveillance requirements of TS 5.1.2, and the water chemistry and cleanliness control requirements of the licensees Health and Safety Procedure HSP-7.2, for the period of January through May 2007.
b. Observations and Findings All parameters reviewed were consistent with limits specified in HSP-7.2, Sampling of Fuel Element Storage Well. The FESW water level and temperature met the requirements of TS 4.1.2. The FESW water level and temperature had been monitored daily as required by the surveillance requirements of TS 5.1.2.1.
c. Conclusions The inspectors determined that the licensee properly maintained the FESW water level, temperature, chemistry, and cleanliness to ensure the safe wet storage of the spent fuel.
4.0 Other Activities 4.1 (Closed) NRC Temporary Instruction (TI) 2201/001 - Inspection of Additional Security Measures for Radioactive Material Quantities of Concern (RAMQC)
a. Inspection Scope The inspectors reviewed program procedures, implementing procedures and records and conducted interviews with responsible personnel and plant employees to evaluate the implementation of the additional security measures on transportation of radioactive material quantities of concern.
Enclosure
b. Observations and Findings The licensee verified the RPV shipment as the transportation of RAMQC based on the activity of the radioisotope of concern in the RPV. However, it would be essentially impossible to separate or concentrate this distributed radioactivity of concern from the massive package being shipped because of low density cellular concrete filling the void space in RPV.
The inspectors verified that the licensee effectively incorporated the Safeguards Information - Modified Handling (SGI-M) protection requirements specified in NRC Orders EA-06-0243 and EA-06-244, Imposing Requirements for the Protection of Certain Safeguards Information, issued November 15, 2006, into its information protection strategy. The inspectors also verified that the licensee effectively implemented the following additional security measures required for the transportation of RAMQC: (1) licensee verification, (2) background investigations, (3) preplanning and coordination, (4) notifications, (5) communications, (6) train crew and accompanying riders, and (7) procedures, training and control of information.
c. Conclusion The inspectors determined that the licensee effectively implemented the additional security measures imposed by U.S. NRC Order EA-07-014, Imposing Additional Security Measures on the Transportation of RAMQC, issued February 12, 2007.
5.0 Exit Meeting The inspectors presented the inspection results to members of the licensees staff at the conclusion of the inspection on May 30, 2007. The licensee confirmed that licensee structure drawings and documents generated by their contractors in support of the RPV removal project were considered proprietary. It was agreed that all paper copies of these documents would be returned to the licensee or shredded, and all electronic files of these documents would be erased.
Enclosure
SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED
- R. Christians, Plant Manager
- R. Cota, Training/Security Supervisor
- J. Henkelman, Quality Assurance Specialist
- M. Johnsen, Tech Support Engineer
- L. Nelson, Health and Safety Supervisor
- S. Rafferty, Reactor Engineer
- M. Moe, Captain, Burns Security
- D. Egge, Quality Assurance Supervisor
- M. Brasel, Project Engineer/Manager
- J. McRill, Tech Support Engineer
- Persons present at the exit meeting.
INSPECTION PROCEDURES USED IP 37801: Safety Reviews, Design Changes and Modifications IP 40801: Self Assessment, Auditing, and Corrective Actions IP 70801: Decommissioning Performance and Status Review at Permanently Shut Down Reactors IP 83750: Occupational Radiation Exposure IP 84750: Radwaste Treatment, Effluent, and Environmental Monitoring IP 60801: Spent Fuel Pool Safety at Permanently Shutdown Reactors TI 2201/001 Additional Security Measures for Radioactive Material Quantities of Concern ITEMS OPENED, CLOSED, AND DISCUSSED Opened 05000409/2007001-01 NCV Failure to Adequately Evaluate Seismic Response of TLD Structure Closed 05000409/2007001-01 NCV Failure to Adequately Evaluate Seismic Response of TLD Structure Discussed None Attachment
INITIALISMS AND ACRONYMS ACP Administrative Control Procedure ADAMS Agencywide Documents Access and Management System CFR Code of Federal Regulations DNMS Division of Nuclear Materials Safety FESW Fuel element storage well LACBWR La Crosse Boiling Water Reactor mrem Millirem NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual PARS Publicly Available Records RPV Reactor Pressure Vessel TS Technical Specifications LSA Low Specific Activity SCO Surface Contaminated Object RAMQC Radioactive Material Quantities of Concern
µCi/ml micro curies/milliliter NEI Nuclear Energy Institute NCR Non-Conformance Report NCV Non-Cited Violation TLD Temporary Lifting Device FC Facility Change NDE Non-Destructive Examination 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.
Calculations 2150-C10; Strand Jack Trolley Adequacy; Revision 1 2150-C30; TLD Runway Structure; Revision 0 2150-C30; TLD Runway Structure; Revision 2 2150-C30; TLD Runway Structure; Revision 2 2150-C50A; RPV Lift Lug and Misc Below Hook Rigging; Revision 0 2150-C50B; RPV Head Adequacy; Revision 0 2150-CTLD Seismic; TLD Runway & Trolley Structure Seismic; Revision 0 2150-CTLD Seismic; TLD Runway & Trolley Structure Seismic; Revision 1 Attachment
0526301.11-S-001; Regeneration of LACBWR 1982 Containment Building Model for Seismic and Structural Analysis; Revision 0 0526301.11-S-002; Seismic Analysis of Modified LACBWR Containment Building; Revision 0 0526301.11-S-002; Seismic Analysis of Modified LACBWR Containment Building; Revision 2 0526301.11-S-003; Structural Analysis of LACBWR Modified Reactor Building Outer Shield Wall to Support TLD Girder Loads during RPV Removal; Revision 1 0526301.11-S007; Structural Integrity Evaluation of Fuel Element Storage Well and Fuel Racks Due to Reactor Pressure Vessel Removal; Revision 0 0526301.11-S-007; Structural Integrity Evaluation of Fuel Element Storage Well and Fuel Racks Due to Reactor Pressure Vessel Removal; Revision 1 0526301.11-S-008; Seismic Analysis of LACBWRs Temporary Lifting Device (TLD) Structure for Removing RPV from the Reactor Building; Revision 0 0526301.11-S-008; Seismic Analysis of LACBWRs Temporary Lifting Device (TLD) Structure for Removing RPV from the Reactor Building; Revision 1 ST-520; LACBWR Transportation and Disposal Container Weight Calculation; Revision 0 Drawings Drawing No. 10, Sheet 1; TLD Hydraulic Jack Trolley, General Arrangement; Revision 4 Drawing No. 10, Sheet 2; TLD Hydraulic Jack Trolley, General Arrangement; Revision 2 Drawing No. 10, Sheet 3; TLD Hydraulic Jack Trolley, Trolley Frame Assembly; Revision 2 Drawing No. 10, Sheet 4; TLD Hydraulic Jack Trolley, General Arrangement; Revision 3 Drawing No. 30, Sheet 1; TLD Runway System, General Arrangement; Revision 9 Drawing No. 30, Sheet 2; TLD Runway System, General Arrangement - East Girder; Revision 4 Drawing No. 30, Sheet 3; TLD Runway System, General Arrangement - West Girder; Revision 2 Drawing No. 30, Sheet 4; TLD Runway System, General Arrangement - Section View; Revision 4 Drawing No. 30, Sheet 5; TLD Runway System, General Arrangement - Section View; Revision 5 Drawing No. 30, Sheet 6; TLD Runway System, East Runway at RB Wall - Details; Revision 7 Drawing No. 30, Sheet 7; TLD Runway System, West Runway at RB Wall - Details; Revision 7 Attachment
Drawing No. 30, Sheet 8; TLD Runway System, RB Rocker Bearing Details; Revision 8 Drawing No. 30, Sheet 9; TLD Runway System, East Runway Girder Inside Rocker Bearing; Revision 2 Drawing No. 30, Sheet 10; TLD Runway System, West Runway Girder Inside Rocker Bearing; Revision 2 Drawing No. 30, Sheet 11; TLD Runway System, General Arrangement - Crane Mat Layout; Revision 5 Drawing No. 50, Sheet 1; RPV Rigging Components: Lift Lug, Equalizers and Pins; Revision 1 Drawing No. 50, Sheet 2; RPV Rigging Components: Lift Lug, Equalizers and Pins; Revision 1 Drawing No. 50, Sheet 3; RPV Rigging Components: Lift Lug, Equalizers and Pins; Revision 3 Drawing No. 50, Sheet 4; RPV Rigging Components: Lift Lug, Equalizers and Pins; Revision 1 Drawing No. 500, Sheet 1; RPV Rigging Components: Lift Lug; Revision 5 Drawing No. 500, Sheet 2; RPV Rigging Components: Lift Lug; Revision 4 Drawing No. 500, Sheet 3; RPV Rigging Components: Lift Lug; Revision 4 Drawing No. 500, Sheet 4; RPV Rigging Components: Lift Lug; Revision 4 Miscellaneous 0526301.11-002; Phase 2 - LACBWR Reactor Pressure Vessel Removal, Structural Analysis and Design Criteria; Revision 0 2150-D1; Engineering Design Basis: Engineering , Rigging and Onsite Transport Services -
Reactor Pressure Vessel Removal Project; Revision 1 2150-D1; Engineering Design Basis: Engineering , Rigging and Onsite Transport Services -
Reactor Pressure Vessel Removal Project; Revision 2 2150-P6; Procedure: Rig-in, Erection, Load Test, Use and Disassembly of TLD System; Revision 0 2150-P6; Procedure: Rig-in, Erection, Load Test, Use and Disassembly of TLD System; Revision 1 10 CFR 50.59 Evaluation; FC 30-06-08: RPV Supports Disconnection, Lift and Transit; Revision 1 NCR No.07-007; Bigge Nonconformance Report: RPV Lift Lug; Revision 0 NCR No. 07-14; American Tank & Fabricating Company Nonconformance Report: RPV Lift Lug; Revision 0 Attachment
Surveillance Corrective Action Report No. 37-01, Open Item A; Reevaluate Preload Conditions on Special Lifting Device to RPV Attachment; dated April 3, 2007 Surveillance Corrective Action Report No. 37-01, Open Item B; 10 CFR 50.59 Evaluation for Facility Change 30-06-08 for Seismic Event Not Per D-Plan Section 9.8; dated April 3, 2007, including June 6, 2007, Supplement Surveillance Corrective Action Report No. 37-01, Open Item C; Effect of TLD Reaction Loads on Fuel Pool Wall; dated April 3, 2007 Surveillance Corrective Action Report No. 37-02, Open Item A; Justify Simplified TLD Model Results Are Conservative; dated May 10, 2007, including June 6, 2007 Supplement Attachment