IR 05000445/2011011

NRC Inspection Report 50-445/2011-11, 50-446/2011-11, and 72-74/2011-01

ML11297A030
Person / Time
Site:	Comanche Peak
Issue date:	10/21/2011
From:	Spitzberg D NRC/RGN-IV/DNMS/RSFSB
To:	Flores R Luminant Generation Co
References
IR-11-001, IR-11-011
Download: ML11297A030 (173)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ber 21, 2011

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 - NRC INSPECTION OF THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION - INSPECTION REPORT 05000445/2011011, 05000446/2011011, AND 07200074/2011001

Dear Mr. Flores:

A team inspection was conducted of your Independent Spent Fuel Storage Installation (ISFSI)

between May 2, 2011 and July 15, 2011. The purpose of the inspection was to inspect the pre-operational demonstration of your ISFSI required by your license, and other safety requirements associated with its future operations. Subsequent to July 15, 2011, several discussions were held concerning the date for your first planned loading campaign. Because of mechanical problems experienced with the vertical cask transporter, a decision was made by your organization to reschedule loading of the first canisters to early 2012. As your organization worked through the process of arriving at this decision, the NRC noted that safe operations continued to be the highest priority in all aspects of the decision.

On August 10, 2011, an exit briefing was conducted with your staff to review the findings of this inspection. The inspection determined that you had completed all required activities identified in the Holtec Certificate of Compliance for use of the Holtec HI-STORM 100 cask system at your site. No violations of NRC regulations were identified. A team of eight NRC inspectors, over a period of three months, reviewed a broad range of topical areas related to programs required to successfully move spent fuel from your spent fuel pool to dry cask storage at your newly constructed ISFSI storage pad. The inspection examined activities conducted under your license as they relate to public health and safety to confirm compliance with the Commissions rules and regulations, orders, and with the conditions of your license. Within these areas, the inspection consisted of an examination of selected procedures and representative records, observations of pre-operational training activities, and interviews with personnel. The pre-operational testing and training exercise required by License Condition #10 was successfully performed demonstrating that your procedures, programs, and training related to dry cask storage operations had been successfully integrated into your site operations.

Luminant Generation Company LLC -2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response if you choose to provide one, will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.htmI.To the extent possible, your response should not include any personal, privacy or proprietary information so that it can be made available to the public without redaction.

Should you have any questions concerning this inspection, please contact the undersigned at (817) 860-8191 or Mr. Vincent Everett at (817) 860-8198 Sinc~

,

D. Blair Spitzberg, Ph.D., Chief Repository & Spent Fuel Safety Branch Dockets: 50-445, 50-446, 72-74 Licenses: NPF-87, NPF-89 Enclosure:

Inspection Report Nos.:

0500445/2011011, 0500446/2011011 07200074/2011001 Attachment:

1. Supplemental Information 2. Inspector Notes

Luminant Generation Company LLC -3-Electronic distribution by RIV:

(Elmo.Collins@nrc.gov ) (Stephanie.Bush-Goddard@nrc.gov )

(Art.Howell@nrc.gov ) (Dale.Powers@nrc.gov)

(Roy, Caniano@nrc.gov) (Rob.Temps@nrc.gov)

(Vivian. Campbell@nrc.gov) (Vincent. Everett@nrc.gov)

(Blair. Spitzberg@nrc.gov) (Lee. Brookhart@nrc.gov)

(Kriss. Kennedy@nrc.gov ) (Clyde.Morell@nrc.gov)

(Troy.Pruett@nrc.gov ) (Jason.Dykert@nrc.gov)

(AntonVegel@nrc.gov ) (Abin. Fairbanks@nrc.gov)

(John.Kramer@nrc.gov ) (Rachel. Browder@nrc.gov)

(Brian.Tindell@nrc.gov ) (Gerald.Schlapper@nrc.gov)

(Wayne.Walker@nrc.gov) (Ashley. Riffle@nrc.gov)

(David.Proulx@nrc.gov ) (Lara. Uselding@nrc.gov )

(Jason.Dykert@nrc.gov ) (Balwant.Singal@nrc.gov )

(Sue.Sanner@nrc.gov ) (MichaeIHay@nrc.gov )

(Victor. Dricks@nrc.gov) (Marisa. Herrera@nrc.gov )

(MarkHenry.Salley@nrc.gov) (Karla.Fuller@nrc.gov )

(Chris.Staab@nrc.gov) (Jenny.Weil@nrc.gov )

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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 50-445, 50-446, 72-74 Licenses: NPF-87, NPF-89 Report Nos.: 05000445/2011011, 05000446/2011011, and 07200074/2011001 Licensee: Luminant Generation Company LLC Facility: Comanche Peak Nuclear Power Plant, Units 1 and 2 Independent Spent Fuel Storage Installation (ISFSI)

Location: FM-56 Glen Rose, Texas Dates: May 2 - 4, 2011, Welding Dry Run May 31 - June 2, 2011; MPC Fluid Operations Dry Run June 6 - 30, 2011, Program Review June 20 - 22, 2011, Transporter Operations to ISFSI Pad Dry Run June 29 - July 1, 2011, Stack-up Operations Dry Run July 5 - 8, 2011, Fuel Building Heavy Loads Dry Run July 15, 2011, Meetings on Vertical Cask Transporter Team Leader: Vincent Everett, Senior Inspector, RIV Repository and Spent Fuel Safety Branch Inspectors: Lee Brookhart, Health Physicist, RIV Rob Temps, Senior Inspector, NMSS Clyde Morell, Storage & Transport Safety Inspector, NMSS Gerald Schlapper, Health Physicist, RIV Rachel Browder, State Agreement Officer, RIV Abin Fairbanks, Reactor Inspector, RIV Jason Dykert, Project Engineer, RIV Accompanying Ashley Riffle, Project Manager, FSME/DWMEP Personnel:

Approved By: D. Blair Spitzberg, Ph.D., Branch Chief Repository and Spent Fuel Safety Branch Division of Nuclear Materials Safety

EXECUTIVE SUMMARY Comanche Peak Nuclear Generating Station NRC Inspection Report 50-445/2011-11, 50-446/2011-11, and 72-74/2011-01 The Comanche Peak Nuclear Power Plant had completed the pre-operational testing and training exercise requirements of the Holtec Certificate of Compliance #1014 for use of the Holtec HI-STORM 100 cask system at their sites Independent Spent Fuel Storage Installation (ISFSI). Comanche Peak had constructed an ISFSI pad to hold 84 HI-STORM 100S storage casks (overpacks) containing the MPC-32 multipurpose canister design. The ISFSI was licensed by the NRC under the general license provisions of 10 CFR Part 72. The licensee had originally planned to load three canisters for placement on the ISFSI pad in the summer 2011.

But due to mechanical delays associated with the vertical cask transporter used to move the loaded casks to the ISFSI, the loading campaign was rescheduled for early 2012. The 2012 loading campaign is scheduled to load twelve casks.

The inspections conducted by the NRC of Comanche Peaks dry cask storage project provided a comprehensive evaluation of the licensee=s compliance with the requirements in the Holtec Certificate of Compliance No. 72-1014 and Technical Specifications, Amendment 7; the Final Safety Analysis Report, Revision 9; the NRC=s Safety Evaluation Report, Amendment 7; and 10 CFR Part 72. The inspection consisted of a team of eight NRC inspectors performing inspections of various phases of activities over a period from May 2 through July 15, 2011.

Twenty three technical areas were reviewed during the inspections including such topical areas as crane design, crane operations, loading operations, fuel verification, radiological programs, quality assurance, heavy loads, welding, and others. During the inspections, the licensee conducted numerous demonstrations for NRC observation related to the operations of equipment and the implementation of procedures to verify that all operations required by the technical specifications could be performed safely. The program review conducted by the NRC concluded that the licensing requirements related to dry cask storage had been adequately incorporated into the sites programs and procedures. During the various pre-operational demonstrations, the Comanche Peak workers demonstrated a comprehensive knowledge of the technical requirements related to the loading and operations of an ISFSI.

Details related to the technical areas reviewed during this inspection are provided as Attachment 2 Comanche Peak Inspector Notes to this inspection report. The following provides a summary of the observations of this inspection.

Canister Drying/Inerting

! Forced helium dehydration dryness limits established in Technical Specification A.3.1.1.1 and Table 3-1 had been incorporated into the licensee=s procedures. The licensee planned to use the forced helium dehydration system for drying all canisters loaded at the site. Operation of the forced helium dehydration system was demonstrated during the pre-operational dry run exercises.

! The licensee had incorporated the use of a cask supplemental cooling system during forced helium dehydration as required by the technical specifications. This system had been previously used at other sites, where an analysis was performed to document the thermal validation of the cooling system. This analysis was submitted to the NRC in-2- Enclosure

September 2009 satisfying the requirements of License Condition #9 to Certificate of Compliance 1014.

! Helium backfill pressure requirements established in Technical Specification A.3.1.1.2 had been incorporated into the licensee=s procedures. High purity helium had been procured for backfilling the canister.

Crane Design

! The Ederer X-SAM crane used in the fuel handling building to lift the spent fuel casks had been accepted by the NRC in 1980 as a single failure proof crane. Specific aspects of the crane which included: the bridge and trolley brakes, main hoist safety devices, emergency stop features, crane two-block protection, and dual rope reeving system met the requirements of NUREG 0554 and NUREG 0612.

! The crane was designed to retain control of and hold the load during a design basis seismic event at the Comanche Peak site. Calculations demonstrated that the forces from a seismic event in the upward and horizontal directions would not exceed the strength of the seismic restraints on the crane.

! The wire rope breaking strength criteria used for the wire rope on the Ederer X-Sam crane had been evaluated by the NRC in the Ederer Generic Licensing Topical Report Safety Evaluation Report, Revision 3. In this 1983 report, the NRC accepted the use of yield strength limits as a conservative and alternative method to requirements listed in NUREG 0554 for assuring wire rope integrity.

Crane Inspection

! The 130 ton fuel building crane was subjected to a daily prior to use inspection that satisfied the requirements of ASME B30.2, Section 2-2.1.2 Frequent Inspection. On an annual basis the crane was subjected to a more rigorous inspection that met the requirements of ASME B30.2, Section 2-2.1.3 Periodic Inspection.

! The cranes hook was inspected annually as required by ASME B30.10, Sections 10-1.4.2 through 10-1.4.6. The cranes wire rope was inspected daily with a full length inspection annually as required by ASME B30.2, Sections 2-2.4.1 and 2-2.4.2, respectively.

Crane Load Testing

! The 130 ton fuel building crane was pre-operationally tested, as required by NUREG 0554, in April and June of 1983. The tests included a full performance test with 100% of the maximum critical load (130 tons) and a 125% static load test (162.5 ton). The fuel building cranes hook was subjected to a 200% hook load test of 260 tons in June of 1979.

! The fuel building crane had a minimum operating temperature of 40 degrees F, which was determined by a Charpy V-Notch impact test that met the requirements of NUREG 0554 and NUREG 0612.

-3- Enclosure

Crane Operation

! The maximum weight the 130 ton fuel building crane would lift during the cask loading campaign was 127.9 tons when lifting the HI-TRAC transfer cask containing the MPC-32 canister loaded with spent fuel out of the spent fuel pool.

! Pre-use inspection and testing of the crane and special lifting devices had been incorporated into procedures. Prior to each shift, a crane operator tested the upper limit switch as required by ASME B30.2, Section 2-3.2.4. The crane was verified to have sufficient wraps of rope around the hoist drum when the cask was located at its lowest point in the fuel handling building.

! Records related to the crane operators assigned to the first loading campaign were reviewed. The crane operators met the physical and experience qualification requirements of ASME B30.2, Section 2-3.2.4.

Dry Run Demonstration

! The licensee successfully completed all the required pre-operational tests specified by License Condition #10 of the Certificate of Compliance. This included welding, drying, and backfilling of a canister and the unloading of a sealed canister. A weighted canister was used to demonstrate heavy load activities inside the fuel handling building, transport between the fuel handling building and the ISFSI, and movement back into the fuel handling building for unloading purposes.

Emergency Planning

! Emergency planning provisions for the ISFSI had been incorporated into the site-wide emergency plan. This included adding a specific emergency action level for an event damaging a loaded cask. Part 50 emergency action levels applicable to the ISFSI included fires, security threats, and events involving a radiological release from a canister.

! The ISFSI design provided for accessibility by emergency personnel and vehicles. The licensee maintained provisions for offsite emergency organization support, which included fire protection, security, and medical support.

Fire Protection

! A Fire Hazards Analysis had been performed specific to the Comanche Peak ISFSI.

Administrative controls were established to limit the quantity of combustible and flammable liquids around the ISFSI and near the transport path during movement of the canister.

! Site specific fire and explosion hazards had been evaluated to determine the effect on the ISFSI and to confirm that the location of the ISFSI, location of the transport route, and the design of the vertical cask transporter were adequate. Twenty two nearby facilities were evaluated that included a 4,000 gallon gasoline tank, 2,000 gallon waste oil tank, 8,000 gallon diesel storage tank, hydrogen bulk storage tanks, and several transformers.

-4- Enclosure

! Additional fire analysis was required for the use of the wheeled vertical cask transporter to account for the fire loading of the tires and the hydraulic fluids. Holtec provided calculations that showed the postulated fire involving the transporter would not result in a significant increase in the temperature of the spent fuel inside a loaded cask being transported.

Fuel Selection/Verification

! For the initial loading campaign, the licensee planned to load intact fuel assemblies that met the requirements of Technical Specification Appendix B, Section 2.1.1, Section 2.4, and the associated tables. The fuel assemblies selected met the limits for length, width, weight, irradiation cooling time, average burn-up, cladding, decay heat, and fuel enrichment.

! The licensee planned to load fuel in the canisters using the regionalized fuel loading concept allowed in Technical Specification Appendix B, Section 2.1.3 and Section 2.4.2.

For the initial loading campaign, the licensee selected the option to load cooler spent fuel into the outer canister locations to provide shielding to the hotter fuel assemblies that were placed in the inner locations of the canister.

! The licensee had developed their own computer code, TARPIT (Thermal Assembly Repository Pad Inventory Tracker), for calculating decay heat values and calculating other limits for restricting spent fuel for dry cask loading. The fuel assembly burn-up values were limited by Comanche Peak to 50,000 megawatt-days (MWD) per metric ton uranium (MTU). Technical Specification Appendix B, Section 2.4.3.4 allowed up to 65,000 MWD/MTU burn-up values. The cask decay heat values for the first three casks planned for loading were 16.07 kW, 21.39 kW, and 23.34 kW. The maximum allowed cask decay heat level specified per the technical specifications was 34 kW.

! The licensee had established provisions for independent verification of the correct loading of spent fuel assemblies into the canister. This included videotaping the location of the spent fuel after loading was complete.

! The licensee had incorporated the requirements to utilize fuel spacers to ensure that the active fuel region of the assemblies remained within the neutron poison region of the canister basket while leaving sufficient room for fuel assembly growth between the fuel and the canister closure lid. Upper and lower spacers were required to be installed to keep the fuel assemblies properly aligned during worst case scenarios.

General License Requirements

! The licensee evaluated the bounding environmental conditions specified in the Holtec Final Safety Analysis Report and Certificate of Compliance No. 1014 Technical Specifications against the conditions at the site. This included: tornados/high winds, flood, seismic events, tsunamis, hurricanes, lightning, burial of the ISFSI under debris, snow, normal and abnormal temperatures, collapse of nearby facilities, and fires/explosions. The site environmental conditions at Comanche Peak were bounded by the Holtec cask design parameters except for tornado driven missiles. A separate analysis showed that the Holtec casks could withstand Comanche Peaks site specific tornado driven missile.

-5- Enclosure

! Projected radiation levels at the ISFSI were calculated for an assumed individual located at the owner controlled area boundary to determine the dose to this individual. The analysis assumed that the ISFSI was fully loaded with all 84 casks with fuel characteristics that bounded the fuel currently stored at the sites spent fuel pool. The calculation concluded that the individual would have to stand at the owner control boundary for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, 365 days to receive 1 mrem of dose from the 84 loaded casks on the ISFSI pad. Doses at the controlled area boundary were projected to be a total of 11 mrem/year, when including very conservative assumptions related to the operations from the nuclear plant. The calculated doses were below the 10 CFR 72.104 limit of 25 mrem/year.

! The licensee performed an evaluation of the Part 50 reactor programs that could be impacted by the addition of an ISFSI. The evaluation included the radiation protection program, emergency planning program, quality assurance program, training program, reactor technical specifications, and the Part 50 license. Revisions to the programs to incorporate the ISFSI were identified and implemented. None of the changes required an amendment to the plants Part 50 operating license or technical specifications.

! The Holtec Certificate of Compliance and Final Safety Analysis Report had been reviewed by the licensee to verify that the design basis for the Holtec system and the conditions and requirements in the Certificate of Compliance and Final Safety Analysis Report were met.

! Comanche Peak had developed procedures for controlling all work associated with cask handling, loading, movement, surveillance, maintenance, and testing. Procedures had been developed specific to the ISFSI activities. Numerous other procedures developed for the Part 50 reactor programs were being adequately applied to the ISFSI program.

Heavy Loads

! The licensee had incorporated the special requirements related to the ISFSI project into the site heavy loads programs and procedures. Crane operators interviewed were knowledgeable of the special handling requirements related to the spent fuel casks.

! Special lifting device height limits and temperature restrictions during movement of the casks had been incorporated into the licensee=s procedures consistent with the requirements in the Certificate of Compliance.

! A safe loads path had been identified and analyzed for moving the spent fuel from the spent fuel pool. Provisions were established in procedures to prevent the crane from moving the loaded cask outside the boundaries of the safe load path while in the fuel building.

! The adequacy of the vertical cask transporter for the expected weight of a loaded cask and the ability of the transporter to safely secure and move the cask to the ISFSI was evaluated by transporting a cask loaded with a concrete filled canister back and forth between the ISFSI and the plant. The design of the wheel hubs on the vertical cask transporter was found to be inadequate for the slopes associated with the Comanche Peak haul path. During testing of the vertical cask transporter, a wheel hub failed causing the wheel to become disconnected from the transporter. As a result, the cask-6- Enclosure

loading campaign was delayed until 2012 to allow for design modifications and repairs of the transporter wheel hubs.

! The licensee=s heavy loads procedural requirements related to the testing and inspection of the transfer cask trunnions were verified against industry standards and manufacturer requirements for load tests, safety margins, and inspection/maintenance.

Loading Operations

! Requirements in the Final Safety Analysis Report related to pre-operational inspections and annual maintenance of equipment were being implemented through the licensees procedures.

! Technical specifications and Final Safety Analysis Report requirements related to spent fuel boron concentration, fuel cladding not being exposed to air, handling of damaged fuel containers, and time-to-boil limits were implemented in the licensees procedures.

Non-Destructive Examination

! The requirements for helium leak testing of a canister were incorporated into the licensees procedures. The helium leak testing equipment used during the dry run demonstration was verified to meet the minimum sensitivity level specified in ANSI N14.5. A review of the helium leak testing specialists qualifications identified that he was properly qualified as a Level III examiner.

! A review of the visual and liquid penetrant examination specialists qualifications identified that he was properly qualified as a Level III examiner.

! Holtecs visual and liquid penetrant examination procedures implemented all the applicable requirements from ASME Section III, Section IV, and the Final Safety Analysis Report in regards to non-destructive examination of welds.

Pressure Testing

! The requirements for canister hydrostatic testing had been incorporated into the licensees procedure and were consistent with the requirements of ASME Section III Subsection NB, Article NB-6000.

! The hydrostatic testing sequence and criteria described in the Final Safety Analysis Report had been incorporated into the licensees procedures.

Quality Assurance

! The licensee had implemented their approved reactor facility Part 50 quality assurance program for the activities associated with the ISFSI. Selected QA activities were reviewed related to calibrations, operating status, and receipt inspections.

! The licensee had identified the components that were important to safety using a graded approach. The licensees list was consistent with Holtecs identification of important to safety equipment listed in the Final Safety Analysis Report.

-7- Enclosure

! A corrective action program that documented issues and classified problems according to their impact on quality and safety was being effectively used by the licensee.

Selected condition reports were reviewed to verify adequate resolution of the issues.

! The licensee=s quality assurance organization had implemented a comprehensive quality assurance audit and inspection program for the ISFSI. The licensee had conducted audits and surveillances of the cask vendor, which included engineering design activities and cask manufacturing quality controls.

Radiation Protection

! The licensee had incorporated As Low As Reasonably Achievable (ALARA) planning into the cask loading program. This included developing reasonable dose goals, obtaining review and concurrence of procedures from the stations ALARA committee, requiring ALARA training for individuals, and conducting radiation pre-job briefings that identify expected radiological conditions for the different work evolutions.

! Requirements for radiological and contamination surveys described in the Final Safety Analysis Report and technical specifications had been incorporated into the licensee=s health physics program for the loading of the casks. This included decontamination of the transfer cask and canister lid prior to welding, performing required surveys of the transfer cask and storage cask, and establishing site-specific dose rate limits for the top and sides to the storage casks.

! The licensee incorporated proper neutron dose consideration into the health physics monitoring program for higher energy neutrons that would be present around the canister when empty of water. This consideration included the use of appropriate personnel dosimetry that could measure neutron doses and applying a correction factor based on survey readings during the initial loading.

Records

! The licensee was maintaining the ISFSI records in their quality related records system.

Records required for retention by 10 CFR 72.174, 10 CFR 72.212, 10 CFR 72.234, and the Final Safety Analysis Report had been identified in the licensee=s program as required records for retention.

Safety Reviews

! Changes to the site related to the construction and operation of the ISFSI were being evaluated in accordance with 10 CFR 72.48 and 10 CFR 50.59 requirements. No issues were identified during the review of selected safety screenings.

Slings

! The slings for downloading a canister met the requirements of NUREG 0612.

Operations required dual/redundant slings that had a rated capacity of twice the sum of the static and dynamic loads.

-8- Enclosure

! The licensees sling inspection program complied with ASME B30.9 in regards to daily sling inspections, annual sling inspections, and proof loading.

Special Lifting Devices

! The licensees special lifting device program complied with ANSI N14.6 in regards to stress design, daily inspections, annual inspections, and 300% proof loadings for the lift yoke, lift yoke extensions, and HI-STORM brackets.

Storage Operations

! The licensee established the cask spacing criteria for positioning the casks on the ISFSI pad which met the thermal criteria specified in the Final Safety Analysis Report.

! Inspection of the HI-STORM 100 storage cask air vents or verification that the temperature monitoring system was operational and the temperature limits were not exceeded was identified in the appropriate procedures to be performed daily as required by Technical Specification A.3.1.2.

Unloading

! The licensee procured the equipment and developed procedures to perform gas sampling if a canister was required to be unloaded. The licensee demonstrated the gas sampling process to the NRC during the dry run demonstrations.

! Canister re-flooding for unloading was demonstrated to the NRC during the dry run demonstrations. The procedure controlling canister re-flooding contained all of the applicable requirements from the Final Safety Analysis Report and the technical specifications.

Welding

! Requirements for hydrogen monitoring during welding of the inner cask lid had been incorporated into the procedures. The licensee stopped welding if levels reached 50 percent of the lower explosive limit for hydrogen.

! All welding procedures contained the required essential, non-essential, and supplemental variables specified in ASME Section IX for gas tungsten arc welding.

! The welding procedures qualification test coupons all satisfactorily passed the required bend and tension tests to qualify the welding procedures and thus the lid to shell weld.

! All weld grinding, machining, and repairs were controlled through approved procedures which contained the applicable requirements of ASME Section III.

! The welders performance qualification test records were reviewed and documented that the welders had met the qualification testing requirements for manual and machine welding of the canister lid. The testing requirements complied with the requirements of ASME Section IX.

-9- Enclosure

SUPPLEMENTAL INSPECTION INFORMATION PARTIAL LIST OF PERSONES CONTACTED Licensee Personnel S. Bernhoft, Project Engineering Manager C. Davis, Radiation Protection R. Fishencord, Emergency Planner D. Fuller, Emergency Planning Manager R. Garcia, Radiation Protection Supervisor B. Henley, Project Manger J. Hull, Fuel Handling Supervisor K. Kilgarif, Radiation Protection C. Lemons, Core Performance Engineer C. Montgomery, Project Engineering Manager L. Neuburger, Summer Intern T. Norman, Radiation Protection D. OConner, Health Physics Supervisor J. Seawright, Regulatory Affairs T. Smith, Maintenance Supervisor J. Simmons, Senior Quality Assurance Auditor Holtec International J. Ciesielski, Welder T. Ciesielski, Welder K. Eggar, Welder J. Fosdick, Construction Manager T. Hagner, Welder J. McMahon, Welding Supervisor M. Mangan, Non-Destructive Examination T. Morin, Holtec Licensing Manager M. Ragan, Cask Loading Supervisor J. Sloane, Field Supervisor S. Soler, Site Services Manager Shaw Group D. Brown, Project Manager C. Hill, Records Management Supervisor D. Keating, Procedure Writer R. Plunkett, Fire Protection Program Engineer Leak Test Specialists D. Hecksel, Level III Examiner Attachment

INSPECTION PROCEDURES USED IP 60854.1 Preoperational Testing of ISFSIs at Operating Plants IP 60856 Review of 10 CFR 72.212(b) Evaluations IP 60857 Review of 10 CFR 72.48 Evaluations LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Discussed None Closed None LIST OF ACRONYMS abs absolute AC alternating current ALARA As Low As Reasonably Achievable ANSI American National Standards Institute ASME American Society of Mechanical Engineers ASPHALT Assembly Selection Planner for Heat and Associated Limitations Tracking ASTM American Society for Testing and Materials atm atmosphere AWS automatic welding system B&W Babcock & Wilcox Btu/hr British thermal unit per hour cc/sec cubic centimeters per sec CFR Code of Federal Regulations cm centimeter CMAA Crane Manufacturers Association of America CMTR certified materials test report CoC Certificate of Compliance CPNPP Comanche Peak Nuclear Power Plant CPSES Comanche Peak Steam Electric Station DCS dry cask storage dpm disintegrations per minute EAB exclusion area boundary EAL emergency action level EATL energy absorbing torque limiter EPD electronic pocket dosimeter EPP emergency plan procedure F Fahrenheit-2- Attachment

FHD forced helium dehydration FSAR Final Safety Analysis Report GE General Electric gpm gallons per minute GTAW gas tungsten arc welding GWD/MTU Giga Watt Day per Metric Ton Uranium HMSLD helium mass spectrometer leak detector ICA item control area ISFSI Independent Spent Fuel Storage Installation ITS important to safety IWRC independent wire rope core kg kilogram kW killo-watt LBD licensing basis document LCO limiting condition for operation LPT low profile transporter m/sec meters per second MMH Morris Material Handling MPC multi-purpose canister mrem MilliRoentgen Equivalent Man NDE non-destructive examination NRC Nuclear Regulatory Commission NIST National Institute of Science and Technology NITS not important to safety NUPIC Nuclear Utility Procurement Issues Committee OCA owner controlled area OSL optically stimulated luminescence PM preventative maintenance PQR procedure qualification record psig pounds per square inch gauge PT liquid penetrant exam PWR pressurized water reactor QA quality assurance QC quality control REMP Radiological Environmental Monitoring Program RVOA removable valve operator assembly RWP radiation work permit SCS supplemental cooling system SER Safety Evaluation Report SNM special nuclear material SPARCS station process & records control system SSE safe shutdown earthquake TARPIT Thermal Assembly Repository Pad Inventory Tracker TLD thermo-luminescent dosimetry TS technical specification TVA Tennessee Valley Authority U-235 Uranium 235 UFSAR Updated Final Safety Analysis Report VCT vertical cask transporter WCP wet cask pit-3- Attachment

WOPQ welder operator performance qualification WPQ welder performance qualification WPS welding procedure specification-4- Attachment

ATTACHMENT 2 COMANCHE PEAK INSPECTOR NOTES Category Topic Page #

Canister Drying/Inerting Dryness Levels 1 Canister Drying/Inerting Forced Helium Dehydration System (FHD) 1 Canister Drying/Inerting Helium Backfill Pressure 2 Canister Drying/Inerting Helium Purity 2 Canister Drying/Inerting Supplemental Cooling System 3 Crane Design Bridge and Trolley Brakes 3 Crane Design Drum Safety Devices 4 Crane Design Emergency Stop Feature 5 Crane Design Seismic Events During Cask Movement 5 Crane Design Seismically Induced Load Swing 6 Crane Design Single Failure Proof Crane 6 Crane Design Two-Block Protection 7 Crane Design Wire Rope Breaking Strength 8 Crane Design Wire Rope Configuration 10 Crane Inspection Crane Inspection - Frequent 11 Crane Inspection Crane Inspection - Periodic 12 Crane Inspection Crane Operational Testing 13 Crane Inspection Hoist Overload Testing 13 Crane Inspection Hoist Two-Block Testing - Limit Device Method 14 Crane Inspection Hook Inspections - Frequent/Periodic 15 Crane Inspection Wire Rope Inspection - Frequent 15 Crane Inspection Wire Rope Replacement Criteria 16 Crane Load Testing Cold Proof Testing 16 Crane Load Testing Dynamic Load Testing (100%) 17 Crane Load Testing Hook Load Testing 17 Crane Load Testing NDE Exams Following Cold-Proof Testing 18 Crane Load Testing Rated Load Marking 18 Page 1 of 7

Category Topic Page #

Crane Load Testing Static Load Testing (125%) 19 Crane Operation Brake Test Prior to Lift 19 Crane Operation Height Limit During Cask Movement 19 Crane Operation Hoist Limit Switch Tested Each Shift 20 Crane Operation Minimum of Two Wraps of Rope 21 Crane Operation Qualification For Crane Operator 21 Crane Operations Maximum Weight of Canister 22 Crane Operations Provisions For Manual Operation 22 Dry Run Demonstration Fuel Loading and Verification Demonstration 23 Dry Run Demonstration MPC Pressure Test, Drying, and Helium Backfill 25 Dry Run Demonstration MPC Removal from Spent Fuel Pool 25 Dry Run Demonstration MPC Transfer to HI-STORM 27 Dry Run Demonstration MPC Welding and NDE 27 Dry Run Demonstration Placement of HI-STORM on ISFSI 28 Dry Run Demonstration Placement of MPC in Spent Fuel Pool 28 Dry Run Demonstration Unloading a Canister 29 Emergency Planning Emergency Drills 30 Emergency Planning ISFSI Emergency Plan 30 Emergency Planning Offsite Emergency Support 31 Fire Protection Fire Accident Response 31 Fire Protection Fire and Explosion Hazards Analysis 32 Fire Protection Fire Protection Plan 36 Fuel Selection/Verification Authorized Contents For Storage 37 Fuel Selection/Verification Damaged Fuel Classification 39 Fuel Selection/Verification Decay Heat, Burnup & Cooling Time Limits 39 Fuel Selection/Verification Fuel Loading Error 43 Fuel Selection/Verification Fuel Spacers 43 Fuel Selection/Verification Material Balance, Inventory, and Records 44 Fuel Selection/Verification Post Loading Verification 45 General License Evaluation of Effluents/Direct Radiation 45 Page 2 of 7

Category Topic Page #

General License Flood Conditions 46 General License Initial Compliance Evaluation Against CoC 47 General License Initial Compliance Evaluation Against FSAR 49 General License Initial Evaluation Against Part 50 License 55 General License Limiting Site Temperatures 56 General License Program Review - RP, EP, QA, and Training 56 General License Revisions to 72.212 Analysis 57 General License Storage Cask Blocked Inlet or Outlet Air Vents 58 General License Written Procedures Required 59 Heavy Loads Component Weights for Heavy Lifts 61 Heavy Loads Licensed Facility Heavy Loads Requirements 61 Heavy Loads Procedures 62 Heavy Loads Safe Load Paths 63 Heavy Loads Site Temperature Limit for Cask Handling 63 Heavy Loads Storage Cask Maximum Lifting Height 63 Heavy Loads Transport Route Surface 65 Heavy Loads Transporter Wheel Failure 65 Heavy Loads Trunnion Annual Testing 69 Heavy Loads Trunnion Initial Load Testing 70 Loading Operations Canister Lid Fit Test 71 Loading Operations Cask System Annual Maintenance 71 Loading Operations Cask System Inspections Prior to Use 73 Loading Operations Fuel Cladding Not Exposed to Air 74 Loading Operations Handling Damaged Fuel Containers 75 Loading Operations Pressure Relief Valves 75 Loading Operations Spent Fuel Pool Boron Concentration 76 Loading Operations Time-to-Boil Time Limits 77 NDE-Helium Leak Testing Helium Leak Test-Vent/Drain Covers 78 NDE-Helium Leak Testing HMSLD Minimum Sensitivity 79 NDE-Liquid Penetrant Acceptance Criteria 79 Page 3 of 7

Category Topic Page #

NDE-Liquid Penetrant Contaminants 80 NDE-Liquid Penetrant Final Interpretation 80 NDE-Liquid Penetrant Lid-To-Shell Weld PT 81 NDE-Liquid Penetrant Liquid Penetrant Testing - Permanent Record 82 NDE-Liquid Penetrant Minimum Elements 82 NDE-Liquid Penetrant Removing Excess Penetrant 83 NDE-Liquid Penetrant Surface Preparation 83 NDE-Personnel Qualification Certification Records 84 NDE-Personnel Qualification Level III Candidates 84 NDE-Personnel Qualification Level III Exam Grading 85 NDE-Personnel Qualification Recertification of Personnel 85 NDE-Personnel Qualification Visual Acuity 86 NDE-Personnel Qualification Written Practice 86 NDE-Visual Examination Acceptance Criteria - Arc Strikes 87 NDE-Visual Examination Acceptance Criteria - Cracks 87 NDE-Visual Examination Acceptance Criteria - Craters 87 NDE-Visual Examination Acceptance Criteria - Fusion 88 NDE-Visual Examination Acceptance Criteria - Lengths 88 NDE-Visual Examination Acceptance Criteria - Overlap 88 NDE-Visual Examination Acceptance Criteria - Porosity 89 NDE-Visual Examination Acceptance Criteria - Slag 89 NDE-Visual Examination Acceptance Criteria - Thickness 89 NDE-Visual Examination Acceptance Criteria - Undercut 90 NDE-Visual Examination Eye Position and Lighting 90 NDE-Visual Examination Minimum Elements 91 NDE-Visual Examination Procedure Validation 91 Pressure Testing Governing Code 92 Pressure Testing Hydrostatic Testing Sequence 92 Pressure Testing Pressure Gauge Calibration 93 Pressure Testing Pressure Gauge Installation 93 Page 4 of 7

Category Topic Page #

Pressure Testing Pressure Gauge Ranges 94 Pressure Testing Thermal Expansion 94 Quality Assurance Approved QA Program 94 Quality Assurance Corrective Actions 95 Quality Assurance Important to Safety Components - Ancillaries 97 Quality Assurance Important to Safety Components - Cask System 98 Quality Assurance Instruments Requiring Calibration 98 Quality Assurance Operating Status 99 Quality Assurance QA Audits 99 Quality Assurance Receipt Inspection Checklists 102 Radiation Protection ALARA Program 102 Radiation Protection Contamination Control 104 Radiation Protection Controlled Area Boundary Dose Rate Analysis 105 Radiation Protection Controlled Area Radiological Doses 106 Radiation Protection Dose Rate Survey - Storage Cask Air Vents 107 Radiation Protection Dose Rate Survey - Storage Cask Side 108 Radiation Protection Dose Rate Survey - Storage Cask Top 109 Radiation Protection Dose Rate Survey - Transfer Cask 110 Radiation Protection Neutron Dosimetry 111 Radiation Protection Shielding Effectiveness Test 113 Radiation Protection Site-Specific Dose Rate Limits - Storage Cask 114 Radiation Protection Temporary Shielding 115 Radiation Protection Transfer Cask Surface Contamination Limit 116 Records Cask Records 117 Records Maintaining a Copy of the CoC and Documents 118 Records Notice of Initial Loading 118 Records Record Retention for 72.212 Analysis 119 Records Registration of Casks with NRC 119 Safety Reviews Changes, Tests, and Experiments 119 Slings Sling Heavy Load Requirements 121 Page 5 of 7

Category Topic Page #

Slings Sling Identification 122 Slings Sling Inspections - Frequent 122 Slings Sling Inspections - Periodic 123 Slings Sling Load Rating 123 Slings Sling Proof Loading 124 Slings Sling Temperature Limits 124 Slings Synthetic Round Sling Removal from Service 124 Slings Synthetic Webbing Sling Removal From Service 125 Slings Wire Rope Sling Removal From Service 125 Special Lifting Device Transporter Annual Testing 126 Special Lifting Device Transporter Initial Acceptance Testing 127 Special Lifting Device Transporter Inspection - Quarterly 128 Special Lifting Device Transporter Lift Bracket Inspection Prior to Use 128 Special Lifting Device Yoke Annual Testing 129 Special Lifting Device Yoke Initial Acceptance Testing 129 Special Lifting Device Yoke Inspection Prior to Use 130 Special Lifting Device Yoke Records of Annual Testing 130 Special Lifting Device Yoke Stress Design -Dual-Load-Path 131 Storage Operations Cask Spacing 131 Storage Operations Heat Transfer Validation Test 132 Storage Operations Overpack Vent Screen Inspections 132 Storage Operations Storage Cask Temperature Monitoring 133 Unloading Operations Canister Gas Sampling 133 Unloading Operations Canister Reflooding 134 Unloading Operations Canister Shell Cooling - High Burnup Fuel 135 Unloading Operations Cavity Reflooding 136 Unloading Operations Hydrogen Monitoring 137 Welding Combustible Gas Monitoring 137 Welding GTAW Essential Variables 138 Welding GTAW Non Essential Variables (1-12) 138 Page 6 of 7

Category Topic Page #

Welding GTAW Non Essential Variables (13-26) 138 Welding GTAW Supplementary Essential Variables 139 Welding Material Specifications 139 Welding Minimum Delta Ferrite Content 140 Welding NDE Inspection Documentation 140 Welding Procedure Qualification Record (PQR) 141 Welding Procedure Qualification Tests 142 Welding Tack Welds 143 Welding Vent and Drain Port Cover Plate Weld PT 144 Welding Weld Grinding and Machining 144 Welding Weld Repairs - Base Metal Defects 145 Welding Weld Repairs - Surface Defects 145 Welding Weld Types for Canister Lid 145 Welding Welder Performance Qualification Test 146 Welding Welding Operator Performance Qualification 147 Welding Welding Procedure Specification (WPS) 148 Page 7 of 7

Intentionally Left Blank COMANCHE PEAK INSPECTOR NOTES Category: Canister Drying/Inerting Topic: Dryness Levels Reference: CoC 1014, Tech Spec A.3.1.1.1 and Table 3-1 Amendment 7 Requirement: When using the vacuum drying process for moisture removal, canister cavity pressure must be 3 torr or less for 30 minutes or more. When using the forced helium dehydration (FHD) system for moisture removal, the gas temperature exiting the demoisturizer shall be 21 degrees F or less, for 30 minutes or more, or the gas dew point exiting the canister shall be 22.9 degrees F or less, for 30 minutes or more.

Observation: The forced helium dehydration (FHD) system was used at Comanche Peak to achieve the dryness level required by Technical Specification A.3.1.1.1 and Table 3-1. The operation of the system was described in Procedure DCS-204, Sections 8.7 through 8.13.

The system operation was demonstrated as part of the dry run activities the week of May 31, 2011. Steps 8.11.8 through 8.11.11 of the procedure required verification that the helium temperature exiting the freeze dryer was less than or equal to 16 degrees F for greater than or equal to 30 minutes. Procedure DCS-204 used a lower temperature value of 16 degrees F than the required 21 degrees F in Technical Specification A.3.1.1.1 to address equipment uncertainty. The technical specification acceptance requirement was also stated in Step 2.1.2 and the preceding note of Section 2.0 Acceptance Criteria.

Final verification that the technical specification limit had been met required sign-off by the cask loading supervisor and the dry cask storage project manager.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 Category: Canister Drying/Inerting Topic: Forced Helium Dehydration System (FHD)

Reference: CoC 1014, Appendix B, Section 3.6.1 Amendment 7 Requirement: For canisters containing one or more fuel assemblies with burnup values greater than 45 GWD/MTU, forced helium dehydration must be used for canister drying. For all other canisters, either forced helium dehydration or vacuum drying may be used for canister drying, unless canister heat load is greater than 26 kW for MPC-32, then only forced helium dehydration shall be used.

Observation: Use of the forced helium dehydration system was planned for all three MPC-32 canisters scheduled for loading in the first campaign. The licensee demonstrated the use of the forced helium dehydration system during the dry run the week of May 31, 2011.

Procedure DSC-204, Sections 8.7 through 8.13 provided instructions for use of the system. Use of the forced helium dehydration system was not optional in Procedure DCS-204, and as such, the criteria in Technical Specification B.3.6.1 was not specified in the procedure. The heat load of the first three casks was 16.07 kW, 21.39 kW, and 23.338 kW.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) NUC-212-4 "Cask Acceptability Report for Cask #1," provided to the NRC August 10, 2011 (c) NUC-212-4 "Cask Acceptability Report for Cask #2,"

provided to the NRC August 10, 2011 (d) NUC-212-4 "Cask Acceptability Report for Page 1 of 149

Cask #3," provided to the NRC August 10, 2011 Category: Canister Drying/Inerting Topic: Helium Backfill Pressure Reference: CoC 1014, Tech Spec A.3.1.1.2 and Table 3-2 Amendment 7 Requirement: For the MPC-32 canister, backfill pressure shall be as follows: a) for cask heat loads less than or equal to 28.74 kW, helium backfill shall be equal to or greater than 29.3 psig up to 48.5 psig; b) for cask heat loads greater than 28.74 kW, helium backfill shall be equal to or greater than 45.5 psig up to 48.5 psig. The pressure range is at a reference temperature of 70 degree F.

Observation: Helium backfill pressure requirements were incorporated into Procedure DCS-204 consistent with the requirements in Technical Specification A.3.1.1.2 and Table 3-2.

Procedure DCS-204, Section 8.12 "Forced Helium Dehydration System Helium Backfill Operations," provided instructions for the helium backfill operations. Step 2.1.3 listed the required helium backfill pressures based on whether the cask heat load was above the 28.74 kW value or below it. The heat load requirements specified in the procedure were identical to the values in Table 3-2. As a prerequisite for loading a cask, Procedure DCS-204, Step 6.17 required certain information to be collected from Report NUC-212-4

"Cask Acceptability Report" generated by the core performance engineer in accordance with Procedure NUC-212. One of the values generated and documented on the cask acceptability report was the lower and upper helium backfill limits based on the Technical Specification A.3.1.1.2 criteria and the actual heat load for the canister planned for loading. Step 8.2.11of Procedure DCS-204 used Attachment 10.2.15 "FHD Helium Backfill Pressure Chart" to determine the acceptable pressure range based on a 70 degree F reference temperature. Step 8.12.15 documented the final helium backfill pressure for the loaded canister. If the pressure requirement was not met, Step 8.12.18 directed the user to return to the beginning of Section 8.12 and start the process over.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) Procedure NUC-212 "Spent Fuel Limits for Dry Cask Storage Operations," Revision 1 Category: Canister Drying/Inerting Topic: Helium Purity Reference: CoC 1014, Tech Spec A.3.1.1.2, Table 3-2 footnote Amendment 7 Requirement: Helium used for backfilling the canister shall have a purity of 99.995% or higher.

Observation: The licensee had purchased 99.995% purity helium to backfill the canisters loaded at Comanche Peak. The licensee's Purchase Order #S 0707970 6D6 was for the purchase of three six-pack helium assemblies at 99.995% purity. The lab test analysis from Matheson Tri-Gas showed the helium assemblies were tested to be 99.999% helium.

Procedure DCS-204, Step 8.7.2 required the helium supply connected into the forced helium dehydration system to be at least 99.995% grade helium. This was the helium that was used for both the helium drying process and the helium backfill of the canister.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) Purchase Order #S 0707970 6D6, dated May 5, 2011 (c) Matheson Tri Gas "Certificate of Batch Analysis," dated May 11, 2011 Page 2 of 149

Category: Canister Drying/Inerting Topic: Supplemental Cooling System Reference: CoC 1014, License Condition 9 Amendment 7 Requirement: Each first time user of a cask supplemental cooling system (SCS), which has not been previously tested and documented with the NRC, shall measure and record coolant temperatures for the inlet and outlet of cooling provided to the annulus between the HI-TRAC transfer cask and canister and the coolant flow rate. The user shall also record the canister operating pressure and decay heat. An analysis shall be performed, using this information that validates the thermal methods described in the FSAR which were used to determine the type and amount of supplemental cooling necessary. Letter reports summarizing the results of each thermal validation test and the supplemental cooling system validation test and analysis shall be submitted to the NRC in accordance with 10 CFR 72.4. Cask users may satisfy these requirements by referencing validation test reports submitted to the NRC by other cask users.

Observation: The supplemental cooling system planned for use at Comanche Peak was the same system that had been used and previously tested at the Arkansas Nuclear One station.

Entergy Operations, Inc. had sent a letter to the NRC in September of 2009 documenting the results of the supplemental cooling system validation test and analysis conducted at the Arkansas Nuclear One site on August 5, 2005. The test was performed on a canister with a heat load of 19.992 kW and was required because at least one fuel assembly placed in the canister was a high burn-up assembly. The results of the test confirmed the thermal analysis used in the Holtec Final Safety Analysis Report for the supplemental cooling system and concluded that peak fuel cladding temperature was 437 degree F, well below the 752 degree F limit in Paragraph 3.7.2.5 of Appendix B of Certificate of Compliance 1014, Amendment 2.

Documents (a) Letter (OCAN090902) from D.B. Brice, Entergy Operations, Inc. to USNRC Reviewed: Document Control Desk entitled, "HI-STORM-100 Cask System Supplemental Cooling System (SCS) Validation Test - Arkansas Nuclear One Units 1 and 2," dated September 29, 2009 [NRC ADAMS Accession No. ML092810250] (b) Holtec Report HI-2094415

"Validation of First Use of Supplemental Cooling System at Arkansas Nuclear One,"

Revision 0 [proprietary-not publically available]

Category: Crane Design Topic: Bridge and Trolley Brakes Reference: NUREG 0554, Section 5.1 Published May 1979 Requirement: Bridge and trolley control and holding brakes should be: a) rated at 100% of maximum drive torque that can be developed at the point of application; b) adjusted with one brake in each system leading the other; and c) automatically actuate on interruption of power and overspeed. The holding brakes should be designed so that they cannot be used as foot-operated slowdown brakes. Drag brakes should not be used.

Observation: The holding brakes on the bridge and trolley were rated at a torque value exceeding the maximum that could be developed at the point of application, automatically actuated on interruption of power, and did not require overspeed protection based on their design.

The holding brakes were designed so they could not be used as foot operated slowdown brakes. The brakes on both the bridge and trolley were General Electric (GE) Model IC9516-461 electric brakes. GE Instruction IC9516 listed on page 5 a rated torque value Page 3 of 149

for these brakes of 25-35 foot-pound. Gibbs and Hill Calculation Job No. F-1115 determined that the maximum torque that would develop on either the bridge and trolley brakes would be 22.83 foot-pounds. This was based on a maximum 130 ton load on the crane and included a trolley dead load of 52 tons and a bridge dead load of 116.75 tons.

Comanche Peak Document TXX-3659, Table C-1, sheet 4 of 8, stated that overspeed sensors were not provided for the trolley and bridge drive brakes because the AC motors used to move the bridge and trolley inherently cannot overspeed. The bridge and trolley brakes would engage if power was lost to the crane. Drag brakes were not used on the X-SAM crane.

Documents (a) General Electric (GE) Instruction IC9516 "AC and DC Brakes - Trolley & Bridge,"

Reviewed: version GEH-64IN (b) Gibbs and Hill Calculation Job No. F-1115 for the Trolley Drive and Bridge Drive, dated June 22, 1978 (c) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (d) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8,1982 (e)

Gibbs & Hill Calculations S.O. No. E1115 for Trolley Drive and for Bridge Drive, dated June 22, 1978 (f) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Design Topic: Drum Safety Devices Reference: NUREG 0554, Section 4.2 Published May 1979 Requirement: The hoist drum should be provided with structural and mechanical safety devices to limit its drop during a shaft or bearing failure. The devices should prevent disengaging from the holding brake.

Observation: The hoist drum was provided with the structural and mechanical safety devices to limit its drop during a shaft or bearing failure. The devices would also prevent disengaging from the holding brake. Ederer Topical Report EDR-I (P)-A,Section III.B.1.b, stated The emergency drum brake system provides an independent means for reliably and safely stopping and holding the load following a failure in the hoist machinery. Hoist machinery failures included shaft or bearing failures.Section III.E.4 of the topical report stated When a significant discontinuity between the motor shaft and the wire rope drum, e.g., failure of a key, shaft, coupling or gear, or actuation of the energy absorbing torque limiter is detected, the failure detection system sets the emergency drum brake system, interrupting the power to the hoist motor. Comanche Peak Document TXX-3659, Table C-1, Sheet 3 of 8, stated (in reference to Section III.C (C.3.k) of the topical report) that As shown in Figure C-1, the drum safety support, which is the same on both ends of the drum, complies with C.3.k of the Regulatory Guide [Regulatory Guide 1.104], in that it will limit the drop of the drum and thereby prevent it from disengaging its holding brake system should the drum shaft or bearing fail or fracture.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8, 1982 (b) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Nuclear Reactor Regulations, U. S. Nuclear Regulatory Page 4 of 149

Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (c) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 (d) Regulatory Guide 1.104 "Overhead Crane Handling Systems for Nuclear Power Plants," Revision 1 (Draft 3), withdrawn July 27, 1981 and republished as NUREG 0554 Category: Crane Design Topic: Emergency Stop Feature Reference: NUREG 0554, Sections 3.3, 6.1, and 6.6 Published May 1979 Requirement: An emergency stop feature should be installed at the control station. For cranes remotely operated using radio control stations, a second emergency stop feature should be provided at ground level to remove power from the crane, independent of the controller.

Cranes that use more than one control station should be provided with electrical interlocks that permit only one control station to be operated at a time.

Observation: An emergency stop feature was installed at both the crane cab and on the remote controller. A crane power shutoff was located at the 838 foot level on the crane motor control center panel. The shutoff location was easily accessible. Ederer Topical Report EDR-I (P)-A,Section III.1.B.f, stated An emergency stop button at each control station removes power from the crane and sets the emergency drum brake system as soon as the load starts to lower. Two control stations existed for the Comanche Peak crane. Both the control station in the cab and the portable radio control station had emergency stop buttons. An electrical interlock was provided in the cab to permit only one control station to be operated at a time.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8, 1982 (b) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (c) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Design Topic: Seismic Events During Cask Movement Reference: NUREG 0554, Section 2.5 Published May 1979 Requirement: The crane should be designed to retain control of and hold the load, and the bridge and trolley should be designed to remain in place on their respective runways with their wheels prevented from leaving the tracks during a seismic event.

Observation: The crane was designed to retain control of and hold the load during a seismic event.

The bridge and trolley were designed to remain in place on their respective runways with their wheels prevented from leaving the tracks during a seismic event. Equipment Qualification Summary SEQSP-MS41A-01, Section 2, Discussion, page 5, stated Trenchard Associates in Addendum A indicates that the requirements of Spec. MS-41A regarding the crane remaining in place was met by using seismic restraints to prevent the trolley and bridge from leaving the rail during a seismic event. Calculation 16345-CS(B)-836A1 demonstrated that during a safe shutdown earthquake (SSE), the assumed seismic loads in the upward and horizontal directions would not exceed the strength of Page 5 of 149

the seismic restraints.

Documents (a) Impell Inc. Equipment Qualification Summary Package No. SEQSP-MS41A-01 Fuel Reviewed: Building Overhead Crane (130 Tons Capacity) by Ederer Inc. Tag No. CPX-MESCFC-01," dated January 10, 1989 (b) Letter from Gibbs and Hill, Inc. to Comanche Peak Steam Electric Station dated August 17, 1979, with two (2) enclosures: Trenchard Associates Seismic Qualification of 17 Ton, 27' Span Overhead Crane and 130 Ton, 67'

Span Overhead Crane for Texas Utilities Services, Inc., certified by Ederer Inc., dated April 30, 1979 and the Gibbs & Hill Operability Calculation for 130 Ton Crane," dated October 25, 1978 (c) Stone and Webster Calculation 16345-CS(B)-836A1 130 T Overhead Crane Rail Anchorage, Revision 0 (d) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Design Topic: Seismically Induced Load Swing Reference: NUREG 0554, Section 2.5; Reg Guide 1.29 Published May 1979 Requirement: The maximum critical load plus operational and seismically induced pendulum and swing load effects on the crane should be considered in the design of the trolley and should be added to the trolley weight for the design of the bridge.

Observation: The licensee addressed the maximum critical load plus operational and seismically induced pendulum and swing load effects on the crane and concluded that the swing load effects during a safe shutdown earthquake (SSE) did not need to be considered in the design of the trolley and did not need to be added to the trolley weight for the design of the bridge. A seismic qualification evaluation performed by Trenchard Associates and attached to a Gibbs and Hill's August 17, 1979 letter to Comanche Peak stated For lateral motions the pendulum effect of the load on the end of the cable has a low frequency compared to the girder. This allows the cable motion and load to be ignored in the analysis of the girder. This statement was considered applicable to the Comanche Peak crane because the girder was a concrete corbel that was integral with the 3-foot thick fuel building wall; therefore, the frequency of the girder was infinite because of the corbel rigidity.

Documents (a) Equipment Qualification Summary SEQSP-MS41A-01 Fuel Building Overhead Reviewed: Crane (130 Tons Capacity)" by Ederer Inc. Tag No. CPX-MESCFC-01, dated January 10, 1989 (b) Letter from Gibbs and Hill, Inc. to Comanche Peak Steam Electric Station dated August 17, 1979, with two (2) enclosures: Trenchard Associates Seismic Qualification of 17 Ton, 27' Span Overhead Crane and 130 Ton, 67' Span Overhead Crane for Texas Utilities Services, Inc., certified by Ederer Inc., dated April 30, 1979 and the Gibbs & Hill Operability Calculation for 130 Ton Crane," dated October 25, 1978 (c) Drawing Number 2323-S-0816 Fuel Building Outline Sections, Revision 3 (d)

NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 (e) Regulatory Guide 1.29 "Seismic Design Classification," Revision 4 Category: Crane Design Topic: Single Failure Proof Crane Reference: NUREG 0554, Section 1.0 Published May 1979 Requirement: When reliance for the safe handling of critical loads is placed on the crane system itself, the system should be designed so that a single failure will not result in the loss of the Page 6 of 149

capability of the system to safely retain the load.

Observation: In 1983, Texas Utilities installed and tested an Ederer X-SAM single failure proof crane in their fuel handling building. The Ederer X-SAM crane had been accepted by the NRC as a single failure proof crane in 1980 after the review of the Ederer Topical Report EDR-I(P)-A, Revision 1 and reconfirmed in 1983 after the NRC reviewed Revision 3 of the Ederer Topical Report. The X-SAM crane will be the crane used at Comanche Peak for lifting the loaded casks in the fuel handling building as part of the ISFSI project.

Documents (a) Letter from Robert L. Baer, U. S. Nuclear Regulatory Commission to William Clark, Reviewed: Jr. Ederer Inc. entitled "Review and Acceptance of Topical Report EDR-I, Ederer's Nuclear Safety Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 1,"

dated January 2, 1980 (b) Letter from Cecil Thomas, U. S. Nuclear Regulatory Commission to William Clark, Jr., Ederer Inc. entitled "Acceptance for Referencing of Licensing Topical Report EDR-I (P), Revision 3, Ederer Nuclear Safety Related Extra Safety and Monitoring (X-SAM) Cranes," dated August 26, 1983 (c) Ederer Inc. Generic Licensing Topical Report EDR-I(P)-A "Ederer's Nuclear Safety Related Extra Safety And Monitoring (X_SAM) Cranes," Revision 3 (d) Letter from H. C. Schmidt, Texas Utilities Services Inc. to B. J. Youngblood, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos.50-445 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (e) Purchase Order No. CP-0041A

"Ederer Inc. Procedure No. 251-Special Site Pre-Operational Test Procedures for X-SAM Crane," dated February 22, 1984 Category: Crane Design Topic: Two-Block Protection Reference: NUREG 0554, Section 4.5 Published May 1979 Requirement: The complete hoisting system should have the required strength to resist failure during two-blocking. As an alternative, a system of upper travel limit switches may be used to prevent two-blocking. The system should include two independent travel limit devices of different designs and activated by separate mechanical means. These devices should de-energize the hoist drive motor and the main power supply. The auxiliary hoist, if used for critical lifts, should also be equipped with two independent travel limit switches to prevent two-blocking.

Observation: The complete hoisting system had the required strength to resist failure during two-blocking. Comanche Peak Document TXX-3659, Table C-1, Sheet 3 of 8, stated The energy absorbing torque limiter (EATL) was designed such that the maximum machinery load, which would result in the event of a high speed two blocking that allows the full breakdown torque of the motor to be applied to the drive shaft, will not exceed twice the design rated loading. In addition, the energy absorbing torque limiter design does not allow the maximum cable loading to exceed the acceptance criteria established in Section III.C (C.3.e) [Ederer Topical Report] during the above described two blocking.

Ederer Topical Report EDR-I (P)-A,Section III.A, stated The energy absorbing torque limiter allows the hoist to safely withstand two blocking, overloading, or load hang-up, and still retain the load even if the drive motor is not de-energized. Not only are the loads controlled following a two-blocking, load hang-up, etc., but the hoists components are also protected, throughout their life, from being overstressed by these incidents. To provide this protection, the energy absorbing torque limiter directly converts the hoists Page 7 of 149

high speed kinetic energy to heat during an overloading incident.

The crane also utilizes a system of upper travel limit switches to prevent two-blocking.

Ederer Topical Report EDR-I (P)-A,Section III.B.2.a, stated Two separate and independent limit switches sequentially activate as the load block reaches its upper limit of travel. The primary rotary limit switch on the drum shaft senses both the upper and lower positions of the load block travel. The primary upper limit switch de-energizes the hoist controls. If the hoisting is not stopped by the rotary limit switch, a secondary, lever operated, limit switch is tripped by the lower block. The secondary switch actuates the failure detection system, since it can be tripped only if there has been a primary limit switch or control system failure. The failure detection system set the emergency drum brake, which removes all power from the hoist. Therefore, the system of upper limit switches included the required two independent travel limit devices of different designs activated by separate mechanical means. The travel limit devices were also designed to de-energize the hoist drive motor and main power supply.

The auxiliary hoist and cantilever hoists were equipped with two independent travel limit switches to prevent two-blocking. Comanche Peak Document TXX-3659, Table C-1, Sheet 3 of 8, stated The auxiliary and cantilever hoists have two independent travel limit switches to prevent two blocking.

The 130, 17, and 5 ton hoist upper and lower limit switches were tested on April 20, 2011, using Procedure MSM-PX-2017. Procedure MSM-PX-2017 was completed through Work Order 3991951 and included Step 8.9.5, which stated Slowly raise and lower 130 ton hoist block to upper and lower limit switches, and ensure the limit switches operate properly. Step 8.10.3 of the procedure stated Slowly raise and lower 17 ton hoist block to upper and lower limit switches, and ensure the limit switches operate properly. Similarly, Step 8.11.3 stated Slowly raise and lower 5 ton hoist block to upper and lower limit switches, and ensure the limit switches operate properly.

All tests were successfully completed.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8,1982 (b) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (c) Procedure MSM-PX-2017 Fuel Building Overhead Crane Mechanical Inspection, Revision 2 (d) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Design Topic: Wire Rope Breaking Strength Reference: NUREG 0554, Section 4.1 Published May 1979 Requirement: The maximum load (including static and inertia forces) on each individual wire rope in the dual reeving system with the maximum critical load attached should not exceed 10%

of the manufacturer's published breaking strength.

Observation: The Ederer X-SAM crane installed at Comanche Peak used a wire rope yield strength concept to keep the rope stresses within the elastic range of the stress-strain curve instead Page 8 of 149

of meeting the NUREG 0554 criteria of 10 percent (10 to 1 safety factor). Comanche Peak's Design Basis Document DBS-ME-006, Attachment 2, Table 1-8 "Fuel Building Overhead Crane Data" identified the two wire ropes used on the fuel building dual reeving system crane as 1-1/4 inch 6 x 37 class independent wire rope core (IWRC)

stainless steel rope with a break strength of 146,000 pounds and a yield strength of 116,800 pounds. Purchase Order 76085 from Ederer Inc. listed the purchase of two ropes of 1-1/4 inch 6 x 37, 18-8 type 304 super tensile stainless steel wire rope. The lengths of rope purchased were 815 feet and 795 feet. In addition, a test sample of the rope was purchased for conducting the strength test. The required break strength criteria established by Ederer was 146,000 pounds. United States Steel Corp. performed a

"break test to failure" on the rope and documented a break strength of 176,500 pounds, exceeding the 146,000 pound break strength specified by the manufacturer. The Comanche Peak crane was a dual reeving system with eight (8) sheaves and 16 parts of rope. The crane was rated at 130 tons (260,000 pounds). As such, each reeve of rope would be expected to hold a maximum of 16,250 pounds (260,000/16). Ten percent of the manufacturer's published break strength (146,000 pounds) would be 14,600 pounds.

The 14,600 pounds equated to a static load factor of 9 to 1. This met the 5 to 1 required safety factor in ANSI B30.2 - 1976, Section 2-1.11.2(a) for wire ropes, but did not meet the 10 to 1 guidance in NUREG 0554. However, when considering the actual break test performed on the rope, the break strength that was actually measured during the testing was 176,500 pounds. Ten percent of this value would be 17,600 pounds, which did exceed the weight that would be experienced on each of the 16 ropes (16,250 pounds)

while lifting the maximum load of 130 tons and provided for a 10.8 to 1 safety factor.

The NRC had reviewed the safety features of the X-SAM crane and issued a Safety Evaluation Report in January 2, 1980, related to Ederer's Generic Licensing Topical Report EDR-I(P), Revision 1 and on August 26, 1983, related to Revision 3. In the 1980 letter, the NRC stated that the design features presented in the topical report for the Ederer X-SAM crane were acceptable for assuring that a single failure would not result in the loss of capability to safely retain a critical load. In the 1983 letter, the NRC Safety Evaluation Report discussed the features of the wire rope used for the X-SAM crane and noted that Ederer had departed from the commonly used approach of relating wire rope strength to the break strength of the rope and instead related it to the yield strength to provide a more conservative approach on wire rope safety by keeping the stress within the elastic range of the stress-strain curve. The NRC staff considered the use of yield strength limits as a conservative and acceptable method of assuring wire rope integrity during abnormal operations. This concept used an energy absorbing torque limiter (EATL) to control the overload condition placed on the rope. The EATL was incorporated into the hoist gear case using a disc spring loaded clutch which slipped at a preselected torque, thereby reducing the load stresses on the rope. As such, the safety criteria for the wire rope was met using a different concept for the X-SAM crane than the NUREG 0554 criteria and was found acceptable to the NRC. Ederer Topical Report EDR-I(P)-A,Section III.C (C.3.e) described the Ederer design for the wire rope and provided the formula for calculating the minimum wire rope break strength for various design rated loads for the crane. Each wire rope system was designed to withstand the peak static and dynamic loads imposed by a single wire rope failure without exceeding 90 percent of the yield strength of the cable with allowances for cable wear and fatigue.

Cable wear was limited to the criteria established in ANSI B30.2, which required the Page 9 of 149

rope to be replaced when the wear on the outer wire exceeded one-third the original diameter of the outside individual wires. Ederer Topical Report, Appendix E "Analysis of Load Motion and Cable Loading Following a Single Drive Train Failure or Single Wire Rope Failure in an X-SAM Type Crane" provided a detail description of the analytical and numerical techniques used to evaluate the failure of a wire rope to demonstrate the load would not be dropped due to the second wire rope being able to absorb the stresses. Appendix I "Analysis of Load Motion and Cable Loading Following a Single Wire Rope Failure in an X-SAM Type Crane Equipped with a Totally Mechanical Drive Train Continuity Detector and Emergency Drum Brake Actuator or a Continuously Engaged Emergency Drum Brake System" also provided information relative to a wire rope break by accounting for the additional load motion and kinetic energy associated with a small amount of drum rotation that would occur prior to the setting of the emergency drum brake. This accounted for the situation where one of the wire ropes broke while lowering the load at the design rated speed as opposed to the drum being stationary when the failure occurred.

Documents (a) Letter from Robert L. Baer, U. S. Nuclear Regulatory Commission to William Clark, Reviewed: Jr., Ederer Inc. entitled "Review and Acceptance of Topical Report EDR-I, Ederer's Nuclear Safety Related Extra Safety and Monitoring (X_SAM) Cranes, Revision 1,"

dated January 2, 1980 (b) Letter from Cecil Thomas, U. S. Nuclear Regulatory Commission to William Clark, Jr., Ederer Inc. entitled "Acceptance for Referencing of Licensing Topical Report EDR-I(P), Revision 3, Ederer Nuclear Safety-Related Extra Safety And Monitoring (X_SAM) Cranes," dated August 26, 1983 (c) Ederer Inc.

Generic Licensing Topical Report EDR-I(P)-A "Ederer's Nuclear Safety Related Extra Safety and Monitoring (X_SAM) Cranes," Revision 3 (d) United States Steel Corporation Report of Tests for Customer Order No. 76085 for 1-1/4 inch Diameter 6 x 64 SFW Super Tensile, Type 304, X-Lay, Right Regular Lay, IWRC Stainless Wire Rope Assemblies, dated November 6, 1979 (e) Ederer Inc. Purchase Order 76085 to United States Steel Corporation, dated February 15, 1979 (f) Comanche Peak Design Basis Document DBD-ME-006 "Control of Heavy Loads at Nuclear Plants," issued December 22, 2009 (g) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 (h) American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Design Topic: Wire Rope Configuration Reference: NUREG 0554, Section 4.1 Published May 1979 Requirement: A dual rope reeving system with individual attaching points and a load balancing system will permit either rope system to hold and transfer the critical load without excessive shock in case of failure of the other rope system. The dual reeving system may be a single rope from each end of the drum terminating at one of the blocks or equalizer with provisions for equalizing beam-type load and rope stretch, with each rope designated for the total load. Alternatively, a 2-rope system may be used from each drum or separate drums using a sheave equalizer or beam equalizer or other combination that provides two separate and complete reeving systems.

Observation: The fuel building overhead crane used a dual rope reeving system with individual attaching points and a load balancing system to hold and transfer the critical load without excessive shock in case of failure of one of the rope systems. Ederer Topical Report Page 10 of 149

EDR-I (P)-A, Figure III.C.3.e, illustrated the dual rope reeving system with two individual attaching points. Table III.F.I of the topical report identified the dual reeving system as a 2-rope system. Section, III.B.3.a, stated A standard reeving scheme has been modified to provide a balanced load path using two independent sets of reeving.

Section III.B.3.b, stated The dead ends of the two independent sets of reeving are attached to the hydraulic load equalization system. This system allows equalization of the two sets of reeving during normal operations, but retards any sudden motion caused by a broken rope.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8, 1982 (b) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (c) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Inspection Topic: Crane Inspection - Frequent Reference: ASME B30.2; Section 2-2.1.2 Revision 1976 Requirement: Cranes in regular use shall be subjected to a frequent crane inspection monthly during normal service, weekly to monthly during heavy service, and daily to weekly during severe service. The frequent inspection points shall include: a) operating mechanisms for proper operation daily; b) all limit switches should be checked at the beginning of each work shift by inching, or running at slow speeds, each motion into its limit switch; c) leakage in lines, tanks, valves, pumps, and other parts of the air or hydraulic systems; d) hooks for cracks, more than 15% of normal throat opening, or more than 10 degrees of twist; e) hook latches for proper operation; f) hoist ropes including end clamps; and g)

the rope reeving system.

Observation: The frequent crane inspections were performed prior to use daily on the fuel building overhead crane. Procedure MDA-402 included a checklist for the crane operators to complete prior to operating the crane. Form MDA-402-1 "Documented Prior to Use Inspection of Crane and Hoists" required the operator to perform the following inspections: 1) inspect control stations for unreadable control function labels; 2) inspect hoist ropes for damage and check for loose end clamps and clips; 3) inspect for improper rope reeving (e.g., ropes not properly seated on the drum or in sheaves); 4) inspect air and hydraulic systems for leakage; 5) inspect hook safety latches for improper operation; 6) inspect hooks for deformation and cracks and verify free rotation; 7) check power and control cords for damage (stand-alone hoists only); 8) inspect hoist chains, including end connections, for damage or defects (stand-alone hoists only); and 9) for chain hoists, raise and lower the load slightly and ensure the load chain is not binding or jumping.

The crane operator was also required to complete the following functional checks prior to using the crane that day: 1) energize the equipment and check all functioning operating mechanisms for mis-adjustment interfering with proper operation and unusual sounds; 2) under no load conditions, carefully check all limit switches and the upper limit device of each hoist. Extreme care shall be exercised. Each hoist/motion shall be inched into the limit or run at slow speed. If the device does not operate properly, the Page 11 of 149

operator shall immediately notify the responsible Supervisor.

Documents (a) Procedure MDA-402 Control of Load Handling Equipment, Revision 11 (b) Form Reviewed: MDA-402-1 "Documented Prior to Use Inspection of Crane and Hoists," Revision 8 (c)

American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Inspection Topic: Crane Inspection - Periodic Reference: ASME B30.2; Section 2-2.1.3 Revision 1976 Requirement: Cranes in regular use shall be subjected to a periodic crane inspection annually during normal and heavy service, and quarterly during severe service. The periodic inspection includes checking for: a) deformed, cracked or corroded members; b) loose bolts or rivets; c) cracked or worn sheaves and drums; d) worn, cracked or distorted pins, bearings, shafts, gears, and rollers; e) excessive brake system wear; f) load, wind, and other indicators over their full range for any significant inaccuracies; g) gasoline, diesel, electric, or other power plants for improper performance; h) excessive drive chain sprocket wear and chain stretch; and i) deterioration of controllers, master switches, contacts, limit switches and pushbutton stations.

Observation: The periodic crane inspection was performed as required annually per Procedure MSM-PX-2017. The last inspection was completed on April 20, 2011. The licensee's crane was designated as an "infrequent use" crane and was inspected annually. Procedure MSM-PX-2017, Step 8.2.1 and Step 8.3.1 required inspection of the bridge and trolley for deformed, cracked, corroded or damaged members; evidence of loose bolts, rivets; and evidence of broken welds. Steps 8.4.6 through 8.4.7 required inspection of the hoist sheaves for wear and cracks. Step 8.9.14 required inspection of the drum for wear and cracks. Steps 8.7.6.3 and 8.8.7.3 required inspection of the bridge and trolley bearings for cracks or damage. Step 8.7.6.1 required inspection of the bridge drive wheel gears for worn or damaged teeth, loose set screws and keys. Step 8.8.7.1 required inspection of the trolley drive wheel gears for worn or damaged teeth, loose set screws and keys.

Step 8.9.13 required inspection of the 130 ton lower block for damaged plates and loose pins or bolts. Step 8.2.10.10 required inspection of the hydraulic bridge brake wheel accessible surface for cracking, scoring, heat checked and blued outer surface. Step 8.2.12.3 required inspection of the bridge electric brake wheel for damage and the accessible surface for scoring, heat checked and a blued outer surface. Step 8.3.14.6 required inspection of the trolley brake wheel accessible surface for scoring, damage, heat checked and a blued outer surface. Step 8.4.9 required visual inspection of the 130 ton hoist electric brakes for wear, glazing, damage or unusual conditions. Step 8.3.11 required measuring the sag in the trolley inching drive chain. Step 8.3.12 required inspection of the trolley inching drive chain sprockets for wear and damage.

The electrical operation and maintenance inspection was conducted using Procedure MSE-PX-2017 and was completed annually. Step 8.2.2 of Procedure MSE-PX-2017 required visual inspection of the contacts on the controller for excessive pitting or wear and required replacement, if necessary. Step 8.2.3 ensured all wiring and electrical components were free of damage and all electrical connections were tight.

Page 12 of 149

Documents (a) Procedure MSM-PX-2017 "Fuel Building Overhead Crane Mechanical Inspection,"

Reviewed: Revision 2 (b) Procedure MSE-PX-2017 "Fuel Building Overhead Crane Electrical Inspection," Revision 3 (c) American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Inspection Topic: Crane Operational Testing Reference: ASME B30.2; Sect 2-2.2.1 Revision 1976 Requirement: Prior to initial use, all new, reinstalled, extensively repaired, or modified cranes shall have the following functions tested: (a) lifting and lowering, (b) trolley travel, (c) bridge travel, (d) limit switches, and (e) locking and safety devices. The trip setting of the hoist limit devices shall be determined by tests with an empty hook traveling in increasing speeds up to the maximum speed. The actuating mechanism of the limit device shall be located so that it will trip the device under all conditions in sufficient time to prevent contact of the hook or load block with any part of the trolley or crane.

Observation: The "prior to initial use" crane testing had been satisfactorily completed in 1983. The crane was tested for lowering, raising, travel, limit switches, and safety devices per Procedure No. 251 in April and June 1983. The following items were tested: X-SAM limit switch functional tests, emergency drum brake timing, emergency drum brake dynamic torque measurement, hydraulic equalization system, wire rope spooling monitor, drive train continuity detector, two blocking test, overload system test, overspeed detection test, emergency lower test, emergency trolley test, and the emergency bridge test.

Documents (a) Ederer Inc. Procedure No. 251 "Special Site Pre-operational Testing Procedures for X-Reviewed: SAM Cranes," Revision 4 (b) American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Inspection Topic: Hoist Overload Testing Reference: NUREG 0554, Section 8.3; NUREG 0612, C-4, (9) Published 1979/1980 Requirement: If the hoisting system is designed with adequate strength to resist failure during load hang-up, the hoisting system should be tested by securing the load-block-attaching points to a fixed anchor and applying the maximum critical load. Alternately, if a load cell system, a motor current-sensing device, or a mechanical load-limiting device is provided to prevent load hang-up, the device(s) should be tested to verify operability.

Observation: The hoisting system was designed with adequate strength to resist failure during load hang-up. Ederer Topical Report EDR-I (P)-A,Section III.A, stated The energy absorbing torque limiter (EATL) allows the hoist to safely withstand two blocking, overloading, or load hang-up, and still retain the load, even if the drive motor is not de-energized. Not only are the loads controlled following a two blocking, load hang-up, etc., but the hoists components are also protected, throughout their life, from being overstressed by these incidents. To provide this protection, the energy absorbing torque limiter directly converts the hoists high speed kinetic energy to heat during an overloading incident.

Additionally, the crane utilizes a load cell system to prevent load hang-up. Ederer Page 13 of 149

Topical Report EDR-I (P)-A,Section III.B.2.b, stated A load cell is installed in the hoist reeving. Exceeding the load limit setting shuts down the hoist, but does not actuate the failure detection system. The load cell senses overloads that result from two blocking or load hang-up which de-energizing the hoist controls and sets the conventional holding brakes on the high speed shafting.

A 125% static load test was completed in 1983 per Traveler Number RI83-1139-8100 with the hoist overload alarm and motor trip tested per Steps 5b and 5c. The hoist motor trip setting was confirmed during the test at 273,000 pounds +/- 2,000 pounds. The alarm and warning light were verified to activate at 267,000 pounds +/- 2,000 pounds.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8, 1982; (b) Traveler Number RI83-1139-8100, Perform Load Test on Fuel Handling Bridge Crane, final review and sign-off April 21, 1983 (c) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Inspection Topic: Hoist Two-Block Testing - Limit Device Method Reference: NUREG 0612, C-4, (8) Published July 1980 Requirement: If the hoisting system is provided with a system of upper travel limit switches to prevent two-blocking, the travel limit switches should be tested to verify operability. If the crane is equipped with a load limiter (strain gage, etc.) the load limiter should be tested to verify operability.

Observation: The hoisting system of the 130 ton crane is provided with a system of upper travel limit switches to prevent two-blocking. Ederer Topical Report EDR-I (P)-A,Section III.B.2.a, stated Two separate and independent limit switches sequentially activate as the load block reaches its upper limit of travel. The primary, rotary limit switch on the drum shaft senses both the upper and lower positions of the load block travel. The primary upper limit switch de-energizes the hoist controls. If the hoisting is not stopped by the rotary limit switch, a secondary, lever operated, limit switch is tripped by the lower block. The secondary switch actuates the failure detection system, since it can be tripped only if there has been a primary limit switch or control system failure. The failure detection system activates the emergency drum brake, which removes all power from the hoist.

The 130 ton hoist upper and lower limit switches were tested to verify operability on April 20, 2011, using Procedure MSM-PX-2017. Procedure MSM-PX-2017 was completed through Work Order 3991951 and included Step 8.9.5, which stated Slowly raise and lower 130 ton hoist block to upper and lower limit switches and ensure the limit switches operate properly.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8,1982 (b)Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (c) Procedure MSM-PX-2017 Fuel Building Overhead Crane Mechanical Inspection, Revision 2 (d) NUREG 0612 Page 14 of 149

Control of Heavy Loads at Nuclear Power Plants, issued July 1980 Category: Crane Inspection Topic: Hook Inspections - Frequent/Periodic Reference: ASME B30.10, Sections 10-1.4.2 and 10-1.4.6 Published 1975 Requirement: Hooks shall be inspected monthly during normal service, weekly to monthly during heavy service and daily to weekly during severe service. Hooks should be inspected for:

a) distortion such as bending, twisting or increased throat opening; b) wear; c) cracks, severe nicks, or gouges; d) latch engagement, damaged or malfunctioning latch (if provided); and e) hook attachment and securing means. Hooks having any of the following deficiencies shall be removed from service unless a qualified person approves their continued use and initiates corrective action: a) cracks; b) wear exceeding 10% of the original sectional dimension; c) bend or twist exceeding 10 degrees from the plane of an unbent hook; and d) an increase in throat opening of 15% (for hooks without latches).

Observation: The hook of the 130 ton crane was inspected to the criteria of ASME B30.10 annually.

The 130 ton crane used at Comanche Peak was designated as an "infrequent use" crane.

The Maintenance Department Administrative Manual Procedure MDA-402 established the hook inspection criteria. The "prior to use" inspection requirements were provided in Attachment 8B. Steps 5 and 6 required inspection of the hook safety latches for improper operation and inspection of the hooks for deformation, cracks and free rotation. The last periodic inspection was completed on April 20, 2011 in accordance with Maintenance Inspection Procedure MSM-PX-2017. Step 8.9.9 of Procedure MSM-PX-2017 documented the inspection of the hook for chemical damage, wear, deformation, cracks, obvious twist from plane of unbent hook, inoperable hook safety latch, evidence of damaged or loose connections and obvious increase in throat opening.

No unacceptable conditions were found.

Documents (a) Procedure MDA-402 "Control of Load Handling Equipment," Revision 11 (b)

Reviewed: Procedure MSM-PX-2017 "Fuel Building Overhead Crane Mechanical Inspection,"

Revision 2 (c) American National Standard [American Society of Mechanical Engineers]

(ASME) B30.10 Hooks, Revision 1975 Category: Crane Inspection Topic: Wire Rope Inspection - Frequent Reference: ASME B30.2, Section 2-2.4.1 (a) Revision 1976 Requirement: All ropes shall be visually inspected once each working day.

Observation: The wire ropes on the 130 ton crane were visually inspected prior to use each working day. Procedure MDA-402, Attachment 8B, Step 2 required the inspection of hoist ropes for damage and to check for loose end clamps and clips. This procedure was completed each day prior to using the 130 ton crane.

Documents (a) Procedure MDA-402 "Control of Load Handling Equipment," Revision 11 (b)

Reviewed: American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976

.

Page 15 of 149

Category: Crane Inspection Topic: Wire Rope Replacement Criteria Reference: ASME B30.2, Section 2-2.4.2 Revision 1976 Requirement: Conditions such as the following should be sufficient reason for questioning continued use of the rope, or increasing the frequency of inspection: a) twelve randomly distributed broken wires in one lay; b) wear of one-third of the original diameter of outside individual wires; c) kinking, crushing, bird caging or any other damage resulting in distortion of the rope structure; d) evidence of heat damage; and e) reduction in diameter in excess of nominal.

Observation: The wire rope on the 130 ton crane was inspected to the ASME B30.2, Section 2-2.4.2 criteria annually per Procedure MSM-PX-2017. The last annual inspection was conducted on April 20, 2011. Step 8.9.7 of the procedure required visual inspection of the following items to verify the wire rope had: (1) less than twelve randomly distributed broken wires in one rope lay or less than four broken wires in one strand of one rope lay; (2) no evidence of kinking, crushing, bird-caging, cutting, unstranding or any other damage resulting in distortion of rope structure; (3) no evidence of heat or corrosion damage; (4) no evidence that end connections (drum and dead-ends) are loose, corroded, cracked, bent, worn or improperly applied, if visible; (5) no evidence that wires in end connections (drum and dead-ends) are corroded, or broken, if visible; and (6) no wear on individual wires on the outside surface of the rope that is equal to or greater than one third of the original wire diameter by visual estimation. Step 8.9.8 required measuring the 130 ton hoist wire rope diameter and replacing the wire rope if the diameter measured less than 1.157 inches. The original purchase order requirement specified by Ederer Inc. for the wire rope currently on the 130 ton crane was for a diameter of 1 1/4 inches (1.250 inches).

Documents (a) Procedure MSM-PX-2017 "Fuel Building Overhead Crane Mechanical Inspection,"

Reviewed: Revision 2 (b) American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 (c) Ederer Inc. Purchase Order 76085 to United States Steel Corporation, dated February 15, 1979 Category: Crane Load Testing Topic: Cold Proof Testing Reference: NUREG 0554, Section 2.4; NUREG 0612, C-2 (8) Published 1979/1980 Requirement: Minimum operating temperatures for the crane should be specified to reduce the possibility of brittle fracture of the ferritic load-carrying members of the crane. The minimum temperature can be determined by: 1) a drop weight test per ASTM E-208, 2)

a Charpy test per ASTM A-370, or 3) a 125% cold proof test. If the crane is made of low alloy steel such as ASTM A514, cold proof testing should be done. If cold proof testing is omitted, the default minimum crane operating temperature is 70 degrees F.

For crane operation at temperatures below 70 degrees F, cold proof testing must be performed and the ambient temperature at which the testing is conducted becomes the minimum crane operating temperature.

Observation: A minimum operating temperature of 40 degree F was specified for the crane.

Comanche Peak Document TXX-3659, Table C-1, sheet 1 of 8, stated The crane was designed and fabricated for a minimum operating temperature of 40 degrees F. The licensee stated that the minimum temperature was determined by a Charpy V-Notch impact test per ASTM A-370 and that documentation related to the testing was stored in Page 16 of 149

the licensee's records system. Procedure MDA-304, Step 6.1.6 stated the 40 degree F minimum operating temperature. The Comanche Peak fuel building was an enclosed structure that was naturally heated by the heat from the spent fuel pools. The building had limited air conditioning and as such typical temperatures in the fuel building remained well over 70 degree F throughout the year. The fuel building overhead crane could not be moved to a location outside the fuel building.

Documents (a) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Reviewed: Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (b) Procedure MDA-304 "Control of Heavy Loads and Critical Lifts," Revision 6 (c) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Load Testing Topic: Dynamic Load Testing (100%)

Reference: NUREG 0554, Section 8.2 Published May 1979 Requirement: After the 125% static load test, the crane should be given a full performance test with 100% of the maximum critical load attached, for all speeds and motions for which the system is designed. This should include verifying all limiting and safety control devices.

Observation: After the 125% static load test, the crane was given a full performance test with 100% of the maximum critical load attached, for all speeds and motions for which the system was designed. Traveler Number RI83-1139-8100, page 2 of 6, Steps 8 through 14, documented that the licensee satisfactorily: 1) ensured a test load of 260,000 +/- 2000 pounds (100 percent of the maximum critical load) had been properly attached to the 130 ton hook; 2) raised the hoist with load attached until it stopped by the geared limit switch trip; 3) lowered the hoist about two feet below the upper limit switch trip; 4) operated the trolley north to south; 5) operated the bridge east to west; 6) ran bridge at full speed and then placed control lever in drift position (first speed position); 7) observed that the foot operated brake would stop the bridge within approximately five feet; and 8) lowered the test load to the railroad loading area. The hoist overload alarm and motor trip was tested during the 125% load test on step 5b and 5c.

Documents (a) Traveler Number RI83-1139-8100, Perform Load Test on Fuel Handling Bridge Reviewed: Crane, April 14,1983 (b) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Load Testing Topic: Hook Load Testing Reference: NUREG 0554, Sect 4.3; ASME B30.10, Sect 10-1.1.2 Published 1975 Requirement: A 200% static load test should be performed for each load-attaching hook. For a duplex (sister) hook, the proof load shall be shared by the two sisters unless the hook is designed for unbalanced loading. Measurements of the geometric configuration of the hooks should be made before and after the test and the acceptance criteria is no permanent increase in throat opening in excess of 0.5% or 0.010 inches (0.25 mm). The load testing should be followed by a nondestructive examination that should consist of volumetric and surface examinations to verify the soundness of fabrication and ensure integrity of the hooks.

Page 17 of 149

Observation: The hook of the licensees fuel building overhead crane was load tested to 200% of the rated load. Ederer Document #F-1115 documented that the hook for the fuel building crane, rated to 130 tons, was load tested to 260 tons. The load test on the hook was completed prior to the installation of the crane at the facility.

Documents (a) Ederer S.O. Document # F-1115 Critical Items List, dated June 27, 1979 (b)

Reviewed: NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Load Testing Topic: NDE Exams Following Cold-Proof Testing Reference: NUREG 0554, Section 2.4 and 2.6 Published May 1979 Requirement: Following the 125% cold-proof testing, a nondestructive examination of the welds whose failure could result in the drop of a critical load should be performed. If any of these weld joint geometries would be susceptible to lamellar tearing, the base metal at the joints should be nondestructively examined. Nondestructive examination of critical areas should be repeated at 4-year intervals or less.

Observation: The Ederer Topical Report -I(P)-A, Section C.4.d, stated that cold proof testing of the X-SAM crane was not required, since the required material testing of Regulatory Position C.1.b(2) was provided." Section C.1.b(2) of the Ederer Topical Report stated that nil ductility transition testing was performed in accordance with the load bearing structural members fabricated from rolled materials as shown on Appendix A "Critical Items List."

Therefore, nondestructive examination of welds was not required to be performed. The licensee also addressed the potential susceptibility of weld joint geometries to lamellar tearing in Design Basis Document DBD-ME-006, Table 1-9, Page 2, which stated The weld geometries used in the existing bridge structure are not considered to be susceptible to lamellar tearing.

Documents (a) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Reviewed: Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (b) Design Basis Document DBD-ME-006 Control of Heavy Loads at Nuclear Plants, Revision 25 (c) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Related Extra Safety And Monitoring (X-SAM)

Cranes, Revision 3, Amendment 3, dated October 8, 1982 (d) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Load Testing Topic: Rated Load Marking Reference: NUREG 0554, Section 8.5; ASME B30.2, Sect 2- Revision 1976 Requirement: The rated load shall be marked on each side of the crane and, if the crane has more than one hoisting unit, each hoist shall have its rated load marked on it or on its load block.

This marking shall be legible from the ground or floor.

Observation: The rated load of the fuel building overhead crane at Comanche Peak was marked on each side of the crane and on each of the hoists. The fuel building crane was rated as 130 tons. Two other hoists were located on the crane as well, a 17 ton and a 5 ton hoist.

Both the 17 and 5 ton hoists were also properly marked.

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Documents (a) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published Reviewed: May 1979 Category: Crane Load Testing Topic: Static Load Testing (125%)

Reference: NUREG 0554, Section 8.2 Published May 1979 Requirement: The crane should be static load tested at 125 percent of the maximum critical load. The test should be conducted at all positions generating maximum strain in the bridge and trolley structures and other positions as recommended by the designer or manufacturer.

Observation: The crane was static load tested at 125 percent of the maximum critical load. The test was conducted at the intersection of the railroad track centerline and the fuel building center line at the 810 6 level so as to generate the maximum strain in the bridge. The strain generated in the trolley structure would be the same regardless of the position of the trolley. Traveler Number RI83-1139-8100, page 1 of 6, Steps 3 through 7, documented that the licensee satisfactorily: 1) ensured a test load of 325,000 +/- 3250 pounds (125 percent of the maximum critical load) had been properly attached to the 130 ton hook; 2) began hoisting the test load at the slowest speed; 3) observed the hoist overload alarm and warning light energize when the weighing system readout indicated 267,000 +/- 2000 pounds; 4) observed the hoist motor trip when the weighing system readout indicated 273,000 +/- 2000 pounds; 5) lowered load till alarms cleared and then bypassed the hoist overload interlock; 6) hoisted the test load at the slowest speed until it was suspended approximately one foot above the 8106 floor level and allowed the load to remain suspended; and 7) lowered the test load back to the 8106 floor level at the slowest speed. The actual load used for the 125 percent load test was 325,430 pounds (162.7 tons) on April 14, 1983 and documented in Traveler No. RI83-1139-8100.

Documents (a) Traveler Number RI83-1139-8100 Perform Load Test on Fuel Handling Bridge Reviewed: Crane, dated April 14,1983 (b) NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Crane Operation Topic: Brake Test Prior to Lift Reference: ASME B30.2, Section 2-3.2.3 (g) Revision 1976 Requirement: The operator shall check the hoist brakes at least once each shift if a load approaching the rated load is to be handled. This shall be done by lifting the load a short distance and applying the brakes.

Observation: Procedure MDA-316, Attachment 8.A.4 "General Requirements for Crane Operators,"

stated in Step 11 "Operators of all cranes shall test the brakes each time a load approaching the rated load is handled by raising it a few inches and applying the brakes."

Documents (a) Procedure MDA-316 "Control of Load Handling," Revision 1 (b) American National Reviewed: Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Operation Topic: Height Limit During Cask Movement Reference: No Reference Provided Requirement: For single failure proof cranes, the cask height during movement should be sufficiently Page 19 of 149

high to allow for engaging of the brakes during an uncontrolled descent before the load would impact the floor.

Observation: The minimum cask height for the 130 ton Ederer single failure proof crane used at Comanche Peak was 1 foot. This minimum height limit had been incorporated into the site heavy lift procedures. Table C-1 of the June 8, 1983, letter (TXX-3659) to the NRC stated on page 1 of 8 under Regulatory Position C.2.b that "The crane was designed such that a maximum load motion following a drive train failure is less than 1 foot." Page 2 of 8 of Table C-1 under Regulatory Position C.3.h stated "The crane was designed such that the maximum load motion following a single wire rope failure is less than 1 foot." The Ederer Topical Report EDR-I(P)-A, Appendix E "Analysis of Load Motion and Cable Loading Following a Single Drive Train Failure or Wire Rope Failure in an X-SAM Crane," Revision 2, provided the analysis to support the 1 foot minimum height limit. A drive train failure while handling a 130 ton load could result in a displacement of the load of a maximum of 5 inches before the drum brakes fully engaged. If a 50% time delay occurred prior to full brake engagement, the load could drop 10 inches. For a single wire rope failure, the remaining rope could stretch as much as 4 inches due to the increased load. In both the drive train failure and the wire rope failure events, the 1 foot limit was adequate to prevent the load from impacting the floor.

The 1 foot minimum height limit had been incorporated into site procedures related to the heavy lifts using the fuel building overhead crane. Procedure DCS-205, Step 5.10 stated that heavy lifts using the fuel building overhead crane were required to be suspended a minimum of one foot above the operating floor or other structures, systems and components along the path of travel to allow for limited decent of the load due to a single failure event. Step 5.10 referenced Procedure MDA-304, Section 6.1 "Heavy Loads Lifted by Cranes or Hoists" which also contained the 1 foot minimum height limit. Procedure DCS-206 used to lift a loaded canister that had been returned to the fuel building from the ISFSI pad and place it into the dry cask pit for unloading, included the one foot minimum height requirement in Step 5.11 and referenced Procedure MDA-304.

Documents (a) Letter (TXX-3659) from H. C Schmidt, Texas Utilities Services to Director of Reviewed: Nuclear Reactor Regulations, U. S. Nuclear Regulatory Commission entitled "Comanche Peak Steam Electric Station Docket Nos. 50-455 and 50-446 Final Response to NUREG-0612," dated June 8, 1983 (b) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8, 1982 (c) Procedure DCS-205 "Stack-up and Transfer of Loaded MPC," Revision 2 (d) Procedure DCS-206 "`Transporting and Transferring a Loaded MPC for Unloading," Revision 2 (e) Procedure MDA-304

"Control of Heavy Loads and Critical Lifts," Revision 6, PCN-9 Category: Crane Operation Topic: Hoist Limit Switch Tested Each Shift Reference: ASME B30.2, Section 2-3.2.4 (a) Revision 1976 Requirement: At the beginning of each shift, the operator shall try out the upper limit device of each hoist under no-load. Care shall be exercised. The block shall be inched into the limit or run in at a slow speed.

Observation: At the beginning of a shift the crane operator tested the upper limit device of the hoist Page 20 of 149

that was to be used that day. Procedure MDA-402 on Form MDA-402-1 "Documented Prior to Use Inspection of Crane and Hoists," Step B. 2 stated "Under no load conditions, carefully check all limit switches and the upper limit device of each hoist. Extreme care shall be exercised. Each hoist/motion shall be inched into the limit or run at slow speed. If the device does not operate properly, the operator shall immediately notify the responsible supervisor."

Documents (a) Procedure MDA-402 Control of Load Handling Equipment," Revision 11 (b) Form Reviewed: MDA-402-1 "Documented Prior to Use Inspection of Crane and Hoists," Revision 8 (c)

American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Operation Topic: Minimum of Two Wraps of Rope Reference: ASME B30.2, Section 2-3.2.3 (h) Revision 1976 Requirement: The load shall not be lowered below the point where two wraps of rope remain on each anchorage of the hoisting drum unless a lower-limit device is provided, in which case, no less than one wrap shall remain.

Observation: The crane used at Comanche Peak for loading operations had sufficient wraps of rope when lifting and lowering the load. The HI-TRAC transfer cask, when placed in the dry cask operations area (elevation 810'), was at the lowest position of any of the cask movement activities. At this position the NRC inspector counted at least twelve wraps of wire rope on the crane drum.

Documents (a) American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Reviewed: Revision 1976 Category: Crane Operation Topic: Qualification For Crane Operator Reference: ASME B30.2, Sections 2-3.1.2 and 2-3.1.6 Revision 1976 Requirement: Qualification to operate a cab operated or remote operated crane, requires the operator to pass a written or oral examination and a practical operating examination specific to the type of crane to be operated unless able to furnish evidence of previous qualification and experience. In addition, the operator shall: a) have vision of at least 20/30 Snellon in one eye and 20/50 in the other with or without corrective lenses; b) be able to distinguish colors regardless of their position; c) have sufficient hearing capability for the specific operation with or without hearing aids; d) have sufficient strength, endurance, agility, coordination and reaction speed for the specific operation; e) have evidence of not having physical defects or emotional instability that would interfere with the operation; and f) not be subject to seizures, loss of control, or dizziness.

Observation: Crane operators assigned to the canister loading activities were required to be qualified to the requirements listed in ASME B30.2. The crane operators assigned to the first cask loading campaign met the requirements. Procedure MDA-308, Step 6.1.9 required passing a written test (score of 80% or better) and passing a practical evaluation.

Procedure MDA-308, Form MDA-308-13 "Crane Operator Physical Qualification Form," required the following physical qualifications for the crane operator or trainee:

(1) have vision of at least 20/30 Snellen in one eye, and 20/50 Snellen in the other, with Page 21 of 149

or without corrective lenses; (2) be able to distinguish colors, regardless of position of colors; (3) be able to hear, with or without hearing aid, adequately for a specific operation; (4) have sufficient strength, endurance, agility, coordination and speed of reaction to meet the demands of equipment; (5) not have evidence of physical defects, or emotional instability that could render a hazard to the operator or others, or which, in the operation of the examiner, could interfere with the operators safe performance.

Evidence of such conditions may be cause for disqualification. In such cases, specialized clinical or medical judgments and tests may be required; (6) not have evidence of being subject to seizures or loss of physical controls. Evidence of such conditions or a history of epilepsy or a disabling heart condition shall be sufficient reason for disqualification.

Specialized medical tests may be required to determine these conditions; and (7) have normal depth perception, field of vision, reaction time, manual dexterity, coordination and no tendencies to dizziness or similar undesirable conditions.

Documents (a) Procedure MDA-308 Requirements for Load Handling Personnel," Revision10 (b)

Reviewed: Form MDA-308-13 "Crane Operator Physical Qualification Form," Revision 0 (c)

American National Standard Institute (ANSI) B30.2 Overhead and Gantry Cranes, Revision 1976 Category: Crane Operations Topic: Maximum Weight of Canister Reference: No Reference Provided Requirement: The maximum weight of the transfer cask containing the canister filled with water and fuel (including dynamic loads) that will be lifted by the crane is to be verified to be within the crane's rated capacity.

Observation: The lift weights for the Holtec storage system components were calculated to not exceed the rated load capacity of the licensee's crane. Holtec Report Number HI-2104639, page 3 of 12, stated The maximum lifted weight during the performance of a HI-STORM system loading at Comanche Peak is during the removal of the HI-TRAC transfer cask from the spent fuel pool with the water jacket full (see Case 2 in Table 7.0.1). Case 2 of Table 7.0.1 showed a total lift weight of 255,814 pounds (127.9 tons), which was less than the 130 ton rating of the Comanche Peak fuel building crane. The 255,814 pounds included the weight of the loaded Hi-TRAC transfer cask, the lift yoke, and the lift yoke extension.

Documents (a) Holtec Report Number HI-2104639 Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak, Revision 0 Category: Crane Operations Topic: Provisions For Manual Operation Reference: NUREG 0554, Sections 3.4; 4.9 Published May 1979 Requirement: A crane that has been immobilized because of failure of controls or components while holding a critical load should be able to hold the load or set the load down while repairs or adjustments are made. This can be accomplished by inclusion of features that will permit manual operation of the hoisting system and the bridge and trolley transfer mechanisms by means of appropriate emergency devices.

Observation: In the event that the fuel building crane was immobilized because of failure of controls Page 22 of 149

or components while holding a critical load, the crane was able to hold the load or lower the load while repairs or adjustments were made. Ederer Topical Report EDR-I (P)-A,Section III.E.7, stated that a total loss of electrical power sets the conventional high speed holding brake and if the load starts to lower, the emergency drum brake is set as well, thereby stopping the load. The capability for lowering a load to a safe resting place, following a total loss of electrical power, was provided by the hoists integrated protective system (HIPS). Its emergency drum brake system provided a large margin of safety under these conditions because of the brakes substantial thermal capacity. The rated load can be safely lowered continuously from maximum hook height to the floor without exceeding the temperature limits of the brakes. It is not necessary to stop the load frequently, as is required when only conventional holding brakes are used.

The bridge and trolley can be manually moved. Ederer Procedure 251 was used to perform the pre-operational tests of the fuel building crane. Section 251.11.4

"Emergency Trolley Movement Test" and Section 251.11.5 "Emergency Bridge Movement Test" involved tagging out all power with the design rated load attached and moving the trolley and the bridge. The brakes were clamped open and either a manual or a power operator can be attached to the high speed shaft of the trolley or the bridge to move them during loss of power. Comanche Peak was in the process of developing specific procedural steps in Procedure MDA-402 for lowering the crane during a power failure or loss of control units. At the time of this inspection, the licensee was waiting for additional information from the crane vendor in order to complete the procedure.

The procedure will be finalized and approved prior to the first loading campaign.

Documents (a) Ederer Inc. Generic Licensing Topical Report EDR-I (P)-A Ederers Nuclear Safety Reviewed: Related Extra Safety And Monitoring (X-SAM) Cranes, Revision 3, Amendment 3, dated October 8,1982 (b) Procedure MDA-402 "Control of Load Handling Equipment,"

Revision 11 (c) Purchase Order No. CP-0041A "Ederer Inc. Procedure No. 251-Special Site Pre-Operational Test Procedures for X-SAM Crane," dated February 22, 1984 (d)

NUREG 0554 Single Failure Proof Cranes for Nuclear Power Plants, published May 1979 Category: Dry Run Demonstration Topic: Fuel Loading and Verification Demonstration Reference: CoC 1014, Condition 10. c, d Amendment 7 Requirement: The dry run shall include selection and verification of specific fuel assemblies to ensure type conformance and the loading of specific assemblies into the canister (using a dummy fuel assembly), including appropriate independent verification.

Observation: The dry run demonstration for fuel selection and verification was performed on July 7, 2011 and was observed by the NRC. The demonstration included a walk-through of the selection and verification process of the spent fuel and the actual loading a dummy fuel assembly into several positions in the canister basket. The licensee provided an explanation of the fuel loading and verification process using Procedures NUC-212, RFO-106, RFO-204, and RFO-302. A tour of the Unit 1 spent fuel pool included a discussion of how the spent fuel was numbered in the storage racks and how the operators will ensure the correct fuel assembly has been selected for movement into the canister. The independent verification will be performed by a person independent of those who loaded the canister after all spent fuel has been placed in the canister.

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The licensee utilized procedure NUC-212 to select the fuel assemblies for loading into the canisters. The procedure established the selection criteria for the fuel assemblies to ensure conformance with the Holtec technical specifications related to the fuel approved for storage in the HI-STORM 100 system. The licensee's core performance engineer pre-selected the fuel for loading into the canisters, completed Form RFO-106 -2 "Fuel Shuffle Sequence Plan" and generated an approved canister loading map. This map included the serial numbers of the selected fuel assemblies and the location in the canister where the fuel assembly was assigned. The selection process was explained as follows: A fuel handling supervisor positioned at the boundary of the pool directed the fuel handler and the spotter (a second fuel handler) on the fuel handling bridge to grapple a specific fuel assembly from the spent fuel pool racks that was approved by Procedure NUC-212 and listed on Form RFO-106-2. Before grappling, all three individuals concurred that the grapple was above the correct spent fuel assembly based on the rack location. All three individuals had the spent fuel rack map, the canister map, and Form RFO-106-2 on hand. Once the correct position was agreed upon, the fuel assembly was grappled and lifted from the rack. Prior to placing the fuel assembly into the canister, the assembly was inspected on all four sides for damage using an underwater camera. The camera was connected to a recording device and the fuel assembly inspection was recorded. Once the fuel assembly had been moved to above the canister, the supervisor directed the two fuel handlers on which fuel basket cell in the canister to place the fuel assembly into. Once above the correct cell, all three would concur that the fuel assembly could be lowered into the canister and disengaged. When all the fuel assemblies were in the canister, a core performance engineer and a quality control inspector would use the underwater camera to record that each fuel assembly was in the correct location and correct orientation in the canister by viewing the serial numbers of each fuel assembly.

The core performance engineer and quality control inspector then signed Form RFO-204-3 "MPC Loading Verification" verifying that all the fuel assemblies were in the proper orientation. The quality control inspectors check was the independent verification.

Both the recording of the serial numbers and the visual inspection for damage were kept with the associated sheets from Procedures NUC-212, RFO-106, and RFO-204 and stored in the licensees records system for the life of the of ISFSI.

A dummy fuel assembly was loaded into the canister for the demonstration on July 7, 2011, per Procedure RFO-302. The dummy fuel assembly was inserted into and out of the four corners of the canister to demonstrate that the fuel handling machine could access all 32 cells. To demonstrate that a fuel assembly could be grappled from inside the canister to perform offloading, the dummy fuel assembly was un-grappled during one of the four insertions, the crane off-set, and then the fuel assembly re-grappled and removed from the canister .

Documents (a) Procedure NUC-212 Spent Fuel Limits for Dry Cask Operations, Revision 1 (b)

Reviewed: Procedure RFO-204 Verification of Core and MPC Loading Patterns, Revision 14 (c)

Procedure RFO-106 Development and Implementation of Fuel Shuffle Sequence Plans, Revision 18 (d) Procedure RFO-302 Handling of Fuel Assemblies, Revision 12

.

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Category: Dry Run Demonstration Topic: MPC Pressure Test, Drying, and Helium Backfill Reference: CoC 1014, Condition 10. f, g Amendment 7 Requirement: The dry run shall include pressure testing, draining, moisture removal (by vacuum drying or forced helium dehydration, as applicable), and helium backfilling. A mockup may be used for this dry-run exercise. The operation of the supplemental cooling system (SCS)

shall be demonstrated, if applicable.

Observation: The canister pressure testing, draining, moisture removal by forced helium dehydration (FHD), helium backfilling, and supplemental cooling system (SCS) demonstrations were performed on May 31, 2011 through June 2, 2011 on a mock-up canister and were observed by the NRC. All the demonstrations were successfully performed using Procedure DCS-204. The pressure testing was demonstrated using Sections 8.4

"Hydrostatic Testing System Set-Up" and 8.5 "Hydrostatic Testing." The canister draining was demonstrated using Sections 8.7 "FHD System Final Setup," Section 8.8

"FHD Purge," and Section 8.9 "FHD System MPC Blowdown Operations." The forced helium dehydration system was demonstrated for drying the canister using Section 8.10

"FHD System Operation for Phase 1," and Section 8.11 "FHD System Operation for Phase 2." The helium backfill process was demonstrated using Section 8.12 "FHD System Helium Backfill Operations." The supplemental cooling system operations was demonstrated using Section 8.16 "Supplemental Cooling System Operations" and the associated attachments.

Issues identified during the demonstration included: the pressure relief valve was located on the drain side removable valve operator assembly (RVOA) verses the vent side as shown in the Holtec Final Safety Analysis Report (FSAR), Figure 8.1.20; the licensee's tracking database for equipment requiring calibration (MAXIMO) had not included all equipment requiring calibration; several steps in Procedure DCS-204 required clarification; and Procedure DCS-204 did not contain a requirement that the operator pressurizing the canister must have view of the pressure gauge during the hydrostatic test in accordance with ASME Section III Article NB-6411. All open items identified were captured in the licensees corrective action system in Condition Report CR-2011-6393.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (c) Condition Report CR-2011-6393 "NRC Observations During Forced Helium Dehydration Dry Runs," created May 31, 2011 (d) Holtec Report HI-2002444

"Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System,"

Revision 9 Category: Dry Run Demonstration Topic: MPC Removal from Spent Fuel Pool Reference: CoC 1014, Condition 10. e Amendment 7 Requirement: The dry run shall include remote installation of the canister lid and removal of the canister and transfer cask from the spent fuel pool or cask loading pool.

Observation: The dry run to demonstrate remote installation of the canister lid and removal of a canister from the spent fuel pool was performed on July 8, 2011 and was observed by the NRC. The licensee followed Procedure DCS-203, Section 8.6 "Install MPC Lid,"

Section 8.7 "Remove HI-TRAC from Wet Cask Pit," and Section 8.8 "Move HI-TRAC to Dry Cask Pit." During the demonstration, the licensee determined that due to safety Page 25 of 149

concerns, the procedural steps involving the connecting of the drain line to the lid would need to be revised. Procedure DCS-203 was modified as the dry run demonstration proceeded. The licensee utilized a lift yoke with a lift yoke extension to lift the HI-TRAC transfer cask with an empty canister from the lowest level of the wet cask pit to the intermediate level. The wet cask pit was located between the Unit 1 and Unit 2 spent fuel pools. Transfer canals connect the two spent fuel pools to the wet cask pit. The use of the lift yoke extension to lift the cask out of the lowest level of the wet cask pit and onto the intermediate level ensured the cranes hook block would not enter the water.

When the canister was lifted from the lowest level and placed on the intermediate level, approximately 8 feet of water covered the top of the canister. The licensee removed the extension for the lift yoke and connected the canister lid to the lift yoke. The lid was moved over the wet cask pit and the drain line attached to the underside of the lid. The lid was placed on the canister, the transfer cask re-grappled using the lift yoke, lifted up and out of the spent fuel pool, and placed in the dry cask pit. The lift of the canister and HI-TRAC transfer cask out of the pool during this demonstration, while filled with water (but no spent fuel), represented the heaviest lift performed during the dry run of approximately 101 tons. The crane operated smoothly and all operations during the demonstration proceeded without delay.

Originally, the licensee had planned to place the canister lid on the canister while in the lower level of the wet cask pit. This required the use of the yoke extension. With the yoke extension connected to the crane, the bottom of the lid cleared the spent fuel pool edge with only a few inches when being moved from it's storage location to the wet cask pit. Once the lid was over the wet cask pit, the drain line had to be installed, requiring the worker to stretch down under the lid while balancing on the edge of the spent fuel pool. Another option considered was to place a raft in the wet cask pit. But then the worker would be fully under the 5 ton lid while installing the drain line. As an alternative approach (which was incorporated into the revised Procedure DCS-203), the license evaluated placing the lid on the canister while on the intermediate level in the spent fuel pool. Radiation protection staff performed calculations to determine that the eight feet of water between the spent fuel in the canister and the top of the pool water when the canister was setting on the intermediate level in the wet cask pit was equivalent to more steel than the lid represented, and as such would provide adequate shielding while the canister was on the intermediate level. Additional access controls could also be established to keep personnel away from the area. If the canister was staged at the intermediate level for lid installation, the yoke extension was not required and the lid could be raised to a height of several feet above the floor where the drain line could be safely installed. When considering the safety issues related to the workers being under the lid versus the radiation levels that would be encountered, the licensee determined that the safer process was to not use the yoke extension and to place the canister on the intermediate level for lid installation. The NRC inspectors agreed that the eight feet of water would provide adequate shielding and that the issue of worker safety while installing the drain line was of higher priority.

Documents (a) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 Reviewed: .

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Category: Dry Run Demonstration Topic: MPC Transfer to HI-STORM Reference: CoC 1014, Condition 10. h, i Amendment 7 Requirement: The dry run shall include transfer cask upending/downending on the horizontal transfer trailer or other transfer device, as applicable to the sites cask handling arrangement and transfer of the canister from the transfer cask to the overpack.

Observation: The stack-up and transfer of the canister from the HI-TRAC transfer cask to the HI-STORM storage cask was demonstrated on June 30, 2011 using a canister filled with concrete. The stack-up and transfer activities were observed by the NRC. The licensee used Procedure DCS-205, Section 8.2 "Prepare MPC and HI-TRAC for Transfer,"

Section 8.3 "Prepare HI-STORM for MPC Transfer," and Section 8.4 "Transport HI-TRAC/MPC from Dry Cask Pit to HI-STORM and Transfer MPC to HI-STORM." The demonstration included the use of the recently installed seismic restraint. The seismic restraint was a steel framework bolted to the floor of the fuel building at the 810'

elevation that surrounded the HI-STORM storage cask with screw jacks used to secure the HI-STORM in place to prevent movement during a seismic event. The mating device was bolted onto the HI-STORM storage cask with a torque value of 8,500 pounds/square inch (psi). The HI-TRAC transfer cask was then bolted onto the mating device prior to disconnecting the yoke. Once bolted, the yoke was disconnected and slings were used to slightly raise the canister to allow the unbolting and retraction of the bottom lid of the HI-TRAC transfer cask. The lowering of the canister was performed using two 55 ft slings, rated at 100,000 pounds each, configured in a basket hitch. The canister used for this demonstration was filled with concrete to simulate the weight of a loaded canister. The 130 ton crane then successfully lowered the canister from the HI-TRAC transfer cask into the HI-STORM storage cask.

Documents (a) Procedure DCS-205 Stack-up and Transfer of Loaded MPC, Revision 2 Reviewed:

Category: Dry Run Demonstration Topic: MPC Welding and NDE Reference: CoC 1014, Condition 10. f Amendment 7 Requirement: The dry run shall include canister welding and non-destructive examination (NDE) of the canister lid.

Observation: The dry run to demonstrate canister welding and NDE examinations was performed on May 2, 2011 through May 5, 2011 and observed by the NRC. The licensee followed Holtec Procedure HSP-504 to perform the welding on a mock-up canister. The welding demonstration included the welding of the canister lid to shell, the welding of the vent and drain line cover plates, the welding of the plug on the cover plates, the welding of the canister closure ring, and demonstration of the in-line hydrogen monitor. The visual NDE examinations were performed using Procedure HSP-507 and the liquid penetrant examinations were performed using Procedure HSP-506. The licensee successfully demonstrated all required welding and the NDE examinations. There were no indications found on any of the welds completed during the demonstration. Several minor issues were identified during the welding including: Procedure HSP-506 for liquid penetrant examinations did not include how the procedure was qualified as required by ASME Section V Article T-653.1; the visual testing Procedure HSP-507 did not contain the minimum acceptance criteria as required by ASME Section V Article T-980.2; and Procedure HSP-504 did not have instructions to run the hydrogen monitor for at least one Page 27 of 149

hour prior to use as specified in the manufacture's manual for the hydrogen monitors. All issues identified were entered into the licensee corrective action program under Condition Report CR-2011-5573 and adequately resolved.

Documents (a) Holtec Procedure HSP-506 "Liquid Penetrant Examination for MPC Closure Reviewed: Welding," Revision 2 (b) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Welding," Revision 2 (c) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC," Revision 4 (d) Condition Report CR-2011-5573

"NRC Observations During Welding Dry Runs," dated May 2, 2011 Category: Dry Run Demonstration Topic: Placement of HI-STORM on ISFSI Reference: CoC 1014, Condition 10. j Amendment 7 Requirement: The dry run shall include placement of the HI-STORM 100 cask system at the ISFSI.

Observation: The placement of a HI-STORM 100 concrete storage cask on the ISFSI pad was demonstrated on June 21, 2011 and observed by the NRC. A HI-STORM storage cask loaded with a canister filled with concrete simulated a fully load cask and weighted approximately 348,800 pounds (174.4) tons. An actual cask loaded with spent fuel was estimated to weigh 348,275 pounds (174.1 tons). The HI-STORM storage cask was transported using the sites vertical cask transporter (VCT) and moved from outside the fuel handling building to the ISFSI pad. The cask was positioned at location #1 on the ISFSI pad in accordance with Attachment 10.1.9 "ISFSI General layout" of Procedure DCS-201. The NRC inspectors followed the transporter, which traveled 0.4 miles per hour, down the length of the heavy haul path and onto the ISFSI pad. The transport took approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from pickup to set-down. The distance traveled from the fuel building to the ISFSI was approximately 3/4 of a mile. During the transportation activities, the ISFSI project manager performed the fire hazardous analysis walk down and identified to the NRC inspectors where all the fire hazardous material was located.

The fire hazards observed were consistent with those that were identified in the Fire Hazards Analysis, Figure 1 Comanche Peak ISFSI Hazard Equipment Plan. After placement of the cask onto the ISFSI pad, the HI-STORM storage cask was returned to the plant protected area along the heavy haul path to demonstrate that a cask could be retrieved from the ISFSI and returned to the plant, should retrieving the spent fuel be necessary.

Documents (a) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 Reviewed: (b) Comanche Peak Document No. 13769701-R-M00002 "Fire Hazards Analysis,"

Revision 0 Category: Dry Run Demonstration Topic: Placement of MPC in Spent Fuel Pool Reference: CoC 1014, Condition 10. a, b Amendment 7 Requirement: The dry run shall include moving the canister and the transfer cask into the spent fuel pool or cask loading pool and preparing the HI-STORM 100 cask system for fuel loading.

Observation: The placement of the HI-TRAC transfer cask containing an empty MPC canister into the spent fuel pool wet cask pit was performed by the licensee the week of July 6, 2011. The cask had been placed into the spent fuel pool on two occasions. On July 8, 2011, the Page 28 of 149

NRC observed the removal of the cask from the spent fuel pool to the dry cask pit. The NRC inspectors had toured the spent fuel pool area earlier and had observed some of the practice activities associated with placement of the cask into the spent fuel pool, but had not watched the entire evolution. Discussions were held with the licensee and procedures reviewed concerning the process to place a canister into the spent fuel pool, the use of the yoke extension to avoid the crane hook and block from being lowered into the pool water, and activities associated with loading spent fuel into the canister. The use of the yoke and yoke extension to place the canister into the spent fuel pool was demonstrated during the removal of the canister for the NRC observers. The placement of the cask into the spent fuel pool was performed using Procedure DCS-203, Section 8.3

"Transport HI-TRAC from Dry Cask Pit to Wet Cask Pit" and Section 8.4 "Move HI-TRAC into Wet Cask Pit."

Documents (a) Procedure DCS-203 "MPC Handling and Fuel Loading Operations," Revision 3 Reviewed:

Category: Dry Run Demonstration Topic: Unloading a Canister Reference: CoC 1014, Condition 10. k Amendment 7 Requirement: The dry run shall include HI-STORM 100 cask system unloading, including flooding the canister cavity and removing canister lid welds. A mockup may be used for this dry-run exercise.

Observation: The dry run to demonstrate unloading an MPC-32 canister was conducted on May 31, 2011 and was observed by the NRC. Procedure DCS-207, Section 8.5 Reflood System Staging and Section 8.6 MPC Water Reflood were demonstrated. During the dry run, workers aligned the valves and reflooded a mock-up canister with water. The removal of the canister lid welds was demonstrated by providing the NRC with a videotape of a welded lid being removed. The video tape, dated May 2011, demonstrated successful removal of the welds on the vent and port covers and the lid weld of an MPC-32 canister. Removing a lid would be performed in accordance with Procedure DCS-207, Section 8.7 "Cutting MPC Lid Free from Shell" using the equipment demonstrated in the video. The process involved installing a drilling template onto the canister lid and using a drill bit to remove the drain port cover plate and the vent port cover plate. This provided access to the vent and drain ports to allow sampling of the gas inside the canister and reflooding. The video tape then demonstrated the removal of the lid weld using a lathe cutter. The lathe cutter had two cutting tools located 180 degree apart. The video showed the successful removal of the lid weld and the lifting of the lid off the canister.

During the stack-up demonstration on June 30, 2011 and July 1, 2011, a canister filled with concrete to simulate a loaded canister positioned inside the fuel building was retrieved from a HI-STORM storage cask into the HI-TRAC transfer cask to demonstrate that a canister could be successfully returned to the fuel building and prepared for return to the spent fuel pool for unloading. During the dry run demonstration on June 21, 2011, a weighted canister inside a HI-STORM storage cask was transported from the ISFSI pad to the plant's protected area to demonstrate that a canister loaded with spent fuel could be returned to the plant for unloading.

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Documents (a) Procedure DCS-201 "Transporting Loaded and Unloaded HI-STORM," Revision 2 Reviewed: (b) Procedure DCS-206 "Transporting and Transferring a Loaded MPC for Unloading,"

Revision 2 (c) Procedure DCS-207 "Unloading a Loaded MPC," Revision 2 Category: Emergency Planning Topic: Emergency Drills Reference: 10 CFR Part 50, App E, Section F.1 Published 2011 Requirement: The emergency program shall provide for the training of employees and exercising, by periodic drills, of radiation emergency plans to ensure that employees are familiar with their specific emergency response duties.

Observation: No emergency drills had been conducted at the site specific to the ISFSI. Training had been provided to site personnel on the new emergency action level scheme in November 2010, which included the emergency action level for the ISFSI. Site personnel conducted drills annually related to plant operations that included all aspects of an emergency response that would be applicable to the ISFSI including fire drills, medical response drill, and radiological drills.

Documents (a) Comanche Peak Emergency Plan, Revision 38 Reviewed:

Category: Emergency Planning Topic: ISFSI Emergency Plan Reference: 10 CFR 72.32(c) Published 2011 Requirement: For an ISFSI that is located on the site of a nuclear power plant licensed for operation, the emergency plan required by 10 CFR 50.47 shall be deemed to satisfy the requirements of this section.

Observation: The licensee was using their Part 50 emergency plan for the ISFSI. Emergency Plan, Table 2.1 Initiating Conditions for EAL Classification, had incorporated the ISFSI and identified an unusual event as the classification for a problem at the ISFSI. Procedure EPP-201 had incorporated the ISFSI into the emergency action level scheme and identified damage to a loaded cask confinement boundary as an unusual event. Several other emergency action levels could be related to the ISFSI. A fire, such as a fire involving the transporter, would be classified as an unusual event if the fire occurred at the ISFSI pad (i.e. ISFSI protected area) and lasted more than 15 minutes. If a radiological release occurred from a canister, there were no alert emergency action levels that were appropriate to the ISFSI. However, the site area emergency classification would be reached if radiation levels offsite were measured in excess of 100 mrem. If the levels exceeded 1,000 mrem, a general emergency would be declared. These readings would be based on surveys taken in the field. For a security event, a hostile action in the owner controlled area was an alert. A hostile action within the protected area, including the ISFSI protected area, would be a site area emergency.

Documents (a) Comanche Peak Emergency Plan, Revision 38 (b) Emergency Plan Procedure (EPP)-

Reviewed: 201 Assessment of Emergency Action Levels, Emergency Classification, and Plan Activation, Revision 12

.

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Category: Emergency Planning Topic: Offsite Emergency Support Reference: 10 CFR 72.122(g) Published 2011 Requirement: The ISFSI must provide for accessibility to the equipment of onsite and available offsite emergency facilities and services such as hospitals, fire and police departments, ambulance services, and other emergency agencies.

Observation: The ISFSI was accessible to emergency agencies including fire, medical and security support from offsite agencies. Provisions were incorporated into the emergency planning program for the Comanche Peak site that included arrangements with offsite emergency organizations including provisions for fire protection support, emergency transport of an injured and contaminated person, and treatment of a contaminated person at the Lake Granbury Medical Center or Texas Health-Cleburne. Arrangements were also made with the Somervell County Sheriffs Department to provide assistance. These agencies had participated in past site emergency exercises at Comanche Peak.

Documents (a) Letter (CP-201001106) from David Fuller, Luminant Power to Mark Crawford, Reviewed: Somerville County Fire Chief with Somervell County Volunteer Fire, Rescue, and Emergency Medical Support (EMS) entitled Annual Review of Agreement Letters, dated August 10, 2010 (b) Letter from Stephen Willis, Somervell County Assistance Chief to David Fuller, Emergency Planning Manager, Luminant Power providing the memorandum of understanding between Somervell County Fire Department and the Comanche Peak Nuclear Power Plant, dated January 1, 2010 (c) Letter from Greg Doyle, Somervell County Sheriff to Luminant Power entitled EOC Operations Assistance Agreement, dated June 2, 2011 (d) Letter from David Fuller, Emergency Planning Manager, Luminant Power to Somervell County Sheriffs Department entitled Annual Review of Agreement letters, dated May 18, 2011 (e) Letter (CP-201001110) from David Fuller, Luminant Power to Blake Kretz, Texas Health-Cleburne entitled Annual Review of Agreement Letters, dated August 10, 2010, with attached letter from Blake Kretz, Texas Health-Cleburne to David Fuller, Luminant Power providing the memorandum of understanding between Texas Health-Cleburne and Luminant Power, dated November 4, 2009 (f) Letter (CP-201001103) from David Fuller, Luminant Power to David Orcutt, Lake Granbury Medical Center entitled Annual Review of Agreement Letter, dated August 10, 2010, with attached letter from David Orcutt, Lake Granbury Medical Center to Matt Bozeman, Luminant Power providing the memorandum of understanding between Lake Granbury Medical Center and Luminant Power, dated November 26, 2007 Category: Fire Protection Topic: Fire Accident Response Reference: FSAR 1014, Section 11.2.4.4 Revision 9 Requirement: Upon detection of a fire adjacent to a loaded HI-TRAC transfer cask or HI-STORM storage cask, the ISFSI operator shall take the appropriate immediate actions necessary to extinguish the fire. Fire fighting personnel should take appropriate radiological precautions, particularly with the HI-TRAC as the pressure relief valves may have opened and water loss from the water jacket may have occurred resulting in an increase in radiation dose. Following the termination of the fire, a visual and radiological inspection of the equipment shall be performed.

Observation: The site fire brigade would respond to a fire involving the cask if the cask is still in the Page 31 of 149

plant protected area of Unit 1 & 2. The site fire brigade would not leave the plant protected area to respond to a fire involving the cask on the heavy haul path or at the ISFSI, as this would leave the plant protected area without a fire brigade. When the cask was being transported on the heavy haul path or being placed in the ISFSI, a fire watch with fire extinguishers accompanied the cask. This requirement was specified in Procedure DCS-201, Attachment 10.1.3 Loaded HI-STORM Pre-Movement Checklist, as Step 1.3. If the fire watch was not able to handle the fire, the fire watch personnel notified the control room and the Somervell County Fire Department was called. An agreement was in place between Luminant and the Somervell County Fire Department to provide fire fighting services outside the plant protected area. If a fire were to occur at the ISFSI during storage operations, the Somervell County Fire Department would be requested to respond to the fire.

Procedure DCS-301, Section 8.9 Damage to a HI-STORM/HI-TRAC/MPC as a Result of a Fire, provided instructions for responding to a fire emergency involving a loaded cask. The control room was notified, personnel were directed to fight the fire, and radiological protection was requested to perform a radiological survey and visual inspection to determine the extent of damage to the cask. If the fire involved the HI-TRAC transfer cask and had caused water to be released from the relief valve on the HI-TRAC water jacket, then the water jacket was refilled. As necessary, temporary shielding was placed around the cask to reduce radiation dose rates. Procedure DCS-301 noted that supplemental cooling may be required to comply with Technical Specification 3.1.4. This would be applicable to the HI-TRAC transfer cask.

Documents (a) Letter (CP-201001106) from David Fuller, Luminant Power to Mark Crawford, Reviewed: Somerville County Fire Chief with Somervell County Volunteer Fire, Rescue, and Emergency Medical Support (EMS) entitled Annual Review of Agreement Letters, dated August 10, 2010 (b) Letter from Stephen Willis, Somervell County Assistance Chief to David Fuller, Emergency Planning Manager, Luminant Power providing the memorandum of understanding between Somervell County Fire Department and the Comanche Peak Nuclear Power Plant, dated January 1, 2010 (c) Procedure DCS-301 Dry Cask Storage Equipment Malfunction, LOOP, LOCA, and Contingencies Guidance, Revision 0 Category: Fire Protection Topic: Fire and Explosion Hazards Analysis Reference: CoC 1014, App. B.3.4.5; FSAR 1014, Sect 2.2.3.3 Amendment 7/Revision 9 Requirement: The potential for fire or explosion shall be addressed, based on site specific considerations. This includes the condition that the onsite transporter fuel tank will contain no more than 50 gallons of diesel fuel while handling a loaded storage cask or transfer cask.

Observation: A Fire Hazards Analysis was completed for the ISFSI and the heavy haul path to evaluate potential site specific fire and explosion hazards. The analysis considered potential hazards beginning at the fuel building, along the heavy haul path, and at the ISFSI. The analysis considered a fire associated with the vertical cask transporter and the prime mover, brush fires from the nearby vegetation along the heavy haul path, fires and explosions from nearby facilities, and fires from nearby parked vehicles. Twenty-two nearby facilities were identified along the heavy haul path and ISFSI that were Page 32 of 149

included in the analysis.

There were no storage facilities near the heavy haul path that included explosives or flammable hazards that could create a significant hazard. The Fire Hazards Analysis, Section 7.2.1.3 included a table of the fire and explosive hazards along the heavy haul path. Section 7.2.1.4 evaluated the hazards including the underground diesel oil storage tank located 130 feet from the heavy haul path, the 2,000 gallon waste oil tank 650 feet away, hydrogen bulk storage tanks 1,085 feet away, several transformers at varying distances, the paint shop 205 feet away, chemical storage building 480 feet away, the 4,000 gallon gasoline storage tank 350 feet away, the 8,000 gallon diesel storage tank 360 feet away, transient combustibles that could be in the vicinity of the heavy haul path, and the affects of a nearby brush fire. Either the amount of flammable, combustible or explosive material produced a hazard less than the design basis accidents analyzed in the Holtec Final Safety Analysis Report (FSAR), Section 4.6.2.1 Fire Accidents or the material was far enough away from the heavy haul path that the impact was bounded by the accidents that could be associated with the vertical cask transporter fire. Most of the trees and shrubs had been cleared away from the heavy haul path to reduce the fire hazard.

The prime mover was used to pull the loaded cask out of the fuel building. Once out of the fuel building, the cask was lifted by the vertical cask transporter for movement to the ISFSI. A 50 gallon limit applied to the prime mover and transporter and had been evaluated in the FSAR, Section 4.6.2.1 Fire Accidents and found to be an acceptable limit for the prime mover and transporter. The prime mover fuel tank had a capacity of 60 gallons of diesel. Procedure DCS-201, Step 6.8 required that the fuel level in the prime mover be verified as less than 3/4 full as part of the pre-operational checks. The transporter fuel tank capacity was 45 gallons. This specification was listed in the vertical cask transporter Operations and Maintenance Manual, Section 7.7.

There were other combustibles associated with the prime mover and vertical cask transporter including tires, lubricating oils, and hydraulic fluid. The transporter had a high potential fire loading, mainly due to the tires and hydraulic fluid reserve. The transporter had a hydraulic fluid reservoir with a capacity of 80 gallons. The Vertical Cask Transporter Operations and Maintenance Manual, Section 7.7 specified the hydraulic fluid to be Quintolubric 888-46. The ISFSI Fire Hazards Analysis, Section 7.2.1.1 identified that the Quintolubric 888-46 synthetic, fire resistant hydraulic fluid met the Factory Mutual Research Approval Standard 6930 for less flammable hydraulic fluids. This type of liquid was self-extinguishing once the ignition source of the fire was removed. The Quintolubric 888-46 had a flash point of 572 degree F compared to diesel fuel which has a flash point of 120 degree F.

Holtec provided Comanche Peak with an analysis (Holtec Report HI-2084156) that had been performed for the transporter used at the Fermi plant. This transporter was identical to the one at Comanche Peak. The Fermi analysis considered 16 tires, 45 gallons of diesel fuel in the diesel tank, 80 gallons of hydraulic fluid, and several gallons of various other flammables such as drive motor oils, steering motor oils, gear pump oil, etc. The quantities of flammable fluids were taken from Section 7.7 Fluids and Capacities, of the Vertical Cask Transporter Operations and Maintenance Manual. The Page 33 of 149

analysis determined that there could be an equivalent of 555 gallons of combustibles associated with the transporter. The tires were the highest fire loading estimated to be equivalent to 422 gallons. The acceptance criteria for the analysis was to demonstrate the cladding temperature of the fuel in the canister would remain below the 1058 degree F accident limit. This limit was stated in the Holtec FSAR, Section 2.0.1 "MPC Design Criteria" and listed in Table 2.0.1 MPC Design Criteria Summary as the maximum permissible fuel cladding temperature limit for off-normal and accident events. The Holtec calculations determined that the fire involving 555 gallons of combustibles, lasting for 40 minutes, would raise the temperature of the spent fuel inside the cask by only 12.2 degree F. When assuming a peak cladding temperature of 711 degree F under normal storage operations, the 12.2 degree F would not raise the cladding temperature to a level of concern.

The prime mover also had combustible hydraulic fluids, but a smaller amount than the transporter. The prime mover connected to the cask using a 6 foot tow bar. A fire involving the prime mover was evaluated in the Fire Hazards Analysis and found to not exceed the design basis fire evaluated in the FSAR Section 4.6.2.1.

Section 6.3 Wildland Fires, of the Fire Hazards Analysis provided a discussion of the potential impact of a fire due to trees and shrubs along the heavy haul path. Douglas fir trees were the predominant trees along the heavy haul path. The licensee had removed the majority of trees and shrubs along the heavy haul path. Those that remained were found to have minimal impact on the loaded cask being moved on the heavy haul path.

The ISFSI pad had a wide area that was cleared of shrubs and trees to comply with security requirements.

At the ISFSI, a fire involving the transporter was the bounding fire. An electrical equipment building and associated step down transporter, located north of the electrical building, were analyzed. The electrical building was a concrete block structure with a poured concrete roof and contained only electrical equipment and cables. The transformer located nearby had 170 gallons of mineral oil. The electrical building was between the transformer and the casks, with the nearest cask approximately 70 feet away. Procedures STA-729 and WCI-606, Attachment 8.J DCS and ISFSI Special Requirements, restricted flammables and combustibles from being taken into the ISFSI without proper authority. A sign was posted on the ISFSI fence that stated Access Through This Fence Is Controlled In Accordance With WCI-606, Attachment 8.J."

Section 10 of the Fire Hazards Analysis identified eleven issues that required administrative controls in order to reduce the risk of a fire or explosion that could adversely affect a loaded HI-STORM cask along the heavy haul path or at the ISFSI.

Condition Report CR-2011-6459 was initiated to track the resolution of the first ten items. The original ten issues included such items as establishing procedures and processes to control combustibles around the cask during movement on the heavy haul path, establishing a fire watch during movement of the cask including availability of fire extinguishers, walking down the heavy haul path prior to moving the cask, not moving the cask if severe weather was predicted, and ensuring arrangements were established for offsite fire support. The eleventh item, which was new in Revision 1 of the Fire Hazards Analysis, required verification that the sprinklers and fire alarms in the RP building Page 34 of 149

along the heavy haul path were operational. The eleven issues were resolved and actions incorporated into Procedure DSC-201, including Attachment 10.1.3 Loaded HI-STORM Pre-Movement Checklist and Procedure STA-729.

Examples of procedural controls in Procedure DCS-201 related to implementing fire controls during cask movement included Step 5.14 which required vehicles operated within 55 feet of the loaded cask to be continuously attended. Step 5.15 required vehicles within 45 feet of a loaded cask to be limited to 50 gallons of flammable or combustible liquids. Step 5.16 required refueling of the transporter or prime mover to be limited to a maximum of a 5-gallon container when transporting a loaded cask. Step 6.7, under Precautions, required a walk-down of the areas involving dry cask storage activities be performed to ensure compliance with transient combustible stand-off zones as defined in Procedure STA-729. Step 6.1.14 of Procedure STA-729 stated that combustible and explosive material shall not be stored within the ISFSI site boundary.

Combustible and explosive material allowed to enter the ISFSI site boundary during spent fuel transfer and storage operations shall be controlled to avoid any potential fire that could exceed the ISFSI design basis fire. Step 6.1.15 provided the stand-off distances for gasoline powered and diesel powered vehicles. The table was consistent with Section 9.0 of the Fire Hazards Analysis for stand-off distances which required vehicles with fuel tanks up to 25 gallons to remain at least 35 feet away from the heavy haul path, fuel tanks up to 50 gallons required a minimum stand-off distance of 45 feet and fuel tanks up to 100 gallons were required to remain at least 55 feet away.

Conversations with the ISFSI project manager indicated the distances were measured from the edge of the heavy haul path. Step 6.4.1.1.C of Procedure STA-729 required a continuous fire watch during dry cask storage operations involving either the prime mover or the transporter when a loaded cask was being moved from the fuel building to the ISFSI pad or back. The fire watch was required to have continuous communications with the control room and have fire extinguishers readily available.

Procedure DCS-201, Attachment 10.1.3 Loaded HI-STORM Pre-Movement Checklist, Step 1.3 required a continuous fire watch while moving a loaded cask from the fuel building to the ISFSI pad. The fire watch was required to have fire extinguishers. Direct communications with the control room was required by Step 1.4. Steps 1.6 thru 1.8 required a walk-down of the heavy haul path and the area around the ISFSI prior to movement of the cask to ensure no significant fire hazards or explosion hazards were within 100 feet. This included inspection of the transformers near the fuel building to verify they were not leaking oil and verification that flammables and combustibles were stored properly in any nearby buildings and designated storage areas. Step 1.10 required controls to be established along the heavy haul path to ensure vehicles met the required stand-off distances. A table with stand-off distances was provided following Step 1.10 consistent with the stand-off distances specified in Section 9.0 of the Fire Hazards Analysis. Step 1.11 required that controls be established to ensure delivery trucks hauling bulk quantities of combustibles such as liquid fuels and combustible gases be restricted so that the deliveries would not occur at the same time as loaded casks were being transported to the ISFSI. Step 1.13 provided for a walk-down of the heavy haul path and the establishment of a controlled zone at approximately 100 feet radius around the loaded cask by establishing barricades or personnel assigned to restrict traffic. Step 1.14 restricted vehicles from being fueled at the automobile service terminal while casks Page 35 of 149

were being transported. The automobile service terminal was listed as Item #9 (gasoline)

and Item #10 (diesel) in the Fire Hazards Analysis table of hazardous equipment items.

This facility was 350 feet from the ISFSI heavy haul path and stored up to 4,000 gallons of gasoline and 8,000 gallons of diesel. Movement of a cask was also restricted during potential bad weather. Step 1.2 of Procedure DCS-201, Attachment 10.1.3 required verification of acceptable weather conditions before transporting the cask to the ISFSI.

This included verification that the outside ambient temperature was greater than 0 degree F, the heavy haul path was not covered with snow or ice, and that no tornado, severe storm, or high wind warning existed. The steps listed above for Procedure DCS-201 were also included in Attachment 10.1.3 of Procedure DCS-206, which was used for returning a cask from the ISFSI to the fuel building, should unloading or other type of maintenance be necessary.

A review of the fire hazards along the heavy haul path listed in the Fire Hazards Analysis was discussed with the licensee during the dry run on June 21, 2011, which moved a HI-STORM storage cask from the fuel building to the ISFSI. An enlarged Figure 1 of the Fire Hazards Analysis, which showed the location of the 21 fire hazards along the heavy haul path was used to review each hazard and confirm the accuracy of the Fire Hazards Analysis. Revision 1 of the Fire Hazards Analysis was issued with an updated Figure 1 adding a 22nd hazard. This new addition was two 38 Kw standby power generators located outside the perimeter of the ISFSI pad. No additional fire hazards were observed by the NRC inspectors during the tours of the haul path and adjacent areas.

Documents (a) Procedure DCS-201 Transporting Loaded and Un-Loaded HI-STORM, Revision 2 Reviewed: (b) Procedure DCS-206 Transporting and Transferring a loaded MPC for Unloading, Revision 2 (c) Procedure STA-729 Control of Transient Combustibles, Ignition Sources, and Fire Watches, Revision 8 (d) Procedure WCI-606 Work Control Process, Revision 14 (e) Condition Report CR-2011-6459 Implementation of Administrative Controls Identified in the Fire Hazards Analysis, initiated June 1, 2011(f) Shaw Technical Report 13769701-R-M-00002 Evaluation of Fire Hazards for the ISFSI, Revision 1 (g) Morris Material Handling Manual DOC036304-10 Vertical Cask Transporter Operations and Maintenance Manual Serial # 036304-036309, Revision 2 (h) Holtec Report HI-2084156, Appendix G Evaluation of VCT Hydraulic Fluid and Tires Fire , Revision 5 (i) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Fire Protection Topic: Fire Protection Plan Reference: 10 CFR 50.48(a)(1) Published 2011 Requirement: Each operating nuclear power plant must have a fire protection plan that satisfies Criterion 3 of Appendix A to Part 50. This fire protection plan must describe the overall fire protection program for the facility.

Observation: The Comanche Peak Fire Protection Report described the program and plans for preventing, detecting and suppressing a fire at the Comanche Peak site. A Fire Hazards Analysis was included in the fire protection report that identified the potential fire hazards for the nuclear plants. The fire protection report was supplemented by a Fire Hazards Analysis developed specific to the ISFSI site. The ISFSI specific Fire Hazards Analysis provided a description of the various potential fires/explosions that could occur Page 36 of 149

at the ISFSI and evaluated the fire or explosion effects on the cask to verify that the events were bounded by the Holtec Final Safety Analysis Report (FSAR).

Documents (a) Comanche Peak Nuclear Power Plant Fire Protection Report, Unit 1 and Unit 2, Reviewed: Revision 28 (b) (b) Shaw Technical Report 13769701-R-M-00002 Evaluation of Fire Hazards for the ISFSI, Revision 1 (c) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 Category: Fuel Selection/Verification Topic: Authorized Contents For Storage Reference: CoC 1014, Appendix B, Section 2.1.1 Amendment 7 Requirement: The HI-STORM 100 cask system canister is authorized for storage of fuel assemblies, fuel debris, and non-fuel hardware meeting the requirements of Appendix B, Section 2.1.1 and Tables 2.1-1 through 2.1-8.

Observation: For the initial loading campaign, the licensee planned to limit loading to intact fuel assemblies as listed in Appendix B, Section 2.1.1.a, of Certificate of Compliance 1014.

All spent fuel selected for loading in the first three canisters met the requirements of Appendix B, Section 2.1.1 and the associated tables. Intact fuel was defined in Certificate of Compliance 1014, Appendix A, Section 1.1 "Definitions" and Procedure NUC-212, Section 3.7 as fuel assemblies without known or suspected cladding defects greater than pinhole leaks or hairline cracks and which can be handled by normal means.

Fuel assemblies without fuel rods in fuel rod locations were not considered intact fuel unless dummy fuel rods were used to displace an amount of water greater than or equal to that displaced by the fuel rod.

Comanche Peak Unit 1 had been operating since 1990 and was currently in Cycle 15.

Unit 2 had been operating since 1993 and was currently in Cycle 12. Of the fuel currently stored in the Comanche Peak spent fuel pool, there were 1191 pressurized water reactor (PWR) fuel assemblies from Unit 1 of which 987 were considered intact and 1079 PWR fuel assemblies from Unit 2 of which 1002 were considered intact. The remaining spent fuel elements were restricted from loading and were included in a restricted fuel assembly list (CP-201001255) maintained by the licensee. The licensee performed a detail evaluation of each operating cycle and documented in Evaluation EV-CR-2010-009331-8 the condition of the fuel for the first fourteen cycles of Unit 1 and eleven cycles of Unit 2. The effort involved reviewing Document CP-201001255

"Comanche Peak Unit 1 and Unit 2 Restricted Fuel Assembly List - September 2010" to confirm that these assemblies should also be restricted from inclusion in the dry cask storage project. The fuel performance reports for each unit/cycle were reviewed to verify the relevant information from the reports had been completely and accurately captured in the restricted fuel assembly list. Once verified, the restricted fuel assembly list could be used to determine a complete and accurate list of fuel assemblies that should not be classified as intact fuel assemblies as defined by Certificate of Compliance 1014.

Not all assemblies listed in the restricted fuel assembly list were required to be restricted from loading into a canister. Those assemblies that experienced an underload and only required inspection prior to reuse could be considered for loading into a canister after further inspection. Evaluation EV-CR-2010-009331-8 discussed restricted assemblies and described the issue associated with each fuel assembly. The majority of the restricted assemblies was due to a top nozzle stress corrosion cracking issue which Page 37 of 149

limited the ability to handle these assemblies by normal means until further evaluations and corrective measures, for example pinning, were completed. The remainder of the assemblies were limited due to leaking fuel rods, known presence of debris or other fuel damage which had not been evaluated. Spent fuel planned for loading in the first three casks was Unit 1 fuel assemblies.

Certificate of Compliance 1014, Appendix B, Section 2.1.1.a, specified that intact fuel that met the limits of Table 2.1-1 "Fuel Assembly Limits" could be loaded into the HI-STORM-100 cask system. Table 2.1-1,Section V "MPC Model MPC-32 and MPC-32F" applied to Comanche Peak.Section V.A.1.a allowed zirconium clad fuel to be loaded in the MPC-32 canister and referenced Table 2.1-2 "PWR Fuel Assembly Characteristics."

All the fuel at Comanche Peak was zirconium clad fuel. Table 2.1-2 provided the fuel assembly specifications for the 17 x 17 A/B/C fuel used at Comanche Peak. Table 2.1-2 values had been incorporated into Procedure NUC-212, Attachment 1 "Fuel Classification" for the key parameters. This included the limit of 5% initial U-235 enrichment and the various dimensional limits for the fuel rods and pellets. Procedure NUC-212, Attachment 1 was required to be completed for each selected spent fuel assembly as part of the documentation package to verify the parameters in Appendix B, Table 2.1-1 and Table 2.1-2 were met for the spent fuel that had been selected. None of the other Tables 2.1-3 through 2.1-8 of the Certificate of Compliance applied to the first canisters planned for loading at Comanche Peak. For the three canisters planned for the first loading campaign, initial U-235 enrichment for Canister #1 was 3% to 4%. For Canister #2, U-235 initial enrichment ranged from 1.9% to 4.8%. For Canister #3, U-235 initial enrichment ranged from 3.6% to 4.8%.

Table 2.1-1,Section V.A.1.e through V.A.1.g, provided fuel assembly limits for length, width, and weight. Attachment 1 of Procedure NUC-212 had incorporated limits for these parameters that were more restrictive.Section V.A.1.c.ii, specified limits for post-irradiation cooling time and average burnup per fuel assembly.Section V.A.1.d.ii, specified decay heat limits per fuel storage location. Both of these sections referenced Section 2.4 of Appendix B as the limits for the 17 x 17 spent fuel. A detail comparison of these limits to the Comanche Peak spent fuel is provided in the Inspector Notes of this inspection report under the Category: Fuel Selection/Verification and the Topic: Decay Heat, Burnup, and Cooling Time Limits.

Documents (a) Procedure NUC-212, Spent Fuel Limits For Dry Cask Operations, Revision 1 (b)

Reviewed: Restricted Fuel Assembly List (RFAL) CP-201001255 "Comanche Peak Restricted Fuel Assembly List," dated September 2010 (c) Form NUC 212-2 "Assembly Selection Planner for Heat and Associated Limitations Tracking," for Fuel Campaign UFO-01, dated June 1, 2011 (d) Form NUC-212-3 "Cask Acceptability Report and Comprehensive Assembly Specifications Supplement," for Canisters No. 1, 2, and 3, provided August 10, 2011 (e) Form NUC 212-4 "Cask Acceptability Report," for Canisters No. 1, 2, and 3, provided August 10, 2011 (f) Form NUC-212-5 "Comprehensive Assembly Specifications Supplement," for all 96 assemblies planned for loading into Canisters No.

1, 2, and 3, provided on August 10, 2011 (g) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 (h)

Condition Report Evaluation CR-2010-009331-8 "Evaluation of CPNPP Fuel Inventory to Create Restricted List," created October 12, 2010 Page 38 of 149

Category: Fuel Selection/Verification Topic: Damaged Fuel Classification Reference: FSAR 1014, Table 1.0.1; ISG-1, Rev. 2 Revision 9 Requirement: A damaged fuel assembly is a fuel assembly with known or suspected cladding defects, as determined by review of records, greater than pinhole leaks or hairline cracks, empty fuel rod locations that are not replaced with dummy fuel rods, missing structural components such as grid spacers, whose structural integrity has been impaired such that geometric rearrangement of fuel or gross failure of the cladding is expected based on engineering evaluations, or those that cannot be handled by normal means. Fuel assemblies that cannot be handled by normal means due to fuel cladding damage are considered fuel debris.

Observation: For the first loading campaign, only intact fuel assemblies will be loaded for storage at the ISFSI. Procedure NUC-212, Section 3.0 "Definitions/Acronyms" defined intact fuel assemblies as "fuel assemblies without known or suspected cladding defects greater than pinhole leaks or hairline cracks and which can be handled by normal means. Fuel assemblies without fuel rods in fuel rod locations shall not be classified as intact fuel assemblies unless dummy fuel rods are used to displace an amount of water greater than or equal to that displaced by the fuel rod." The definition in Procedure NUC-212 was identical to that in the Holtec Final Safety Analysis Report (FSAR), Table 1.0.1

"Terminology and Notation." The licensee's software used to determine the acceptability of the spent fuel assemblies selected for placement in a canister had been hard coded to accept only intact fuel assemblies. Future loading campaigns will address the handling of damaged fuel assemblies and the use of damaged fuel containers.

Documents (a) Procedure NUC-212 Spent Fuel Limits for Dry Cask Operations Revision 1 (b)

Reviewed: Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Fuel Selection/Verification Topic: Decay Heat, Burnup & Cooling Time Limits Reference: CoC 1014, License Condition 6, App. B, Sect. 2.4 Amendment 7 Requirement: Fuel assemblies stored in the HI-STORM 100 cask system canister must meet the decay heat, burnup and cooling time limits specified in Appendix B, Section 2.4 of the Certificate of Compliance.

Observation: Spent fuel selected for loading in the first three canisters had been selected in accordance with the limitations specified in Certificate of Compliance 1014, Appendix B, Section 2.4 "Decay Heat, Burnup, and Cooling Time for Zirconium Clad Fuel." Procedure NUC-212 had incorporated the requirements from Section 2.4 and described the licensees overall approach for selecting spent fuel in compliance with the requirements. Procedure NUC-212, Step 5.6 stated that the numerical limits and data tables from Certificate of Compliance 1014 and the Holtec Final Safety Analysis Report (FSAR) were hard coded into the computer code utilized by the licensee for determining individual fuel assembly qualification for loading. The licensee had developed their own computer code, given the name TARPIT (Thermal Assembly Repository Pad Inventory Tracker), for calculating decay heat values and conducting other analysis as required. For cask loading, all assemblies were required to be intact. Storage of non-intact fuel was not supported by the licensees current analysis code. The decay heat load calculation in Page 39 of 149

TARPIT was based on methods and equations in NRC Regulatory Guide 3.54. The licensee had performed an independent validation and verification of the TARPIT code using Procedure NUC- 006 "Control of Computer Programs." This included a code verification review conducted by an independent verification engineer and an independent reviewer. A verification test report was issued. The verification test report was very detail and tested the numerous parameters associated with the TARPIT computer code. Numerous test cases using data that would not meet the Certificate of Compliance 1014 criteria were tested to verify that TARPIT would reject the fuel assembly. This included examples of decay heat levels that were within allowed limits but were too high for the region of loading proposed, assemblies that were not intact and therefore on a restricted list, and assemblies with inadequate cooling times. The code performed as required for these cases. The NRC inspector provided additional examples of input to test the licensee's software. The data was run using TARPIT and the examples that should be rejected were successfully recognized by the computer code as not complying with the Certificate of Compliance 1014 requirements.

Procedure NUC-212, Section 8.3 "Cask Load Planning" outlined the use of the Assembly Selection Planner for Heat and Associated Limitations Tracking (ASPHALT Report)

which documented a listing of acceptable candidate assemblies and also noted which assemblies required a pre-load inspection. Proposed cask loadings were reviewed by the licensee's core performance engineer and results presented in a Cask Acceptability Report and Comprehensive Assembly Specifications Supplement Report (Forms NUC-212-3, 212-4, and 212-5). This report outlined the software and data files used to produce the report. The report also contained a listing of the acceptance criteria outlined in Procedure NUC-212 and reasons for acceptance or rejection of the proposed cask loading configuration. Form NUC-212-4 provided a graphical drawing (map) of the canister internal fuel basket cells showing the preselected spent fuel assemblies by fuel ID number assigned to each of the locations. Form NUC-212-5 provided an individual data sheet for each of the 32 fuel assemblies proposed for loading in the canister that provided detail information for the assembly including fuel history, burnup, enrichment, cooling time, fuel classification, spacer requirement, decay heat, and the canister location assigned to the assembly.

The licensee planned to load fuel in the canisters using the regionalized fuel loading concept allowed in Certificate of Compliance 1014, Appendix B, Section 2.1.3

"Regionalized Fuel Loading" and Section 2.4.2 "Regionalized Fuel Loading Decay Heat Limits for Zirconium Clad Fuel. This concept utilized an inner Region 1 and an outer Region 2 and limited the total canister heat load to 34 kW. Region 1 provided the option to place hotter spent fuel in the interior basket locations to provide shielding by the cooler assemblies placed in the outer Region 2. Appendix B, Table 2.4-2 "Fuel Storage Regions per MPC" listed the number of locations in the canister that could be allocated for each of the two regions. For the MPC-32 canister, 12 locations could be assigned to Region 1 and 20 assigned to Region 2. The licensee had incorporated this limitations into Procedure NUC-212, Step 5.9. Certificate of Compliance 1014, Appendix B, Section 2,4.2 "Regionalized Fuel Loading Decay Heat Limits for Zirconium Clad Fuel" provided four equations to determine the maximum allowable decay heat per fuel storage location. The equations involved the maximum allowed heat load for the canister (34 kW), the number of inner (12) versus outer (20) storage locations used for the Page 40 of 149

regionalized concept, and a ratio factor between the decay heat of the inner and outer region. This ratio was referred to as "X". Section 2.4.2 limited the value of X to between 0.5 to 3. Procedure NUC-212, Step 8.3.3 was more conservative and limited X to between 1 and 3. The higher the value of X, the greater the differential in decay heat allowed between the inner and outer regions. An X value of 1 resulted in the inner and outer regions having the same decay heat limit. An X factor of 3 resulted in the inner region decay heat limit being three times that of the outer region. The value of X was selected by the licensee. For the first three canisters, the licensee, in consultation with Holtec, selected an X value of 1.3. This value provided for a slightly hotter Region 1 than the fuel in Region 2. Based on this selection of X, the decay heat limit calculated for the outer region (Region 2) was 0.927 kW. The decay heat limit calculated for the inner region (Region 1) was 1.205 kW. Form NUC-212-4 was reviewed for the first three canisters to be loaded. Each form listed the correct Region 1 and Region 2 limits and listed the highest values for the spent fuel assemblies selected for each of the regions. For Canister #1, the highest fuel assembly assigned to Region 1 had a decay heat of 0.579 kW. For Region 2, the fuel assembly with the highest decay heat was 0.515 kW. For Canister #2, the highest assemblies were 0.883 kW for Region 1 and 0.708 kW for Region 2. For Canister #3. the highest assemblies were 0.996 kW for Region 1 and 0.772 kW for Region 2. The overall heat load for the entire canister was limited by the Certificate of Compliance 1014, Appendix B, Section 2.4.2 to 34 kW. The total heat load for each canister was calculated as part of Form NUC-212-4. For Canister

1. 1, the total heat load was 16.07 kW, for Canister #2 the total heat load was 21.39 kW, and for Canister #3 the total heat load was 23.338 kW.

Certificate of Compliance 1014, Appendix B, Section 2.4.3 "Burnup Limits as a Function of Cooling Time for Zirconium Clad Fuel" provided the equations for calculating the burnup limit per fuel assembly allowed in the canister. Burnup was a function of minimum cooling time, maximum decay heat, and minimum average enrichment. Appendix B, Section 2.4.3.4 limited the calculated burnup value to 65,000 megawatt-days (MWD) per metric ton uranium (MTU) for PWR fuel assemblies.

Procedure NUC-212, Step 5.4 limited the selection of acceptable spent fuel assemblies for storage in the Comanche Peak canisters to a burnup value of less than 50,000 MWD/MTU. The equation in Appendix B, Section 2.4.3 incorporated a number of parameters that were provided in Table 2.4-3 "PWR Fuel Assembly Cooling Time-Dependent Coefficients." The calculation also depended on the decay heat limits calculated for Region 1 and 2 discussed above. Each of the two regions had a different burnup limit based on the differences in the two decay heat limits. The burnup limit for each individual spent fuel assembly was calculated and provided on Form NUC-212-5.

Each of the 32 assemblies had a separate Form NUC-212-5 and a burnup value calculated for the individual assembly based on it's initial average U-235 enrichment, which region (1 or 2) the fuel assembly would be placed in, the fuel classification of 17 x 17 A, B or C, and the cooling time. The software then selected the fuel assembly with the smallest delta between the actual burnup value and the burnup limit, by region, and listed the Region 1 and Region 2 values on Form NUC-212-4. For Canister #1, Form NUC-212-4 listed the fuel assembly with the smallest delta value in Region 1 as having a burnup value of 38,753 MWD/MTU. A review of the Form NUC-212-5 sheets for the 12 assemblies assigned to Region 1 verified that this was the smallest delta value and applied to Fuel ID No. F32 placed in fuel storage location 19 of the fuel basket. The Page 41 of 149

delta value was 26,037 MWD/MTU based on a calculated limit for the particular fuel assembly of 64,790 MWD/MTU. For Region 2 of Canister #1, fuel storage location 32 had the spent fuel assembly with the smallest delta value between the burnup limit and the calculated burnup for the 20 assemblies assigned to Region 2. Fuel ID No. D47 was in this fuel slot location with a burnup of 32,199 MWD/MTU. The limit for this fuel assembly was calculated to be 49,788 MWD/MTU. This gave a delta value of 17,589 MWD/MTU. A review of the individual NUC-212-5 forms for all three canisters was completed to determine the fuel assembly in each of the three canister with the highest burnup value. For Canister #1, the highest burnup value of any of the 32 assemblies was 38,753 MWD/MTU. For Canister #2, the highest burnup value was 42,596 MWD/MTU. For Canister #3, the highest burnup value was 42,724 MWD/MTU. All were below the 50,000 MWD/MTU limit specified in Procedure NUC-212, Step 5.4.

The burnup limit for individual spent fuel assemblies was provided on each of the 32 NUC-212-5 forms. This limit was calculated using the equation provided in Certificate of Compliance 1014, Appendix B, Section 2.4.3 and the associated tables. A burnup limit for each fuel assembly was calculated based on which region the fuel was scheduled to be placed, the cooling time, and the minimum fuel assembly average enrichment. Several individual assemblies were selected for review by the NRC inspectors and the calculations manually performed to verify the burnup limits listed on Form NUC-212-5 were consistent with the values derived from the equation in Appendix B, Section 2.4.3. The licensee used several conservative factors as described in Procedure NUC-212, Step 5.11 in their calculations. These included subtracting 0.05 from the initial average enrichment used in the calculations and not applying the option to linearly extrapolate between the cooling time values on Table 2.4-3. Instead, the actual cooling time for the fuel assembly was rounded down to the nearest whole year.

This produced a more conservative calculation. The licensee also applied a 5%

conservative factor to the calculated burnup limit by using 0.95 times the calculated burnup value as the limit listed on Form NUC-212-5.

The licensee described the conservatism used in the calculations for determining the acceptability of individual spent fuel assemblies to be accepted for placement into the Comanche Peak canisters in Procedure NUC-212, Step 5.11. Step 5.11.A stated that a 5% uncertainty was applied to the burnup calculations. The burnup of the individual spent fuel assemblies equaled the measured burnup x 1.05. Step 5.11.B stated that when utilizing the central zone enrichment data (to compare to the limits in Appendix B, Table 2.1-2 for initial enrichment) and for determining the spent fuel boron concentration limits, 0.05 was added to the central zone enrichment value to account for uncertainty.

Step 5.11.C stated that average initial enrichment, as opposed to central zone enrichment values, were used to determine the decay heat burnup limits for Appendix B, Section 2.4.2 calculations. The average initial enrichment plus 0.05 was used to account for uncertainty. This uncertainty was not used when calculating the decay heat values in accordance with Regulatory Guide 3.54.

A review of the associated Forms NUC-212-4 and NUC-212-5 for all three canisters found that the licensee's software correctly determined that the spent fuel selected was in compliance with the burnup limits of the Certificate of Compliance.

Page 42 of 149

Documents (a) Procedure NUC- 212 Spent Fuel Limits For Dry Cask Operations, Revision 1 (b)

Reviewed: Verification Test Report "TARPIT, Rev 0," Revision 0 (c) Procedure NUC- 006-3 Verification Test Report, Revision 0 (d) Form NUC 212-2 "Assembly Selection Planner for Heat and Associated Limitations Tracking," for Fuel Campaign UFO-01, dated June 1, 2011 (d) Form NUC-212-3 "Cask Acceptability Report and Comprehensive Assembly Specifications Supplement," for Canisters No. 1, 2, and 3, provided August 10, 2011 (e) Form NUC-212-4 "Cask Acceptability Report," for Canisters No. 1, 2, and 3, provided August 10, 2011 (f) Form NUC-212-5 "Comprehensive Assembly Specifications Supplement," for all 96 assemblies planned for loading into Canisters No.

1, 2, and 3, provided on August 10, 2011 (g) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 (h)

NRC Regulatory Guide 3.54 "Spent Fuel heat Generation in an Independent Spent Fuel Storage Installation," Revision 1 Category: Fuel Selection/Verification Topic: Fuel Loading Error Reference: CoC 1014, Appendix B, Section 2.2 Amendment 7 Requirement: If any loading condition of Appendix B, Section 2.1 is violated, the affected fuel assemblies shall be placed in a safe condition, the NRC Operations Center shall be notified within 24 hrs, and a special report describing the cause of the violation and actions taken to restore compliance and to prevent recurrence shall be submitted to the NRC within 30 days.

Observation: Procedure NUC-212, Section 8.6 "Violations" had incorporated the notification requirements specified in Certificate of Compliance 1014, Appendix B, Section 2.2 to place the affected fuel assemblies in a safe condition, notify the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and submit a special report within 30 days describing the cause of the violation and actions taken to restore compliance and prevent recurrence.

Documents (a) Procedure NUC-212 Spent Fuel Limits For Dry Cask Operations, Revision 1 Reviewed:

Category: Fuel Selection/Verification Topic: Fuel Spacers Reference: FSAR 1014, Tables 2.1.9 Amendment 7 Requirement: Each user shall specify the fuel spacer length based on fuel assembly length and the presence of a damaged fuel container, allowing an approximate 2 to 2 1/2 inch gap underneath the canister lid. Fuel spacers shall be sized to ensure that the active fuel region of intact fuel assemblies remains within the neutron poison region of the canister basket.

Observation: The licensee had incorporated the fuel spacer requirements in a note in Procedure NUC-212 following Step 8.3.6. The note provided specific lengths for the upper and lower spacers and categorized them as "A" for spent fuel assemblies not containing an insert,

"B" for assemblies containing a burnable poison rod assembly (BPRA) or plug, "C" for assemblies containing a rod cluster control assembly (RCCA), and "D" for assemblies containing a neutron source. All lower spacers were 5 inches. The required upper spacers ranged from 3.75 inches to 14 inches in length based on which type of assembly (A through D) was placed in the canister basket fuel storage location in order to meet the 2 to 2 1/2 inch gap underneath the canister lid and ensure that the active fuel region Page 43 of 149

remained within the neutron poison region of the canister basket. Procedure NUC-212, Section 4.0 "References" listed the Holtec Document 1937-4 "Fuel Spacer Lengths for CPNPP," Revision1, which provided the specific fuel spacer lengths for the Comanche Peak spent fuel assemblies. Attachment A to Holtec Document 1937-4 stated "The fuel basket and fuel assemblies are assumed to move independently in the axial direction due to axial clearance inside the canister cavity. Upper and lower fuel spacers are sized to keep the active fuel region within the neutron absorber region under the worst case alignment scenarios. Another important dimension to consider when sizing fuel spacers is the clearance under the canister lid. This clearance must be approximately 2 to 2 1/2 inches to allow for potential fuel assembly growth." Attachment A to Document 1937-4 evaluated the Comanche Peak fuel dimensions in relation to the canister dimensions.

Spacer lengths were evaluated for the various types of assemblies to ensure the active fuel region remained within the neutron absorber region of the canister basket and the 2 to 2 1/2 inch requirement was met. The spacer lengths determined acceptable were the same values listed in Procedure NUC-212 in the note preceding Step 8.3.7. The upper and lower fuel spacer length requirement for each individual assembly was listed in Forms NUC-212-2, NUC-212-4 and NUC-212-5. For the first three canisters to be loaded, all upper spacers were "B" with a length of 9.75 inches.

Documents (a) Procedure NUC-212 Spent Fuel Limits For Dry Cask Operations, Revision 1 (b)

Reviewed: Letter from Frayne D. Ronkowski, Holtec Int. to Craig Montgomery, Luminant Power Company entitled "Fuel Spacer Lengths for CPNPP - Action Requested," dated November 3, 2010 with attachment entitled "Attachment A to Document ID: 1937-4 R1" Category: Fuel Selection/Verification Topic: Material Balance, Inventory, and Records Reference: 10 CFR 72.72(a) Published 2011 Requirement: Each licensee shall keep records showing the receipt, inventory (including location),

disposal, acquisition, and transfer of all SNM with quantities specified in 10 CFR 74.13(a)(1).

Observation: The licensees Special Nuclear Material (SNM) accountability plan, as supplemented with procedures detailing the development of fuel movement plans, met the records requirement in 10 CFR 72.72(a). Details of the SNM accountability plan were presented in Procedure NUC-020. The licensee utilized nine item control areas (ICA) to designate the location of special nuclear material items. The nine areas were defined in Procedure NUC-020, Section 6.1.2 "Item Control Areas." Item Control Area-9 included the wet cask loading pit, fuel building cask decontamination area, fuel building railroad loading bay, the ISFSI heavy haul path and the ISFSI pad. Item Control Area-9 applied to the activities involving the loaded canisters. Fuel assemblies were loaded into the canister in the wet cask loading pit in accordance with Procedure RFO-106. Step 6.3.3 of Procedure RFO-106 required that prior to movement of any fuel, the requirements of Procedure NUC-212 must be satisfied before approval of the RFO-106 move plan. This ensured that fuel loaded into the canister complied with the limits for dry cask operations. Step 8.4.3 of NUC-212 required verification of canister loading as directed in Procedure RFO-204. Step 8.4.4 of Procedure NUC-212 required completion of SNM transfer forms prior to moving the canister from the spent fuel pool to the ISFSI. The loaded canister was then moved to the ISFSI pad for long term storage and the SNM accounting software/database was updated.

Page 44 of 149

Documents (a) Procedure DCS-203, MPC Handling and Fuel Loading Operations, Revision 1 (b)

Reviewed: Procedure NUC-020, Special Nuclear Material Accountability Plan, Revision 15 (c)

Procedure NUC-021, Special Nuclear Material Transfer Operations, Revision 12 (d)

Procedure NUC-212, Spent Fuel Limits For Dry Cask Operations, Revision 1 (e)

Procedure RFO-106, Development and Implementation of Fuel Shuffle Sequence Plans, Revision 18 (f) Procedure RFO-204, Verification of Core and MPC Loading Patterns, Revision 14 Category: Fuel Selection/Verification Topic: Post Loading Verification Reference: FSAR 1014, Section 8.1.4.3 Revision 9 Requirement: Perform a post-loading visual verification of the assembly identification markings to confirm that the serial numbers match the approved fuel loading pattern.

Observation: The licensee's procedures had incorporated the requirement to perform a post-loading visual verification to confirm that the fuel assembly serial numbers matched the loading plan. The note preceding Step 8.4.23 of Procedure DCS-203 stated that the canister was to be loaded with spent fuel assemblies per Procedure RFO-302 and in accordance with Procedure RFO-106. Fuel verification was to be completed in accordance with Procedure RFO-204. Procedure DCS-203, Step 8.4.24 required the core performance engineer to verify that loading and video inspection had been completed in accordance with Procedure RFO-204. Procedure RFO-204 described the process of verifying that the correct assemblies had been placed in the correct basket locations in the canister.

Section 8.5 "MPC Load Verification" required verification by two individuals, one of which should be the core performance engineer. The two must concur that the fuel assembly was in the correct location and in the correct orientation. If the second individual was not a quality control (QC) inspector, then the video recordings made during the verification process were required to be reviewed by QC. The MPC map from the canister loading paperwork associated with Procedure RFO-106 was used for the verification. Upon completion of the fuel verification, the core performance engineer completed Form RFO-204-3 "MPC Loading Verification."

Documents (a) Procedure NUC-212 Spent Fuel Limits For Dry Cask Operations, Revision 1 (b)

Reviewed: Procedure RFO-106 "Development and Implementation of Fuel Shuffle Sequence Plans,"

Revision 18 (c) Procedure RFO-204 Verification of Core and MPC Loading Patterns, Revision 14 (d) Procedure RFO-302 "Handling of Fuel Assemblies," Revision 12 (e)

Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 Category: General License Topic: Evaluation of Effluents/Direct Radiation Reference: 10 CFR 72.212(b)(5)(iii) & 10 CFR 72.104(a) Published 2011 Requirement: The general licensee shall perform a written evaluation prior to use that establishes that the requirements of 10 CFR 72.104 "Criteria for Radioactive Materials in Effluents and Direct Radiation from an ISFSI" have been met. 10 CFR 72.104 requires the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 25 mrem to the whole body, 75 mrem to the thyroid and 25 mrem to any other critical organ during normal operations and anticipated occurrences, Observation: The requirement in 10 CFR 72.212(b)(5)(iii) was the same as the requirement in Page 45 of 149

Certificate of Compliance 1014, Technical Specification A.5.7.2 to verify compliance for the Comanche Peak site against 10 CFR 72.104(a). This was presented in the licensee's 10 CFR 72.212 Evaluation Report, Section 5.11 "10 CFR 72.212(b)(5)(iii) - Radiological Evaluation Pursuant to 10 CFR 72.104" to demonstrate compliance for the Comanche Peak site. A detail discussion of how Comanche Peak was shown to comply with the annual dose to a real individual located beyond the controlled area boundary can be found in the Inspector Notes of this inspection report under the Category: Radiation Protection and the Topic: Controlled Area Boundary Dose Rate Analysis.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) Holtec Report No. HI-2104636 Dose Versus Distance from HI-STORM 100S Version B Containing MPC-32 For Comanche Peak, Revision 4 Category: General License Topic: Flood Conditions Reference: CoC 1014, Appendix B, Section 3.4.4 Amendment 7 Requirement: Verify that the site analysis for flooding does not exceed the Certificate of Compliance limits of 15 fps water velocity and a height of 125 feet of water.

Observation: Section 5.12.1.3 "Flood" of the licensee's 72.212 Evaluation Report documented that the ISFSI pad elevation ranged from 830 to 831 feet elevation (above sea level) and that the probable maximum flood together with maximum wave run-up could attain a level of 794.7 feet at the plant site. This was based on a probable maximum flooding level of 789.7 feet with a 5 foot wave run-up and wind tide. The Squaw Creek reservoir adjacent to the ISFSI pad was fed by the Brazos River and had a service spillway at elevation 775.0 feet. The 775 foot lake level was used as the pre-flood lake level. Flooding in the area around the Comanche Peak site was due to precipitation runoff into streams and rivers. The site was not subject to surges, tsunamis, or ice jams and there have been no floods due to a dam failure. Potential dam failures were included in the flooding analysis and discussed in the Comanche Peak Updated Final Safety Analysis Report (UFSAR),

Section 2.4.4 "Potential Dam Failure." The Morris Sheppard and the DeCordova Bend dams were evaluated for the effects of a dam failure on the Comanche Peak site. The domino-type failure of the dams were shown to have no affect on the ISFSI. A wind velocity of 40 mph was assumed for the wave run-up calculations. The high winds were assumed to occur at the time of a probable maximum flood. This flooding condition was discussed in more detail in the Comanche Peak UFSAR, Section 2.4 "Hydraulic Engineering." Section 2.4.3 described the analysis performed to assess the probable maximum flood. The ISFSI pad was designed for the 25-year, 24-hour rain event of 7.45 inches. The ISFSI design included additional freeboard that was sufficient for the 100-year storm event to ensure adequate drainage of water off the ISFSI pad to prevent blockage of the cask air inlet ducts. As stated in Section 5.12.1.3 of the 72.212 Evaluation Report, Since the ISFSI pad will not experience any flooding, the HI-STORM 100 cask system design basis flood condition of 15 feet per second water velocity and a height of 125 feet of water (full submergence of the loaded cask)

identified in Section 2.2.3.4 and Table 2.2.8 of the Holtec FSAR, and analyzed in Section 3.4.6 of the Holtec FSAR, is bounding.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) Comanche Peak Steam Electric Station Updated Final Safety Analysis Report Page 46 of 149

(UFSAR), Amendment 103b (c) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: General License Topic: Initial Compliance Evaluation Against CoC Reference: 10 CFR 72.212(b)(5) Published 2011 Requirement: A general licensee shall perform written evaluations, prior to use and before applying the changes authorized by an amended Certificate of Compliance to a cask loaded under the initial Certificate of Compliance or an earlier amended Certificate of Compliance, which establishes that the cask, once loaded with spent fuel or once the changes authorized by an amended Certificate of Compliance have been applied, will conform to the terms, conditions, and specifications of the Certificate of Compliance or amended Certificate of Compliance listed in 10 CFR 72.214.

Observation: The Comanche Peak 72.212 Evaluation Report evaluated the terms, conditions and specifications in Certificate of Compliance 1014, Amendment 7, and documented that the conditions as set forth had been met at the Comanche Peak site. Section 5.9 "10 CFR 72.212(b)(5)(i) - Conformance with Certificate of Compliance Terms, Conditions, and Specifications" of the 72.212 Evaluation Report and Appendix 1 "HI-STORM 100 Cask System Certificate of Compliance Evaluation" provided a detail comparison of the requirements in the Certificate of Compliance against the procedures and programs established at the Comanche Peak nuclear power plant. The licensee was using a combination of already existing Part 50 programs and procedures plus newly developed procedures and documents specifically developed for the ISFSI. The HI-STORM 100S storage cask, the HI-TRAC 125 transfer cask, and the MPC-32 canister listed in Certificate of Compliance 1014, License Condition #1 were planned for use at Comanche Peak to store the pressurized water reactor (PWR) spent fuel. The MPC-32 canister holds 32 spent fuel assemblies. The HI-STORM 100S, Version B and the HI-TRAC 125D will be used.

Appendix 1 of the 72.212 Evaluation Report provided a detail list of each of the licensing conditions in the Certificate of Compliance, the technical specifications in Appendix A and the approved content and design requirements in Appendix B and how Comanche Peak complied with these licensing requirements. License Condition 2 and 3 required the licensee to develop written procedures for handling, loading, movement, surveillance, and maintenance. The licensee had developed a series of procedures as dry cask storage (DCS) procedures that covered the required topics in License Condition 2 and 3. The procedures included required activities associated with both loading and unloading a cask. Additional site procedures were also identified related to shift surveillances, conduct of maintenance, crane operations and rigging, health physics, quality assurance, and fuel handling. Procedures that would be used by contract personnel, such as the Holtec welders and non-destructive testing personnel, were provided by Holtec and included HSP procedures or MSLT-MPC-Holtec procedures.

License Condition 4 required activities being performed related to structures, systems, and components designated as important to safety to be conducted under an NRC approved quality assurance program. Comanche Peak was using their Part 50 approved QA program. License Condition 5 established heavy load requirements associated with handling a loaded cask. Comanche Peak was using their Part 50 heavy loads program Page 47 of 149

and their single failure proof crane in the fuel handling building. No lifting operations outside the Part 50 facility were required. License Condition 6 required that the fuel loaded in the casks meet the required fuel specifications in Appendix B of the Certificate of Compliance 1014. During this NRC inspection, a detailed review of the licensee's program related to fuel selection was completed to verify that processes and procedures had been put in place by the licensee to ensure that only spent fuel consistent with Appendix B of Certificate of Compliance 1014 would be selected for storage. More detail related to selection of the spent fuel is found in the Inspector Notes of this inspection report under the Category "Fuel Selection and Verification." License Condition 7 required that features and characteristics for the site, cask, and ancillary equipment must be in accordance with Appendix B of the Certificate of Compliance. A discussion of each of the design features in Appendix B were discussed in Appendix 1, Table 3 of the 72.212 Evaluation Report. References were provided to various Comanche Peak procedures and documents to show compliance with the design features listed in Appendix B. License Condition 8 discussed making changes to the Certificate of Compliance. Only Holtec can make requests to the NRC to change Certificate of Compliance 1014, since they are the certificate holder. Comanche Peak is a general licensee, and as such cannot make change requests directly to the NRC. License Condition 9 discussed two special requirements for the first canister system placed in service. The requirement to document the performance of the supplemental cooling system had been satisfied by Arkansas Nuclear One and a letter sent to the NRC dated September 29, 2009. This satisfied the supplemental cooling system report requirement in License Condition 9 and Comanche Peak was not required to provide any additional information to the NRC. The requirement related to the heat transfer validation test, which was discussed more in Step 23 of the Holtec Final Safety Analysis Report (FSAR), Section 8.1.7 "Placement of HI-STORM in Service," had not been submitted to the NRC by another licensee, and as such, the required test and analysis would have to be performed by Comanche Peak. License Condition 10 listed the required pre-operational testing and training exercises. Comanche Peak completed all the required testing and training exercises. Details related to each of these can be found in the Inspector Notes of this inspection report under the Category "Dry Run Demonstration." License Condition 11 approved an exemption related to the HI-STORM 100 underground system and was not applicable to Comanche Peak. License Condition 12 also applied to the underground system and was not applicable to Comanche Peak. License Condition 13 authorized use of the Holtec HI-STORM 100 system as a general license if the user possessed a Part 50 license. Comanche Peak was using the HI-STORM 100 system as a general license and currently held a Part 50 reactor license for the two unit Comanche Peak Nuclear Power Plant, Docket Numbers 50-445 and 50-446.

Appendix 1, Table 2 "CoC Appendix A - Technical Specifications" and Table 3 "CoC Appendix B - Approved Content and Design Features" of the 72.212 Evaluation Report provided detail information related to compliance of the Comanche Peak program with the requirements in Appendix A and B of Certificate of Compliance 1014. Most of these sections referenced back into specific sections of the 72.212 Evaluation Report or to specific site procedures.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) Letter (OCAN090902) from D.B. Brice, Entergy to USNRC Document Control Desk Page 48 of 149

entitled, "HI-STORM-100 Cask System Supplemental Cooling System (SCS) Validation Test - Arkansas Nuclear One Units 1 and 2," dated September 29, 2009 [NRC ADAMS Accession No. ML092810250] (c) Holtec Report HI-2094415 "Validation of First Use of Supplemental Cooling System at Arkansas Nuclear One," Revision 0 [proprietary - not publically available] (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: General License Topic: Initial Compliance Evaluation Against FSAR Reference: 10 CFR 72.212(b)(6) Published 2011 Requirement: The general licensee shall review the FSAR referenced in the Certificate of Compliance or amended Certificate of Compliance and the related NRC Safety Evaluation Report, prior to use of the general license, to determine whether or not the reactor site parameters, including analysis of earthquake intensity and tornado missiles, are enveloped by the cask design basis considered in these reports. The results of this review must be documented in the evaluation made in 10 CFR 72.212(b)(5).

Observation: The licensee documented the required written evaluations in the 72.212 Evaluation Report as Section 5.12 "10 CFR 72.212(b)(6) Reactor Site Parameters Review of the Cask FSAR and SER." Section 5.12 included site specific analysis of fires and explosions, tornados, floods, tsunamis and hurricanes, earthquakes, lightning, burial of the ISFSI under debris, environmental temperatures, snow, and collapse of nearby facilities. Each topical area was reviewed against the Comanche Peak Updated Final Safety Analysis Report (UFSAR) or other site specific documents. For flooding, the licensee determined that the ISFSI pad, at elevation 830 feet, was above the probable maximum flood level evaluated in the Comanche Peak UFSAR, Section 2.4 "Hydraulic Engineering." The elevation of the nearby Squaw Creek reservoir spillway was 775 feet. When considering the maximum potential rain fall, the drainage design of the ISFSI pad, and potential wind driven waves on the reservoir, it was shown that ISFSI pad flooding was not feasible. This was documented in the 72.212 Evaluation Report as Section 5.12.1.3 "Flooding." For a tsunami, the Comanche Peak site was well inland, over 300 miles from the Gulf of Mexico, such that a tsunami was not a possible environmental event. This was stated in the Comanche Peak UFSAR, Section 2.4.6

"Probable Maximum Tsunami Flooding" and documented in the 72.212 Evaluation Report in Section 5.12.1.4 "Tsunami and Hurricane." A hurricane could affect the Comanche Peak site. Section 5.12.1.4 of the 72.212 Evaluation Report stated that the probable maximum hurricane could reach winds of 81 miles/hour (mph) at 30 feet above ground. The Comanche Peak UFSAR, Section 2.3.1.1.2 "Hurricanes" discussed the potential impact of a hurricane on the Comanche Peak site. The impact was bounded by the Holtec FSAR, Table 2.2.4 "Tornado Characteristics" which listed the HI-STORM 100 storage cask design basis as a maximum wind speed of 360 mph.

Earthquake effects on the ISFSI were discussed in the 72.212 Evaluation Report, Section 5.10 "Cask Storage Pad and Areas" and Section 5.12.1.5 "Earthquake Intensity." The earthquake design basis for the Holtec HI-STORM storage cask discussed in the Holtec FSAR, Section 3.4.7.1 "Seismic Events" was compared to the potential Comanche Peak site seismic events discussed in the Comanche Peak UFSAR, Section 2.5 "Geology and Seismology." The design basis earthquake for the Comanche Peak site was a horizontal Page 49 of 149

ground motion of 0.12 g and a vertical acceleration of 0.08 g. The ISFSI pad was built to Parameter Set "B" requirements in the Holtec FSAR, Table 2.2.9 "Examples of Acceptable ISFSI Pad Design Parameters." Geotechnical investigations were performed of the subsoil under the ISFSI pad based on analysis of borehole drilling samples.

Impact of a seismic event on the ISFSI pad was performed based on two configurations.

One with all the casks on the pad and one with half the casks present creating an unbalanced condition on the pad. The results of the calculations determined that the ISFSI pad was adequately designed to support the static and dynamic loads of the HI-STORM casks and would not experience excessive sliding requiring anchoring in accordance with Certificate of Compliance 1014, Appendix B, Section 3.4 "Site Specific Parameters and Analysis." Additional information related to the ISFSI pad design and construction can be found in the NRC inspection report of the ISFSI pad construction in December 2010. Inspection Report 72-74/2010-01 was issued March 17, 2011 as Adams Accession No. ML110760665.

The licensee discussed the potential effects of lightning on a loaded HI-STORM storage casks while on the ISFSI pad in the 72.212 Evaluation Report, Section 5.12.1.6

"Lightning." The licensee referenced the Holtec FSAR, Section 11.2.12.2 "Lightning Analysis" which stated that a lightning strike on a HI-STORM cask had been analyzed to show there would be no adverse affect on the stored spent fuel and that the storage cask's carbon steel outer shell would provide the path for the lightning to the ground.

Burial of the storage casks under debris from shifting soils was evaluated in the 72.212 Evaluation Report, Section 5.12.1.7 "Burial Under Debris." The ISFSI pad was not located near shifting soil. The area around the pad, at an elevation above the pad, was analyzed to verify the soil would remain stable. The area around the pad was higher on all sides except for the southern edge. Ditches had been constructed to direct all storm water flows from the higher sides away from the ISFSI pad and to the southeast.

Environmental temperatures affecting the site were evaluated in the 72.212 Evaluation Report, Section 5.12.1.8 "Environmental Temperatures." The Holtec FSAR, Section 2.2.1.2 "Handling" established a minimum working area ambient temperature of 0 degree F when moving a loaded HI-TRAC transfer cask or a HI-STORM storage cask. This limit was established to provided a sufficient safety margin before brittle fracture might occur during handling operations. Loading, transport and unloading operations were restricted in the Comanche Peak procedures such that ambient work area temperatures must be greater than 0 degrees F. Step 5.18 of Procedure DCS-201 and Step 5.18 of Procedure DCS-206 required that transport of a loaded canister in the HI-TRAC transfer cask or HI-STORM storage cask be limited to working area ambient temperature of greater than 0 degrees F. Moving a loaded canister inside the fuel building was controlled through Procedure DCS-203. Step 5.13 of Procedure DCS-203 stated

"Transport of a loaded canister in a HI-TRAC or HI-STORM shall be limited to working area ambient temperatures of greater than 0 degrees F." In addition, the fuel building crane was limited to a minimum temperature of 40 degree F for any heavy lifts per Procedure MDA-304, Step 6.1.6.

Holtec FSAR, Section 2.2.1.4 "Environmental Temperatures" and Table 2.2.2

"Environmental Temperatures" established acceptable annual average ambient Page 50 of 149

temperatures that were used in the calculations for the thermal analysis of the spent fuel while in storage. As long as these values were not exceeded as a time averaged yearly mean, there would be no long term detrimental affect on the stored spent fuel. The Comanche Peak site had an annual mean temperature of 66 degree F and an average annual daily maximum temperature of 77 degree F as shown in the Comanche Peak UFSAR, Table 2.3-15 "Values of Mean, Average and Extreme Daily Maximum, and Average and Extreme Daily Minimum Surface Temperatures at Fort Worth." This data was for the time period 1931 to 1960. Both values were below the 80 degree F Holtec limit in the Holtec FSAR, Table 2.2.2. Comanche Peak's UFSAR, Section 2.3.2.1.3

"Temperatures" provided a discussion of the mean and average temperatures experienced at the site. The licensee also evaluated Holtec's limits for the annual average soil temperatures (77 degree F), off-normal ambient 3-day average temperatures (-40 to 100 degree F), and the extreme accident level ambient 3-day average temperatures (125 degree F) listed in the Holtec FSAR, Table 2.2.2. In each of these cases, the Comanche Peak site could be shown to meet the Holtec limits. Comanche Peak UFSAR, Table 2.3-15 provided the various values for the Comanche Peak site that were compared to the Holtec limits. These values for Fort Worth were actually slightly higher than the values experienced at the Comanche Peak site. The Comanche Peak UFSAR, Table 2.3-16

"Statistics and Diurnal Variation of Meteorological Parameters" provided values actually measured at the Comanche Peak site meteorological station for 1973 to 1976 and were compared to the values for the same time period for the Fort Worth data. The Comanche Peak site was 3 degree F lower for the annual mean temperature. Because the summer of 2011 was the hottest 3-month period (June - August) on record for any state in the U.S, selected data from the National Weather Service was reviewed by the NRC inspector to verify that the Holtec limits were still bounding when considering the recent record temperatures. This record period of 92 days had an average temperature of 86.8 degree F. This value was the average of the high (anywhere in the state) and the low (anywhere in the state) for each day, then averaged over the 92 days in the summer. This value was what established 2011 in Texas as the hottest summer on record in the nation, breaking the old record previously held by Oklahoma. Of the major cities in Texas, Wichita Falls was the city with the highest summer average of 91.9 degree F. For the National Weather Service data for Fort Worth, Texas, June had a high of 104 degrees on June 18, 2011, July had a high of 106 on July 25, 2011 and August had a high of 110 on August 2, 2011. The July high on July 26, 2011 of 106 degree F had a low for that day of 85 degree F giving a 96 degree F average. For August 2, 2011, the high was 110 degree F with the low for that day of 83 degree F and an average of 97 degree F. As can be seen from these recent 2011 values representing the hottest summer on record, the Holtec limit from Table 2.2.2 of 100 degree F (average of the high and low for the day, then averaged over a 3-day period) was still bounding when compared to even the hottest individual day of the month.

The Holtec FSAR, Section 2.2.1.6 "Snow and Ice" and Table 2.2.8 "Additional Design Input Data for Normal, Off-Normal, and Accident Conditions" listed a snow pressure loading of 100 pounds/square foot as the bounding pressure loading that the HI-STORM storage cask was designed for. The 72.212 Evaluation Report, Section 5.12.1.9 "Snow" referenced the Comanche Peak UFSAR, Section 2.3.1.2.8 "Precipitation" for data related to snowfall. The maximum 24-hour and seasonal snowfall records between 1898 and 1970 in Fort Worth was 12 inches and 15.3 inches, respectively. During 1970 to 1978, Page 51 of 149

the maximum seasonal snowfall in Fort worth was 17.6 inches. If the maximum winter snowfall occurred in one storm on top of an already present 100-year snowpack of 13 inches, the total weight of the snow was calculated to be approximately 16 pounds per square foot, well below the Holtec limit of 100 pounds/square foot.

The licensee also evaluated whether the collapse of any nearby facilities could affect the ISFSI. The nearby old steam generator storage facility and switchyard towers were all far enough away that their collapse would not affect the casks on the ISFSI pad. There were no other nearby facilities that could impact the ISFSI pad.

The effects of a tornado on the ISFSI were evaluated in the 72.212 Evaluation Report, Section 5.12.1.2 "Tornado." The potential effects of a tornado on a loaded cask that required evaluation were discussed in the Holtec FSAR, Section 3.1.2.1.1.5 "Tornado."

The characteristics of the design basis tornado was listed in the Holtec FSAR, Table 2.2.4 "Tornado Characteristics." The design basis for the rotational wind speed was 290 mph; the maximum translational wind speed was 70 mph; and the maximum total wind speed was 360 mph. The Comanche Peak UFSAR, Section 3.3.2 "Tornado Loadings" used a design basis tornado of 300 mph rotational and 60 mph translational for a total of 360 mph maximum total wind speed. Tornado generated missiles were discussed in several of the Holtec FSAR sections including Section 2.2.3.5 "Tornado," Section 3.4.8

"Tornado Wind and Missile Impact," Section 11.2.3 "Tip-Over," and Section 11.2.6

"Tornado." These Holtec FSAR sections discussed the generic design basis missiles that had been analyzed for the HI-STORM storage casks, the assumptions used in the calculations, and the results. The impact of a missile on the HI-TRAC transfer cask was provided in Holtec FSAR, Section 3.4.8.2 "HI-TRAC Transfer Cask." However, for the activities at the Comanche Peak site, the transfer cask was always protected inside the fuel building when loaded with spent fuel. NUREG-0800 and Regulatory Guide 1.76 provided acceptable postulated missiles that should be evaluated due to a tornado event.

The Holtec FSAR evaluated three tornado generated missile types consistent with NUREG-0800 and Regulatory Guide 1.76. These were classified as a large missile, a penetrant missile, and a micro-missile traveling at an impact velocity that was 35% of the maximum wind speed (360 mph) for the design basis tornado. Holtec FSAR Table 2.2.5

"Tornado Generated Missiles" provided a description of the three missile types. They included an 1800 kilogram (kg) automobile traveling at 126 mph, a rigid solid steel cylinder of 8-inches in diameter weighing 125 kg traveling at 126 mph, and a solid 1 inch diameter sphere weighing 0.22 kg traveling at 126 mph. The missiles were assumed to impact the cask in a manner to produce maximum damage, resulting in damage to the cask and reducing the shielding. Holtec FSAR, Section 3.4.8.1 "HI-STORM Storage Overpack" evaluated the structural impact of the tornado driven missiles on the HI-STORM storage cask and the potential for penetration. Section 11.2.3 "Tip-Over" evaluated the potential for a tornado driven missile to tip the HI-STORM storage cask over. The analysis determined that the tornado driven missiles could not tip over a loaded cask and the missiles would not penetrate the inner canister holding the spent fuel.

Comanche Peak's Part 50 licensing basis included an analysis for tornado driven missiles. This was described in Design Basis Document DBD-CS-081 "General Structural Design Criteria" and included an automobile weighing 1810 lbs traveling at 59 Page 52 of 149

meters/second (m/sec) [132 mph], a utility pole 35 feet long weighing 510 kg traveling at 55 m/sec ( 123 mph), a 12 inch diameter pipe 40 feet long weighing 340 kg traveling at 47 m/sec (105 mph), a 6 inch diameter pipe 15 feet long weighing 130 kg traveling at 52 m/sec (116 mph), a 1 inch steel rod 3 feet long weighing 4 kg traveling at 51 m/sec (114 mph), and a 4 x 12 wood plank 12 feet long weighing 52 kg and traveling at 83 m/sec (186 mph). The comparison of the missiles analyzed in the Holtec FSAR versus those analyzed in the Comanche Peak Design Basis Document found that the Comanche Peak missile assumptions were not bounded by the Holtec missiles related to size and velocity. The assumed large missile of an automobile analyzed in the Holtec FSAR was 1800 kg traveling at 126 mph and was smaller and slower than the Comanche Peak assumed automobile of 1810 kg traveling at 132 mph. Holtec performed an evaluation of the Comanche Peak missiles and documented the evaluation in Holtec Report HI-2104637 "Environmental Hazard Evaluations for Comanche Peak HI-STORM," Revision 0. The analysis performed by Holtec used the same methodology as described in the Holtec FSAR. The analysis determined that the Comanche Peak missile assumptions would not cause the HI-STORM cask to tip over, result in significant deformation or damage to the cask, and could not penetrate the inner canister. The results of the analysis were consistent with the conclusions described in the Holtec FSAR. The licensee performed a 72.48 evaluation (Evaluation No. EV-CR-2011-007002-15) and determined that the differences related to the Comanche Peak missile assumptions from the missiles assumed in the generic Holtec FSAR did not require NRC approval. The analysis showed that the Holtec casks could withstand the effects of the Comanche Peak site specific missiles without any changes to the design of the casks.

A site specific Fire Hazards Analysis was performed for the ISFSI activities to verify that the fire conditions at the Comanche Peak site were bounded by the Holtec FSAR. The fire analysis reviewed the site specific fire hazards associated with Comanche Peak's ISFSI and heavy haul path. Specific fire hazards were identified and analyzed to verify that the fire and explosion effects on the loaded cask would be bounded by the analysis that had been conducted in the Holtec FSAR. Section 2.2.3.3 "Fire" of the Holtec FSAR stated that a fire accident near an ISFSI was considered extremely remote due to the absence of significant combustible material. The Comanche Peak Fire Hazards Analysis confirmed that combustible material near the heavy haul path and the ISFSI pad at Comanche Peak would not present a fire hazard to the cask. The Fire Hazards Analysis established requirements to verify that no changes to the storage of combustible material had occurred, as analyzed in the Fire Hazards Analysis, prior to each transport of a loaded canister from the plant to the ISFSI. Procedure DCS-201, Attachment 10.1.3

"Loaded HI-STORM Pre-Movement Checklist" included the requirement to walk-down the heavy haul path and the ISFSI area using Attachment 1, Figure 1 "Comanche Peak ISFSI Hazards Equipment Plan" of the Fire Hazards Analysis to verify there were no significant fire or explosive hazards nearby. Holtec FSAR, Section 2.2.3.3 also stated that the only credible fire concern related to the transport vehicle fuel tank resulting in a fire engulfing the loaded cask while being moved to the ISFSI. The Holtec FSAR assumed that a maximum of 50 gallons of diesel would be in the transported fuel tank.

For the transporter used at Comanche Peak, the fuel tank was sized at 45 gallons. Holtec FSAR, Section 4.6.2.1 "Fire Accidents" provided the calculations for the assumed fire of 3.62 minutes (217 seconds) at an average temperature of 1475 degree F. The effect of the fire on the spent fuel inside the canister was calculated to be an increase of 1.1 degree Page 53 of 149

F. This increase was insignificant to the overall temperature of the spent fuel canister shell of 469 degree F or the spent fuel cladding at 711 degree F as listed in the Holtec FSAR, Table 4.4.6 "Maximum MPC Temperatures for Long Term Normal Storage Conditions" for a design basis loaded cask. The evaluation of fire hazards in the Comanche Peak Fire Hazards Analysis, Section 7.2.1.1 "Fire Hazards Associated with the Vertical Cask Transporter (VCT) and/or Low Profile Transporter (LPT) Prime Mover Specific to the Comanche Peak Nuclear Power Plant (CPNPP)" identified that the 50 gallon fuel tank fire on the transporter could cause failure of the hydraulic lines and leakage of the hydraulic fluid, with approximately 80 gallons of additional combustible liquid contributing to the fire. Section 7.2.1.1 also stated that the liquid pool fire could ignite the 16 tires on the transporter. For this postulated fire, the Fire Hazards Analysis, Revision 0 stated that it would be difficult to demonstrate that the effects of a fire involving the transporter combustibles, including the tires, would be bounded by the design basis storage cask fire analyzed in Section 4.6.2.1 of the Holtec FSAR. The Fire Hazards Analysis identified compensatory actions that would be required consisting of a continuous fire watch with adequate fire suppression equipment. The Certificate of Compliance1014, Appendix B, Section 3.4.5 allowed for analysis of site specific fire considerations in cases where a fire was not bounded by the fire conditions analyzed by the certificate of compliance holder. In Comanche Peaks case, rather than analyzing the potential to exceed the 50 gallon bounding fire, the fire analysis stated that administrative actions would be taken to preclude the occurrence of a fire that could exceed the design basis fire. This position in Revision 0 of the Fire Hazards Analysis was not consistent with the provisions of the Certificate of Compliance or the methodology described in 10 CFR 72.48 where actions can be taken without NRC approval. Holtec provided Comanche Peak with Holtec Report HI-2084156, Appendix G

"Evaluation of VCT Hydraulic Fluid and Tire Fire," which analyzed a more comprehensive fire scenario for the Fermi Nuclear Plant transporter. The Fermi transporter and the Comanche Peak transporter had been built to the same specification (HI-84065R4). The analysis calculated that when considering all combustible fluids on the transporter, including hydraulic fluids and motor oils when filled to capacity, and the transporter tires, with a fire lasting 40.2 minutes, the temperature increase on the fuel cladding would be 12.2 degree F, raising the design basis loaded fuel temperature from 711 degree F to 732.2 degree F. This value was well below the 1058 degree F maximum cladding temperature limit for off-normal and accident conditions listed in the Holtec FSAR, Section 2.0.1 "MPC Design Criteria." The helium pressure inside the canister was calculated to increase by 1.4 pounds per square inch-gauge (psig), going from 99 psig to 100.4 psig. This was well below the 200 psig accident limit listed in the Holtec FSAR, Table 2.2.1 "Design Pressures." By adding the additional combustible fluids and the tires, an aggregate combustible volume was determined to be 555 gallons. Comanche Peak issued Revision 1 of the Fire Hazards Analysis which incorporated the new information into Section 7.2.1.1 and Holtec completed a 72.48 analysis (Evaluation No.

EV-CR-2011-007002-14) that determined that the conditions of 10 CFR 72.48 had been met to allow use of the transporter.

For explosions, Certificate of Compliance 1014, Appendix B, Section 3.4.4 stated that the potential for explosions while handling a loaded storage cask or transfer cask shall be addressed. The transfer cask, while loaded with spent fuel was always inside the fuel building. The HI-STORM storage cask, while being moved to the ISFSI pad or located Page 54 of 149

on the ISFSI pad was subject to nearby explosion hazards. Holtec FSAR, Section 2.2.3.10 "Explosions" and Section 2.3.6 "Fire and Explosion Protection" stated that there were no credible internal explosive events associated with the HI-STORM 100 system.

FSAR Section 3.4.7.2 "Explosion" provided analysis for acceptable external pressures that would not result in a cask tipping over. These values were also provided in Table 2.2.1 "Design Pressures." For the storage cask, a 5 psi design basis steady state pressure differential across the storage cask diameter or a 10 psi pulse with duration of less than or equal to 1 second was considered an acceptable value that would not tip over the cask. The storage cask was also qualified to sustain a lateral impulse of 60 psi differential pressure for 85 milliseconds without tipping over. The Comanche Peak Fire Hazards Analysis, Section 8.0 "Evaluation of Potential Explosion Hazards" evaluated the various fire and explosion sources located along the heavy haul path and near the ISFSI. The 8,000 gallon diesel fuel storage tank at 360 feet from the heavy haul path and the 4,000 gallon gasoline storage tank at 350 feet from the heavy haul path were evaluated. The diesel fuel was excluded because of it's high flash point and would not be expected to form an explosive mixture. The gasoline tank was assumed to rupture and not catch fire, but to form a flammable gas-air mixture, then ignite. The calculation determined that the 10 psig limit in the Holtec FSAR, Table 2.2.1 would be exceeded at 178 feet, which was less than the 350 foot distance between the gasoline tank and the heavy haul path.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) Comanche Peak Steam Electric Station Updated Final Safety Analysis Report (UFSAR), Amendment 103b (c) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 (d)

Comanche Peak Document No. 13769701-R-M00002 "Fire Hazards Analysis," Revision 0 and Revision 1 (e) Comanche Peak 72.48 Evaluation No. EV-CR-2011-007002-14

"Site Specific Fire Hazards Evaluation (13769701-R-M-00002, Rev. 1 Comanche Peak ISFSI Project Evaluation of Fire Hazards"), Revision 0 (f) Comanche Peak 72.48 Evaluation No. EV-CR-2011-007002-15 "Site Specific Tornado Missile Evaluation (HI-2104637, Rev. 0 Environmental Hazards Evaluation for Comanche Peak HI-STORM"),

Revision 0 (g) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 (h) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (i) Procedure DCS-206 Transporting and Transferring a Loaded MPC for Unloading, Revision 2 (j) Holtec Report HI-2084156 "Fermi Fire Hazards Analysis,"

[Appendix G only], Revision 5 (k) NUREG-0800 "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Revision 3 (l) Regulatory Guide 1.76 "Design Basis Tornado and Tornado Missiles for Nuclear Power Plants," Revision 1 (m) NRC Inspection Report 72-74/2010-01 [ADAMS Accession No. ML110760665],

issued March 17, 2011 (n) National Weather Website "www.srh.noaa.gov/fwd" Category: General License Topic: Initial Evaluation Against Part 50 License Reference: 10 CFR 72.212(b)(8) Published 2011 Requirement: Prior to use of the general license, determine whether activities related to storage of spent fuel involve a change in the facility technical specifications or require a license amendment for the facility pursuant to 10 CFR 50.59(c). Results of this determination must be documented in the evaluation made in 10 CFR 72.212(b)(5).

Page 55 of 149

Observation: The licensee performed an evaluation of the ISFSI activities related to the storage of spent fuel at the Comanche Peak site using the 10 CFR 50.59 process and determined that no changes were required to the facility technical specifications or licensee. The 72.212 Evaluation Report, Section 5.14 "10 CFR 72.212(b)(8) - 10 CFR 50.59 Evaluation of ISFSI Activities" documented the reviews performed by the licensee. A number of 10 CFR 50.59 screenings/evaluations were performed in areas such as cranes, ISFSI pad, electrical systems related to the ISFSI, modifications needed to the fuel building, and security. None of the changes required any amendments to the plant's Part 50 operating license or technical specifications. Table 5.14.1 "ISFSI Activities Evaluated in Accordance with 10 CFR 50.59" provided a list of the screenings and evaluations performed.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) National Weather Website "www.srh.noaa.gov/fwd" Category: General License Topic: Limiting Site Temperatures Reference: CoC 1014, Appendix B, Section 3.4.1; 3.4.2 Amendment 7 Requirement: The maximum average yearly temperature at the site shall be verified as 80 degrees F.

The temperature extremes, averaged over a 3-day period, shall be greater than -40 degrees F and less than 125 degrees F.

Observation: The Comanche Peak site was below the maximum average yearly temperature of 80 degree F and within the -40 to 125 degree F temperature extremes averaged over any 3-day period. The Comanche Peak UFSAR, Section 2.3.2.1.3 "Temperatures" and Table 2.3-15 "Values of Mean, Average and Extreme Daily Maximum, and Average and Extreme Daily Minimum Surface Temperatures at Fort Worth" provided information related to temperatures in nearby Fort Worth, Texas. Values in the table ranged from 1931 to 1973. The average daily maximum for each month averaged for a year was 77 degree F. This met the 80 degree F limit. The extreme minimum for the worst month was in January 1964 at 4 degree F. This was above the -40 degree F limit. The extreme maximum was 108 degree F recorded in August 1964. The summer of 2011 was one of the hottest summers on record in Texas. The July and August 2011 maximum daily highs for Granbury, Texas, just several miles from the Comanche Peak site, recorded all but 2 days above 100 degree F. The highest temperature reached was 109 degree F.

These temperatures were below the 125 degree F limit.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed:

Category: General License Topic: Program Review - RP, EP, QA, and Training Reference: 10 CFR 72.212(b)(10) Published 2011 Requirement: The general licensee shall review the reactor emergency plan, quality assurance program, training program and radiation protection program to determine if their effectiveness is decreased and, if so, prepare the necessary changes and seek and obtain the necessary approvals.

Observation: The licensee performed a review of the reactor emergency plan, quality assurance program, training program and radiation protection program and documented that review Page 56 of 149

in the 72.212 Evaluation Report, Section 5.16. Changes were made to the programs to incorporate the dry cask storage project. The emergency plan review was discussed in Section 5.16.1.1 and identified that an emergency action level for an unusual event had been added to the emergency plan and procedures. The quality assurance program review was documented in Section 5.16.1.2 and stated that the Part 50 quality assurance plan would be used for ISFSI activities. Luminant Power had notified the NRC of the intent to use the Part 50 quality assurance plan by letter December 22, 2009 in accordance with the requirement in 10 CFR 72.140(d). The quality assurance plan had been revised to incorporate the ISFS as Appendix F Dry Cask Storage System Quality Assurance Program. The quality assurance program applied a graded approach consistent with NUREG/CR-6407 Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Important to Safety. Design Basis Document DBD-ME-080-01 provided the basis for classifying the various structures, systems, and components according to the important to safety classification system described in NUREG/CR-6407. The NRC approved Holtec quality assurance program was also identified as being applicable to those activities associated with work performed by Holtec, such as the canister fabrication.

The training program was updated to incorporate specific training and qualification requirements for personnel assigned to the ISFSI. This statement was included in Section 5.16.1.3 of the 72.212 Evaluation Report. Section 5.16.1.4 discussed the radiation protection program and stated that the existing program had been revised to extend the existing radiological controls used for the Part 50 program to the dry cask storage activities. These controls included radiation work permits, as-low-as-reasonably-achievable (ALARA) processes, and requirements related to surveys, postings, access control, personnel monitoring, etc.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) DBD-ME-080-01 Dry Cask Storage System, Revision 1 (c) Letter (CP-200901726)

from Luminant Power to USNRC Document Control Desk entitled Comanche Peak Steam Electric Station Docket NOS. 50-445, 50-446, and 72-74 ISFSI Quality Assurance Program, dated December 22, 2009 Category: General License Topic: Revisions to 72.212 Analysis Reference: 10 CFR 72.212(b)(7) Published 2011 Requirement: The general licensee shall evaluate any changes to the written evaluations required by 10 CFR 72.212(b)(5) and 10 CFR 72.212(b)(6) using the requirements of 10 CFR 72.48(c).

A copy of this record shall be retained until spent fuel is no longer stored under the general license issued under 10 CFR 72.210.

Observation: The 10 CFR 72.212 Evaluation Report, Section 5.13.2 stated that the 72.212 Evaluation Report would be controlled as a licensing basis document in accordance with Procedure STA-116. This procedure required the use of the 10 CFR 72.48 process for making changes to the 72.212 Evaluation Report. Revision 1 of the 72.212 Evaluation Report was the latest version. A review was performed of the changes made since Revision 0 to verify the 72.48 process had been implemented. Procedure STA-707 described Comanche Peak's 72.48 review process and included Form STA-707-4 "Applicability Determination," Form STA-707-3 "72.48 Screen," and Form STA-707-5 "72.48 Page 57 of 149

Evaluation." There had been nine applicability determinations, two screenings, and two evaluations performed related to the 72.212 Evaluation Report. Section 5.13.1

"Analysis" of the 72.212 Evaluation Report (Revision 1) listed two deviations from the Holtec Final Safety Analysis Report (FSAR) that had been identified. These related to the fire associated with the vertical cask transporter and the tornado driven missiles.

Both were found to not require a change to the Holtec Certificate of Compliance through the use of the 72.48 process.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 0 Reviewed: and Revision 1 (b) Procedure STA-116, Maintenance of CPNPP Licensing Basis Documents Operating License Conditions and Technical Specifications, Revision 11 (c) Procedure STA-707 "10 CFR 50.59 and 10 CFR 72.48 Reviews," Revision 18 (d)

Form STA-707-4 Applicability Determination "CPNPP 10CFR72.212 Evaluation Report," closed as Condition Report AI-CR-2011-007002, Action Tasks 1 through 5 and Action Tasks 9 through 11 (e) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: General License Topic: Storage Cask Blocked Inlet or Outlet Air Vents Reference: CoC 1014, Appendix B, Section 3.4.9 Amendment 7 Requirement: For those users whose site specific design basis includes an event that results in blockage of the storage cask inlet or outlet air vents for an extended period of time longer than the completion time in LCO 3.1.2 (i.e. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for heat loads less than or equal to 28.74 kW and 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> for heat loads greater than 28.74 kW), an analysis may be performed to demonstrate adequate heat removal for the duration of the event. If the analysis is not performed or adequate heat removal cannot be verified, alternate methods of cooling must be established.

Observation: The Comanche Peak ISFSI pad had been designed and located such that flooding and burial due to debris would not occur. The 72.212 Evaluation Report, Section 5.12.1.3

"Flooding" and Section 5.12.1.7 "Burial under Debris" discussed the features of the ISFSI pad that made these events very unlikely. The ISFSI pad was located at elevation 830 feet. The adjacent Squaw Creek reservoir was at elevation 775 feet. The area around the ISFSI pad had been designed with a drainage pattern to direct excessive water flow during heavy rains away from the pad area and toward the reservoir. The soil in the vicinity of the ISFSI pad was stable and not subject to mudslides. There were no active volcanoes in Texas. The potential for blockage of the inlet and outlet vents as discussed in the Holtec Final Safety Analysis Report (FSAR), Section 2.2.3.12 "Burial Under Debris" was not credible. Therefore, further analysis of cask vent blockage, as specified in Certificate of Compliance 1014, Appendix B, Section 3.4.9, was not required for the ISFSI at Comanche Peak.

Documents (a) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Reviewed: (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9

.

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Category: General License Topic: Written Procedures Required Reference: CoC 1014, License Condition 2 & 3 Amendment 7 Requirement: Written operating procedures shall be prepared for cask handling, loading, movement, surveillance and maintenance. The user's site-specific written procedures shall be consistent with the technical basis described in Chapter 8 of the Holtec Final Safety Analysis Report (FSAR). Written cask acceptance tests and maintenance program shall be prepared consistent with the technical basis described in Chapter 9 of the Holtec FSAR.

Observation: Comanche Peak had developed procedures for controlling all work associated with cask handling, loading, movement, surveillance, maintenance, and testing consistent with the Holtec Final Safety Analysis Report (FSAR) Chapter 8 "Operating Procedures" and Chapter 9 "Acceptance Criteria and Maintenance Program." Procedures had been developed specific to the ISFSI activities with numerous other procedures used for Part 50 programs adequately applied to the ISFSI program. Procedures included those for fuel selection, handling, and accountability (RFO-106, RFO-204, RFO-207, RFO-302, NUC-020, NUC-021, NUC-212, and NUC-302), canister loading and transport to the ISFSI (DCS-201, DCS-202, DCS-203, DCS-204, and DCS-205), canister unloading, should a problem require removal of the spent fuel after the canister had been sealed (DCS-201, DCS-206, and DCS-207), heavy loads and rigging (DCS-107, MDA-304, MDA-308, MDA-402, MXE-PX-2017, and MSM-PX-2017), radiation protection activities (RPI-606, RPI-627, RPI-710, RPI-790, RPI-791, RPI-792, and STA-651),

sampling of the boron in the canister during loading or unloading of a canister and analysis of gas samples taken prior to removing the lid during unloading operations (DCS-507), welding (HSP-504, HSP-505, HSP-508, HSP-509, HSP-513, and HQP-9.2),

non-destructive examination of the welds (HSP-506, HSP-507, and MSLT-MPC-HOLTEC), dealing with abnormal events such as crane problems, problems during transport to the ISFSI pad, tip-over, fire or explosion, damage of fuel while loading the canister, etc (DCS-301), operational inspections of the storage cask vents and the temperature monitoring system (OPT-102A series), receipt inspections, annual inspections/tests and periodic maintenance activities associated with the various ISFSI equipment (DCS-101, DCS-104, DCS-105, DCS-106, DCS-109, DCS-110, DCS-111, and DCS-112), performing safety evaluations (STA-707), requirements related to periodic reporting or for notifying the NRC of nonroutine issues (STA-501 and STA-502), and requirements related to records and documentation (STA-116 and STA-302).

Documents (a) Procedure DCS-101 "HI-STORM, MPC, and HI-TRAC Storage and Pre-Use Reviewed: Inspection," Revision 0 (b) Procedure DCS-104 "MPC Receipt Inspection," Revision 1 (c) Procedure DCS-105 "HI-STORM Receipt Inspection," Revision 1 (d) Procedure DCS-106 "HI-TRAC Annual Inspection and Maintenance," Revision 0 (e) Procedure DCS-107

"Dry Cask Storage Rigging Plan," Revision 0 (f) Procedure DCS-109 "Vertical Cask Transporter Maintenance," Revision 0 (g) Procedure DCS-110 "HI-STORM In-Service Annual Inspection and Maintenance," Revision 0 (h) Procedure DCS-111 "Inspection and Testing of Dry Cask Storage Lifting Devices," Revision 2 (i) Procedure DCS-112

"Inspection and Testing of the 125 Ton Lift Yoke and Lift Yoke Extension," Revision 3 (j) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 (k) Procedure DCS-202 "MPC Preparation for Loading," Revision 2 (l) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (m) Procedure DCS-204 MPC Closure Operations (Sealing, Drying, and Backfill), Revision 2, PCN-1 (n)

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Procedure DCS-205 "Stack-Up and Transfer of Loaded MPC," Revision 2 (o) Procedure DCS-206 Transporting and Transferring a Loaded MPC for Unloading, Revision 2 (p)

Procedure DCS-207 Unloading a Loaded MPC, Revision 2 (q) Procedure DCS-301

"Dry Cask Storage Equipment Malfunction, LOOP, LOCA, and Contingencies Guidance," Revision 0, PCN 1 (r) Procedure DCS-507 "Chemistry Sampling of the Multipurpose Canisters (MPC)," Revision 0, PCN 1 (s) Procedure MDA-304 "Control of Heavy Loads and Critical Lifts," Revision 6 (t) Procedure MDA-308 "Requirements for Load Handling Personnel," Revision 9 (u) Procedure MDA-402 "Control of Load Handling Equipment," Revision 11 (v) Procedure MSM-PX-2017 "Fuel Building Overhead Crane Mechanical Inspection," Revision 2 (w) Procedure MSE-PX-2017 "Fuel Building Overhead Crane Electrical Inspection," Revision 3 (x) Procedure NUC-020

"Special Nuclear Material Accountability Plan," Revision 15 (y) Procedure NUC-021

"Special Nuclear Material Transfer Operations," Revision 12 (z) Procedure NUC-212 Spent Fuel Limits for Dry Cask Storage, Revision 1(aa) Procedure NUC-302

"Handling of Fuel Assemblies," Revision 12 (bb) Shift Surveillance OPT-102A-1 Mode 1 and 2 Shiftly Surveillances, Revision 37 (cc) Shift Surveillance OPT-102A-3 Mode 3 Shiftly Surveillances, Revision 25 (dd) Shift Surveillance OPT-102A-4 Mode 4 Shiftly Surveillances, Revision 22 (ee) Shift Surveillance OPT-102A-5 Mode 5 Shiftly Surveillances, Revision 22 (ff) Shift Surveillance OPT-102A-6 Mode 6 Shiftly Surveillances, Revision 24 (gg) Procedure RFO-106 "Development and Implementation of Fuel Shuffling Sequence Plans," Revision 18 (hh) Procedure RFO-204 "Verification of Core and MPC Loading Patterns," Revision 14 (ii) Procedure RFO-207 "Surveillance of Fuel Assemblies and Insert Components," Revision 8 (jj) Procedure RPI-606

"Radiation Work and General Access Permits," Revision 22 (kk) Procedure RPI-627

"Job Coverage for Dry Fuel Storage," Revision 1 (ll) Procedure RPI-710 Radiological, Environmental Monitoring, Sampling, and Analysis Program, Revision 16 (mm)

Procedure RPI-790 MPC Accessible Surface Contamination Survey, Revision 1 (nn)

Procedure RPI-791 HI-TRAC Transfer Cask Surface Dose Rates, Revision 1 (oo)

Procedure RPI-792 HI-STORM Overpack Surface Dose Rate, Revision 1 (pp)

Procedure STA-116 "Maintenance of CPNPP Licensing Basis Documents, Operating License Conditions, and Technical Specifications," Revision 11 (qq) Procedure STA-302

"Station Records," Revision 22 (rr) Procedure STA-501 "Nonroutine Reporting,"

Revision 15 (ss) Procedure STA-502 "Routine Reporting," Revision 13 (tt) Procedure STA-651 "ALARA Program," Revision 10, PCN 1 (uu) Procedure STA-707 "10 CFR 50.59 and 10 CFR 72.48 Reviews," Revision 18 and associated 72.48 Resource Manual, Revision 0 (vv) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on MPC and MPC Lid," Revision 6 (ww) Holtec Procedure HSP-505 "Control and Issuance of Weld Filler Metal for MPC Site Welding Services," Revision 2 (xx) Holtec Procedure HSP-506 "Liquid Penetrant Examination for MPC Field Closure Welding," Revision 2 (yy) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Welding," Revision 2 (zz) Holtec Procedure HSP-508 "Repair of Deposited Weld Metal for MPC Closure Welding," Revision 1 (aaa) Holtec Procedure HSP-509 "Procedure for MPC Seal Weld Removal in the Field," Revision 0 (bbb) Holtec Procedure HSP-513

"Base Metal Repair Procedure for MPC Field Closure Welding," Revision 1 (ccc)

Holtec Procedure HQP-9.2 "Welder Qualification Requirements," Revision 6 (ddd)

Holtec Procedure MSLT-MPC-HOLTEC "Helium Mass Spectrometer Leak Test Procedure," Revision 3660-CP-00 Page 60 of 149

Category: Heavy Loads Topic: Component Weights for Heavy Lifts Reference: FSAR 1014, Section 8.0; Tables 8.1.1 through 8.1.4 Revision 9 Requirement: The handling weights for the Holtec storage system components are provided in Tables 8.1.1 through 8.1.4 of the FSAR. The user shall take appropriate actions to ensure the lift weights do not exceed the rated loads of all user-supplied lifting equipment.

Observation: The licensee had verified that the weight of the Holtec storage system components would not exceed the rated load for the Comanche Peak lifting equipment. Holtec Report No.

HI-2104639, page 3 of 12, stated The maximum lifted weight during the performance of a HI-STORM System loading at Comanche Peak is during the removal of the HI-TRAC transfer cask from the spent fuel pool with the water jacket full [which] is less than 130 tons (see Case 2 in Table 7.0.1). Case 2 of Table 7.0.1 showed a total lift weight of 255,814 pounds (127.9 tons), which was less than the 130-ton rating of the fuel building crane. The individual weights shown in Table 7.0.1 were consistent with those in Holtec Final Safety Analysis Report (FSAR) Table 8.1.1 "Estimated Handling Weights of HI-STORM 100 System Components with 125 Ton HI-TRAC" and Table 8.1.2 "Estimated Handling Weights with 125 Ton HI-TRAC" and accounted for the as-built condition of the Comanche Peak components and the weight of the Comanche Peak spent fuel.

Documents (a) Holtec Report Number HI-2104639 Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak, Revision 0 (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Heavy Loads Topic: Licensed Facility Heavy Loads Requirements Reference: CoC 1014, License Condition 5 Amendment 7 Requirement: Each lift of a canister, transfer cask, or storage cask must be made in accordance with the existing heavy loads requirements and procedures of the licensed facility at which the lift is made. A plant specific review (under 50.59 or 72.48, if applicable) is required to show operational compliance with existing plant specific heavy loads requirements.

Observation: The heavy lifts and crane operations associated with the dry cask storage operations were performed in accordance with the plant's maintenance department procedures used for all heavy lift activities in the plant. Procedure DCS-205 used to place a loaded canister and HI-TRAC transfer cask onto the HI-STORM and lower the canister, specified in Step 5.10 that all lifting and rigging was to be controlled by SOER 06-01 "Rigging, Lifting, and Material Handling," and Procedures MDA-304 and MDA-402. Procedure DCS-203, which was used for moving the canister into the spent fuel pool for loading and then removing the loaded canister from the spent fuel pool to the dry cask pit, required the use of Procedure MDA-304 for rigging and lifting operations. Step 5.7 stated that all rigging and lifting operations shall be conducted in accordance with MDA-304. Heavy loads lifted by the fuel building overhead crane shall be transported within the safe load areas shown on Procedure MDA-304, Attachment 8.C "Fuel Building Overhead Crane Safe Load Area." If not possible, then engineering approval was required prior to the lift.

This same requirement was included in Procedure DCS-206, Step 5.10. Procedure DCS-206 was used to bring a loaded cask into the fuel building that had been returned from the ISFSI pad, then move the loaded canister to the dry cask pit for unloading.

The licensees crane was rated to 130 tons. The maximum lift weight during the dry Page 61 of 149

cask storage operations was calculated at approximately 128 tons. The licensee performed a 50.59 evaluation to determine if the seismic and structural analyses performed by Holtec met the Comanche Peak Updated Final Safety Analysis Report (UFSAR) requirements for the plant. The Holtec seismic analysis had used a newer version of ANSYS (versions 12.0, 12.1, and 13.0) than the licensees UFSAR had used (version 5.4). The conclusion of the 50.59 evaluation determined that the analysis performed by Holtec produced essentially the same results as the licensees ANSYS version 5.4.

Documents (a) 50.59 Evaluation No - 59EV-2009-000859-01-00 Dry Cask Storage System (DCSS)

Reviewed: Equipment Incorporation into CPNPP dated July 13, 2011 (b) Procedure MDA-304

"Control of Heavy Loads and Critical Lift," Revision 6 (c) Procedure MDA-402 "Control of Load Handling Equipment," Revision 11 (d) Procedure DCS-203 "MPC Handling and Fuel Loading Operations," Revision 3 (e) Procedure DCS-205 "Stack-up and Transfer of Loaded MPC," Revision 2 (f) Procedure DCS-206 "Transporting and Transferring a Loaded MPC for Unloading," Revision 2 Category: Heavy Loads Topic: Procedures Reference: NUREG 0612, Section 5.1.1 (2) Issued July 1980 Requirement: Procedures should be developed to cover load handling operations for heavy loads that are or could be handled over or in proximity to irradiated fuel or safe shutdown equipment. The procedures should include: a) identification of the required equipment; b) inspections and acceptance criteria required before movement of the load; c) the steps and proper sequence to be followed in handling the load; d) defining the safe load path; and e) special precautions.

Observation: The procedures for heavy loads handled over or in proximity to irradiated fuel or safe shutdown equipment contained the NUREG 0612 recommendations. Procedure DCS-203 provided instructions for transferring the empty canister inside the HI-TRAC transfer cask to the spent fuel pool. Once the spent fuel was loaded into the canister, Procedure DCS-203 was used to move the canister to the dry cask pit for welding and backfilling. Upon completion of welding, the canister was transferred to the HI-STORM storage cask by placing the HI-TRAC transfer cask on top of the HI-STORM storage cask and the canister downloaded in accordance with Procedure DCS-205. The identification of required equipment was listed in Sections 7.0 of each procedure. Prior to each use, the yoke/yoke extensions and HI-TRAC trunnions were inspected for damage (Steps 8.3.3 and 8.6.3 of Procedure DCS-203 and Step 8.4.2 of Procedure DCS-205). Procedure DCS-203 and Procedure DCS-205 contained the proper sequence to lift the load. Procedure MDA-304 contained a safe load path for moving the HI-TRAC transfer cask in the fuel handling building in Attachment 8.C "Fuel Building Overhead Crane Safe Load Area." Special precautions were either documented as steps in Procedures DCS-203 and DCS-205 or were identified as CAUTION statements throughout each procedure.

Documents (a) Procedure MDA-304 Control of Heavy Loads and Critical Lifts, Revision 6 (b)

Reviewed: Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (c)

Procedure DCS-205 Stack-up and Transfer of Loaded MPC, Revision 2 (d) NUREG 0612 Control of Heavy Loads at Nuclear Power Plants, issued July 1980 Page 62 of 149

Category: Heavy Loads Topic: Safe Load Paths Reference: NUREG 0612, Section 5.1.1 (1) Issued July 1980 Requirement: Safe load paths should be defined for the movement of heavy loads to minimize the potential for heavy loads, if dropped, to impact irradiated fuel in the reactor vessel and in the spent fuel pool, or to impact safe shutdown equipment. The path should follow, to the extent practical, structural floor members, beams, etc., such that if the load is dropped, the structure is more likely to withstand the impact.

Observation: The safe load path for heavy lifts within the licensees fuel handling building was incorporated into Procedure MDA-304. The licensee had analyzed and established a safe load path for handling heavy loads and critical lifts in the fuel handling building and documented the acceptable paths in Procedure MDA-304, Attachment 8.C "Fuel Building Overhead Crane Safe Load Area." During the dry run demonstrations, the inspectors noted that the licensee was following the safe load path, as required.

Documents (a) Procedure MDA-304 Control of Heavy Loads and Critical Lifts, Revision 6 (b)

Reviewed: NUREG 0612 Control of Heavy Loads at Nuclear Power Plants, issued July 1980 Category: Heavy Loads Topic: Site Temperature Limit for Cask Handling Reference: CoC 1014, Appendix B, Section 3.4.8 Amendment 7 Requirement: Loading, transport, and unloading operations shall only be conducted with working area ambient temperatures of 0 degrees F or higher.

Observation: Loading, transport and unloading operations were restricted in procedures such that ambient work area temperatures must be greater than 0 degrees F. The transport of the HI-STORM storage cask to the ISFSI pad for storage or from the ISFSI pad to the fuel building for unloading was controlled by Procedures DCS-201 and DCS-206. Step 5.18 of both procedures required that transport of a loaded canister in the HI-TRAC transfer cask or HI-STORM storage cask be limited to working area ambient temperature of greater than 0 degrees F. Moving a loaded canister inside the fuel building was controlled through Procedure DCS-203. Step 5.13 of Procedure DCS-203 stated

"Transport of a loaded canister in a HI-TRAC or HI-STORM shall be limited to working area ambient temperatures of greater than 0 degrees F." In addition, the fuel building crane was limited to a minimum operating temperature of 40 degree F for any heavy lifts per Procedure MDA-304, Step 6.1.6.

Documents (a) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 Reviewed: (b) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (c)

Procedure DCS-206 "Transporting and Transferring a Loaded MPC for Unloading,"

Revision 2 (d) Procedure MDA-304 "Control of Heavy Loads and Critical Lifts,"

Revision 6 Category: Heavy Loads Topic: Storage Cask Maximum Lifting Height Reference: CoC 1014, Tech Spec A.5.5, Table 5.1 Amendment 7 Requirement: Between the fuel building and the ISFSI pad, the loaded overpack may be handled at any height necessary if: a) the lifting device provides support from underneath (i.e. on a rail Page 63 of 149

car, heavy haul trailer, air pads, etc.), or b) is being handled by a device that was designed in accordance with the increased safety factors of ANSI N14.6 and has redundant drop protection features. If the requirements above are not met, the following conditions apply. Free standing overpacks are limited to a vertical lift height of: a) 11 inches when the transporter route conditions (i.e. surface hardness and pad thickness) are equivalent to or less than either Set A or Set B of FSAR Table 2.2.9, or b) the vertical lift height is determined by site specific analysis to ensure that the impact loading due to a design basis drop event does not exceed 45 g's at the top of the canister fuel basket.

Observation: The vertical cask transporter, used to move the loaded cask from the fuel building to the ISFSI pad, was designed with lifting components to the requirements of ANSI N14.6, CMAA-70, or NOG-1 with redundant drop protection features. Holtec Report HI-2084065 described the specifications for the vertical cask transporter that were used at several nuclear sites. This same specification applied to the Comanche Peak transporter.

Step 4.5 stated that the cross beams on the transporter, attached to the lifting tower masts, met the more limiting applicable stress limits of either ANSI N14.6 (1993) or ASME BTH-1 "Design of Below the Hook Lifting Devices," Revision 2005. Step 4.6 stated that the lifting towers met the stress limits of NOG-1 "Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girders), Revision 2004.

Holtec Report HI-2084065, Section 11 "Testing Requirements" stated that the transporter was static load tested to 125% of the rated load. The rated load of the transporter was 205 tons. A 125% load test would require a load of 256.25 tons. MMH Factory Acceptance Test Procedure 36304-05 documented the test weights used for the Comanche Peak transporter load test as 535,100 pounds (267.550 tons) raised to the maximum height and held for 10 minutes. The hydraulic pressure was maintained on the lift cylinders during the load test as the sole means of preventing load settling. The hydraulic pressure was measured during the load test to ensure the pressure limits were not exceeded. All accessible welds on essential components were visually inspected after the test. During the lifting and transport operations, the HI-STORM storage cask was connected to the transport's cross beams by two brackets with increased safety factors in accordance with ANSI N14.6 (1993). The calculation package for the HI-STORM brackets (Holtec Document HI-992272) was reviewed to verify the material used for fabrication required the six times yield strength and ten times ultimate strength specified in ANSI N14.6.

Step 4.3 of the Holtec Report HI-2084065 stated that the transporter was designed with redundant drop protection. This was accomplished through the use of a safety catcher in each lifting tower that provided redundant drop protection to prevent an uncontrolled load drop in the event of a hydraulic system failure. The safety catchers were a spring-closed wedge system that could only be disengaged by hydraulic pressure. If the transporter system lost hydraulic pressure (the first drop protection system), then the safety catchers would engage to prevent the load from being dropped. As long as hydraulic pressure was maintained during transporter raising and lowering operations, the safety catcher would remain disengaged to allow free movement of the load.

Procedure DCS-210, Step 5.32 stated "When activated, the transporter safety catchers will allow the load to lower approximately 1/2" before full clamping force is attained.

To release the safety catchers, the load must be raised this same distance while the safety catcher is pressurized by hydraulic pressure." Cask lifting and lowering could only be performed while the vehicle was stationary and the operator's chair locked in position Page 64 of 149

facing the cask. During the dry run demonstration on June 21, 2011, the HI-STORM storage cask was moved from outside the fuel building to the ISFSI pad at a height of approximately 8 inches from the ground using the vertical cask transporter. Procedure DCS-201, Step 8.1.16 specified the height limit of 4 to 8 inches above ground during transit to the ISFSI pad.

Documents (a) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 Reviewed: (b) Holtec Document No. HI-992272 Calculation Package for Cask Miscellaneous Items, Revision 13 (c) Morris Material Handling (MMH) Document 036304-10 Vertical Cask Transporter (VCT) Operations and Maintenance Manual Serial #: 036304-036309, Revision 2 (d) Holtec Report HI-2084065 "Conformed Specification for the Design and Fabrication of the Holtec Vertical Cask Transporter for Use at the Braidwood, Byron, Cook, Fermi, LaSalle, and Perry Nuclear Plants," Revision 4 (e)

Morris Material Handling (MMH) Factory Acceptance Test Procedure 36304-05 "Holtec Vertical Cask Transporter Factory Acceptance Test Procedure," Revision 5 (f) American National Standards Institute (ANSI) N14.6 "Special Lifting Devices for Shipping Containers Weighting 10,000 Pounds or More," Revision 1993 (g) Crane Manufacturers Association of America, Inc. (CMAA) -70 "Specification for Top Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes" (h) American Society of Mechanical Engineers NOG-1-2004 "Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girders)," Revision 2004 Category: Heavy Loads Topic: Transport Route Surface Reference: CoC 1014, Tech Spec A.5.5.a.1 Amendment 7 Requirement: The transport route surface condition (surface hardness and pad thickness) shall be equivalent to or less limiting than either the Set A or Set B parameters in Table 2.2.9 of the FSAR. (This does not apply if handled by a device designed in accordance with increased safety factors of ANSI N14.6 and having redundant drop protection.)

Observation: The transport route between the fuel building and the ISFSI pad did not meet the Set A or Set B parameters in Table 2.2.9 of the Holtec Final Safety Analysis Report. As such, Comanche Peak employed a transporter that incorporated increased safety factors of ANSI N14.6 and had redundant drop features. See Category: Heavy Loads and Topic:

Storage Cask Maximum Lifting Height in the Inspector Notes of this inspection report for a detail description of the single failure proof aspects of the vertical cask transporter.

Documents (a) Holtec Document No. HI-992272 Calculation Package for Cask Miscellaneous Reviewed: Items, Revision 13 (b) Morris Material Handling (MMH) Document 036304-10 Vertical Cask Transporter (VCT) Operations and Maintenance Manual Serial #: 036304-036309, Revision 2 Category: Heavy Loads Topic: Transporter Wheel Failure Reference: No Reference Provided Requirement: During testing of the vertical cask transporter at the Comanche Peak site, an inner wheel hub assembly failed.

Observation: On June 27, 2011, test maneuvers were being conducted on the vertical cask transporter Page 65 of 149

inside the plant protected area, at turning pad #2. A HI-STORM 100 storage cask loaded with a canister filled with concrete, to simulate an actual loaded canister, was being transported on the heavy haul path during a training session. During a turning maneuver, a loud pop was heard by the driver. He was directed to stop by his ground assistant, who had observed that an inner wheel hub had broken off and the wheel was no longer attached to the transporter. Work was stopped and the cask lowered to the ground.

Condition Report CR-2011-007348 was initiated.

The transporter was in the process of acceptance testing by Comanche Peak and was scheduled for use in the upcoming loading campaign to start July 11, 2011. The transporter would be used to lift a loaded HI-STORM storage cask (approximately 174 tons) several inches from the ground and move the cask from outside the fuel building to the ISFSI pad, approximately 3/4 mile away. The transporter design load (static) was 205 tons. The transporter had made five complete trips between the plant and the ISFSI plus seven trips between turning pad 3 (plant protected area) to turning pad 8 (entrance to ISFSI), which would be 80% of the heavy haul path length, before the wheel failed. A concrete filled canister inside a HI-STORM storage cask, weighing approximately the same (174 tons) as a storage cask loaded with a canister containing spent fuel, was used during the trips. The heavy haul path was a compacted dirt road with a portion of the heavy haul path at an approximate 5% incline. There were eight concrete turning pads that had been built at turning locations along the heavy haul path. These turning locations were necessary because the path from the plant to the ISFSI pad was not a straight path. The transporter had 16 wheels. These were large aircraft landing wheels filled with foam. Eight wheels were on the outside of the transporter and contained motor driver units to move the transporter. Eight wheels were on the inside as "idler'

wheels. The idler wheels did not have drive units. The wheel that broke was an idler wheel on the back inside on the driver's side. The wheel hub completely sheared.

Comanche Peak removed the other seven idler wheels and found several that also showed cracking of the wheel hub, but had not completely failed. The outer drive wheel adjacent to the inner wheel that failed was also removed and examined. It did not show any signs of failure.

There were seven wheeled transporters that had been built and delivered to various ISFSI sites. Two were currently in use to move casks. These sites had relatively flat, short haul paths with a couple of turning pads. The LaSalle Nuclear Plant vertical cask transporter had the most extensive use (six cask movements over a 1/2 mile level haul path) and was examined for the potential for failure of the inner wheels. All eight idler wheels were removed from the transporter and examined. No cracking was found on any of the eight wheel hubs and the transporter was returned to service.

The failed hub assembly was delivered to a laboratory testing facility in Richardson, Texas by Comanche Peak. The laboratory performed several tests including visual examination and stereomicroscopy of the failed area. Three sampled areas were cold-sectioned, mounted in radial cross sections, wet ground and polished. Metallography and energy dispersive X-ray spectroscopy (EDX) was performed. Chemical composition was determined on the fractured steel using a SpectroMaxx Optical Emission Spectrograph and a LECO C-200 analyzer. Tensile testing was performed on a 0.5 inch wide flat section of the fractured steel sample. Knoop microhardness testing was performed on a Page 66 of 149

mounted sample. Rockwell C hardness testing was performed at the hub flange surface.

Charpy impact testing was performed on full size, V-notch transverse specimens machined from the axle at the hub. The conclusion reached by the analysis was that the bearing hub failed as a result of inadequate resistance to fatigue fracture initiation at the axle/hub fillet weakened by the presence of excessive manganese-sulfide inclusions uncovered by machining scratches. The low-energy impact steel had been embrittled by concentrated inclusions in its critical axle-hub fillet. The presence of the inclusions and machining scratches provided multiple surface stress raisers. During transporter operations, multiple fatigue cracks occurred within the axle-hub fillet. These offset micro cracks slowly grew inward and joined in the subsurface. The circumferential fracture front grew deeper becoming a single transverse "ring" of fatigue crack penetration with each significant load. In time, the growing fracture reduced the cross-sectional area of the axle-hub joint so that an overload condition was created on the remaining intact area. Upon reaching the yield strength of the steel, the joint rapidly fractured along the many embrittling inclusions. Final fracture was swift and sudden.

Holtec made arrangements for a new transporter to be delivered to Comanche Peak for use in the first loading campaign. Instead of using the new transporter, it was decided to remove all eight brand new idler wheel assemblies and install them on the Comanche Peak transporter, since the Comanche Peak transporter had already undergone extensive testing and modifications to address system overheating problems due to the extreme summer temperatures. New wheel assemblies had been ordered from the Italian manufacturing plant and would be installed on the new transporter. Holtec had performed a mean time to failure analysis and concluded that the movement of two casks from the plant to the ISFSI pad using the transporter with new wheel hubs could be done safely with little risk of failure. In addition, provisions could be established to repair a wheel, should it fail during transport.

On July 15, 2011, the station operations review committee (SORC) met to review the situation to determine if the loading campaign should proceed, with a possible July 25th loading date and a reduction from three to two casks for the first loading. The station operations review committee consisted of senior site managers from the various plant departments with the plant manager as the chairman. Holtec and the transporter manufacturer provided information on the wheel failure and the plant staff discussed the results of the recently completed failure analysis. The capability of the transporter to continue to operate in an impaired condition and the contingency plans to address failure of a wheel, including provisions for disengaging the transporter from the cask along the heavy haul path in order to repair the transporter, was discussed. It was recognized that even in normal situations, the potential always existed for the failure of the transporter, requiring it to be repaired before it had delivered the cask to the ISFSI pad. Comanche Peak would also have the spare transporter available as a backup.

Holtec and the plant staff discussed issues related to disengaging the cask from the transporter while on the heavy haul path. If placed on a turning pad, the cask would be level and relatively stable, but the turning pad had been built to a concrete hardness upper limit that exceeded that allowed for the ISFSI pad by the Holtec Final Safety Analysis Report. Concrete compressive strength limits for a Set "B" pad of 6,000 pounds/square inch (psi), found in the Holtec Final Safety Analysis Report (FSAR),

Page 67 of 149

Table 2.2.9 "Examples of Acceptable ISFSI Pad Design Parameters," were required to prevent damage to the spent fuel if the cask tipped over. The turning pads exceeded these upper limits. Sand could be placed on the turning pad and sandbags placed around the cask could reduce the tip over impact to acceptable levels. If the cask was placed on the heavy haul path, the tip over accident would not be a concern due to the softness of the heavy haul path, which was compacted soil. However, due to the slope of the heavy haul path, cribbing would be needed to level the cask to prevent tipping over and rolling down the heavy haul path. The heavy haul path was close to the lake at many points and retrieving the cask from the lake would be challenging. Sand bags could be placed around the cask to reduce the likelihood of the cask rolling away.

After discussions by the various managers and the vendors, the station operations review committee determined that additional information was needed before a determination of no safety impact could be made. The root cause analysis by the transporter manufacturer had not been completed and plant management wanted to review the results of the analysis and have further discussions of the impact of delaying the loading campaign.

One of the concerns related to delaying the planned cask loading was the loss of capacity of the spent fuel pool during the next refueling outage. Because of commitments made by the plant to the NRC to maintain certain spacing of the spent fuel in the racks, 300 fuel shuffles would be necessary to make room for the fuel from the fall outage if the cask loading campaign was delayed. Each handling of a spent fuel assembly presented some small risk of mis-handling and damaging the assembly. This risk had to be balanced against the consequences of the risk of failure of the transporter while moving a cask.

The commitments made by Comanche Peak to the NRC related to spacing requirements for spent fuel assemblies in the spent fuel pool to meet NRC issued EA-02-026 "Order for Interim Safeguards and Security Compensatory Measures," dated February 25, 2002.

This document, which is a safeguards document and not publically available, included requirements in Section B.5.b related to mitigation strategies for restoring core cooling, containment, and spent fuel pool cooling capability to cope with the loss of large areas of the facility due to large fires and explosions from any cause. This requirement was also specified in 10 CFR 50.54(hh)(2). NRC Bulletin 2011-01 was issued May 11, 2011 to highlight the recent Fukushima Daiichi facility accident and to require licensee's to provide verification of their compliance with 10 CFR 50.54(hh)(2) and Section B.5.b requirements. Comanche Peak management had implemented a strong "nuclear safety culture" program at the site. Worker perception of the decisions made by management and how they related to management's commitment to a safety culture was a key consideration in decisions being made by management. The use of a transporter with a known problem that could result in failure of the transporter to perform it's safety function versus the need to reshuffle spent fuel assemblies in the pool if the cask loading campaign was delayed presented a challenging decision for the station operations review committee.

Comanche Peak had planned for their second loading campaign in 2012 to load nine casks. If the current loading campaign was delayed, the loading campaign in 2012 would be twelve casks. No decision was made at the July 15, 2011, station operations review committee meeting and more information was requested from Holtec including (1) a tip-Page 68 of 149

over analysis for a free standing cask on the heavy haul path, (2) compensatory actions that would be needed if a transporter failure occurred, (3) test requirements that should be performed on the backup transporter before it is used, balancing the need to truly test the device versus knowing that testing was only contributing to it's eventual failure, and (4) a root cause failure analysis of the wheel failure.

A preliminary root cause analysis was performed by the original equipment manufacturer on the axle and hub assembly that failed. Information was provided to Comanche Peak on July 22, 2011. The analysis concluded that the current design of the axle and hub assembly was not compatible with the Comanche Peak heavy haul path due to its compound angles (slope of heavy haul path (5%) plus the heavy haul path had a slight sideway slope (3%) to provide for drainage ). The root cause concluded that the side way slope on parts of the heavy haul path placed additional stresses on the axle and hub assembly causing cyclic fatigue in the assembly. Cyclic fatigue eventually lead to failure of the component. Two corrective actions were identified as options. One was to change the design of the heavy haul path. The other was to change the design of the axle and hub assembly, which was selected as the appropriate option. This required a re-design effort by the vendor and the manufacture and purchase of new wheel components which would delay readiness of the transporter until early next year. On July 25, 2011, the station operations review committee decided to delay the cask loading campaign to 2012.

Documents (a) Condition Report CR-2011-7348 "VCT Wheel Hub Failure," created on June 27, Reviewed: 2011 (b) Morris Material Handling (MMH) Document 036304-10 Vertical Cask Transporter (VCT) Operations and Maintenance Manual Serial # 036304-036309, Revision 2 (c) Morris Material Handling (MMH) Factory Acceptance Test Procedure

1. 36304-05 "Holtec Vertical Cask Transporter Factory Acceptance Test Procedure 36304-05 P&H Device Serial # CN-36309," approved January 22, 2010 (d) Metallurgical Engineering Services, Inc. Report "Failure Analysis of Bearing Hub Assembly," dated July 6, 2011 (e) Morris Material Handling P&H Calculation No. 36304-1 "Seismic Analysis of Vertical Cask Transporter," approved May 18, 2009 (f) Holtec Report HI-2084065 "Conformed Specification for the Design and Fabrication of the Holtec Vertical Cask Transporter for Use at the Braidwood, Byron, Cook, Fermi, LaSalle, and Perry Nuclear Plants," Revision 4 (g) Comanche Peak Station Operations Review Committee (SORC) Review UFO #1 "VCT Release for Transport," dated July 13, 2011 (h) Stone &

Webster Drawing 13769701-05000-C-CVL-402-2 "Haul Path Plan and Profile," Sheet 2, Revision 1 (i) NRC Bulletin 2011-01 "Mitigating Strategies," dated May 11, 2011

[Adams Accession # ML111250360] (j) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Heavy Loads Topic: Trunnion Annual Testing Reference: FSAR 1014, Sections 9.2.1, 9.1.2.1; Table 9.1.3 Revision 9 Requirement: Testing of the lifting trunnions shall be performed annually, or prior to use if the transfer cask has not been used for greater than one year. The licensee may elect to repeat the initial 300% load testing described in Sections 9.1.2.1 of the FSAR or may elect to perform dimensional testing, visual inspection, and nondestructive testing of major load-carrying welds and critical areas in accordance with ANSI N14.6, Section 6.3.1 (1993).

Observation: The lifting trunnions of the HI-TRACK 125 transfer cask used at Comanche Peak were Page 69 of 149

initially tested to 300% of the designed maximum load on December 10, 2010. Annual testing planned by Comanche Peak would not include a 300% load test, but would be the ANSI 14.6 alternate inspection process. Procedure DCS-101 required a pre-use inspections of the HI-STORM storage casks, canisters, and HI-TRAC transfer cask if they had been in storage. Step 6.13 referenced the user to Procedure DCS-106 as part of the pre-operational activities prior to using the HI-TRAC if the annual inspection was needed. Procedure DCS-106 provided instructions for the HI-TRAC annual maintenance and inspections. Procedure DCS-106, Section 8.2 "Lifting Trunnion #1 Inspection at 90 Degree Axis" and Section 8.3 "Lifting Trunnion #2 Inspection at 270 Degree Axis" provided the instructions for performing the annual inspection. Step 8.2.1 for Trunnion

1. 1 and Step 8.3.1 for Trunnion #2 required a visual inspection of the trunnions. Step 8.2.3 and Step 8.3.3 required a surface examination of the accessible areas of the two lifting trunnions using either liquid penetrant or magnetic particle. Steps 8.2.7 and 8.2.8 for trunnion #1 and Steps 8.3.7 and 8.3.8 for trunnion #2 required measuring the diameter and length of the trunnions to satisfy the dimensional test. Procedure DCS-106, Section 2.0 "Acceptance Criteria" provided the acceptance criteria for the trunnion testing and referenced Holtec FSAR, Section 9.2.1 "Structural and Pressure Parts" and Table 8.1.10 "HI-TRAC Overpack Inspection Checklist." Step 2.1 of Procedure DCS-106 listed the acceptance criteria for the lifting trunnions as no detectable deformations, cracks, wear, corrosion or physical damage on the surfaces. Also listed was no obvious damage such as cracked, broken, or loose components; no signs of thread damage; no signs of water damage or corrosion; no chipped, cracked, blistered, or missing paint.

Documents (a) Holtec Report No. HI-2104639 "Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak," Revision 0 (b) Holtec Procedure PS-113 "Trunnion Load Test Procedure for HI-TRACK 100 and HI-TRACK 125 Systems," Revision 8 (c)

Procedure DCS-101 HI-STORM, MPC, and HI-TRAC Storage and Pre-use Inspection, Revision 0 (d) Procedure DCS-106 HI-TRACK Annual Inspection and Maintenance, Revision 0 (e) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Category: Heavy Loads Topic: Trunnion Initial Load Testing Reference: FSAR 1014, Section 9.1.2.1 Revision 9 Requirement: The lifting trunnions shall be tested to 300% of the maximum design lifting load (750,000 lbs for the 125 ton transfer cask and 600,000 lbs for the 100 ton transfer cask.

The load shall be applied for a minimum of 10 minutes, after which the accessible parts of the trunnions and trunnion attachment areas shall be visually examined to verify no deformation, distortion or cracking has occurred.

Observation: The lifting trunnions of the HI-TRACK 125 transfer cask used at Comanche Peak were initially tested to 300% of the designed maximum load on December 10, 2010. This test satisfied the annual requirement until December 10, 2011. Holtec Report No. HI-2104639 calculated the maximum weight of the HI-TRACK 125 while coming out of the spent fuel pool loaded with fuel in Case #2 as 255,814 pounds (127.9 tons). When subtracting the lift yoke (5,895 pounds) and the lift yoke extension (5,273 pounds), the resulting weight on the lifting trunnions was 244,646 pounds (122.3 tons). The HI-TRACK 125 trunnions were designed to handle 250,000 pounds (125 tons). The test load applied to the HI-TRACK trunnions per Step 9.3.5 of Holtec Procedure PS-113 was Page 70 of 149

770,651 pounds (385.3 tons) for ten minutes. The test was documented in Procedure SP-113, Exhibit 3.4 "Trunnion/Support Lug Load Test Data Record" and included in Holtec Report DP-0832-006. After the load was applied a visual examination of the trunnions confirmed that no deformation, distortion or cracking had occurred.

Documents (a) Holtec Document No HI-2104639 "Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak," Revision 0 (b) Holtec Procedure PS-113 "Trunnion Load Test Procedure for HI-TRACK 100 and HI-TRACK 125 Systems," Revision 8 (c)

Holtec Report DP 0832-006 "HI-TRAC 125D Vecasp for Comanche Peak (Luminant),"

Revision 0 Category: Loading Operations Topic: Canister Lid Fit Test Reference: FSAR 1014, Table 9.1.1 Revision 9 Requirement: As part of the Holtec inspection and test acceptance criteria, the canister lid, closure ring, and vent and drain port cover plates shall be fit tested prior to canister operation.

Observation: The canister lid, closure ring, and vent and drain port cover plates were required to be fit tested prior to canister operation in accordance with Procedures DCS-104 and DCS-202.

After a new canister and associated parts was received at the warehouse, receipt inspection was performed using Procedure DCS-104, Section 8.1 "Component Parts Verification." After completion of this activity, the canister and associated parts were placed in a designated storage area. At this time, the work effort could continue with the fit-up tests in accordance with Procedure DCS-104 or a protective cover placed over the canister, lid, closure rings, and hardware and the fit-up test performed later under Procedure DCS-202 prior to actual use. Both procedures had identical directions for conducting the fit-up tests. For Procedure DCS-104, Section 8.2 "MPC Lid Fit-Up" installed the lid onto the shell and verified that the lid fit. If the lid did not fit, Step 8.2.11 required that a condition report be issued. Section 8.3 "MPC Lid-to-Shell Joint Fit-Up" completed the lid to shell joint fit-up test to verify that the height of the lid properly matched the height of the shell. Step 8.3.1 measured the depth from the bottom of the grove weld prep area on the lid to the top of the canister shell using a calibrated ruler. A quality control inspector verified the measured depth was acceptable. The vent/drain port cover plates were fit tested in Section 8.4 "Port Cap and Cover Plate Fit-Up." The closure ring fit check was performed in Section 8.5 "Closure Ring Fit Check."

The equivalent sections in Procedure DCS-202 were Section 8.3 "MPC Lid Fit-Up,"

Section 8.4 "MPC Lid-to-Shell Joint Fit-Up," Section 8.5 "Port Cap and Cover Plate Fit-Up," and Section 8.6 "Closure Ring Fit Check." Wording for these sections were the same in both procedures.

Documents (a) Procedure DCS-104 MPC Receipt Inspection Revision 1 (b) Procedure DCS-202 Reviewed: "MPC Preparation for Loading," Revision 2 Category: Loading Operations Topic: Cask System Annual Maintenance Reference: FSAR 1014, Table 9.2.1 Revision 9 Requirement: The following cask system maintenance shall be performed annually, or prior to use if out of service for greater than 1 year: a) overpack [storage cask] external surface visual examination, b) transfer cask and overpack visual inspection of identification markings Page 71 of 149

c) transfer cask internal and external visual inspection for compliance with design drawings, d) load testing of the transfer cask trunnions, and e) transfer cask shield tank pressure relief valve calibration.

Observation: The required annual inspections of the HI-STORM storage cask and the HI-TRAC transfer cask were incorporated into Procedures DCS-106 for the HI-TRAC and Procedure DCS-110 for the HI-STORM. The annual storage cask visual inspection of the identification markings for the HI-STORM storage cask was incorporated into Procedure DCS-110, Section 8.1 HI-STORM Identification Marking's Inspection.

Section 8.1 required verification that the HI-STORM serial number and tag number were legibly inscribed on the cask. The external surface visual examinations of the HI-STORM storage cask were described in Section 8.2 "HI-STORM External Visual Inspection." The annual visual inspection included any observable damage to the cask body or lid including dents, scratches, gouges, rust, corrosion, or problems with the paint. If any issues were observed, Steps 8.1.2 and 8.2.3 required that a condition report be issued.

The annual HI-TRAC internal and external visual inspection for compliance with design drawings was incorporated into Procedure DCS-106, Section 8.6 HI-TRAC Internal and External Visual Inspection. The annual inspection criteria included looking for corrosion, chipping, cracks. blistering of paint, gouges, scratches and an overall assessment of the condition of the HI-TRAC transfer cask. Acceptance criteria was provided in Step 2.5 as no obvious damage such as cracked, broken, or loose components, no signs of thread damage, no signs of water damage or corrosion, no chipped, cracked, blistered, or missing paint." Procedure DCS-106, Section 8.7 "HI-TRAC Cask Identification Markings and Inspections" provided directions for performing the annual inspection of the identification markings on the HI-TRAC transfer cask. This included verifying the legibility of the model number, serial number, tag number, empty weight value, azimuth markings, and radiation monitoring targets. Any issues found were required to be documented on a condition report in accordance with Step 5.8.

The load testing of the HI-TRAC lifting trunnions was required annually by the Holtec Final Safety Analysis Report (FSAR), Section 9.2.1 "Structural and Pressure Parts" and Table 9.1.3 "HI-TRAC Transfer Cask Inspection and Test Acceptance Criteria," or prior to use if the period exceeded one year. Table 9.1.3 stated that the annual load testing of the lifting trunnions shall be performed per ANSI N14.6. Section 6.3 Testing to Verify Continuing Compliance of ANSI N14.6 provided instructions for the annual load test and allowed the load test to be omitted for cases where surface cleanliness and conditions permit dimensional testing, visual inspection and non-destructive testing of the major load carrying welds and critical areas. Procedure DCS-106, Section 8.2 Lifting Trunnion #1 Inspection at 90 Degree Axis, and Section 8.3 Lifting Trunnion

1. 2 Inspection at 270 Degree Axis, provided instructions for conducting the annual HI-TRAC lifting trunnion examinations. The examination included a visual exam looking for deformation, cracks, wear marks, corrosion, physical damage and cracks on the machined lip of the trunnion. A liquid penetrant surface examination was then performed on the two trunnions. Step 2.2 of Procedure DCS-106 identified ANSI/ASME Boiler and Pressure Vessel Code, 1995 Edition through 1997 Addends,Section III, Division 1, Paragraph NF-5350 as the liquid penetrant acceptance standard.

Dimensional testing was performed with acceptable dimensions and tolerances listed in Page 72 of 149

Sections 8.2 and 8.3 of the procedure and a drawing of the trunnions provided to aid in performing the dimensional test. The HI-TRACs original load test was performed December 10, 2010 and documented in Holtec Report DP 0832-006.

The HI-TRAC pressure relief valve annual calibration requirement was incorporated into Procedure DCS-106, Section 8.5 HI-TRAC Water Jacket Relief Valve Verification/

Calibration or Replacement. The set point requirement was 60 psig (60 to 62 psig) for the water jacket relief valve #1 and 65 psig (65 to 67 psig) for water jacket relief valve

1. 2. These set points were consistent with the values specified in Holtec FSAR Section 9.1.4.1 "Valves, Rupture Discs, and Fluid Transport Devices."

Documents (a) Procedure-106 HI-TRAC Annual Inspection and Maintenance, Revision 0 (b)

Reviewed: Procedure DCS 110 HI-STORM In-Service Annual Inspection and Maintenance, Revision 0 (c) Holtec Report DP 0832-006 "HI-TRAC 125D Vecasp for Comanche Peak (Luminant)," Revision 0 (d) American National Standards Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More,"

1993 (e) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Loading Operations Topic: Cask System Inspections Prior to Use Reference: FSAR 1014, Table 9.2.1 Revision 9 Requirement: The following cask system components shall be visually inspected prior to each loading campaign: a) overpack cavity, b) transfer cask cavity, and the c) transfer cask lifting trunnion and pocket trunnion recesses. In addition, the overpack bolts shall be inspected prior to installation.

Observation: Cask system component inspections to meet the requirements in the Holtec Final Safety Analysis Report (FSAR), Table 9.2.1 were incorporated into several procedures. The HI-STORM storage cask cavity visual inspection prior to fuel loading was incorporated into Procedure DCS-205, Section 8.3 "Prepare HI-STORM for MPC Transfer" and Procedure DCS-101, Attachment 10.1.3 HI-STORM Pre-Use Inspection. Procedure DCS-101, Attachment 10.1.3, Step 3.0 required inspection of the inside of the HI-STORM body for dirt, debris, tools or other material present inside the HI-STORM body. Procedure DCS-205, Step 8.3.1 stated Ensure there is no foreign material in the HI-STORM. The storage cask bolt visual inspection prior to installation during each use was described in Procedure DCS-101, Attachment 10.1.3 HI-STORM Pre-Use Inspection, Step 4.0 Lid Studs, Hex Nuts, and Washers. This section required verification that all the studs, nuts, and washers were present and the lid stud thread condition was smooth and uniform. The HI-TRAC transfer cask cavity visual inspection prior to each loading campaign was incorporated into Procedure DCS-203, Section 8.2.3 which stated Visually verify there is no foreign material inside the HI-TRAC. The HI-TRAC lifting trunnion and pocket trunnion recess visual inspection prior to each loading campaign was included in Procedure DCS-101, Attachment 10.1.5 HI-TRAC Inspection. The trunnions were inspected for corrosions, dents, scratches, gouges, or other visible damage or deformation. The HI-TRAC water jacket water level was required to be visually checked during each loading campaign. This requirement was incorporated into Procedure DCS-101, Attachment 10.1.5 HI-TRAC Inspection, Section 9.0 which required the water jacket level to be verified to within one inch of the top of the water Page 73 of 149

jacket prior to use.

Documents (a) Procedure DCS-101 HI-STORM, MPC, and HI-TRAC Storage and Pre-Use Reviewed: Inspections, Revision 0 (b) Procedure-106 HI-TRAC Annual Inspection and Maintenance, Revision 0 (c) Procedure DCS 110 HI-STORM In-Service Annual Inspection and Maintenance, Revision 0 (d) Procedure DCS-203 "MPC Handling and Fuel Loading Operations," Revision 3 (e) Procedure DCS-205 Stack-up and Transfer of Loaded MPC," Revision 2 (f) Procedure RPI-710 Radiological, Environmental Monitoring, Sampling, and Analysis Program, Revision 16 (g) Shift Surveillance OPT-102A-1 Mode 1 and 2 Shiftly Surveillances, Revision 37 (h) Shift Surveillance OPT-102A-3 Mode 3 Shiftly Surveillances, Revision 25 (i) Shift Surveillance OPT-102A-4 Mode 4 Shiftly Surveillances, Revision 22 (j) Shift Surveillance OPT-102A-5 Mode 5 Shiftly Surveillances, Revision 22 (k) Shift Surveillance OPT-102A-6 Mode 6 Shiftly Surveillances, Revision 24 (l) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Loading Operations Topic: Fuel Cladding Not Exposed to Air Reference: CoC 1014, Appendix B, Section 3.4.10 Amendment 7 Requirement: Procedures and/or mechanical barriers shall be established to ensure that during loading operations (and unloading) that either the fuel cladding is covered by water or the canister is filled with an inert gas.

Observation: Procedures and mechanical barriers were established to ensure the fuel cladding was covered by water or filled with an inert gas at all times. When water was being removed from the canister, two methods were used in the procedures to ensure the fuel was not exposed to air. In Procedure DCS-203, prior to removing the canister fully out of the spent fuel pool, water was removed from inside the canister in Steps 8.7.12 through 8.7.24 by inserting a short tube into the drain port that extended approximately 8" below the water level. The water was drained out until the tube lost suction. The eight inches was well above the top of the spent fuel in the canister. In this way, water could be removed to reduce spillage during canister movement from the spent fuel pool to the dry cask pit. Removing the water was also required for welding to prevent water contact with the lid resulting in a heat sink. During welding in the dry cask pit, the water remained in the canister covering the spent fuel and an argon purge was used to displace the air in the gap between the water level and the lid. The argon purging requirement was incorporated into Procedure HSP-504, Section 6.3 "MPC Lid Fit and Argon Gas Purge." Later in Procedure DCS-204, after the lid was welded in place, water was totally removed from the canister in Section 8.9 "FHD System MPC Blowdown Operations."

This was accomplished by using 99.995% purity helium to force the water out of the canister. After successful drying of the spent fuel in a helium environment, the canister was backfilled with helium in Procedure DCS-204, Section 8.12 "FHD System Helium Backfill Operations." At no time throughout the canister loading and sealing process was air allowed to come into contact with the spent fuel.

Documents (a) Procedure DCS-203 MPC Handling and Fuel Loading Operations Revision 1 (b)

Reviewed: Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Revision 2 (c) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on MPC and MPC Lid," Revision 6 Page 74 of 149

Category: Loading Operations Topic: Handling Damaged Fuel Containers Reference: FSAR 1014, Sections 6.4.4 and 8.0 Revision 9 Requirement: Damaged fuel assemblies and fuel debris shall be loaded into damaged fuel containers (DFCs) prior to being loaded into the canister.

Observation: The planned loading of the first three canisters did not include provisions for damaged fuel or damaged fuel containers. The fuel selection Procedure NUC-212, Step 5.1 stated

"Do not store fuel which is not classified as intact (this restricts any known or suspected leaking fuel) into a dry cask MPC." Step 5.2 stated "Any known or suspected leaking fuel assemblies should not be considered intact." Step 3.7 defined intact fuel assemblies as "Fuel assemblies without known or suspected cladding defects greater than pinhole leaks or hairline cracks and which can be handled by normal means." This definition was consistent with the definition of intact fuel in the Certificate of Compliance 1014 Technical Specifications (Appendix A), Section 1.1 "Definitions."

Documents (a) Procedure NUC-212 Spent Fuel Limits for Dry Cask Operations, Revision 1 Reviewed:

Category: Loading Operations Topic: Pressure Relief Valves Reference: FSAR 1014, Table 8.0.1 Revision 9 Requirement: Pressure relief valves in the water and gas processing systems limit the canister pressure to acceptable levels. FSAR Figures 8.1.20, 8.1.21 and 8.1.23 provide drawings showing the pressure relief valves for the pressure testing, canister blowdown, and helium backfill.

Observation: The required pressure relief valves were in place for pressure testing, blow down, and helium backfill operations. For the hydrostatic pressure testing operations, a 140 pounds per square inch-gauge (psig) relief valve (FSV-1) was located on the drain line removable valve operator assembly (RVOA) in accordance with Procedure DCS-204, Attachment 10.1.6 "Hydrostatic Test System Set-Up." The Holtec Final Safety Analysis Report (FSAR), Figure 8.1.20 "MPC Lid-to-Shell Pressure Testing Example P&ID" showed a relief valve on the vent port RVOA. This discrepancy was brought to the attention of the licensee during the canister fluid operations dry run. On June 1, 2011, a phone conversation with Holtec's Licensing Manager and engineering personnel confirmed that the difference was acceptable and the relief valve on either side of the system would provide the same protection. For the blow down operations, Procedure DCS-204, Attachment 10.1.7 "FHD System Piping Diagram" showed three pressure relief valves on the system. On the drain line RVOA was the 140 psig relief valve (FSV-1), on the vent line RVOA was a 95 psig relief valve (FSV-2), and on the line to the helium bottles was a 3,000 psig relief valve (VS-2). This arrangement met the requirements of Holtec FSAR, Figure 8.1.21 "MPC Blowdown Example P&ID" for the blow down operations. The helium backfilling operation had the same relief valve arrangement as the blowdown arrangement. This arrangement of relief valves met the requirement of Holtec FSAR, Figure 8.1.23 "Helium Backfill System Example P&ID."

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Page 75 of 149

Category: Loading Operations Topic: Spent Fuel Pool Boron Concentration Reference: CoC 1014, Tech Spec A.3.3.1 Amendment 7 Requirement: The boron concentration in the canister shall meet the limits specified in Technical Specification 3.3.1 for the applicable canister model and the most limiting fuel assembly array and class. The boron concentration must be verified within limits using two independent measurements taken within four hours of filling the canister with water, and every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> thereafter while fuel and water are in the canister. If the boron is not within limits, the licensee must suspend loading or unloading operations, suspend positive reactivity additions, and initiate actions to restore boron concentration to within limits.

Observation: The spent fuel pool boron concentration limits from Technical Specification A.3.3.1 had been incorporated into Procedures DCS-203 and DCS-204 for cask loading and DCS-207 for cask unloading. Step 5.11 of both loading procedures provided a table for the boron limits consistent with Technical Specification A.3.3.1. This same table was incorporated into Step 5.18 of Procedure DCS-207 for unloading a canister. Procedure DCS-203, Step 8.2.10 and 8.2.11 documented the sampling and analysis of the spent fuel pool boron levels prior to filling the canister with spent fuel pool water and lowering it into the wet cask loading pit. Two independent spent fuel pool water samples were required within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of placing the fuel into the canister. Step 8.2.12 documented that the cask loading supervisor had verified that the sample results met the minimum boron concentrations of Step 5.11 and as listed in Form NUC-212-4 "Cask Acceptability Report." The cask acceptability report was a document issued by the core performance engineer prior to loading each canister and listed the key parameters applicable to the particular canister being loaded. Under the heading "CoC Tech Spec Limits based on Cask Loading" of the cask acceptability report, the required minimum boron concentration limit for the individual canister being loaded was stated, consistent with the Technical Specification A.3.3.1 requirements. Procedure DCS-203, Step 8.2.13 required the time for the next boron sampling to be documented on the cask loading supervisor status board. The time listed on the status board was calculated as 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the previous sample, as required by Technical Specification Surveillance Requirement A.3.3.1.1.

Both Procedure DCS-203 and Procedure DCS-204 directed chemistry personnel to perform the boron sampling and to ensure the boron levels were above the technical specification limits. Both procedures referenced Procedure DSC-507 for conducting the sampling. Procedure DCS-507, Step 2.5.3 directed chemistry personnel to notify the operations shift manager and the cask loading supervisor if any of the acceptance criteria was not met. Attachment 1 "Appendix A Certification of Compliance No. 1014" provided a table listing the Technical Specification A.3.3.1 requirements including the statement that if the boron concentrations were not within limits, then loading (or unloading) operations were to be suspended, any positive reactivity additions were to be suspended, and actions were to be initiated to restore boron concentrations to within the limits.

Procedure DCS-204 included instructions for conducting the hydrostatic test of the canister. The hydrostatic test was performed using water from the wet cask loading pit.

As such, the boron concentrations used for the test required confirmation that the requirements of Technical Specification A.3.3.1 for boron minimum levels were met.

Page 76 of 149

Procedure DCS-204, Steps 6.15 through 6.19 and Steps 8.4.1 through 8.4.4 included the requirements to take boron samples from the wet cask loading pit, confirm they were within the technical specification limits, and record the next sample time on the cask loading supervisors status board prior to starting the hydrostatic test. Step 5.25 identified Attachment 10.1.22 "Additional Boron Concentration Sampling" as the attachment to use if additional boron sampling was necessary due to extended or prolonged time frames related to delays in the cask operations.

Boron sampling of the wet cask loading pit was also required if the water circulation system was used. The water circulation system took water from the wet cask loading pit to provide cooling to the canister if the time-to-boil limit was approaching. Procedure DCS-204, Step 8.2 directed the cask loading supervisor to monitor the time-to-boil clock to determine the need for the water circulation system during the welding of the canister lid. Step 5.7 required deployment of the water circulation system if canister blowdown could not be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the time to boil limit and referenced Attachment 10.1.3 "Water Circulation System Set-Up and Operations." Attachment 10.1.3, Step 1.1 required wet cask loading pit sampling for boron and verification that the Technical Specification A.3.3.1 limits were met prior to initiating operation of the system.

For unloading, Procedure DCS-207 had incorporated the Technical Specification A.3.3.1 requirements for sampling, including the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> limits into Step 5.18. Step 5.18. 3 referenced DCS-507 as the applicable procedure for conducting the boron sampling of the spent fuel pool and canister. Section 8.6 "MPC Water Reflooding" referenced Technical Specification A.3.3.1 and included blank spaces to record the boron concentration limits, time for the next sampling, and approval by the cask loading supervisor that the boron concentration limits were met.

Documents (a) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (b)

Reviewed: Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Revision 2 (c) Procedure DCS-207 "Unloading a Loaded MPC," Revision 2 (d)

Procedure DCS-507 "Chemistry Sampling of the Multi-Purpose Canister (MPC),"

Revision 0 (e) Procedure NUC-212 Spent Fuel Limits for Dry Cask Operations Revision 2 Category: Loading Operations Topic: Time-to-Boil Time Limits Reference: FSAR 1014, Section 4.5.2 and Table 4.5.3 Revision 9 Requirement: Wet transfer operations begin when the lid is placed on the canister in the spent fuel pool and end when the canister is blown down following pressure testing. During wet operations, the heat up rate for a canister with a decay heat load of 38 kW (maximum allowed), is 4.99 degrees per hour, assuming the canister cavity and transfer cask annulus and water jacket are full of water. Using this heat up rate, Table 4.5.3 of the FSAR provides the time-to-boil for various initial water temperatures. If wet transfer operations cannot be completed prior to boiling, a forced water circulation shall be initiated and maintained to remove decay heat from the canister cavity. A minimum flow rate of 10.5 gpm shall be established into the canister drain port and out of the vent port.

Observation: The time-to-boil time limits and requirements were incorporated into Procedures DCS-Page 77 of 149

203 and DCS-204. Procedure DCS-203, Step 8.6.25.1 recorded the time the lid was placed on the canister while in the spent fuel pool after the spent fuel had been loaded.

Steps 8.6.25.2 and 8.6.25.3 referenced Attachment 10.1.3 "Time-to-Boil Worksheet."

Attachment 10.1.3 calculated a canister heat load specific time-to-boil deadline by use of the equation: t = [(212 degree F - Pool temperature) x 26,032 Btu/degree F]/[(canister kW x 3414.43)(Btu/hr)]. This equation was consistent with the method used to calculate the Holtec Final Safety Analysis Report (FSAR) values in Table 4.5.3 "Maximum Allowable Time for Wet Transfer Operations." FSAR Section 8.1, on page 8.1-13, allowed the cask user to perform a site specific analysis to determine the time-to-boil limit.

Procedure DCS-204, used for sealing, drying, and backfilling the canister, required the cask loading supervisor to monitor the time-to-boil limit until the water had been removed from the canister following the hydrostatic pressure testing (blow down completed). If blow down could not be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the time-to-boil limit, Attachment 10.1.3 required the systems to be placed in a safe condition and the water circulation system deployed. Attachment 10.1.3 provided set-up instructions for the water circulation system for two scenarios based on whether the root pass weld on the lid had been completed. The difference between the two scenarios was an additional pump and flow meter that was added to the vent port side to provide suction if the root pass had not been completed. This second pump was intended to provide a means to balance the amount of water going in with the amount of water coming out to prevent an overflow condition along the sides of the lid where welding had not been completed. In both scenarios, water was injected into the drain port and removed through the vent port.

For the scenario with the welding of the lid not completed, Steps 3.2.2 through 3.2.5 provided instructions to maintain a higher flow rate out of the canister than the flow rate into the canister. The suction side through the vent port relied on a short suction tube that would continue to draw water until suction was lost. By doing this, a gap the length of the suction tube, approximately 8" below the bottom of the lid, was maintained without creating a situation where water could be drained below the top of the spent fuel. Steps 3.1.4 and 3.2.6 of Attachment 10.1.3 required the canister inlet water flow to be maintained at a flow rate greater than 12 gpm as indicated by flow meter FM-12.

Documents (a) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (b)

Reviewed: Procedure DCS-204 MPC Closure Operations (Sealing, Drying, and Backfilling),

Revision 2 (c) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Category: NDE-Helium Leak Testing Topic: Helium Leak Test-Vent/Drain Covers Reference: CoC 1014, Tech Spec A.3.1.1.3 Amendment 7 Requirement: The helium leak rate through the canister vent and drain port confinement welds shall meet the leak tight criteria of ANSI N14.5 (1997). This degree of containment is achieved by demonstration of a leakage rate less than or equal to 2 X 10(-7) atm*cc/sec of helium at an upstream pressure of 1 atmosphere (atm) absolute (abs) and a downstream pressure of 0.01 atm abs or less.

Observation: The helium leak test was demonstrated during the welding dry run May 2-4, 2011 Page 78 of 149

consistent with the acceptance standards specified in Certificate of Compliance 1014 and ANSI N14.5-1997. Procedure MSLT-MPC-Holtec, Step 8.1 required that the total leak rate of the vent port plus the drain port be less than or equal to 2 x 10(-7) atm cc/sec helium. A Varian Model 959M helium leak detector was used to measure the leak rate on the vent and drain ports. The sensitivity of the helium leak rate detector was calculated to be 2.0 x 10(-9) atm cc/sec. The upstream pressure was 1 atmosphere and the downstream pressure was 10 millitorr [1.3 x 10(-6) atmosphere] during the test.

Documents (a) Leak Testing Specialists Inc. Procedure MSLT-MPC-Holtec "Helium Mass Reviewed: Spectrometer Leak Test Procedure Multipurpose Canister," Revision: 3660-CP-00 (b)

American National Standard Institute (ANSI) N14.5 "Leakage Tests on Packages for Shipment for Radioactive Materials," Revision 1997 Category: NDE-Helium Leak Testing Topic: HMSLD Minimum Sensitivity Reference: ANSI N14-5, Section 8.4 Revision 1997 Requirement: The helium mass spectrometer leak detector (HMSLD) shall have a minimum sensitivity of 1/2 the acceptance leak rate. For example, a package with a leak tight acceptance criteria of 1.0 X 10(-7) ref-cc/sec requires a minimum helium mass spectrometer leak detector sensitivity of 5.0 x 10(-8) ref-cc/sec. This sensitivity requirement applies to both the hood and detector probe methods. The helium mass spectrometer leak detector shall be calibrated to a traceable standard.

Observation: Procedure MSLT-MPC-Holtec, Step 7.1 required the helium mass spectrometer leak detector (MSLD) sensitivity to be less than 1/2 the acceptance criteria leak rate. The minimum acceptable leak rate was stated in Section 8.0 as 2.0 x 10(-7) atm-cc/sec (He)

and referenced Technical Specification 3.1.1.3. Technical Specification 3.1.1.3 required the helium leak rate through the canister vent and drain port confinement welds to meet the leak tight criteria specified in ANSI N14.5-1997. Section 2 "Definitions" of ANSI N14.5-1997defined leak tight as having a leak rate of 1 x 10(-7) ref-cc/sec of air. A note to the definition stated that 1 x 10(-7) ref-cc/sec of air was equivalent to 2 x 10(-7) atm-cc/sec of helium. A calibration standard traceable to the National Institute of Standards and Technology (NIST) with a leak rate between 10(-6) and 10(-8) atm-cc/sec was required by Step 4.3 of Procedure MSLT-MPC-Holtec. Section 5.0 "MSLD Startup and Instrument Calibration" provided instructions on calibrating the helium mass spectrometer.

Documents (a) Leak Testing Specialists Inc. Procedure MSLT-MPC-Holtec "Helium Mass Reviewed: Spectrometer Leak Test Procedure Multipurpose Canister," Revision 3660-CP-00. (b)

American National Standards Institute (ANSI) N14.5 "Leak Testing on Packages for Shipment of Radioactive Materials," 1997 Category: NDE-Liquid Penetrant Topic: Acceptance Criteria Reference: ASME Section III, Article NB-5352 Code Year 1995 Requirement: Only indications with major dimensions greater than 1/16 inch should be considered relevant. The following relevant indications are unacceptable: (1) any cracks or linear indications. Linear indications have a length at least 3 times greater than the width; (2)

rounded indications with dimensions greater than 3/16 inch (4.8 mm); (3) more than Page 79 of 149

four rounded indications in a line, separated by 1/16 inch (1.6 mm) or less edge to edge; and (4) more than ten rounded indications in any 6 square inch area in the most unfavorable location relative to the indications being evaluated.

Observation: Procedure HSP-506, Section 9. Acceptance Criteria provided the criteria to determine if material examinations were acceptable during the liquid penetrant examinations.

Section 9.3 "Weld Examinations" referenced the criteria in NB-5352 and stated that imperfections producing indications with major dimensions greater than 1/16" shall be considered relevant imperfections. Relevant imperfections were unacceptable if they met the four NB-5352 criteria, which were listed in Step 9.3.1.2 of the procedure.

Documents (a) Holtec Procedure HSP-506 Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding, Revision 2 Category: NDE-Liquid Penetrant Topic: Contaminants Reference: ASME Section V, Article 6, T-641 Code Year 1995 Requirement: The user shall obtain certification of contaminant content for all liquid penetrant materials used on austenitic stainless steels. The certifications shall include the manufacturers batch number and sample results. Sub-article T-641(b) limits the total halogen (chlorine plus fluorine) content of each agent (penetrant, cleaner and developer)

to 1.0 weight percent (wt.%) when used on austenitic stainless steels.

Observation: Certified mill test reports with chemical analysis for the materials used for the high temperature liquid penetrant examinations documented that the cleaner solvent, developer, and dye penetrant contained less than 1.0 weight percent halogen. Qualified materials for high temperature applications were listed in Procedure HSP-506, Section 5.0 "Penetrant Materials" and matched the products listed on the certified mill test report. The licensee expected all casks loaded in the first campaign to exceed the 125 degree F limit requiring high temperature materials, and as such, did not plan to use low temperature application solvent, developer, and dye penetrant. Procedure HSP-506, Step 5.4 required chemical analysis for any penetrant materials used. Step 5.4 stated Certification of contaminant content shall be provided by the manufacturer for each batch of penetrant material. Certification records shall be maintained by the Quality Department and made available for review by the Authorized Nuclear Inspector or Customer.

Documents (a) Holtec Procedure HSP-506 Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding, Revision 2 (b) Sherwin Incorporated Certified Mill Test Report with Chemical Analysis for DUBL-CHEK D-350 Batch No. 019-D71, DUBL-CHEK KO-19 Aerosol Batch No. 08-D56, and DUBL-CHEK KO-17 Batch No. 07-B54 Category: NDE-Liquid Penetrant Topic: Final Interpretation Reference: ASME Section V, Article 6, T-676.1 Code Year 1995 Requirement: Final interpretation shall be made after allowing the penetrant to bleed-out for 7-30 minutes under standard temperatures (50 and 125 degrees F). The 7-30 minute clock starts immediately after application of a dry developer.

Observation: Procedure HSP-506 had incorporated the required time frames from the ASME code for Page 80 of 149

the final interpretation using a dry developer. Procedure HSP-506, Step 8.1 stated Inspection and evaluation shall be made no sooner than 10 minutes after the developer coating has dried and no later than 30 minutes after the application of the developer.

This requirement applied to normal temperatures. For the dye penetrant examination of the canisters loaded at Comanche Peak, the temperature of the weld will exceed 125 degree F. As such, high temperature penetrant, solvent, and developer will be required.

Procedure HSP-506 included instructions for conducting the dye penetrant examination using both normal temperature and high temperature penetrant materials. For high temperature applications, a minimum of 2 minutes and maximum of 7 minutes was required by Step 8.1.1. The high temperature liquid penetrant examination described in Procedure HSP-506 had been qualified by Holtec by performing a qualification test in the temperature range of 120 - 350 degree F on May 12, 2011. The qualification test was approved by a qualified Level III examiner. The welding dry run demonstration conducted May 2-4, 2011 used the high temperature materials and time requirements.

Documents (a) Holtec Procedure HSP-506 Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding, Revision 2 (b) Holtec Procedure Qualification Demonstration for HSP-506, approved May 12, 2011 Category: NDE-Liquid Penetrant Topic: Lid-To-Shell Weld PT Reference: CoC 1014, Appendix B, Table 3-1 Amendment 7 Requirement: Only ultrasonic testing or multi-layer liquid penetrant (PT) examination is permitted on the lid-to-shell weld. If PT alone is used, at a minimum, it will include the root and final weld layers and each approximately 3/8 inch of weld depth.

Observation: Liquid penetrant examination was required by Procedure HSP-504 to be performed on the root pass weld, final weld, and any intermediate welds that exceed 3/8" or when combined with preceding welds added to 3/8" weld material deposition since the last liquid penetrant examination. The liquid penetrant examination was required on the root pass weld per Step 6.4.4. Step 6.4.7 required measuring and recording the depth of weld material for the first fill layer after the root pass to ensure less than 3/8" was deposited.

The liquid penetrant examination was required after the first fill layer in Step 6.4.9 if the weld deposit exceeded 3/8". Step 6.4.11 directed the welders to place the second fill layer and Step 6.4.11.2 summed the first and second welds (if a liquid penetrant examination was not performed on the first fill layer) to verify they did not exceed 3/8".

If the 3/8" was exceeded, Step 6.4.14 required that a liquid penetrant examination be performed on the weld. This same concept of adding weld depths to verify the 3/8" was not exceeded between liquid penetrant examinations was included in the following sections of Procedure HSP-504 until the final pass weld was performed. Step 6.4.24 required a liquid penetrant examination of the final weld. In addition, Step 6.5.2 required a liquid penetrant examination of the weld after the hydrostatic test was performed on the canister.

Documents (a) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC,"

Reviewed: Revision 6

.

Page 81 of 149

Category: NDE-Liquid Penetrant Topic: Liquid Penetrant Testing - Permanent Record Reference: CoC 1014, Appendix B, Table 3-1 Amendment 7 Requirement: The inspection results, including findings (indications), shall be made a permanent part of the user's records by video, photographic, or other means which provide an equivalent retrievable record of weld integrity. The video or photographic records should be taken during the final interpretation period described in ASME Section V, Article 6, T-676.

Observation: The final inspection results were designated as a permanent part of the users records.

Procedure HSP-506, Step 4.2 required the results of the liquid penetrant tests be recorded on Exhibit 12.1 "Liquid Penetrant Inspection Test Report." Section 11.0

"Documentation" stated that Exhibit 12.1 shall be included in the completed weld package for retention.

Documents (a) Holtec Procedure HSP-506 "Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Liquid Penetrant Topic: Minimum Elements Reference: ASME Section V, Article 6, T-621 Code Year 1995 Requirement: Each liquid penetrant (PT) procedure shall include the: (1) materials, shapes or sizes to be examined and the extend of the examination; (2) type of each penetrant, remover, emulsifier, and developer; (3) pre-examination cleaning and drying, including the cleaning materials used and minimum time allowed for drying; (4) applying the penetrant, the length of time the penetrant will remain on the surface (dwell time), and the temperature of the surface during examination; (5) removing excess penetrant and drying the surface before applying the developer; (6) details for applying the developer, and length of developing time before interpretation; and (7) post-examination cleaning.

Observation: The minimum elements required for the penetrant testing were incorporated into Procedure HSP-506. Section 1.2 stated: The shapes and material covered under this process are groove welds, fillet welds and base metal repair welds consisting of an acceptable weld joint. The base materials for these processes are stainless steel.

Section 5.0 Penetrant Materials listed the specific penetrant, solvent, and developer to be used by name. Section 6.0 Surface Finish and Cleaning specified pre-examination cleaning (solvent cleaner) and minimum drying times of 5 minutes prior to the application of the penetrant. High temperature applications shall have a minimum drying time of 30 seconds. Section 4.4 defined general application as 60 degree F to 125 degree F. High temperature application was 125 degree F to 350 degree F. The approved solvent cleaner was listed in Section 5.0. Section 7.0 Application provided instructions for both high temperature and regular temperature application and the dwell time for the developer. The dwell time was listed in Step 7.2 as 10 - 60 minutes for normal temperatures and 1 - 3 minutes for high temperatures. Step 7.3 required removal of excess penetrant after the dwell time was met using a clean, dry, lint free cloth prior to applying the developer. Step 7.4 specified a 5 minute drying time after the excess penetrant was removed prior to applying the developer for normal temperatures and immediate application of the developer for high temperatures. Step 8.1 specified a 10 -

30 minute time frame after the developer coating had dried before starting the inspection for normal temperatures. For high temperatures, the final interpretation shall be made after allowing the penetrant to bleed-out for a minimum of 2 minutes to a maximum of 7 Page 82 of 149

minutes. Section 10.0 Post Examination Cleaning required all excess developer be removed and the surface cleaned with a solvent soaked cloth to remove any examination material.

Documents (a) Holtec Procedure HSP-506 Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding, Revision 2 Category: NDE-Liquid Penetrant Topic: Removing Excess Penetrant Reference: ASME Section V, Article 6, T-673.3 Code Year 1995 Requirement: Excess solvent removable penetrants shall be removed by wiping with a cloth or absorbent paper until most traces of the penetrant have been removed. The remaining traces shall be removed by lightly wiping the surface with a cloth or absorbent paper moistened with solvent. Care shall be taken to avoid the use of excess solvent. Flushing the surface with solvent, following the application of the penetrant and prior to developing, is prohibited.

Observation: Procedure HSP-506, Step 7.3 and 7.3.1 incorporated the requirement of ASME Section V, Article 6, T-673.3. The steps stated: After the dwell time has elapsed, the excess penetrant shall be removed using clean, dry and lint-free cloth or absorbent towel. The remaining traces shall be removed by wiping the surface lightly with a solvent dampened cloth or towel. Avoid use of excess solvent as this may remove penetrant from discontinuities. Flushing the surface with solvent is prohibited.

Documents (a) Holtec Procedure HSP-506 Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding, Revision 2 Category: NDE-Liquid Penetrant Topic: Surface Preparation Reference: ASME Section V, Article 6, T-642 (b) Code Year 1995 Requirement: Prior to each liquid penetrant examination, the surface to be examined and all adjacent areas within one inch must be dry and free of all dirt, grease, lint, scale, welding flux, weld spatter, paint, oil, and other extraneous matter that could obscure surface openings or otherwise interfere with the examination.

Observation: Procedure HSP-506, Step 6.2 incorporated the requirement from ASME Section V, Article 6, T-642(b). Step 6.2 stated Prior to liquid penetrant testing, the surface to be inspected as well as all adjacent areas within at least 1" (if applicable) shall be clean, dry and free of any dirt or grease, lint, scale, welding flux, weld spatter, paint, oil, and other extraneous matter that could obscure surface openings. The area to be inspected will be pre-cleaned with solvent cleaner to assure proper surface preparation.

Documents (a) Holtec Procedure HSP-506 Liquid Penetrant Examination for MPC Field Closure Reviewed: Welding, Revision 2 (b) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC," Revision 6

.

Page 83 of 149

Category: NDE-Personnel Qualification Topic: Certification Records Reference: SNT-TC-1A, Section 9 Revision 1992 Requirement: Certification records should contain the name of the certified individual, the certification level and method, the individual's educational background and NDE experience, a statement of satisfactory completion of training per the employers written practice, visual examination results, evidence of successful completion of examinations including grades, date of certification, and the signature of the employer.

Observation: The certification records for the individuals scheduled to perform the nondestructive examinations (NDE) for the first loading campaign contained all the required information in accordance with SNT-TC-1A, Section 9 including education and background, examination scores, certification level, training hours, visual acuity exam results, date of certification and the certification expiration date. The certification records reviewed for the visual and liquid penetrant examiner will expire on January 9, 2016. The certification records reviewed for the helium mass spectrometer leak detector examiner will expire on February 13, 2013.

Documents (a) Certification Record for David Hecksel, Level III HMSLD, expiration date February Reviewed: 13, 2013 (b) Certification Record for Michael Mangan, Level III VT, expiration date January 9, 2016 (c) Certification Record of Michael Mangan, Level III PT, expiration date January 9, 2016 (d) American Society for Nondestructive Testing, Inc. SNT-TC-1A Recommended Practices for Qualification and Certification of nondestructive testing Personnel, Revision 1992 Category: NDE-Personnel Qualification Topic: Level III Candidates Reference: SNT-TC-1A, Section 6 Revision 1992 Requirement: A Level III candidate who has completed less than 2 years of engineering or science study must have 4 years of experience comparable to a Level II. A Level III candidate who has completed 2 years of engineering or science study must have 2 years of experience comparable to a Level II. A Level III candidate who has completed 4 years of engineering or science study must have 1 year of experience comparable to a Level II.

Observation: All Level III examiners met the education and experience requirements as documented in their certification record. The SNT-TC-1A education and experience requirements for Level III examiners was listed in Holtec Procedure HQP-9.1, Step 6.1.3 and were consistent with the SNT-TC-1A, Section 6 requirement. The visual and liquid penetrant examiner had over 11 years of experience. The helium mass spectrometer leak detector examiner had a 2 year degree and over 9 years of leak testing experience.

Documents (a) Holtec Procedure HQP-9.1 "Written Practice for Qualification of NDE Personnel,"

Reviewed: Revision 13 (b) Certification Record for David Hecksel, Level III HMSLD, expiration date February 13, 2013 (c) Certification Record for Michael Mangan, Level III VT, expiration date January 9, 2016 (d) Certification Record of Michael Mangan, Level III PT, expiration date January 9, 2016 (e) American Society for Nondestructive Testing, Inc. SNT-TC-1A Recommended Practices for Qualification and Certification of nondestructive testing Personnel, Revision 1992 Page 84 of 149

Category: NDE-Personnel Qualification Topic: Level III Exam Grading Reference: SNT-TC-1A, Section 8 Revision 1992 Requirement: Level III examiners take 3 examinations: Basic, Method, and Specific. A composite grade should be determined by simple averaging of the results of the 3 examinations. A passing composite grade should be 80% with no one examination below 70%.

Observation: The persons scheduled to perform the nondestructive examinations (NDE) on the first loading campaign had passed their latest three exams with an average composite grade above 80%, with none falling below 70%. Procedure HQP-9.1 established the qualification requirements for the NDE personnel. Section 6.3.6 required all Level III personnel to have passed their three exams with a minimum composite score of 80%

with no test scores less that 70%. The Level III examiner scheduled to perform the visual testing procedure had a composite score of 88.3% (with none under 70%), which was documented on his certification record. The Level III examiner scheduled to perform the penetrant testing procedure had a composite score of 95.6% (with none under 70%), which was documented on his certification record. The Level III examiner scheduled to perform the helium leak testing procedure had a composite score of 93.3%

(with none under 70%), which was documented on his certification record.

Documents (a) Holtec Procedure HQP-9.1 "Written Practice for Qualification of NDE Personnel,"

Reviewed: Revision 13 (b) Certification Record for David Hecksel, Level III HMSLD, expiration date February 13, 2013 (c) Certification Record for Michael Mangan, Level III VT, expiration date January 9, 2016 (d) Certification Record of Michael Mangan, Level III PT, expiration date January 9, 2016 (e) American Society for Nondestructive Testing, Inc. SNT-TC-1A Recommended Practices for Qualification and Certification of nondestructive testing Personnel, Revision 1992 Category: NDE-Personnel Qualification Topic: Recertification of Personnel Reference: SNT-TC-1A, Section 9 Revision 1992 Requirement: Maximum recertification intervals are 3 years for Levels I and II, and 5 years for Level III. Recertification may be granted without testing provided there is documented continuing satisfactory performance. Without documented continuing satisfactory performance, reexamination is required for those sections deemed necessary by the Level III examiner.

Observation: The SNT-TC-1A recertification requirements were incorporated into Procedure HQP-9.1 for certifying nondestructive examination personnel (NDE). The recertification requirements were established in Step 6.6 which required recertification of Level I and II NDE personnel every three years. Recertification could be based on either evidence of continuing satisfactory performance or by reexamination as deemed necessary by the designated Level III individual. Level III personnel were recertified every 5 years through examination. At the discretion of Holtec, for Holtec NDE personnel, reexamination could be required at any time. An individual who had not performed NDE for one year was required to be recertified prior to performing any NDE work.

Documents (a) Holtec Procedure HQP-9.1 "Written Practice for Qualification of NDE Personnel,"

Reviewed: Revision 13 (b) American Society for Nondestructive Testing, Inc. SNT-TC-1A Recommended Practices for Qualification and Certification of nondestructive testing Personnel, Revision 1992 Page 85 of 149

Category: NDE-Personnel Qualification Topic: Visual Acuity Reference: SNT-TC-1A, Section 8.2 Revision 1992 Requirement: The NDE examiner should have natural or corrected near-distance acuity in at least one eye capable of reading Jaeger Number 1 at a distance of not less than 12 inches on a standard Jaeger test chart, or capable of perceiving a minimum of 8 on an Ortho-Rater test pattern. This should be verified annually. The NDE examiner should demonstrate the capability of distinguishing and differentiating contrast among colors used in the applicable method. This should be verified every 3 years.

Observation: The personnel assigned to perform the nondestructive examinations (NDE) during the first loading campaign had been properly tested for visual acuity. Procedure HQP-9.1, Step 6.3.1. required each NDE examiner to pass a natural or corrected near-distance acuity in at least one eye such that the individual was capable of reading a minimum of Jaeger Number 1 or equivalent type and size letters at a distance of not less than 12 inches on a standard Jaeger test chart. Additionally, the NDE examiners were required to demonstrate the capability of distinguishing and differentiating contrast between colors used in the method of examination. All exams were administered on an annual basis. The visual and liquid penetrant examiner's visual acuity record documented that he had passed the visual acuity test on March 3, 2011. The helium leak testing specialist had passed the visual acuity exam on June 20, 2011. Both test records documented that the individuals were capable of distinguishing contrasts between colors.

Documents (a) Holtec Procedure HQP-9.1 "Written Practice for Qualification of NDE Personnel,"

Reviewed: Revision 13 (b) Visual Acuity Exam Record for David Hecksel, dated June 20, 2011 (c)

Visual Acuity Test Record for Michael Mangan, Level III VT, dated March 3, 2011 (d)

American Society for Nondestructive Testing, Inc. SNT-TC-1A Recommended Practices for Qualification and Certification of nondestructive testing Personnel, Revision 1992 Category: NDE-Personnel Qualification Topic: Written Practice Reference: SNT-TC-1A, Section 5 Revision 1992 Requirement: The employer shall establish a written practice for control and administration of non-destructive testing personnel training, examination and certification. The written practice should describe the responsibility of each level of certification for determining the acceptability of material or components. The written practice shall describe the training experience and examination requirements for each level of certification.

Observation: Procedure HQP-9.1 established the controls and administration of the non-destructive testing personnel training, examination, and certification program. Procedure HQP-9.1 described the responsibility, training experience, and examination requirements for each of the three levels of certification. Section 4.0 "Discussion" listed the expected non-destructive examination (NDE) skills of a person qualified at each level. Section 6.1

"Education, Training, and Experience Requirements for initial Qualification" defined the years of formal college education, years of actual experience, and years of training required for each certification level. Exhibit 9.1 "Training and Experience Time" provided a detail list of training hours and experience hours required for each of the Page 86 of 149

types of non-destructive examination techniques. Section 6.3 "Examinations" provided the requirements for the physical exams, written exams and practical exams for each of the three qualification levels. A minimum composite grade of 80% with no individual exam below 70% was required. Exhibit 9.1.2 "Examination Requirements" provided a table with the minimum number of questions required for Level I and Level II exams for each of the NDE certification areas. Level III NDE personnel were required by Step 6.1.3 to meet Level II requirements or equivalent, plus additional training or education.

Documents (a) Holtec Procedure HQP-9.1 "Written Practice for Qualification of NDE Personnel,"

Reviewed: Revision 13 (b) American Society for Nondestructive Testing, Inc. SNT-TC-1A Recommended Practices for Qualification and Certification of nondestructive testing Personnel, Revision 1992 Category: NDE-Visual Examination Topic: Acceptance Criteria - Arc Strikes Reference: ASME Section III, Article NF-5360 (i) Code Year 1995 Requirement: Arc strikes and blemishes in the weld or base material are acceptable, provided no cracking is visually detected.

Observation: The arc strike acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.9 stated "Arc strikes and associated blemishes on the weld or in the base material are acceptable provided no cracking is visually detected."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Cracks Reference: ASME Section III, Article NF-5360 (a) Code Year 1995 Requirement: Cracks are unacceptable.

Observation: The ASME crack criteria was incorporated into the visual weld examination procedure.

Procedure HSP-507, Step 8.2.1 stated that during the visual examination, cracks are unacceptable.

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Craters Reference: ASME Section III, Article NF-5360 (e) Code Year 1995 Requirement: Craters outside the weld area are irrelevant, provided there are no cracks.

Observation: The crater acceptance criteria was incorporated into the visual weld examination procedure. The crater acceptance criteria was incorporated into Procedure HSP-507.

Step 8.2.5 which stated "Craters are acceptable when the criteria for weld size are met.

Craters that occur outside the specified weld length are irrelevant provided there are no cracks."

Page 87 of 149

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Fusion Reference: ASME Section III, Article NF-5360 (c) Code Year 1995 Requirement: For fillet welds, incomplete fusion of more than 3/8" (10 mm) in any 4" (102 mm)

segment is unacceptable. For fillet welds, incomplete fusion of more than 1/4" (6 mm) in welds less than 4" (102 mm) is unacceptable. For groove welds, any incomplete fusion is unacceptable. Rounded end conditions (starts and stops) shall not be considered indications of incomplete fusion.

Observation: The incomplete fusion acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.3 stated "In fillet welds, incomplete fusion of 3/8" in any 4" segment and 1/4" in welds less than 4" long is acceptable. For groove welds, incomplete fusion is not acceptable. For fillet and groove welds, rounded end conditions that occur in welding (starts and stops) shall not be considered indications of incomplete fusion and are irrelevant."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Lengths Reference: ASME Section III, Article NF-5360 (h) Code Year 1995 Requirement: For welds 3" and longer, weld lengths shorter than specified by more than 1/4" (6 mm)

are unacceptable. For welds less than 3" long, weld lengths shorter than specified by more than 1/8" (3.2 mm) are unacceptable. Intermittent welds not spaced within 1" (25 mm) of the specified location are unacceptable.

Observation: The acceptance weld length criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.8 stated "The length and location of welds shall be as specified on the detail drawing, except that weld lengths may be longer than specified. For weld lengths less than 3", the permissible under-length is 1/8" and for welds 3" and longer, the permissible under-length is 1/4". Intermittent welds shall be spaced within 1" of the specified location."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Overlap Reference: ASME Section III, Article NF-5360 (d) Code Year 1995 Requirement: When fusion in the overlap length cannot be verified, an overlap length of greater than 3/8" (10 mm) in any 4" (102 mm) segment, and 1/4" (6 mm) in welds less than 4" (102 mm) long, is unacceptable.

Observation: The overlap acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.4 stated "Overlap is acceptable provided the criteria for weld size and fusion can be satisfied. When fusion in the overlap length Page 88 of 149

cannot be verified, an overlap length of 3/8" in any 4" segment and 1/4" in welds less than 4" long, is acceptable."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Porosity Reference: ASME Section III, Article NF-5360 (g) Code Year 1995 Requirement: The following degrees of random porosity are unacceptable: (1) the sum of the diameters of random porosity exceeding 3/8" (10 mm) in any one linear inch of weld; (2) the sum of the diameters of random porosity exceeding 3/4" (19 mm) in any 12 linear inches (305 mm) of weld; or (3) four or more pores aligned, and the pores separated by 1/16" (1.6 mm) or less edge to edge.

Observation: The porosity acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.7 stated "Only surface porosity whose major surface dimension exceeds 1/16" shall be considered relevant. Fillet and groove welds that contain surface porosity are unacceptable if: the sum of diameters of random porosity exceeds 3/8" in any linear inch of weld or 3/4" in any 12" of weld; or four or more pores are aligned and the pores are separated by 1/16" or less, edge to edge."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Slag Reference: ASME Section III, Article NF-5360 (j) Code Year 1995 Requirement: Slag 1/8" (3.2 mm) or less in size is irrelevant. Slag greater than 1/4" (6 mm) in size after cleaning is unacceptable.

Observation: The slag acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.10 stated "Slag whose major surface dimension is 1/8" or less is irrelevant. Isolated surface slag that remains after weld cleaning and does not exceed 1/4" in its major surface dimension is acceptable. (Slag is considered to be isolated when it does not occur more frequently than once per weld or more than once in a 3" weld segment.) Spatter remaining after the cleaning operation is acceptable."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Thickness Reference: ASME Section III, Article NF-5360 (b) Code Year 1995 Requirement: Welds thinner than specified by greater than 1/16" (1.6 mm) for more than one-fourth the weld length are unacceptable. Welds thicker than specified are unacceptable if they interfere with mating parts.

Observation: The weld thickness acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.2 stated "A fillet weld is permitted Page 89 of 149

to be less than the size specified by 1/16" for one-fourth of the length of the weld.

Oversized fillet welds are acceptable if the oversized weld does not interfere with mating parts."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Acceptance Criteria - Undercut Reference: ASME Section III, Article NF-5360(f)(2) Code Year 1995 Requirement: Undercuts deeper than 1/32" (.8 mm) on one side for the full length of the weld are unacceptable. Undercuts deeper than 1/32" (.8 mm) on one side for one-half the length of the weld AND deeper than 1/16" (1.6 mm) on the same side for one-fourth the length of the weld, are unacceptable.

Observation: The undercut acceptance criteria was incorporated into the visual weld examination procedure. Procedure HSP-507, Step 8.2.6 stated "For material 3/8" and less nominal thickness, undercut depth of 1/32", on one side of the member for the full length of the weld, or 1/32" on one side for one-half the length of the weld, and 1/16" for one-fourth the length of the weld on the same side of the member is acceptable. For members welded on both sides where undercut exists in the same plane of a member, the cumulative lengths of undercut are limited to the lengths of undercut allowed on one side. Melt-through that results in a hole in the base metal is unacceptable. For material greater than 3/8" nominal thickness, undercut depth of 1/32" for the full length of the weld and 1/16" for one fourth the length of the weld on both sides of the member is acceptable. When either welds or undercut exist only on one side of the member or are not in the same plane, the allowable undercut depth of 1/32" may be increased to 1/16" for the full length of the weld."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Eye Position and Lighting Reference: ASME Section V, Article 9, T-952 Code Year 1995 Requirement: Direct visual examinations shall be conducted with the eye within 24" (610 mm) of the surface, at an angle not less than 30 degrees. The minimum light level shall be 50 foot-candles.

Observation: Procedure HSP-507, Step 6.1.1 stated "Direct visual examination may usually be made when access is sufficient to place the eye within 24" of the surface to be examined and at an angle not less than 30° to the surface to be examined. Mirrors may be used to improve the angle of vision." Step 6.2 of the procedure stated " The weld under immediate examination shall be illuminated with a flashlight or other auxiliary lighting, as required, to attain a minimum of 100-foot candles."

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Page 90 of 149

Category: NDE-Visual Examination Topic: Minimum Elements Reference: ASME Section V, Article 9, T-941.2 Code Year 1995 Requirement: Each Visual Testing (VT) procedure shall include the: (1) how visual examination is to be performed; (2) type of surface condition and criteria for surface cleaning; (3)

cleaning instructions or reference to cleaning procedures; (4) method or tool for surface preparation, if any; (5) whether direct or remote viewing is used; (6) special illumination, instruments, or equipment to be used, if any; (7) sequence of performing examination, when applicable; (8) data to be documented; (9) report forms to be completed; Observation: Procedure HSP-507 contained the required elements specified by ASME Section V, Article 9, T-941.2 related to visual examination (testing) requirements. Step 6.1 described how visual testing was to be performed and included a 24" maximum distance between the eye and the surface to be examined and a 30 degree minimum angle requirements. Mirrors and hand held low power magnifying lenses may be used. If remote visual examinations are required, telescopes, borescopes, fiber optics or cameras may be used. Step 6.2 stated that a flashlight or other auxiliary lighting shall be used to attain a minimum of 100-foot candles of illumination. Step 4.3 discussed the surface cleaning instructions and specified that cleaning shall be performed by approved methods such as use of a wire brush or scotch brite pad. The final welds shall be cleaned prior to examination with all visible weld slag, dirt, spatter, etc. removed. Section 8.0

"Visual Weld Examination Acceptance Criteria" provided the list of acceptable weld conditions. Exhibit 10.1 "Weld Inspection Report" provided for documentation of the weld examination results.

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Category: NDE-Visual Examination Topic: Procedure Validation Reference: ASME Section V, Article 9, T-941.4 Code Year 1995 Requirement: The procedure shall contain or reference a report of what was used to demonstrate that the examination procedure was adequate. In general, a fine line 1/32" or less in width, or some other artificial flaw located on the surface or a similar surface to that to be examined, may be considered a test method for this demonstration. The line or artificial flaw should be in the least discernible location on the area examined, to prove the procedure.

Observation: Procedure HSP-507 referenced a procedure qualification record that demonstrated that the examination procedure was adequate. Step 6.3.1 of Procedure HSP-507 referenced Procedure Qualification Record (PQR) -11, "HMD Visual Inspection Procedure and Adequacy Demonstration." Procedure Qualification Record -11 made use of a standard reference that consisted of a fine line 1/32" or less in width by approximately 3/4" long on a steel surface.

Documents (a) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Reviewed: Welding," Revision 2 Page 91 of 149

Category: Pressure Testing Topic: Governing Code Reference: FSAR 1014, Section 9.1.2.2.2 Revision 9 Requirement: Pressure testing (hydrostatic or pneumatic) of the canister confinement boundary shall be performed in accordance with the requirements of ASME Code Section III, Subsection NB, Article NB-6000, when field welding of the canister lid-to-shell weld is completed.

If hydrostatic testing is used, the canister shall be pressure tested to 125% of design pressure.

Observation: The requirements for the canister hydrostatic testing had been incorporated into Procedure DCS-204 consistent with ASME Code Section III, Subsection NB, Article NB-6000. The canister design pressure was 100 pounds per square inch-gauge (psig) per Holtec Final Safety Analysis Report (FSAR), Table 2.2.1 "Design Pressures." Procedure DSC-204, Section 8.4 "Hydrostatic Test System Set-up" and Section 8.5 "Hydrostatic Testing" provided the instructions for performing the canister hydrostatic test. A note preceding Step 8.4.5 required the inlet pressure gauges, P-3 and P-5, to be calibrated within 14 days of the test. These were the gauges that were used to confirm the successful test. The 14-day calibration was required by ASME Section III, Article NB-6413. Procedure DCS-204, Step 8.5.9 required pressurizing the canister to 126-130 psi.

Test results were documented in Step 8.5.10 after the pressure was maintained for 10 minutes per Step 8.5.10.2. Step 8.5.10.6 recorded whether the test was successful. Step 8.5.10.6 and the note above it stated that a minimum internal pressure of 126 psig had to be maintained for the 10 minute period of the test and an internal pressure drop of less than 1 psig was required for a successful test. Procedure DCS-204, Step 8.5.10.6 required a quality control inspector to witness the hydrostatic test and sign off on the test results.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Category: Pressure Testing Topic: Hydrostatic Testing Sequence Reference: FSAR 1014, Table 2.0.1, Sections 8.1.5.4;9.1.2.2.2 Revision 9 Requirement: During hydrostatic testing, demineralized water or spent fuel pool water is admitted to the canister through a supply line connected to the drain port RVOA. The canister is pressurized to 125 +5/-0 psig and held for 10 minutes with no pressure drop. Following the 10-minute hold at test pressure, the canister lid to shell weld is examined to confirm no observable water leakage. The canister is then depressurized through a return line connected to the vent port RVOA and routed back to the spent fuel pool or liquid radwaste system. Once the canister is depressurized, the liquid penetrant examination of the canister lid-to-shell weld is repeated. Any evidence of cracking or deformation is cause for rejection.

Observation: The hydrostatic testing sequence and criteria described in Holtec Final Safety Analysis Report (FSAR), Section 8.1.5.4, Section 9.1.2.2.2, and Table 2.0.1 had been incorporated into Procedure DCS-204. Procedure DCS-204, Attachment 10.1.6 "Hydrostatic Test System Set-Up" provided a schematic of the system connections to perform the hydrostatic test. The hydrostatic test was described in Procedure DSC-204, Section 8.4

"Hydrostatic Test System Set-up" and Section 8.5 "Hydrostatic Testing." The procedural Page 92 of 149

steps to perform the hydrostatic test were reviewed. The water for the test was injected into the canister through the removable valve operator assembly (RVOA) drain port.

The water was drawn from the wet cask pit and was verified as complying with the minimum boron concentrations in Step 8.5.11. Procedure DCS-204, Steps 8.5.8 and 8.5.9 required the canister to be pressurized to 126-130 psi. Step 8.5.10.2 required holding the pressure for 10 minutes while inspecting the canister lid to shell weld for evidence of leakage. Step 8.5.14 depressurized the system via the vent port RVOA back to the wet cask pit. Once the canister was depressurized, Step 8.5.21 required a visual and liquid penetrant examination of the lid-to-shell weld in accordance with Procedures HSP-506 and HSP-507.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) Holtec Procedure HSP-506 "Liquid Penetrant Examination for MPC Field Closure Welding," Revision 2 (c) Holtec Procedure HSP-507 "Visual Weld Examination of MPC Closure Welding," Revision 2 (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Pressure Testing Topic: Pressure Gauge Calibration Reference: ASME Section III, Article NB-6413 Code Year 1995 Requirement: All test gauges shall be calibrated against a standard dead weight tester or a calibrated master gauge. The gauges shall be calibrated before each test or series of tests. A series of tests is that group of tests using the same pressure test gauge or gauges, which is conducted at the same site within a period not exceeding 2 weeks.

Observation: The test gages P-3 and P-5 used to verify compliance for the hydrostatic test of the canister lid weld with the ASME code requirement were required to be calibrated within 14 days of use. This requirement was stated in a note in Procedure DCS-204 proceeding Step 8.4.5 and in Attachment 10.1.19 "Instrument Calibration Verification." Attachment 10.1.19 stated that for hydrostatic testing to comply with ASME BPVC NB-6413, gauges P-3 and P-5 must have valid calibration within 14 days of use.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 Category: Pressure Testing Topic: Pressure Gauge Installation Reference: ASME Section III, Article NB-6411 Code Year 1995 Requirement: Pressure test gauges shall be connected directly to the component and visible to the operator controlling test pressure.

Observation: The pressure gage, P-3, used to measure the pressure during the hydrostatic test was directly attached to the vent port removable valve operator assembly (RVOA). This was shown in Procedure DCS-204, Attachment 10.1.6 "Hydrostatic Test System Set-Up."

Procedure DCS-204, Step 8.5.4 required an assigned individual to constantly monitor the pressure gage during the hydrostatic test. The caution statement preceding Step 8.5.4 allowed the use of a camera for monitoring the pressure gauge to reduce personnel exposure.

Page 93 of 149

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 Category: Pressure Testing Topic: Pressure Gauge Ranges Reference: ASME Section III, Article NB-6412 Code Year 1995 Requirement: Analog type indicating pressure gauges used in testing shall be graduated over a range not less than 1.5 times nor more than 4 times the test pressure. Digital type pressure gauges may be used without range restriction.

Observation: The pressure gage (P-3) used for hydrostatic testing of the canister lid weld was a digital gage and did not have a range restriction.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 Category: Pressure Testing Topic: Thermal Expansion Reference: ASME Section III, Article NB-6126 Code Year 1995 Requirement: If a pressure test is to be maintained for a period of time and the test medium in the system is subject to thermal expansion, precautions shall be taken to avoid excessive pressure.

Observation: Precautions were taken during the hydrostatic test to prevent overpressurization.

Procedure DCS-204, Section 8.5 "Hydrostatic Testing" required the canister pressure to be maintain at 126 - 130 psig (125% of the design pressure) for a period of ten minutes.

Step 8.5.4 required an assigned individual to monitor the pressure gage during pump operation. During the dry run demonstration May 31 through June 3, 2011, the pump operator, the supervisor, and others monitored the pressure gage to prevent inadvertent overpressurization. The test system included a pressure relief valve (FSV-1) rated at 140 psig.

Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 Category: Quality Assurance Topic: Approved QA Program Reference: 10 CFR 72.140(d) Published 2011 Requirement: A QA program previously approved by the Commission as satisfying the requirements of Appendix B to Part 50 will be accepted as satisfying the requirements of Part 72. In filing the description of the QA program required by Part 72.140(c), each licensee shall notify the NRC of it's intent to apply it's previously approved QA program to ISFSI activities. The notification shall identify the previously approved QA program by date of submittal, docket number and date of Commission approval.

Observation: Luminant Power provided the required letter to the NRC on December 22, 2009, stating that the 10 CFR Part 50, Appendix B quality assurance program previously approved by the NRC will be used for the ISFSI activities. The Luminant Power letter stated that the QA program had been approved for use under dockets 50-445 and 50-446 by supplemental Safety Evaluation Report #22, dated January 1990. To support the dry Page 94 of 149

cask storage program, Comanche Peak added Appendix F Dry Cask Storage System QA Program to the Quality Assurance Manual which included information related to the implementation of the quality assurance requirements for activities associated with the ISFSI. A graded approach to the QA program was applied to the ISFSI using Important to Safety categories consistent with NUREG/CR-6407 Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety.

Documents (a) Letter (CP-200901726) from Fred Madden, Luminant to US NRC Document Control Reviewed: Desk entitled Comanche Peak Steam Electric Station Docket NOS 50-445, 50-446 and 72-74 ISFSI Quality Assurance Program. dated December 22, 2009 (b) Luminant Power Comanche Peak Nuclear Power Plant Quality Assurance Manual, Revision 17 (c)

Design Basis Document DBD-ME-080-01 Dry Cask Storage System, Revision 1 Category: Quality Assurance Topic: Corrective Actions Reference: 10 CFR 72.172 Published 2011 Requirement: The licensee shall establish measures to ensure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures must ensure that the cause of the condition is determined and corrective action taken to preclude repetition. This must be documented and reported to appropriate levels of management.

Observation: The licensee had established a corrective action program to document and track conditions adverse to quality. Procedure STA-421 provided instructions for entering an issue into the corrective action system. Once an issue was entered, Procedure STA-422 was used to process, evaluate, trend, and resolve the issue. Procedure STA-421 defined conditions adverse to quality and significant issues adverse to quality. All individuals were encouraged to report any situation or event that had an impact on personal safety, plant safety, plant operability, or involved a fire, radiological release, security condition, injury, or cyber security threat. Events that were reportable to an outside agency, had the potential to affect plant equipment, or involved a missed, near-miss or failed surveillance were also issues identified as warranting a condition report. Procedure STA-421, Attachment 8A Conditions Documented on a Condition Report, provided a list of examples that would be reported using a condition report. Attachment 8B "Initiating a Condition Report" provided directions for electronically filing a condition report. Once the condition report was entered into the condition reporting system, Procedure STA-422 was used to process the issue. A corrective action program group provided initial screening and assigned a condition level as Level A (significantly adverse to quality),

Level B (adverse to quality), Level C (adverse to quality but has minimal impact on safe reliable operations), or Level D (not adverse to quality). Procedure STA-422, Attachment 8A Determination of the Condition Level, provided numerous examples and further definitions for classifying the condition reports. Level A condition reports required a root cause analysis and extent of condition evaluation. Level B condition reports required a cause analysis. Level B condition reports were divided between high tier issues and low tier issues. High tier Level B condition reports also required an extent of condition analysis. A management review committee provided oversight of the corrective action process to ensure proper screening of the issue and appropriate Page 95 of 149

assignment for resolution and closure. If the issue was assigned as Level A or B, then a new condition report was created upon closure of the original condition report that required an effectiveness review in accordance with Procedure STA-422, Attachment 8G

"Performance of an Effectiveness Review." This effectiveness review was performed at some later period of time to evaluate the effectiveness of the corrective actions taken to correct the problem and to verify that the corrective actions had not created negative unintentional consequences.

A list of condition reports related to the ISFSI and the 130-ton fuel building overhead crane dating back to 2008 were reviewed. The licensee had documented a wide range of issues. Most issues were Level C and D. There were no Level A issues. Level B issues included ISFSI pad and heavy haul path issues identified during construction, documentation issues identified by the quality assurance organization, fuel building crane operational issues, and operating experiences reported at other nuclear sites that could affect Comanche Peak. Selected condition reports were reviewed in detail to verify adequate classification of the issue and the basis for closure. These included a variety of issues such as boron requirements for the spent fuel pool, burnup calculations related to spent fuel assemblies selected for cask loading, use of synthetic rigging (Kevlar), transporter wheel failure, fuel building overhead crane issues, quality assurance audit findings, and errors associated with the TARPIT computer code. All issues were adequately classified, resolved, and closed.

Documents (a) Procedure STA-421 Initiating Condition Reports, Revision 17 (b) Procedure STA-Reviewed: 422 Processing Condition Reports, Revision 25 (c) List of ISFSI Related Condition Reports dated June 13, 2011 (d) Condition Report CR-2009-00859 "Modification Activities Associated with the Dry Fuel Storage Activities Need to be Captured," Level D, initiated March 10, 2009 (e) Condition Report CR-2010-03062 "Conflict Between Current Spent Fuel Pool Boron Limits and Holtec Technical Specification A.3.3.1 Minimum Limits," Level D, initiated March 31, 2010 (f) Condition Report CR-2010-08557 "Conservative Heat Load Values Based on Fuel Assembly Specific Levels (Not Average Decay Heat for a Discharge Cycle) are Needed for the Dry Cask Storage Project," Level D, initiated September 14, 2010 (g) Condition Report CR-2010-09109

"Testing and Adjustments Needed to Travel Limit Switches on Fuel Building Overhead Crane," Level D, initiated October 5, 2010 (h) Condition Report CR-2010-10911 "Issues Identified During Testing and Placement of Concrete for the Haul Path," Level C, initiated January 31, 2011 (i) Condition Report CR-2010-10918 "Certification Requirements for Sand Cone Test Technician During Backfill for the ISFSI Pad," Level D, initiated December 3, 2010 (j) Condition Report CR-2010-11562 "ISFSI Pad Nonconformances Related to Final Inspection of Rebar," Level C, initiated December 29, 2010 (k) Condition Report CR-2011-00271 "Comanche Peak Quality Assurance Findings During Review of Holtec Field Notification Change Requests (FNCR) Related to Engineering Design Change Requests (EDCR)," Level C, issued January 11, 2011 (l)

Condition Report CR-2011-00571 "During Work on Heavy Haul Path, a Worker Drilled into an Underground Conduit," Level B, initiated January 17, 2011 (m) Condition Report CR-2011-02323 "Holtec Provided Slings Removed from Service," Level C, initiated February 28, 2011 (n) Condition Report CR-2011-03185 "Evaluation for Use of Kevlar Synthetic Rigging for Dry Cask Storage Operations," Level D, initiated March 22, 2011 (o) Condition Report CR-2011-05573 "NRC Observations During Welding Page 96 of 149

Demonstration," Level D, initiated May 2, 2011 (p) Condition Report CR-2011-05641

"Error Found in Decay Heat Table of TARPIT Code," Level C, initiated May 5, 2011 (q)

Condition Report CR-2011-05642 "Somervell County Sheriff's Department Letter of Agreement May Need Updating Related to ISFSI," Level D, initiated May 5, 2011 (r)

Condition Report CR-2011-05645 "Evaluation Required of the Fuel in the Forklift Used with the Low Profile Transporter While Inside the Fuel Building," Level D, initiated May 5, 2011 (s) Condition Report CR-2011-05819 "Evaluation of NRC Information Notice 2011-10 Related to Thermal Loading Incident at Byron on August 28-29, 2010 and Potential Improvement Needed for Comanche Peak," Level C, initiated May 11, 2011 (t) Condition Report CR-2011-06069 "Quality Assurance Surveillance to Assess Readiness of the Plant for Dry Cask Storage Activities," Level D, initiated May 18, 2011 (u) Condition Report CR-2011-06248 "Fuel Building Overhead Crane Brakes Fail to Fully Open (Release) due to Insufficient Air Pressure," Level C, initiated May 24, 2011 (v) Condition Report CR-2011-06287 "Fuel Building Overhead Crane Tripped Twice While Moving Equipment," Level B, initiated May 25, 2011 (w) Condition Report CR-2011-06350 "Fuel Building Overhead Crane Tripped Numerous Times While Moving the HI-TRAC and Lifting Rig," Level C, initiated May 27, 2011 (x) Condition Report CR-2011-06393 "NRC Observations During Forced Helium Dehydration Dry Runs,"

Level D, initiated May 31, 2011 (y) Condition Report CR-2011-06566 "Failure of Electrical/Hydraulic System on Vertical Cask Transporter During Transport of Empty HI-STORM Storage Casks," Level D, initiated June 3, 2011 (z) Condition Report CR-2011-06683 "Software Error Found in TARPIT Decay Heat Calculations," Level C, initiated June 7, 2011 (aa) Condition Report CR-2011-06800 "Hydraulic System Failure of Vertical Cask Transporter While Moving HI-STORM with Loaded Dummy MPC,"

Level D, initiated June 11, 2011 (bb) Condition Report CR-2011-07348 "Vertical Cask Transporter Wheel Hub Failure," Level D, initiated June 27, 2011 (cc) Condition Report CR-2011-07717 "NRC Observations During Wet Operations Dry Runs," Level D, initiated July 8, 2011 Category: Quality Assurance Topic: Important to Safety Components - Ancillaries Reference: FSAR 1014, Table 8.1.6 Revision 9 Requirement: Ancillary equipment shall be classified as important to safety (ITS) in accordance with FSAR Table 8.1.6 Observation: The important to safety classifications in Table 8.1.6 of the Holtec Final Safety Analysis Report (FSAR) were consistent with the licensees safety classifications in Design Basis Document DBD-ME-080-01, Table 10.1 DCSS Equipment Safety Classification for the ancillary equipment. Most ancillary equipment was not important to safety. Those items that were important to safety included the mating device (Category B), temperature instrumentation on the supplemental cooling system (Category B), instrumentation on the forced helium dehydrator (Category B), canister lift cleats (Category A), lifting yoke and lifting yoke extension (Category A), canister lifting slings (Category A), and seismic stack-up braces (Category A).

Documents (a) Design Basis Document DBD-ME-080-01 Dry Cask Storage System, Revision 1 Reviewed: (c) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Page 97 of 149

Category: Quality Assurance Topic: Important to Safety Components - Cask System Reference: FSAR 1014, Table 2.2.6 Revision 9 Requirement: Structures, systems, and components of the HI-STORM 100 cask system are identified as important to safety (ITS) in accordance with NUREG/CR-6407 "Classification of Transportation and Dry Spent Fuel Storage System Components." Holtec FSAR, Table 2.2.6 provides a summary of the classification of the structures, systems, and components as important to safety A, B, C, and NTIS (not important to safety).

Observation: The important to safety classifications in the Holtec Final Safety Analysis Report (FSAR) 1014, Table 2.2.6 "Materials and Components of the HI-STORM 100 System" were consistent with the licensees safety classifications in Design Basis Document DBD-ME-080-01, Table 10.1 DCSS Equipment Safety Classification for the canister, HI-STORM storage cask and HI-TRAC transfer cask. The canister and HI-TRAC transfer cask were listed in Design Basis Document DBD-ME-080-01, Table 10.1 as Category A components. The HI-STORM storage cask was listed as Category B. Holtec FSAR Table 2.2.6 listed the various individual components of the canister, HI-STORM storage cask, and HI-TRAC transfer cask and classified the components as Category A, B, C or as not important to safety. For the canister and HI-STAR transfer cask, the highest category in FSAR Table 2.2.6 was Category A for any of the individual components. For the HI-STORM storage cask, the highest category in FSAR Table 2.2.6 was Category B for any of the components.

Appendix F Dry Cask Storage System QA Program of the Quality Assurance Manual included information related to the implementation of the quality assurance requirements for activities associated with the ISFSI. A graded approach to the QA program was applied to the ISFSI using Important to Safety categories consistent with NUREG/CR-6407 Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety. Section 3 Program Elements of the NUREG defined the three important to safety categories AS Category A Critical to Safe Operations, Category B Major Impact on Safety, and Category C Minor Impact on Safety. Design Basis Document DBD-ME-080-01described the licensees list of important to safety components and used the NUREG/CR-6407 safety categories.

Documents (a) Luminant Power Comanche Peak Nuclear Power Plant Quality Assurance Manual, Reviewed: Revision 17 (b) Design Basis Document DBD-ME-080-01 Dry Cask Storage System, Revision 1 (c) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Category: Quality Assurance Topic: Instruments Requiring Calibration Reference: FSAR 1014, Table 8.1.7 Revision 9 Requirement: Instruments requiring calibration include flow rate monitors, canister pressure gauges, gas and water temperature gauges, temperature surface pyrometer, vacuum gauge for gas sampling and moisture monitoring instruments for Forced Helium Dehydration (FHD)

operations.

Observation: All of the identified instruments had been calibrated or were identified to be new equipment that had not been previously used and complied with manufacturer's Page 98 of 149

calibration requirements. All the equipment that required calibration in accordance with Table 8.1.7 of the FSAR were placed into either Comanche Peak's Maintenance Program (MAXIMO) for equipment that the licensee's lab could calibrate. Or the equipment that required offsite calibration had conditional driven preventative maintenance work orders (PM 348221, PM 348222, and PM 348223) developed to ensure that equipment was properly calibrated prior to a new loading campaign. The MAXIMO data sheet and PMs contained the required flow rate monitors, pressure gages, temperature gauges, temperature surface pyrometer, and moisture monitoring instruments for FHD operations. Vacuum gauges do not apply since the licensee elected to do FHD instead of vacuum drying.

Documents (a) MAXIMO Data Sheet printout dated June 1, 2011; (b) Condition Driven PM 348221, Reviewed: PM 348222, and PM 348223, dated July 11, 2011 Category: Quality Assurance Topic: Operating Status Reference: 10 CFR 72.168(b) Published 2011 Requirement: The licensee shall establish measures to identify the operating status of structures, systems, and components of the ISFSI such as tagging valves and switches to prevent inadvertent operations.

Observation: The inspector verified that systems and components associated with the ISFSI campaign were properly tagged to prevent inadvertent operations. Systems such as the forced helium dehydration (FHD) system, supplemental cooling system (SCS), hydrostatic pump system, heavy lifting equipment, and the removable valve operator assembly (RVOA) water line systems were all properly tagged. The items tagged included pressure gages, temperature gages, flow meters, valves, switches, slings, and other associated equipment. The tags matched the different procedures to avoid inadvertent use.

Documents No documents reviewed Reviewed:

Category: Quality Assurance Topic: QA Audits Reference: 10 CFR 72.176 Published 2011 Requirement: The licensee shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the QA program and to determine the effectiveness of the program.

Observation: The licensee had developed a comprehensive plan for auditing the spent fuel dry cask storage program consistent with Section 18 Audits of the Quality Assurance Oversight Plan for Comanche Peak Nuclear Power Plant Spent Fuel Dry Cask Storage System Project. The Comanche Peak quality assurance organization was conducting audits and surveillances of the dry cask storage activities at Comanche Peak and of the cask vendor, Holtec, included the engineering design activities and the cask manufacturing process.

Other audits conducted of the Holtec engineering design and manufacturing process that were conducted by third party groups were included in the licensees quality assurance assessment process. This included audits by the Nuclear Utility Procurement Issues Committee (NUPIC) and the Holtec Joint Utility Oversight Group. The Holtec Joint Page 99 of 149

Utility Oversight Group contracted with an outside inspection organization to provide a resident inspector at the Holtec manufacturing facility for witness and verification of critical areas of the manufacturing, inspection, and testing process for important-to-safety components being manufactured for it's members, such as Comanche Peak.

Procedure NQA 3.02 provided the methodology for audits and surveillances performed by the Comanche Peak quality assurance organization. This procedure assigned responsibilities for the licensees quality assurance program, established qualification requirements for auditors, defined requirements for scheduling audits and developing assessment plans, and provided guidance for audit planning, performance, reporting, and follow-up. The procedure also provided guidance for conduction surveillances, reporting results, and conducting follow-up surveillances. Audit frequencies for the different technical areas were provided in Attachment 8A Audit Objectives, Frequencies, and Source Requirements. Dry cask storage activities were included in the audit scopes for nearly all the disciplines.

Procedure NQA 3.14 established the criteria and methodology for the selection, qualification, evaluation, control and audits of vendors providing safety related items and services to Luminant Power. The procedure also applied to commercial grade suppliers that were maintained on the Luminant Powers Approved Vendor List. Vendors approved to supply dry cask storage system equipment and services classified as important to safety Category A or B were listed on the approved vendors list with other safety related vendors. Procedure NQA 3.14 provided instructions for generating and maintaining the approved vendors list, requirements for evaluating a vendor for inclusion on the approved list, and the process for dealing with findings and corrective actions from the audit. Annual vendor evaluations were required by the procedure for all companies listed on the approved vendors list that provided safety related or commercial grade items. A vendor could be placed on the approved vendors list based on one or more of the following: (1) a review of historical data on vendor performance and capability including possession of a current ASME Certificate of Authorization/Quality System Certificate, (2) a review of the vendors quality assurance program as conducted in accordance with Procedure NQA 3.14, or (3) a pre-award audit or commercial grade survey performed at the vendors facility by Luminant Power or an approved third party.

Holtec International was listed on the approved vendors list to provide safety related components. The last evaluation of Holtec was April 26, 2011. Holtec had originally been added to the approved vendors list through a NUPIC third part audit conducted October 2008. Tennessee Valley Authority was the lead utility performing the audit, which included team members from nine different utilities. The audit reviewed Holtec program controls for design and fabrication of wet and dry fuel storage racks and casks and determined that Holtecs programs were acceptable to continue to qualify the company for inclusion on TVAs acceptable suppliers list. Based on this, Luminant Power included Holtec, International on their approved vendors list.

Several selected audits and surveillances were reviewed related to the Comanche Peak dry cask storage program. This included audits and surveillance of Holtec activities, including work being performed at the Holtec manufacturing division, as well as activities being conducted by both Holtec and the licensee's staff at the Comanche Peak Page 100 of 149

site. Audit and surveillance findings were adequately resolved and documented. The audits and surveillances reviewed included such areas as work on the ISFSI pad, security, training program, procedure adequacy, manufacture of ISFSI components, and observation of tests conducted on the ISFSI components. The Luminant Surveillance QAA-10-055 of the Holtec Manufacturing Division on September 8 - 9, 2010 included observation of several weld inspections conducted by the Holtec quality control inspectors and observation of the helium leak testing on the three canister shells to be delivered to Comanche Peak for the first loading campaign. All canisters passed the leak test.

Documents (a) Luminant Power Comanche Peak Nuclear Power Plant Quality Assurance Manual, Reviewed: Revision 17 (b) Quality Assurance Oversight Plan for Comanche Peak Nuclear Power Plant Spent Fuel Dry Cask Storage System Project, Revision 0 (c) Procedure NQA 3.02 Audit and Surveillance Programs, Revision 6 (d) Procedure NQA 3.14 "Control of Vendor Activities, Revision 19 (e) NUPIC Joint Audit #20100 of Holtec Int. conducted October 10 - 20, 2008, issued December 15, 2008 (f) Luminant Surveillance QAA-10-037 of Holtec, Int. conducted June 28 - 30, 2010, issued July 22, 2010 (g) Luminant Correspondence CP-201001139 from D. Volkening to M. Soler, Holtec, Inc. entitled Evaluation of Holtec Deficiency Response to Luminant Surveillance QAA-10-037, dated August 14, 2010 (h) Luminant Correspondence CP-201001248 from D. Volkening to M. Soler, Holtec, Int. entitled Evaluation of Holtec Deficiency Response for Luminant Surveillance QAA-10-037, dated September 16, 2010 (i) Luminant Correspondence CP-201001552 from D. Volkening to M. Soler, Holtec, Int. entitled Evaluation of Holtec Deficiency Response for Luminant Surveillance QAA-10-059 dated November 29, 2010 (j) Luminant Correspondence CP-201001553 from D.

Volkening to M. Soler, Holtec, Int. entitled Evaluation of Holtec Deficiency Response for Luminant Surveillance QAA-10-037, dated November 29, 2010 (k) Luminant Correspondence CP-201001229 from D. Volkening to M. Soler, Holtec, Int. entitled Luminant Power QA Surveillance (QAA-10-055) of Holtec, Int. Manufacturing Division, dated September 27, 2010 (l) Luminant Correspondence CP-201100450 from D. Volkening to M. Soler, Holtec, Int. entitled Closure of Luminant Surveillance QAA-10-059, dated March 28, 2011 (m) Luminant Correspondence CP-201001328 from D.

Volkening to M. Soler, Holtec, Int. entitled Luminant Power QA Surveillance (QAA-10-059) of Holtec, Int. Manufacturing Division, dated October 7, 2010 (n) Luminant Correspondence CP-201001541 from D. Volkening to M. Soler, Holtec, Int. entitled Luminant Power QA Surveillance of Holtec, Int. Manufacturing Division, dated November 24, 2010 (o) Luminant Correspondence CP-201100141 from D. Volkening to M. Soler, Holtec, Int. entitled Evaluation of Holtec Deficiency Response for Luminant Surveillance QAA-10-068, dated January 24, 2011 (p) Luminant Correspondence CP-201100496 from D. Volkening to M. Soler, Holtec, Int. entitled Closure of Luminant Surveillance QAA-10-068, dated April 2, 2011 (q) Luminant Office Memo from John Simmons to Holtec Vendor File (QVE-0454.1.A) entitled Evaluation of Holtec Manufacturing Division Surveillance Report 11-RVS-06 (QAA-11-033), dated April 30, 2011 (r) Surveillance of Ingrams Concrete Batch Plant, conducted October 19, 2010 (s) Surveillance of Concrete Placement on Haul Path Along Side Fuel Building, conducted December 2, 2010 (t) Surveillance of Backfill Activities for ISFSI Pad, conducted November 29 to December 3, 2010 (u) Surveillance of Concrete Placement for Haul Road Inside Protected Area, conducted December 14, 2010 (v) Surveillance of Page 101 of 149

Holtec Quality Control Inspection of ISFSI Pad Rebar, conducted December 27 - 28, 2010 (w) Surveillance of Forms Documenting the ISFSI Rebar Placement Findings, conducted January 11, 2011 (x) Surveillance of Rone Engineering Seven Day Break Tests of Samples Taken of Pour #1 of ISFSI Pad," conducted January 11, 2011 (y)

Surveillance of Second ISFSI Pad Pour," conducted January 26, 2011 (z) Surveillance of ISFSI Pad Concrete Final Pour," conducted on February 14-15, 2011 (aa) Surveillance of the Review of Newly Developed Procedures DCS-201 through DCS-207," conducted April 19-20, 2011 (bb) Surveillance of Security Meeting to Discuss Revision to Security Procedures for ISFSI Activities," conducted April 26, 2011 (cc) Surveillance of Pre-Job Briefing for NRC Observed Welding Dry Run," conducted April 28, 2011 (dd)

Surveillance of Project Readiness as it Relates to Training Activities," conducted June 2, 2011 (ee) Condition Report CR-2011-06069 "Quality Assurance Surveillance to Assess Readiness of the Plant for Dry Cask Storage Activities," Level D, initiated May 18, 2011 Category: Quality Assurance Topic: Receipt Inspection Checklists Reference: FSAR 1014, Table 8.1.8, 8.1.9, 8.1.10 Revision 9 Requirement: Tables 8.1.8, 8.1.9, and 8.1.10 provide sample receipt inspection checklists for the HI-STORM storage cask, canister, and HI-TRAC transfer cask. Users shall develop site-specific receipt inspection checklists.

Observation: The licensee had performed the required receipt inspections for the HI-STORM storage cask, HI-TRAC transfer cask, and the canister in accordance with the requirements of the Holtec Final Safety Analysis Report (FSAR). Procedure DCS-105 contained all the required inspection attributes listed in the FSAR Table 8.1.8 "HI-STORM 100 System Overpack Inspection Checklist" for the HI-STORM 100S storage cask. The receipt inspection included the exterior surface, interior cavity, anchor block stud holes, inlet and outlet air vent openings, vent screens, paint, identification plate, lifting studs and lid bolt holes. The inspection verified cleanliness and general condition. Procedure DCS-106 contained all the inspection attributes listed in FSAR Table 8.1.10 "HI-TRAC Transfer Cask Inspection Checklist" for the HI-TRAC transfer cask. Specific items that were inspected included the lifting trunnions, outside surface of the transfer cask, neutron shield jacket plugs, and the neutron shield jacket pressure relief valves.

Procedure DCS-104 contained all the inspection attributes listed in FSAR Table 8.1.9

"MPC Inspection Checklist" for receipt inspection of the canister. Specific items that were inspected included checking the following: canister body and lid surfaces for cracks, gouges, or dents; interior of the canister for debris; drain line dimensions, cleanliness and condition of the threads; and lifting lugs for general condition.

Documents (a) Procedure DCS-105 HI-STORM Receipt Inspection, Revision 1 (b) Procedure Reviewed: DCS-106 HI-TRAC Annual Inspection and Maintenance, Revision 0 (c) Procedure DCS-104 MPC Receipt Inspection, Revision 1 (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Radiation Protection Topic: ALARA Program Reference: FSAR 1014, Section 10.1.1 Revision 9 Requirement: Licensees using the HI-STORM 100 cask system will utilize and apply their existing site Page 102 of 149

ALARA policies, procedures and practices for ISFSI activities to ensure that personnel exposure requirements of 10 CFR 20 are met.

Observation: The licensee had applied the existing station "As Low As Reasonably Achievable (ALARA)" program into the cask loading program. This included the implementation of ALARA requirements from Procedure RPI-606 for developing radiation work permits (RWP), Procedure RPI-627 for providing radiological job coverage, and Procedure RPI-651 for implementing the site ALARA program. The activities of the dry cask storage program were required to be implemented under the ALARA program described in Procedure RPI-651. This included developing reasonable dose man-rem goals, obtaining review and concurrence from the station ALARA committee, attending ALARA training, and conducting pre-job briefings concerning radiation levels and low dose rate areas during work activities. ALARA job planning, tracking personnel doses by task, and conducting post-job briefings were also part of the station's ALARA program. All personnel were encouraged to provide suggestions to reduce the overall project dose.

Radiation protection coverage by experienced health physics technicians was required for all work activities around the loaded cask. Radiation work permits, such as RWP-20110601 for the dry cask storage activities, were developed in accordance with Procedure RPI-606. Attachment 4 "ALARA Planning Checklist" contained an extensive list of ALARA considerations that must be addressed when developing an RWP. This checklist included reviewing the work planned, radiological conditions that could be present in the work areas, shielding that may be needed, dosimetry requirements, protective clothing requirements, engineering controls that could be used to reduce exposure, the option to use remote telemetry systems, airborne issues, etc. Procedure RPI-627 was developed specifically for the dry cask storage operations. Attachment 1

"ALARA Practices" provided ALARA planning considerations for dry cask storage operations. These planning considerations were specific to cask activities such as use of the shield snake in the canister annulus gap, use of temporary shielding during welding, strict enforcement of personnel remaining in the low dose rate area away from the cask when not working, etc.

The licensee performed several benchmarking site visits at other nuclear plants loading casks, including the Salem nuclear plant, Sequoyah nuclear plant, and the Palo Verde nuclear plant to observe and obtain lessons learned and radiological dose information based on the respective plant's experience in conducting dry cask storage operations.

Experiences from all three plants were utilized in developing the ALARA dose estimates for the initial RWP at Comanche Peak. The licensee had estimated in RWP 20110601 that the person-rem dose for the first cask loading campaign of three casks would be approximately 0.400 person-rem per cask based on information collected during their site visits at the other nuclear plants. The information collected was used for various informal training of the radiation protection technicians that were assigned to the cask loading campaign. The training also included the limitations of the survey instrumentation that will be used at Comanche Peak, the capability of the Siemens electronic personnel dosimeters (EPDs) in the expected neutron energy fields, and the use of the CR-39 chip in the personnel dosimeters for monitoring neutron dose. The radiation protection technicians had participated in all the Comanche Peak site dry run demonstrations performed to met Certificate of Compliance 1014, License Condition 10

"Preoperational Testing and Training Exercise." The NRC had observed the performance of the radiation protection technicians during these dry runs and found them Page 103 of 149

to be knowledgeable of the various activities and the associated dose rates that could be expected. During the dry run demonstrations, the NRC inspectors observed that the radiation protection technicians provided good interface with the workers and had established good controls that will be implemented during the actual cask loadings to keep personnel exposures as low as possible. Health physics controls were implemented during the dry runs to familiarize the workers with the controls that will be in place for the actual cask loading activities. Low dose rate areas were designated and workers were restricted to these locations unless they were participating in the demonstration.

Documents (a) Procedure RPI-606 "Radiation Work and General Access Permits," Revision 22 (b)

Reviewed: Procedure RPI-627 Job Coverage for Dry Fuel Storage, Revision 1 (c) Procedure RPI-651 ALARA Program, Revision 10 (d) Radiation Work Permit RWP-20110601 "Dry Cask Storage Activities," Revision 00 Category: Radiation Protection Topic: Contamination Control Reference: FSAR 1014, Table 8.0.1 Revision 9 Requirement: Procedural guidance is given to warn operators prior to cutting or grinding activities to reduce personnel contamination problems.

Observation: Comanche Peak planned to decontaminate the top of the transfer cask and the canister lid prior to welding to remove contamination. This would result in a clean welding area on the lid and the canister shell where welding and any grinding would be performed.

Procedure RPI-627, Section 6.5 "Installing the MPC Lid and Removing the HI-TRAC from the Wet Cask Pit" directed the workers to rinse the transfer cask, canister and yoke during removal from the spent fuel pool. Then long handled squeegees were used to remove the water after rinsing. Section 6.6 "Decontamination of the HI-TRAC" discussed decontaminating the lid and the top of the transfer cask after placement in the dry cask pit. Decontamination to less than 1,000 disintegrations per minute (dpm) per 100 square centimeters (100 square cm) beta-gamma and 20 dpm/100 square cm alpha were the limits stated in the procedure. Since the spent fuel pool water was relatively clean and the transfer cask was highly polished to allow for easy removal of contamination, no problems were expected related to successful decontamination of the transfer cask and canister lid prior to welding. Procedure RPI-627 included discussions of additional contamination controls, as needed, such as contamination barriers, protective clothing, and use of decontamination agents. Radiation Work Permit 20110601, Task 2, which would be in effect during welding, provided protective clothing requirements if the transfer cask and canister exceeded the contamination limits.

During removal of the canister lid, should opening a sealed canister be necessary, Procedure RPI-627, Attachment 2 "Abnormal Operations", Section 1.5 "Initial Breach of the MPC Seal Weld on the Vent and Drain Ports" and Section 1.8 "MPC Lid to Body Weld Removal" direct that contamination levels be monitored in the work area during any drilling or cutting.

Documents (a) Procedure RPI-627 Job Coverage for Dry Fuel Storage, Revision 1 (b) RWP 2011-Reviewed: 0601 Dry Cask Storage Activities, Revision 00

.

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Category: Radiation Protection Topic: Controlled Area Boundary Dose Rate Analysis Reference: CoC 1014, Tech Spec A.5.7.2 Amendment 7 Requirement: Considering the planned number of casks to be deployed and the cask contents, the licensee shall perform an analysis to confirm the dose limits of 10 CFR 72.104(a) will be satisfied under actual site conditions. 10 CFR 72.104(a) states that the annual dose to any real individual located beyond the controlled area must not exceed 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem to any other critical organ as a result of direct radiation from the ISFSI during normal operations and anticipated occurrences.

The results of the analysis shall be documented in the 10 CFR 72.212 evaluation report.

Observation: Holtec Report HI-2104636 provided site specific calculations for the Comanche Peak ISFSI to evaluate the dose rate at the owner controlled area to verify compliance with 10 CFR 72.104(a). The Holtec report used the HI-STORM 100S Version B storage cask loaded with the MPC-32 canister. The results were documented in the 72.212 Evaluation Report, Section 5.11 "10 CFR 72.212(b)(5)(iii) - Radiological Evaluation Pursuant to 10 CFR 72.104." Only direct radiation exposure calculations were performed. Because of the design of the welded and sealed canisters, there were no effluent pathways associated with the stored canisters under normal conditions. The ISFSI site was located near the southeast portion of the peninsula on the Comanche Peak property approximately 3/4 mile from the nuclear power plant. The nearest owner controlled area boundary was approximately 1323 meters (m) [0.82 miles] from the southeast corner of the ISFSI pad. Calculations in Holtec Report HI-2104636 used 1048.5 meters, which was a value of 80% of the owner controlled area. This conservative distance was stated in Section 4 "Assumptions" of the Holtec report.

The Holtec calculation was based on a full array of 84 casks on the ISFSI pad and an occupancy time of one year (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />) for a person standing at the edge of the owner controlled area. The calculations were performed for one cask, then multiplied by 84.

By doing this, no consideration was given to the self shielding effects of the casks placed in a geometrical array on the pad, which would result in a lower dose rate at distances from the ISFSI pad because the outer concrete storage casks would provide for shielding of the inner casks. The calculation used a B&W 15x15 fuel assembly design with a burnup, cooling time, and enrichment of 55,000 MWD/MTU, 3 years and 3.9% U-235, respectively. Holtec Report HI-2104636, Section 4 "Assumptions" discusses the basis for the calculations to determine the dose rates for the cask. The calculations assumed a B&W 15x15 fuel assembly design. This fuel assembly had been determined by Holtec to be a bounding fuel assembly design. The Holtec Final Safety Analysis Report (FSAR),

Section 5.2 "Source Specifications" discussed the basis for the 15x15 B&W fuel assembly bounding the other fuel assembly designs, including the 17x17 B&W fuel used at Comanche Peak. The bounding values used in the calculations for burnup, cooling time, and enrichment of 55,000 MWD/MTU, 3 years and 3.9% U-235, respectively, were selected based on the fuel inventory at Comanche Peak. The computer codes used for the Holtec calculations were SAS2H, ORIGEN-S, and MCNP 4A.

For conservatism, Holtec used the distance of 1048.5 meters as the distance from the ISFSI site to the closest point of the owner controlled area. The estimated dose rate calculated at this distance was 2.96 mrem/year. The licensee extrapolated this dose rate to the actual owner controlled area boundary of 1323 meters in the 72.212 Evaluation Page 105 of 149

Report Section, 5.11.1.5 "Contribution to Owner Controlled Areas Boundary Dose Rate from ISFSI," and determined that the dose rate would be less than 1.0 mrem/year.

The licensee was required by 10 CFR 72.104(a) to include doses from other nearby fuel cycle activities into the dose calculations. The operating Comanche Peak nuclear power plant, as a source term, was considered in the determination of whether the 10 CFR 72.104(a) dose limit was being met. The licensee reviewed the calculated annual dose estimates based on the environmental monitoring dosimeters located at varying distances from the plant and within the owner controlled area that monitored radiation levels due to plant operations. The annual environmental monitoring data was included in the Comanche Peak Annual Radiological Environmental Operating Report. The licensee evaluated the data over a six-year period from 2004 through 2009. While the data was variable from year to year, the licensee determined that the dose rates at the owner controlled area boundary were statistically not significantly greater than zero each year, once the background control values were subtracted. However, based on data from monitoring stations inside the owner controlled area, the licensee conservatively documented in the 72.212 Evaluation Report that the Comanche Peak plant contributed approximately 10 mrem/year to the radiation dose rates at the owner controlled area boundary. Therefore, the licensee documented that the total annual dose to an individual postulated to have continuous occupancy at the owner controlled area boundary as a result of the Comanche Peak power plant operations and the ISFSI was 11 mrem/year, which was below the 25 mrem/year limit specified in 10 CFR 72.104(a).

Documents (a) Holtec Report No. HI-2104636, Dose Versus Distance from HI-STORM 100S Reviewed: Version B containing MPC-32 for Comanche Peak (b) Comanche Peak Nuclear Power Plant, 10 CFR 72.212 Evaluation Report, Revision 1 Category: Radiation Protection Topic: Controlled Area Radiological Doses Reference: 10 CFR 72.106(a)/(b)/(c) Published 2011 Requirement: For each ISFSI, a controlled area must be established. Any individual located on or beyond the nearest boundary of the controlled area may not receive from any design basis accident 5 rem TEDE for accident conditions. Minimum distance from ISFSI to nearest boundary of controlled area must be 100 meters. Controlled area may include roads, railroads or waterways as long as arrangements are made to control traffic and protect public.

Observation: The licensee's owner controlled area boundary was a sufficient distance from the ISFSI pad that the 5 rem accident dose to the whole body (TEDE - total effective dose equivalent) would not be exceeded during an accident. The 72.212 Evaluation Report, Section 5.11.9 "Dose at Nearest Boundary from Accidents Involving Storage Casks" discussed the dose evaluations for the Comanche Peak ISFSI. The nearest location of the owner controlled area to the ISFSI pad was 1,323 meters [0.82 miles] as stated in the 72.212 Evaluation Report, Section 5.11.1 "Nearest Owner Controlled Area Boundary."

This exceeded the minimum 100 meter requirement in 10 CFR 72.106(b). The Holtec Final Safety Analysis Report (FSAR) had evaluated the various accident scenarios for the 100 meter distance for the HI-STORM 100 casks. FSAR Section 11 "Accident Analysis" had evaluated a wide range of accidents that could occur at the ISFSI including cask tip-over, tornados, floods, fires, earthquakes, explosions, lightning, etc. Of these, Page 106 of 149

none were shown to exceed the 5 rem whole body dose limit at 100 meters. As such, the accident dose at 1,323 meters would be significantly lower and in compliance with the 5 rem limit.

Documents (a) Holtec Report No. HI-2104636 Dose Versus Distance from HI-STORM 100S Reviewed: Version B Containing MPC-32 for Comanche Peak, Revision 4 (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 (c) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Category: Radiation Protection Topic: Dose Rate Survey - Storage Cask Air Vents Reference: CoC 1014, Tech Spec A.5.7.8.e Amendment 7 Requirement: The licensee shall measure the contact dose rates (gamma + neutron) on the surface of each inlet and outlet vent duct screen of the storage cask. The measured dose rates must not exceed the licensees site-specific surface dose rate limits as calculated from Technical Specification A.5.7.3.

Observation: The licensee had established procedural requirements to perform the required radiological surveys to confirm compliance with the surface dose rate limits for the inlet and outlet vent ducts as required in Certificate of Compliance 1014, Technical Specification A.5.7.3.c. Procedure DCS-201, Step 8.4.56 required the cask loading supervisor to notify the radiation protection department that the cask was on the ISFSI pad and ready for the required radiation protection surveys. The surveys were performed in accordance with Procedure RPI-792. Step 4.2.3.3 of Procedure RPI-792 established neutron plus gamma dose rate limits of 32.5 mrem/hr for the top vent (outlet) and 71.9 mrem/hr for the bottom vent (inlet). These values were specific to the Comanche Peak casks and spent fuel and had been calculated in Holtec Report HI-2104635. Table 2

"Surface Dose Rates for the HI-STORM at 55,000 MWD/MTU and 3 Year Cooling Time for Use in Radiation Protection Programs" of the Holtec report was the source of the values listed in Procedure RPI-792, Step 4.2.3.3. The values were calculated based on specific spent fuel limits that applied to the Comanche Peak fuel. The bounding source term was 55,000 MWD/MTU, 3 year cooling time, and 3.9% U-235. Holtec Report HI-2104636 had calculated the dose at the controlled area boundary using this same source term. Table 1 "Dose Rate Versus Distance from the HI-STORM 100S Version B Containing the MPC-32" of Holtec Report HI-2104636 calculated the dose at the 1,048.51 meter controlled area boundary as 4.03 x 10(-6) mrem/hr, well below the 25 mrem/hr limit specified in 10 CFR 72.104(a) and referenced in Technical Specification A.5.7.2. Computer codes used for the calculations were SAS2H, ORIGEN-S, and MCNP 4A.

Procedure RPI-792, Step 6.3 required a survey of all eight HI-STORM storage cask air vents to meet the Certificate of Compliance 1014, Technical Specification A.5.7.8.e requirement. The surveys were conducted at each of the four inlet vent ducts on the bottom of the cask and at each of the four outlet vent ducts at the top of the cask for a total of eight survey points. Neutron plus gamma dose rates at each survey point were taken on contact with the vent and documented on Form 792-1 "HI-STORM Overpack

[Storage Cask] Surface Dose Rates." Procedure RPI-792, Steps 6.3.4 and 6.3.8 stated that if the maximum total dose rate measured for any one of the survey points exceeded Page 107 of 149

the maximum total dose rate limit, then the radiation protection supervisor would be immediately notified and a condition report issued per Step 6.4. An evaluation was required by Step 6.4 to verify that the correct contents had been loaded into the canister and that the annual limit in 10 CFR 72.104(a) of 25 mrem at the site boundary was not exceeded due to the higher dose rates.

Documents (a) Procedure DCS-201 "Transporting Loaded and Unloaded HI-STORM," Revision 2 Reviewed: (b) Procedure RPI-792 HI-STORM Overpack Surface Dose Rate, Revision 1 (c)

Procedure RPI-627 Job Coverage for Dry Fuel Storage, Revision 1 (d) Holtec Report No: HI-2104635 HI-STORM Radiation Protection Program Dose Rate Limits for Comanche Peak Revision 4 (e) Holtec Report HI-2104636 "Dose Versus Distance from HI-STORM 100S Version B Containing MPC-32 for Comanche Peak," Revision 4 Category: Radiation Protection Topic: Dose Rate Survey - Storage Cask Side Reference: CoC 1014, Tech Spec A.5.7.5, A.5.7.8.c Amendment 7 Requirement: The licensee shall measure the surface dose rates (gamma + neutron) on each loaded storage cask. A minimum of 12 dose rate measurements shall taken on the side of the storage cask in 3 sets of 4 measurements each. One set shall be taken at mid-plane, one set shall be taken at approximately 60 inches above mid-plane and one set shall be taken at approximately 60 inches below midplane. The 4 measurements in each set should be taken approximately 90 degrees apart around the circumference. The measured dose rate must not exceed the licensees site-specific surface dose rate limit as calculated from Technical Specification A.5.7.3 or 300 mrem/hour whichever is lower.

Observation: The licensee had established procedural requirements to perform the various required radiological surveys on the side of the HI-STORM storage cask as required by Certificate of Compliance 1014, Technical Specification A.5.7.5 and A.5.7.8.c. Procedure DCS-201, Step 8.4.56 required the cask loading supervisor to notify the radiation protection department that the cask was on the ISFSI pad and ready for the required radiation protection surveys. The surveys were performed in accordance with Procedure RPI-792.

Step 6.1 of Procedure RPI-792 described the requirements for taking the gamma survey readings on the side of the HI-STORM storage cask. Step 6.2 described the requirements for the neutron surveys. Both Step 6.1 and Step 6.2 included detail instructions for taking the measurements which required four measurements at the mid-height, four on the upper horizontal plane approximately 60 inches above the mid-height, and four on the lower horizontal plan approximately 60 inches below the mid-height. All four measurements at each of the three heights were to be taken at 90 degrees apart around the circumference of the cask and in contact with the cask. Both gamma and neutron measurements were required at each of the 12 survey points and were recorded on Form RPI-792-1 "HI-STORM Overpack Surface Dose Rates." The total surface dose rate for each survey point was the sum of the gamma plus neutron dose rate. The requirement for the survey locations in Procedure RPI-792 were identical to the requirement in Certificate of Compliance 1014, Technical Specification A.5.7.8.c. The limits for the sides of the cask were listed in Procedure RPI-792, Step 6.1.4 as 24.5 mrem/hr at the upper level (60 inches above mid-height), 40 mrem/hr at the midplane, and 8.2 mrem/hr below the mid-height. If any of these limits were exceeded, Step 6.1.4 required notification of the radiation protection supervisor and completion of a condition report per Step 6.4. An evaluation was required by Step 6.4 to verify that the correct contents Page 108 of 149

had been loaded into the canister and that the annual limit in 10 CFR 72.104(a) of 25 mrem at the site boundary was not be exceeded due to the higher dose rates.

The cask side surface dose rate limits specified in Procedure RPI-792, Step 6.1.4 were from Holtec Report HI-2104635, Table 2 "Surface Dose Rates for the HI-STORM at 55,000 MWD/MTU and 3 Year Cooling Time for Use in the Radiation Protection Program." These values were based on calculations from Holtec Report HI-2104636 which verified compliance with the 25 mrem annual dose limit in10 CFR 20.104(a).

Documents (a) Procedure DCS-201 "Transporting Loaded and Unloaded HI-STORM," Revision 2 Reviewed: (b) Procedure RPI-792 HI-STORM Overpack Surface Dose Rate, Revision 1 (c)

Procedure RPI-627 Job Coverage for Dry Fuel Storage, Revision 0 (d) Holtec Report No: HI-2104635 HI-STORM Radiation Protection Program Dose Rate Limits for Comanche Peak, Revision 4 (e) Holtec Report HI-2104636 "Dose Versus Distance from HI-STORM 100S Version B Containing MPC-32 for Comanche Peak," Revision 4 Category: Radiation Protection Topic: Dose Rate Survey - Storage Cask Top Reference: CoC 1014, Tech Spec A.5.7.5, A.5.7.8.d Amendment 7 Requirement: The licensee shall measure the surface dose rates (gamma + neutron) on each loaded storage cask. A minimum of 5 dose rate measurements shall taken on the top of the storage cask. One measurement should be taken at the center of the lid. The other 4 measurements shall be taken halfway between the center and edge of the concrete, approximately 90 degrees apart around the circumference. The measured dose rate shall not exceed the licensees site-specific surface dose rate limit as calculated from Technical Specification A.5.7.3 or 30 mrem/hour whichever is lower.

Observation: The licensee had established procedural requirements to perform the various required radiological surveys on the top of the HI-STORM storage cask as required by Certificate of Compliance 1014, Technical Specification A.5.7.5 and A.5.7.8.d. Procedure DCS-201, Step 8.4.56 required the cask loading supervisor to notify the radiation protection department that the cask was on the ISFSI pad and ready for the required radiation protection surveys. Step 6.2 of Procedure RPI-792 described the requirements for taking the gamma survey readings on the top of the HI-STORM storage cask. The requirements in Procedure RPI-792, Step 6.2 were identical to the requirements specified in the Certificate of Compliance 1014, Technical Specification A.5.7.8.d. One survey point was located in the center of the top of the lid. Four survey points were located in the middle of the top lid, approximately half way between the center and the edge of the top concrete shield, 90 degree apart around the circumference of the lid. Both gamma and neutron readings were required. The gamma and neutron readings were then entered into Form RPI-792-1 "HI-STORM Overpack Surface Dose Rate" for each of the five survey points and added together to obtain the total dose rate. The limits for the cask lid surveys was specified in Procedure RPI-792, Step 6.2.4 as 14.6 mrem/hr in the center and 20.5 mrem/hr at the other four locations. If these limits were exceeded, Step 6.2.4 required notification of the radiation protection supervisor and completion of a condition report per Step 6.4. An evaluation was required by Step 6.4 to verify that the correct contents had been loaded into the canister and that the annual limit in 10 CFR 72.104(a) of 25 mrem at the site boundary was not exceeded due to the higher dose rates.

Page 109 of 149

The lid surface dose rate limits specified in Procedure RPI-792, Step 6.2.4 were from Holtec Report HI-2104635, Table 2 "Surface Dose Rates for the HI-STORM at 55,000 MWD/MTU and 3 Year Cooling Time for Use in the Radiation Protection Program."

These values were based on calculations from Holtec Report HI-2104636 which verified compliance with the 25 mrem annual dose limit in10 CFR 20.104(a).

Documents (a) Procedure DCS-201 "Transporting Loaded and Unloaded HI-STORM," Revision 2 Reviewed: (b) Procedure RPI-792 HI-STORM Overpack Surface Dose Rate, Revision 0 (c)

Procedure RPI-627 Job Coverage for Dry Fuel Storage, Revision 0 (d) Holtec Report No: HI-2104635 HI-STORM Radiation Protection Program Dose Rate Limits for Comanche Peak Revision 4 (e) Holtec Report HI-2104636 "Dose Versus Distance from HI-STORM 100S Version B Containing MPC-32 for Comanche Peak," Revision 4 Category: Radiation Protection Topic: Dose Rate Survey - Transfer Cask Reference: CoC 1014, TS A.5.7.3, A.5.7.5 and A.5.7.8.a & b Amendment 7 Requirement: The licensee shall establish site specific dose rate limits (gamma and neutron) for the top and side of the transfer cask. The licensee shall measure the surface dose rates (gamma

+ neutron) for each loaded transfer cask. A minimum of 4 dose rate measurements shall taken on the side of the transfer cask at mid-plane, approximately 90 degrees apart around the circumference, and between the radial ribs of the water jacket. A minimum of 4 dose rates measurements shall be taken on the lid, halfway between the edge of the center opening and the outer edge of the lid, and approximately 90 degrees apart around the circumference. The measured dose rate shall not exceed the licensees site-specific surface dose rate limits as determined from Technical Specification A.5.7.3 or 30 mrem/hr (gamma + neutron) on the top and 300 mrem/hr on the side, whichever is lower.

Observation: The licensee had established the dose rate survey requirements in Procedure RPI-791 consistent with the Certificate of Compliance 1014 requirements. Procedure RPI-791, Step 6.1 established the required dose rate surveys for the side of the HI-TRAC transfer cask. Step 6.2 established the required surveys for the top of the transfer cask. For the sides, four survey points were required and their locations described on Part C of Form RPI-791-1 "HI-TRAC Transfer Cask Surface Dose Rates." The four survey points were at the midplane height and 90 degree apart around the circumference. The dose rates were required to be measured between the radial ribs of the water jacket. Form RPI-791-1 also described the required survey for the lid and showed the survey points on the drawing in Part C of Form RPI-791-1. The dose rates were measured at four points located half way between the edge of the inner hole of the lid and the outer edge of the top lid at 0, 90, 180, and 270 degree. All surveys were contact dose rates. The description of the survey points in Procedure RPI-971 were identical to those stated in Certificate of Compliance 1014, Technical Specification A.5.7.8.a and b. The gamma and neutron values were required to be recorded on Part D of Form RPI-971-1 and were added together to obtain the total dose rate. The maximum dose rate limits for the HI-TRAC transfer cask were specified in Procedure RPI-971, Step 4.2.2 as 458.7 mrem/hr on the sides and 66.1 mrem/hr on the top. These values were also included in Steps 6.1.4 and 6.2.4 as well as listed on Form RPI-791-1, Part D. If the dose rate limits were exceeded, Steps 6.1.4 and 6.2.4 referred to Step 6.3, which required the radiation protection supervisor to be notified and a condition report issued. Step 6.3 also required an evaluation to verify that the correct contents had been loaded into the canister and that Page 110 of 149

the annual limit in 10 CFR 72.104(a) of 25 mrem at the site boundary was not exceeded due to the higher dose rates.

The transfer cask side surface dose rate limits and top dose rate limits specified in Procedure RPI-791 were from Holtec Report HI-2104635, Table 1 "Surface Dose Rates for the HI-TRAC at 55,000 MWD/MTU and 3 Year Cooling Time for Use in the Radiation Protection Program." These values were based on calculations from Holtec Report HI-2104636 which verified compliance with the 25 mrem annual dose limit in 10 CFR 20.104(a).

Documents (a) Procedure RPI-791 HI-TRAC Transfer Cask Surface Dose Rate, Revision 1 (b)

Reviewed: Procedure RPI-627 Job Coverage for Dry Fuel Storage, Revision 0 (c) Holtec Report No: HI-2104635 HI-STORM Radiation Protection Program Dose Rate Limits for Comanche Peak Revision 4 (d) Holtec Report HI-2104636 "Dose Versus Distance from HI-STORM 100S Version B Containing MPC-32 for Comanche Peak," Revision 4 Category: Radiation Protection Topic: Neutron Dosimetry Reference: FSAR 1014, Section 5.2.2 Revision 9 Requirement: Curium (Cm-244) accounts for approximately 92 - 97% of the total number of neutrons produced. Alpha, neutron reactions in isotopes other than Cm-244 account for approximately 0.3 - 2% of the neutrons produced while spontaneous fission in isotopes other than Cm-244 account for 2 - 8% of the neutrons produced within the fuel.

Observation: The licensee had incorporated considerations into the health physics monitoring program for the higher energy neutron spectrum that would be present around the canister when empty of water. When the spent fuel was in the spent fuel pool or in the canister while the canister was filled with water or in the HI-STORM 100 concrete cask, the neutron spectrum encountered by workers would be similar to the moderated neutron spectrum typically found at a nuclear power plant. As such, neutron radiation survey instruments and personnel dosimeters normally used for neutron monitoring would adequately monitor the personnel doses. However, when the water was removed from the canister, the neutron dose would be higher energy and as such would have a higher quality factor (see 10 CFR 20.1004 for energy dependent quality factors) resulting in a higher dose to workers. Some neutron survey instruments and personnel dosimeters are energy dependent and would not properly measure the neutron dose at higher energy levels.

An internal evaluation was performed by the Comanche Peak radiation protection organization to evaluate the neutron measuring instrumentation and the personnel dosimeters used onsite to determine if correction factors would be necessary to account for the variation of the neutron energy spectra during different phases of the loading and storage operations This report was incorporated into the response to Condition Report CR-2010-006437-4. The evaluation documented the methodology the licensee planned to use to monitor radiation workers involved in the spent fuel transfer processes. The licensee planned to utilize conservative dose correction factors for the initial loading effort and determine appropriate dose correction factors for electronic personnel dosimeters for future dose monitoring based on data obtained during the initial loading.

The licensee used the Landauer optically stimulated luminescence (OSL) dosimeter Page 111 of 149

badge for personnel monitoring with processing supplied by Landauer. The badges included a Neutrak dosimeter that employed CR-39 and track etch technology for measuring exposure to neutrons of varied energies. The Neutrak used a polyethylene radiator that recorded recoil protons resulting from neutron interactions for measuring fast neutron and a boron loaded Teflon radiator that recorded alpha particles resulting from neutron interactions for fast, intermediate, and thermal neutron measurements. The fast neutron CR-39 chip has an energy range from 40 keV to 40 MeV and a dose measurement range from 20 mrem to 25 rem. The fast, intermediate, and thermal neutron CR-39 chip has an energy range from 0.25 eV to 40 MeV and a dose measurement range from 10 mrem to 25 rem.

The Siemens electronic personnel dosimeter (EPD) was currently used by the licensee to measure both gamma and neutron exposure. Plans were to transition to the MG DMC2000N electronic personnel dosimeter in the future. The energy range for the Siemens electronic personnel dosimeter was 0.25 eV to 15 MeV with a dose measurement range of 1 mrem to 1,000 rem. The licensee recognized that a limitation for the electronic dosimeter was its energy dependence and that a correction factor for the neutron spectrum for various evolutions was necessary to determine dose to an individual. A correction factor multiplier of 1.7 used during routine activities at the site during normal operations was based on Bonner sphere measurements of the most common neutron spectra encountered at the site and comparison of the electronic personnel dosimeter to the Landauer CR-39 chip. During initial cask loadings, a correction factor/multiplier of 2 will be applied to the worker's electronic personnel dosimeter readings for conservatism. Data obtained from the OSL dosimeter badges, the electronic personnel dosimeter, and the survey instruments during the first loading campaign will be used to better refine the correction factor.

For survey instrument measurements in the field, the licensee planned to use the ASP-1 portable neutron detector (Remball) interfaced with an Eberline radiation detection systems. The ASP-1 was a boron trifluoride detector surrounded by a polyethylene sphere that acted as a moderator. The ASP-1 was expected to under-respond to the neutron energies anticipated during cask loading. Results obtained with this system will be compared to results obtained with a REM-500 tissue equivalent proportional counter.

The licensee's REM-500 was calibrated April 12, 2011 to two NIST traceable Californium-252 (Cf-252) sources. Tests of the REM-500 instrument response had been conducted by the National Institute of Science and Technology (NIST) and showed close agreement between results for bare and for moderated CF-252 sources by the REM-500.

The licensee planned to conduct tests to establish correction factors for the ASP-1 neutron monitoring instrument and the dosimeters based on the REM-500 readings.

Dosimetry packs consisting of two OSL dosimeter badges with CR-39, two Siemens electronic personnel dosimeters, and two DMC2000N electronic personnel dosimeters would be placed in a zip lock bag. One bag would be placed on the side of the canister and one on the top during the time the canister was filled with water. Two new bags of dosimeters would be used when the canister was filled with helium. And two new bags would be used to take measurements when the cask was on the ISFSI pad. Readings at each of these times would also be taken using the REM-500 and the ASP-1 Remball.

The licensee stated that the REM 500 will be used to establish dose rates until Page 112 of 149

performance of the ASP-1 has been evaluated and appropriate correction factors determined.

Documents (a) Condition Report CR-2010-006437-4 "Evaluation of Neutron Dose Rates During Reviewed: Spent Fuel Dry Cask Loading and Storage Evolutions," initiated July 6, 2011 (b) Internal Comanche Peak Report "Neutron Monitoring and Determination of EPD Correction Factors for Dry Cask Storage," dated May 17, 2011 [attached to Condition Report CR-2010-006437-4] (c) Health Physics Instruments Calibration Record for the REM 500, Cal # 15927, Serial # 347, calibrated April 12, 2011 (d) Calibration Certificate from Edward F. Janzow, Frontier Technology Corp. to Far West Technology, Inc. for a CF-252 neutron source, Model 100, Serial No. FTC-CF-199 calibrated on January 28, 1991 (e) Letter from James A. Booth, Frontier Technology Corp. to Far West Technology entitled "Statement of Neutron Emission Rate," dated May 6, 2008 (f) Letter from U. S.

Department of Commerce, Director, National Institute of Standards and Technologies (multiple signatures including Edward Boswell, Physical Scientist) to Far West Technology entitled "Report of Test, NIST Test No. 249769" dated March 8, 1992 Category: Radiation Protection Topic: Shielding Effectiveness Test Reference: FSAR 1014, Section 9.1.5.2, Table 9.2.1 Revision 9 Requirement: Following the first fuel loading of each HI-STORM 100 cask system, a shielding effectiveness test shall be performed at the loading facility site to verify the effectiveness of the radiation shield. This test shall be performed after the HI-STORM storage cask and HI-TRAC transfer cask have been loaded with a canister containing spent fuel assemblies and the canister has been drained, moisture removed, and backfilled with helium.

Observation: The shielding effectiveness test requirements were met by performing the radiological surveys required by Technical Specification A.5.7.8 of Certificate of Compliance 1014.

The surveys required by Technical Specification A.5.7.8 had been incorporated into Procedure RPI-792. The dose rate limits for the HI-STORM storage cask were listed in Step 4.2.3.1 for the sides of the storage cask, Step 4.2.3.2 for the lid, and Step 4.2.3.3 for the inlet and outlet vents. A prescribed number of survey points and their locations were included in Procedure RPI-792 and Form RPI-792-1 "HI-STORM Overpack Surface Dose Rates." Procedure RPI-792 was implemented after the HI-STORM storage cask had been placed on the ISFSI pad and the radiation protection staff notified in accordance with Procedure DCS-201, Step 8.4.56. Procedure RPI-792, Step 6.1.1 required contact gamma radiation surveys at 12 pre-designated points. Four measurements were at the mid-height horizontal plane, 90 degree apart around the circumference, four measurements were on the upper horizontal plane approximately 60 inches above the mid-height plane, 90 degree apart around the circumference, and four measurements were on the lower horizontal plane, approximately 60 inches below the mid-height plane, 90 degree apart around the circumference. Step 6.1.2 required the same locations for contact neutron dose rates. The values were recorded on Form RPI-792-1. This provided for a total of 12 survey points as required by Technical Specification A.5.7.8.c. Procedure RPI-792, Step 6.2.1 required contact gamma survey readings on the top of the HI-STORM lid at five pre-determined survey points. One was in the center of the lid and four points were in the middle of the top lid approximately half way between the center of the lid and the edge of the top concrete shield, 90 degree Page 113 of 149

apart around the circumference. Step 6.2.2 required contact neutron readings at the same survey points. These values were recorded on Form RPI 792-1. This provided for the five survey points required by Technical Specification A.5.7.8.d. Procedure RPI-792, Step 6.3.1 required contact gamma surveys of each of the four bottom (inlet) vent ducts.

Step 6.3.2 required a neutron contact reading of the four bottom (inlet) vent ducts. Step 6.3.5 required contact gamma readings of each of the four upper (outlet) vent ducts. Step 6.3.6 required a neutron contact reading at each of the upper (outlet) vent ducts. The vent duct values were recorded on Form RPI-792-1. This provided for the required measurements on the inlet and outlet vent duct screens required by Technical Specification A.5.7.8.e.

Holtec Report HI-2104635 had calculated acceptable dose rates at each of the survey points established in Procedure RPI-792 and listed the values in Holtec Report HI-2104635, Table 2 "Surface Dose Rates for the HI-STORM at 55,000 MWD/MTU and Three Year Cooling Time for Use in the Radiation Protection Program." These values were incorporated into Procedure RPI-792, Step 4.2.3 and listed on Form RPI-792-1.

The values calculated in the Holtec report were less than the Technical Specification A.5.7.4 upper limits of 300 mrem/hr on the side of the storage cask and 30 mrem/hr on the cask lid.

Documents (a) Procedure RPI-792 "HI-STORM Overpack Surface Dose Rate," Revision 1 (b)

Reviewed: Procedure DCS-201 "Transporting Loaded and Unloaded HI-STORM," Revision 2 (c)

Holtec Report No: HI-2104635 HI-STORM Radiation Protection Program Dose Rate Limits for Comanche Peak Revision 4 Category: Radiation Protection Topic: Site-Specific Dose Rate Limits - Storage Cask Reference: CoC 1014, Tech Spec A.5.7.3, A.5.7.4 Amendment 7 Requirement: The licensee shall establish site-specific surface dose rate limits (gamma + neutron) for the top and sides of the storage cask, and for the inlet and outlet air vents. The surface dose rate limits for the storage cask may be set between the dose rate assumed in the 72.104(a) analysis and the dose rate needed to exceed the 72.104(a) dose limits, but shall not be set greater than 30 mrem/hr on the top and 300 mrem/hr on the side.

Observation: The licensee established site-specific surface dose rate limits for total gamma and neutron for the top and sides of the storage cask, and for the inlet and outlet air vents.

The dose rate limits were calculated in Holtec Report No. HI-2104635. The surface dose rate analysis resulted in a dose rate profile as a function of location on the HI-STORM storage cask. The respective surface dose rate limits for the required survey points were summarized in Table 2 "Surface Dose Rate for the HI-STORM at 55,000 MWD/MTU and 3 Year Cooling Time for Use in the Radiation Protection Program" of the Holtec Report HI-2104635 and were also provided in the output summary in Appendix D "HI-STORM 100S Version B Results" of the report. The maximum calculated site-specific surface dose rate limit for the top of the storage cask was 20.5 mrem/hr and the maximum site-specific dose rate limit for the side of the cask was 40 mrem/hr. The site-specific values were less than the maximum level of 30 mrem/hr for the top of the cask and 300 mrem/hr for the side of the cask specified in Certificate of Compliance 1014, Technical Specification A.5.7.4.

Page 114 of 149

Holtec Report HI-2104635, Section 4 "Assumptions" discusses the basis for the calculations to determine the dose rates for the cask. Holtec Report HI-2104635 used a B&W 15x15 fuel assembly design in the calculations. This fuel assembly had been determined by Holtec to be a bounding assembly design. The Holtec Final Safety Analysis Report (FSAR), Section 5.2 "Source Specifications" discussed the basis for the 15x15 B&W fuel assembly as a bounding design for other fuel assemblies such as the 17x17 B&W fuel used at Comanche Peak. The bounding values for burnup, cooling time, and enrichment of 55,000 MWD/MTU, 3 years and 3.9% U-235, respectively, were selected based on the fuel inventory at Comanche Peak. The computer codes used for the calculations were SAS2H, ORIGEN-S, and MCNP 4A.

The calculation identified four distinct primary radiation source terms, which included neutron source from the decay of spent nuclear fuel, photon source from the decay of the spent nuclear fuel, photons from the decay of Co-60 in the end-fittings of the fuel assemblies, which represents the activation of the steel components in the fuel assemblies, and photons from the decay of Co-60 in the non-fuel hardware. The calculation also included secondary neutrons and photons, which were accounted for during the MCNP calculation by running a coupled neutron-gamma calculation. The results of the calculations were incorporated into Procedure RPI-792, Step 4.2.3 as the limits to be applied to the dose rate measurements taken of the storage cask after placement on the ISFSI pad. Holtec Report HI-2104636 provided calculations based on the source term used in Holtec Report HI-2104635 and concluded that the dose rate at the controlled area boundary at 1048.5 meters would be 2.96 mrem/yr. This was well within the 25 mrem/yr limit in 10 CFR 72.104(a).

Documents (a) Procedure No. RPI-792 HI-STORM Overpack Surface Dose Rate, Revision 1 (b)

Reviewed: Holtec Report No: HI-2104635 HI-STORM Radiation Protection Program Dose Rate Limits for Comanche Peak Revision 4 (c) Holtec Report HI-2104636 "Dose Versus Distance from HI-STORM 100S Version B Containing MPC-32 for Comanche Peak,"

Revision 4 (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Category: Radiation Protection Topic: Temporary Shielding Reference: FSAR 1014, Sect 10.1.4; Tables 10.1.1 and 10.1.2 Revision 9 Requirement: To minimize occupational dose during loading and unloading operations, a specially-designed set of auxiliary shielding is available. The shielding consists of an automated welding system (AWS) baseplate, transfer cask temporary shield ring, annulus shield, storage cask vent duct shield insert, transfer cask transfer step, mating device, and shield panel trim plates. The 125D transfer cask uses the mating device instead of the transfer step. The licensee shall determine the need for the auxiliary shielding.

Observation: Temporary shielding had been adequately incorporated into the licensees procedures for use during loading and unloading operations. Procedure RPI-627, Section 6.7 "Welding of the MPC Lid" discussed the modular shield that remained installed during welding.

Step 6.8 "RVOA Installation and Blowdown" directed the radiation protection staff to place additional shielding blankets around the removable valve operator assembly (RVOA) holes, as needed. Step 6.10 "Final Welding" directed the use of local shield blankets during hand welding, as needed. Procedures DCS-203, DCS-204 and DCS-207 Page 115 of 149

incorporated numerous steps involving shielding including the use of the shield plate, temporary shielding for the annulus, and temporary shielding on the canister lid after the welding machine was removed.

Documents (a) Procedure RPI-627 "Job Coverage for Dry Fuel Storage," Revision 1 (b) Procedure Reviewed: DCS-203 "MPC Handling and Fuel Loading Operations," Revision 3 (c) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling)," Revision 2 (d)

Procedure DCS-207 "Unloading a Loaded MPC," Revision 2 Category: Radiation Protection Topic: Transfer Cask Surface Contamination Limit Reference: CoC 1014, Tech Spec A.3.2.2 Amendment 7 Requirement: Removable contamination on the exterior surfaces of the transfer cask and accessible portions of the canister shall not exceed 1000 disintegrations per minute per 100 square centimeters (dpm/100 square centimeters) from beta and gamma sources and 20 dpm/100 square centimeters from alpha sources. The accessible portion of the canister is the upper portion of the canister external shell wall accessible after the inflatable annulus seal is removed and before the annulus shield ring is installed.

Observation: The contamination limits from Technical Specification A.3.2.2 of Certificate of Compliance 1014 had been incorporated into the licensee's procedures. Procedure RPI-790, Section 4.2 "Limitations" and Procedure DCS-203, Step 8.9.10 specified the 1,000 disintegrations per minute (dpm) per 100 square centimeters (100 square cm)

beta/gamma and 20 dpm/100 square cm alpha as the contamination limits. When the HI-TRAC transfer cask containing the canister was removed from the spent fuel pool and placed in the dry cask pit, the top exterior surfaces were decontaminated per Procedure DCS-203, Step 8.9.3. The annulus seal was removed per Step 8.9.6. Then a survey was performed on the canister lid and the top 3 inches of the canister side per Steps 8.9.9 and 8.9.10 to verify that the contamination limits in Technical Specification A.3.2.2 were met.

Procedure RPI-790 would be used to perform the survey. Section 4.1 "Precautions" stated that the top of the canister and HI-TRAC transfer cask and the canister shell down to the inflatable annulus seal were to be decontaminated prior to the contamination survey, but the area below the seal was not to be decontaminated. The seal was to be removed and water drained approximately 6 inches, then the contamination survey performed. Step 4.3.5 stated that if the contamination limits were not met for the area under the annulus seal, then the inaccessible areas of the canister should be considered contaminated with loose surface contamination. This condition could be resolved by flushing the annulus with clean demin water or unloading the canister and removing it from the HI-TRAC transfer cask in order to decontaminate the canister shell and inside of the HI-TRAC. Procedure RPI-790, Section 6.2 "Contamination Smear Analysis" stated that if the contamination limits were exceeded, to notify the radiation protection supervisor and proceed to Section 6.3 "Corrective Action." Section 6.3 required a condition report to be issued and a plan to decontaminate the canister be developed and completed within 7 days. Controls to mitigate the spread of contamination were required to be established and the affected area posted in accordance with station radiation protection posting requirements. Procedure RPI-627 in the note in Section 6.6

"Decontamination of the HI-TRAC" provided a reminder that decontamination of the Page 116 of 149

area under the annulus seal was not to be performed prior to the contamination survey to verify that the annulus seal had not leaked.

The requirements in the Comanche Peak procedures were consistent with the Holtec Final Safety Analysis Report (FSAR), Section 8.1.5.2 which stated that after decontaminating the canister lid top and the shell area above the annulus seal, to deflate the seal and survey the canister lid top surface and the accessible areas of the top three inches of the canister. An ALARA Note preceding Step 8.1.5.2.f stated The canister exterior shell survey is performed to evaluate the performance of the inflatable annulus seal. Indications of contamination could require the canister to be unloaded. In the event the canister shell is contaminated, users must decontaminate the annulus. If the contamination cannot be reduced to acceptable levels, the canister must be returned to the spent fuel pool and unloaded. The canister may then be removed and the exterior shell decontaminated.

Documents (a) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 1 (b)

Reviewed: Procedure RPI-627 "Job Coverage for Dry Fuel Storage," Revision 0 (c) Procedure RPI-790 MPC Accessible Surface Contamination Survey, Revision 0 (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Records Topic: Cask Records Reference: 10 CFR 72.234(d)(2) & (d)(3) Published 2011 Requirement: A list of records required for each cask is provided in 10 CFR 72.234(d)(2). The certificate holder is required by 10 CFR 72.234(d)(3) to provide an original of these records to the user Observation: The licensee was maintaining the required records in their quality related records system consistent with 10 CFR 72.234. The records system was maintained in accordance with Procedures STA-116 and STA-302, which included specific requirements for the dry cask storage program records. A tour of the records vault was performed by the NRC inspectors and specific records requested to verify their retrievability. The records included the originals received from the vendor, Holtec, International. Hard copies were still available, but would eventually be converted to electronic files (scanned) and maintained in the licensee's electronic file system, SPARCS (Station process & records control system). The hard copies would be destroyed once an electronic copy was made. The licensee had documentation packages for the three casks planned for loading in the first campaign. This included the Hi-Storm 100 storage cask fabrication records for Serial No.s SN465, SN466, and SN467 and the MPC-32 canister fabrication records for Serial No.s SN156, SN157, SN158. The document packages for HI-STORM storage cask #466 and canister #158 were reviewed in detail against the 10 CFR 72.234 requirements. The document packages including the following information: (1) copy of Certificate of Compliance #72-1014, Amendment 7; (2) documentation showing fabrication start dates for SN156, SN157, and SN158 of August 6, 2010, August 11, 2010, and August 17, 2010, respectfully; (3) documentation showing the fabrication completion date of February 28, 2011 for all three canisters SN156, SN157, and SN158; (4) a Certificate of Conformance letter from Holtec stating that the canister was manufactured and processed in accordance with Holtecs QA Manual, Revision, 14 and Page 117 of 149

Comanche Peak's purchase order; (5) a Certificate of Compliance letter from Holtec stating that the material procurement, fabrication, and inspection of the components were in full conformance with the provisions of the USNRC Certificate of Compliance Number 72-1014 issued pursuant to the governing USNRC regulations; and (6) the address of the licensee (Comanche Peak) was found on the table containing the fabrication.

Documents (a) Procedure STA-116 "Maintenance of CPNPP Licensing Basis Documents, Operating Reviewed: License Conditions, and Technical Specifications," Revision 11 (b) Procedure STA-302

"Station Records," Revision 22 (c) Holtec Document Package 1024-466 for Hi-Storm

1. 466, Revision 0 (d) Holtec Document Package 1023-158 for MPC #158, Revision 0 Category: Records Topic: Maintaining a Copy of the CoC and Documents Reference: 10 CFR 72.212(b)(11) Published 2011 Requirement: The general licensee shall maintain a copy of the Certificate of Compliance and, for those casks to which the license has applied the changes of an amended Certificate of Compliance, the amended Certificate of Compliance, and the documents referenced in the certificates, for each cask model used for storage of spent fuel, until use of the cask model is discontinued.

Observation: Procedure STA-116, Section 4.14 "Licensing Basis Documents (LBDs)" listed the following documents for retention by the licensee: (1) Certificate of Compliance 1014, Amendment 7, dated December 28, 2009 (SER attached), (2) technical specifications and bases for the HI-STORM 100 cask system (Appendices A and B to Certificate of Compliance 1014 for above ground systems), and (3) Holtec International Final Safety Analysis Report (FSAR) for the HI-STORM 100 cask system, Revision 9, dated February 13, 2010. The records were maintained in a computerized system called the Electronic Licensing Basis Documents system.

Documents (a) Procedure STA-116 "Maintenance of CPNPP Licensing Basis Documents (LBDS),

Reviewed: Operating License Conditions, and Technical Specifications," Revision 11 (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Records Topic: Notice of Initial Loading Reference: 10 CFR 72.212(b)(1) Published 2011 Requirement: The general licensee shall notify the NRC at least 90 days prior to first storage of spent fuel.

Observation: Luminant Power notified the NRC by letter dated April 4, 2011, of the plans to begin fuel loading at the Comanche Peak site on July 11, 2011. This notification met the requirements of the 90 day notification of initial loading required by 10 CFR 72.212(b)(1).

Documents (a) Letter (CP-201100459) from Fred Madden, Luminant Power to USNRC Document Reviewed: Control Desk entitled "Comanche Peak Nuclear Power Plant (CPNPP) Docket Nos. 50-445, 50-446, and 72-74 Notification of Cask Loading Activities Pursuant to 10 CFR 72.212(b)(1)(i)," dated April 4, 2011 Page 118 of 149

Category: Records Topic: Record Retention for 72.212 Analysis Reference: 10 CFR 72.212(b)(5)(iii) Published 2011 Requirement: A copy of the 10 CFR 72.212 analysis shall be retained until spent fuel is no longer stored under the general license issued under 10 CFR 72.210.

Observation: Procedure STA-116, Section 4.14 "Licensing Basis Documents" required the 72.212 Evaluation Report to be retained as a licensing basis document.

Documents (a) Procedure STA-116 "Maintenance of CPNPP Licensing Basis Documents (LBDS),

Reviewed: Operating License Conditions, and Technical Specifications," Revision 11 (b) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Category: Records Topic: Registration of Casks with NRC Reference: 10 CFR 72.212(b)(2) Published 2011 Requirement: The general licensee shall register the use of each cask with the NRC no later than 30 days after using the cask to store spent fuel.

Observation: The requirement to notify the NRC within 30 days of using a cask to store spent fuel was incorporated into procedure STA-502, Attachment 8.A "List of Routine Reports Assigned to Nuclear Generation" as STA No. RN-113. The Director, Oversight and Regulatory Affairs was assigned the reporting responsibility.

Documents (a) Procedure STA-502 "Routine Reporting," Revision 13 Reviewed:

Category: Safety Reviews Topic: Changes, Tests, and Experiments Reference: 10 CFR 72.48(c)(1) Published 2011 Requirement: A licensee can make changes to their facility or storage cask design if certain criteria are met as listed in 10 CFR 72.48.

Observation: The licensee had combined the 72.48 screening and evaluation process with the 50.59 process used at the site. Procedure STA-707 described the screening and evaluation process for both requirements and used several different forms. Section 6.3 "Scope Review and Applicability Determination" directed the responsible person initiating the review to use Form STA-707-4 "Applicability Determination" to determine if the activity was within the scope of this procedure. This initial applicability determination allowed the filtering out of editorial updates and administrative changes. Form STA-707-4 asked a series of questions and provided guidance on where to proceed if the answer to the question was "yes." Question VI stated "Does the proposed activity involve a change to the ISFSI facility or dry storage cask design or operation as described in the dry storage cask Final Safety Analysis Report (FSAR) or the 72.212 Report?" If the answer was yes, the change required a screening of the issue. Procedure STA-707, Section 6.5.2 "72.48 Screen" provided instructions for conducting a screening. The screening was documented on Form STA-707-3 "72.48 Screen." The screening involved a series of questions consistent with the required questions listed in 10 CFR 72.48. If the change required a change to the Certificate of Compliance or it's appendices, the user was directed to notify the Certificate of Compliance holder and request the change. If the Page 119 of 149

activity required a change to the 72.212 Evaluation Report, the user was directed to proceed to Procedure STA-116. Procedure STA-116, Section 2.0 "Applicability" listed the 72.212 Evaluation Report as a licensing basis document. Step 4.15 stated that changes to licensing basis documents were processed in accordance with Attachment 8.D

"Initiating and Processing a Change to a Licensing Basis Document." Attachment 8.D included a number of processing steps, include Step 1.2.3 which required attaching the 72.48 applicability determination along with any 72.48 screening and evaluations performed related to the change.

If any of the questions on Form STA-707-3 relating to the requirements listed in10 CFR 72.48 were answered yes, then Form STA-707-5 "72.48 Evaluation" was required to be completed. This form required a description of the activity and included the questions from 10 CFR 72.48(c) to determine if the change could be implemented without an amendment to the Certificate of Compliance pursuant to 10 CFR 72.244 (for general licensees). Form STA-707-3 also included a question related to whether the proposed activity created a deviation from or required a change to the 10 CFR 71 Certificate of Compliance or safety analysis report related to the transportation of the spent fuel canister. If so, the Certificate of Compliance holder was to be notified.

The licensee had developed a resource manual and a training module for the 10 CFR 72.48 program. The resource manual provided a good description of the purpose and philosophy of the 10 CFR 72.48 process and how it related to other processes that controlled licensing basis activities at Comanche Peak. Relevant definitions and applicable terms were provided with discussions of what they meant. A detail discussion was provided related to the meaning of the statements in 10 CFR 72.48 that related to the screening process and the evaluation process. The resource manual was well developed and very informative in assisting the user in making decisions related to the 72.48 screening and evaluation process. Examples were provided to further illustrate how to implement the process. The training material included numerous drawings and pictures describing the various components of the dry cask system specific to Comanche Peak to help familiarize the personnel assigned to perform the screenings and evaluations with the key safety components of the various systems. The training module described the forms required and gave examples of what the various screening and evaluation criteria meant. Lessons learned at other sites relevant to the 10 CFR 72.48 process were provided.

Nine applicability determinations were reviewed related to Revision 1 of the 72.212 Evaluation Report to evaluate the 72.48 process being used at Comanche Peak. Issues described in the applicability determination review ranged from adding and correcting references, revising sentences to better reflect the intended meaning, clarifying that the dose rate calculation for the owner controlled area performed by Holtec were at 80% of the actual owner controlled area distance, and adding a description of the radiation detector that will be used for monitoring for a criticality while loading the canister. Two issues met the requirement for a 72.48 screening. These were tornado driven missiles and the fire potential of the vertical cask transporter. Both issues answered "yes" on the applicability determination Form STA-707-4 for question VI "Change to the ISFSI facility or dry storage cask design or operation as described in the dry storage cask FSAR or 72.212 Evaluation Report." The screening was performed using Form STA-707-3 on Page 120 of 149

the two issues and resulted in a "yes" answer to screening question III.1 "Does the proposed activity involve a change to a structure, system, or component that adversely affects a dry storage cask FSAR described design function?" In addition, the issue related to the vertical transporter fire also answered "yes" to screening question III.6

"Does the proposed activity require a change to the dry storage cask Certificate of Compliance, including appendices?" A 72.48 evaluation was performed for each of the two issues using Form STA-707-5. One related to the potential fire associated with the transporter (EV-CR-2011-007002-15) and one related to tornado driven missiles (EV-CR-2011-007002-14). The 72.48 evaluations were performed because the two conditions were not bounded by the Holtec FSAR. The wheeled vertical transporter planned for use at Comanche Peak had a fire potential due to the rubber tires that exceeded the Holtec FSAR design basis accident analysis of 50 gallons. The tornado missiles assumed by Comanche Peak for their Part 50 site analysis were not bounded by the tornado missile assumptions used in the Holtec FSAR. Neither evaluation resulted in the need to amend the certificate of compliance.

Documents (a) Procedure STA-116, Maintenance of CPNPP Licensing Basis Documents Operating Reviewed: License Conditions and Technical Specifications, Revision 11 (b) Procedure STA-707

"10 CFR 50.59 and 10 CFR 72.48 Reviews," Revision 18 (c) CPNPP 72.48 Resource Manual," Revision 0 (d) Training Material PTB1STAXA5 "10 CFR 72.48 Screening and Evaluation Training," Change No. 1 (e) Form STA-707-4 Applicability Determination

"CPNPP 10CFR72.212 Evaluation Report," closed as Condition Report AI-CR-2011-007002, Action Tasks 1 through 11 (f) Form STA-707-3, Activity Document No. EV-CR-2009-000859-00-89 "72.48 Screen-Master FDA Dry Cask Storage Project," prepared July 20, 2011 (g) Form STA-707-3, Activity Document No. EV-CR-2009-000859-00-92

"72.48 Screen-FDA-2009-000859-03-00 Licensed and Ancillary Equipment Tagging and Evaluation," prepared July 27, 2011 (h) Form STA-707-5, 72.48 Evaluation No. EV-CR-2011-007002-14 "Site Specific Fire Hazards Evaluation (13769701-R-M-00002, Rev. 1 Comanche Peak ISFSI Project Evaluation of Fire Hazards)," Revision 0 (i) Form STA-707-5, 72.48 Evaluation No. EV-CR-2011-007002-15 "Site Specific Tornado Missile Evaluation (HI-2104637, Rev. 0 Environmental Hazards Evaluation for Comanche Peak HI-STORM)," Revision 0 (j) Comanche Peak Nuclear Power Plant 10 CFR 72.212 Evaluation Report, Revision 1 Category: Slings Topic: Sling Heavy Load Requirements Reference: NUREG 0612, Section 5.1.6 (1) (b) Issued July 1980 Requirement: Dual or redundant slings should be used such that a single component failure or malfunction in the sling will not result in an uncontrolled lowering of the load, OR the load rating of the sling should be twice the sum of the static and dynamic loads.

Observation: Dual and redundant slings were used to download the canister from the HI-TRACK transfer cask to the HI-STORM storage cask. The load rating on the slings was twice the weight of the static and dynamic loads of the canister. The slings purchased at Comanche Peak had a vertical rating of 50,000 pounds with a rated capacity of 100,000 pounds in the basket formation. Two of these slings were used to download the canister.

The fully loaded canister was calculated to be 87,455 pounds per Holtec Document HI-2104639. The static value of 90,000 pounds was used in the calculation on Attachment A of Purchase Specification PS-1211. A dynamic load of 10% was added to the static Page 121 of 149

load. The sum was multiplied by a safety factor of two and divided by the two slings in the basket formation. The minimum vertical rated capacity of each sling was calculated to be 49,500 pounds. The slings that were purchased by Comanche Peak were 50,000 lb (vertical rated) capacity slings (TPSE-EE5000 - 55ft Twin-Path Sparkeater eye & eye slings).

Documents (a) Holtec Document No. HI-2104639 "Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak," Revision 0; (b) Holtec Purchase Specification. PS-1211 "MPC Lift Sling," Revision 5 (c) NUREG 0612 Control of Heavy Loads at Nuclear Power Plants, issued July 1980 Category: Slings Topic: Sling Identification Reference: ASME B30.9, Section 9-5.3. Revision 1984 Requirement: Each sling should be permanently marked to show: (a) name or trademark of manufacturer; (b) manufacturers code or stock number; (c) rated loads (rated capacities)

for the types of itches used; (d) type of natural or synthetic material; (e) date of manufacture.

Observation: Slings used at Comanche Peak for the dry cask storage project were tagged with the key information specified in ASME B30.9. Procedure MDA-402, Attachment 8.C Prior to Use Inspection of Rigging Equipment included inspection requirements for the synthetic round/web slings and the wire ropes. For the wire ropes, the pre-use inspection required verification that the wire rope was marked or tagged with the manufacturer's name, size, and capacity. For the synthetic slings, Attachment 8.C required that the pre-use inspection verify that the sling was marked or tagged with the manufacturers name, model/stock number, size, capacity, and type web material (also type of cover material on round slings). The NRC inspector reviewed the following synthetic slings to verify the information was current and readable: canister downloading slings, canister lid slings, and HI-STORM storage cask lid slings.

Documents (a) Procedure MDA-402 Control of Load Handling Equipment, Revision 11 (b)

Reviewed: American National Standard [American Society of Mechanical Engineers] (ASME)

B30.9 Slings, Revision 1984 Category: Slings Topic: Sling Inspections - Frequent Reference: ASME B30.9, Section 9-4.7.1 (b) Revision 1984 Requirement: A visual inspection for damage shall be performed each day or shift the sling is used.

Observation: A visual inspection for damage was required on each sling prior to use. Procedure MDA-402, Attachment 8.C Prior-to-Use Inspection of Rigging Equipment required inspection of wire rope slings, synthetic web slings, and round slings prior to use.

Attachment 8.C required ensuring the sling was marked with the size and capacity and the preventive maintenance due date had not expired. Each sling was inspected for damage such as cuts, crushing, kinks, broken wire, corrosion, heat damage, end connections for damage, acid or burns, tears, holes, snags, worn stitching, knots, and the tattle-tails such that their length meets the criteria described on the tag. During the dry run demonstrations involving heavy lifts, the NRC inspectors observed the sling Page 122 of 149

inspections prior to their use.

Documents (a) Procedure MDA-402 Control of Load Handling Equipment, Revision 11 (b)

Reviewed: American National Standard [American Society of Mechanical Engineers] (ASME)

B30.9 Slings, Revision 1984 Category: Slings Topic: Sling Inspections - Periodic Reference: ASME B30.9, Section 9-4.7.1 Revision 1984 Requirement: A complete inspection for damage to the sling shall be conducted at intervals not to exceed one year.

Observation: The slings used for the dry cask storage project dry runs were inspected daily (prior to use) and monthly. Procedure DCS-111 established the periodic inspection requirement for the slings. Section 8.5 "Daily/Monthly Empty MPC Lift Sling Inspection" required inspection of the canister lift slings. Section 8.6 "Daily/Monthly MPC Lid Rigging Sling Inspection" required inspection of the canister lid lift slings. Section 8.11 "Inspection of MPC Loaded Long Lift Slings" required inspection of the canister downloading slings.

All slings were required by Procedure DCS-111 to be inspected prior to use and on a monthly schedule.

Documents (a) Procedure DCS-111 Inspection and Testing of Dry Cask Storage Lifting Devices, Reviewed: Revision 1 (b) American National Standard [American Society of Mechanical Engineers]

(ASME) B30.9 Slings, Revision 1984 Category: Slings Topic: Sling Load Rating Reference: NUREG 0612, Section 5.1.1 (5) Issued July 1980 Requirement: In selecting the proper sling, the load used should be the sum of the static and maximum dynamic load. The rating identified on the sling should be in terms of the "static load" which produces the maximum static and dynamic load.

Observation: The proper slings were selected for use on the dry cask storage project. The heaviest and most critical lift involved the lowering of the loaded and sealed canister from the HI-TRAC transfer cask into the HI-STORM storage cask. Two slings were used for the downloading operation. Calculations documented in Attachment A of Purchase Specification PS-1211 showed that the sum of the static and dynamic loads were used to calculate the minimum vertical rated capacity of each sling. The fully loaded canister was calculated to weigh 87,455 pounds per Holtec Document HI-2104639. The static value used in the calculation on Attachment A of Purchase Specification PS-1211 was 90,000 pounds. A dynamic load of 10% was added to the static load. The sum was multiplied by a safety factor of two and then divided by two to represent the configuration of the slings in a basket formation during the lift. The required minimum vertical rated capacity of each sling was calculated to be 49,500 pounds. The slings that were purchased by Comanche Peak were 50,000 pound vertical rated capacity slings (TPSE-EE5000 - 55ft Twin-Path Sparkeater eye & eye slings).

Documents (a) Holtec Document No. HI-2104639 "Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak," Revision 0 (b) Holtec Purchase Specification. PS-1211

"MPC Lift Sling," Revision 5 (c) NUREG 0612 Control of Heavy Loads at Nuclear Page 123 of 149

Power Plants, issued July 1980 Category: Slings Topic: Sling Proof Loading Reference: ASME B30.9, Section 9-5.4 Revision 1984 Requirement: When specified by the purchaser, web slings of all types shall be proof loaded. The proof load for single leg (branch) slings and endless slings shall be two times the vertical rated load (rated capacity).

Observation: The slings used for the downloading of the canister into the HI-STORM storage cask were proof tested to two times the vertical rated load. The two slings purchased at Comanche Peak for downloading the canister had a rated capacity of 50,000 pounds in the vertical setup. The two 50,000 pounds TPSE-EE5000 - 55ft Twin-Path Sparkeater eye & eye slings were each proof loaded to 100,000 pounds (in a vertical setup) by I&I Sling Inc. per Test#101705 on December 14, 2010.

Documents (a) I&I Sling Inc. Test#101705, dated December 14, 2010 (b) American National Reviewed: Standard [American Society of Mechanical Engineers] (ASME) B30.9 Slings, Revision 1984 Category: Slings Topic: Sling Temperature Limits Reference: No Reference Provided N/A Requirement: Synthetic slings shall not be used in contact with objects that exceed the temperature limit of the sling.

Observation: The downloading slings used to lower the canister into the HI-STORM storage cask were designed to withstand up to 300 degrees F. Holtec Purchase Order PS-1211 stated in Step 5.10 that the temperature rating of the slings shall be no less than 300ºF. Each of the two 50,000 pounds TPSE-EE5000 - 55ft Twin-Path Sparkeater slings contained tags stating that the temperature rating of the sling was 300ºF.

Documents (a) Holtec Purchase Specification. PS-1211 "MPC Lift Sling," Revision 5 Reviewed:

Category: Slings Topic: Synthetic Round Sling Removal from Service Reference: ASME B30.9, Section 9-4.8 Revision 1984 Requirement: A synthetic round sling shall be removed from service if any of the following conditions are present: (a) cuts, gouges, badly abraded spots; (b) seriously worn surface fibers or yarns; (c) considerable filament or fiber breakage along the line where adjacent strands meet (light fuzzing is acceptable); (d) particles of broken filament or fibers inside the rope between the strands (inspect inside the rope); (e) discoloration or harshness that may mean chemical damage or excessive exposure to sunlight. Inspect filaments or fibers for weakness or brittleness. ( f ) kinks or hockles; (g) melting or charring on any part of the sling; (h) excessive pitting or corrosion, or cracked, distorted or broken fittings; (i)

other visible damage that causes doubt as to the strength of the sling Observation: The synthetic round sling inspection criterion, consistent with the key elements of ASME B30.9, was listed in Procedure MDA-402 and was used to inspect the slings prior to use.

Procedure MDA-402, Attachment 8.C Prior to Use Inspection of Rigging Equipment Page 124 of 149

required synthetic round slings to be inspected for the following: (1) ensure equipment is marked or tagged with the manufacturers name, model/stock number, size, capacity, and type web material (also type of cover material on round slings), (2) ensure the preventive maintenance inspection due date has not expired, (3) inspect for acid or caustic burns and melting or charring of any part of the sling, (4) inspect for holes, tears, cuts, snags, and broken or worn stitching in load bearing splices, (5) inspect for damage to the cover which has exposed the core yarns, (6) feel along the length of the sling for knots, (7) ensure the tattle-tails are not chemically degraded and that their length meets the criteria described on the tag, and (8) inspect end connections for damage caused by crushing, cracking, deformation, or wear. The riggers provided a demonstration of their inspection of the slings to the NRC inspectors using the criteria in Procedure MDA-402 during the heavy loads dry run demonstration.

Documents (a) Procedure MDA-402 Control of Load Handling Equipment, Revision 11 (b)

Reviewed: American National Standard [American Society of Mechanical Engineers] (ASME)

B30.9 Slings, Revision 1984 Category: Slings Topic: Synthetic Webbing Sling Removal From Service Reference: ASME B30.9, Section 9-5.7 Revision 1984 Requirement: A synthetic webbing sling shall be removed from service if any of the following conditions are present: (a) acid or caustic burns; (b) melting or charring of any part of the sling; (c) holes, tears, cuts, or snags; (d) broken or worn stitching in load bearing splices; (e) excessive abrasive wear; (f) knots in any part of the sling; (g) excessive pitting or corrosion, or cracked, distorted, or broken fittings; (h) other visible damage that causes doubt as to the strength of the sling Observation: The synthetic webbing sling inspection criterion, consistent with the key elements of ASME B30.9, was listed in Procedure MDA-402 and was used to inspect the slings prior to use. Procedure MDA-402, Attachment 8.C Prior to Use Inspection of Rigging Equipment required synthetic web slings to be inspected for the following: (1) ensure equipment is marked or tagged with the manufacturers name, model/stock number, size, capacity, and type web material, (2) ensure PM inspection due date has not expired, (3)

inspect for acid or caustic burns and melting or charring of any part of the sling, (4)

inspect for holes, tears, cuts, snags, and broken or worn stitching in load bearing splices, and (5) inspect end connections for damage caused by crushing, cracking, deformation, or wear. The NRC inspectors observed the riggers performing the sling inspections in accordance with Procedure MDA-402 during the dry run demonstrations for the heavy loads.

Documents (a) Procedure MDA-402 Control of Load Handling Equipment, Revision 11 (b)

Reviewed: American National Standard [American Society of Mechanical Engineers] (ASME)

B30.9 Slings, Revision 1984 Category: Slings Topic: Wire Rope Sling Removal From Service Reference: ASME B30.9, Section 9-2.8.3 Revision 1984 Requirement: A wire rope sling shall be removed from service if any of the following conditions are present: (a) for strand laid and single part slings ten randomly distributed broken wires in Page 125 of 149

one rope lay, or five broken wires in one strand in one rope lay; (b) severe localized abrasion or scraping; (c) kinking, crushing, birdcaging or any other damage resulting in distortion of the rope structure; (d) evidence of heat damage; (e) end attachments that are cracked, deformed, or worn to the extent that the strength of the sling is substantially affected; (f) severe corrosion of the rope or end attachments; Observation: The wire rope inspection criterion, consistent with the key elements of ASME B30.9, was listed in Procedure MDA-402 and was used to inspect the wire ropes prior to use.

Procedure MDA-402, Attachment 8.C Prior to Use Inspection of Rigging Equipment required wire ropes to be inspected for the following: (1) ensure equipment is marked or tagged with the manufacturers name, size, and capacity, (2) ensure preventive maintenance inspection due date has not expired, (3) inspect for cuts, crushing, kinks, broken wire (no more than five randomly distributed broken wires in one rope lay, or two broken wires in one strand in one rope lay), (4) inspect for corrosion or heat damage, and (5) inspect end connections for damage caused by crushing, cracking, deformation, or wear. The riggers provided a demonstration of their inspection of the wire ropes to the NRC inspectors using the criteria in Procedure MDA-402 during the heavy loads dry run demonstration.

Documents (a) Procedure MDA-402 Control of Load Handling Equipment, Revision 11 (b)

Reviewed: American National Standard [American Society of Mechanical Engineers] (ASME)

B30.9 Slings, Revision 1984 Category: Special Lifting Device Topic: Transporter Annual Testing Reference: ANSI N14.6, Sect 7.3.1; Sect 6.3.1 Revision 1993 Requirement: Annually, not to exceed 14 months, the special lifting device shall be subjected to a test load equal to 300% of the maximum service load if a single component failure on the yoke could result in an uncontrolled lowering of the load. If the design for handling the load incorporates a single-failure proof concept, then each path in the dual-load-path device shall be tested to 150% of the load instead of the 300%. After sustaining the load for a period of not less than 10 minutes, critical areas, including major load bearing welds, shall be subject to visual inspection for defects and all components shall be inspected for permanent deformation. In cases where surface cleanliness and conditions permit, the load testing may be omitted and dimensional testing, visual inspection and nondestructive testing of major load-carrying welds and critical areas shall suffice.

Observation: For the transporter, the lifting brackets which connected the HI-STORM storage cask to the transporter lifting beam were considered special lifting devices and were required by Procedure DCS-111 to be inspected annually. For the annual inspection criteria, Procedure DCS-111, Section 8.8 "Annual Recertification Inspection of the HI-STORM Lifting Brackets" described the process for recertifying the two lifting brackets. Results of the recertification were documented on Attachment 10.1.6 "HI-STORM Lift Brackets Recertification Inspection." Recertification was by dimensional testing, visual inspection, and nondestructive testing of the major load-carrying welds and critical areas. The process included disassembling the lifting bracket to inspect the equipment, piece by piece, by use of a visual exam and liquid penetrant or magnetic particle exam.

This procedure was scheduled for completion prior to each campaign or annually if the brackets would be used for long than a one year period.

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Documents (a) Procedure DCS-111 Inspection and Testing of Dry Cask Storage Lifting Devices, Reviewed: Revision 2 (b) Holtec Procedure HSP-335 HI-STORM Lifting Bracket Load and Functional Test, Revision 2 (c) Holtec Document Package DP 0323-019 "Document Package for Lift Brackets for Comanche Peak," Revision 0 (d) E-mail from Frayne Ronkowski, Holtec, Inc. to Craig Montgomery, Luminant Power entitled "NRC Action Item," providing hydraulic test cylinder effective areas, dated August 17, 2011 (e)

American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Transporter Initial Acceptance Testing Reference: ANSI N14.6, Sect 7.3.1; Sect 6.2.1; Sect 6.5 Revision 1993 Requirement: Prior to initial use, the special lifting device shall be subjected to a test load equal to 300% of the maximum service load if a single component failure on the yoke could result in an uncontrolled lowering of the load. If the design for handling the load incorporates a single-failure proof concept, then each path in the dual-load-path device shall be tested to 150% of the load instead of the 300%. After sustaining the load for a period of not less than 10 minutes, critical areas, including load bearing welds, shall be subject to nondestructive testing using liquid penetrant or magnetic particle examination.

Observation: The two brackets attached to the transporter lifting beam to lift and carry the HI-STORM storage cask with a loaded canister to the ISFSI pad were loaded tested to 300% of the load prior to initial use. The load test of 300% was completed on December 15-16, 2010 and consisted of applying a hydraulic pressure to the lifting bracket apparatus, frame and pins to simulate the 300% load. The test was conducted in accordance with Procedure HSP-335 and documented on Exhibit 3.2 "Load and Functional Test Data Record." One bracket was load tested to 3,750 pounds/square inch (psi) and the second to 3,700 psi.

The tests were performed using two hydraulic cylinders with a total effective area of 173.18 square inches. For a pressure of 3,750 psi x 173.18 square inches, the total pressure on the lifting bracket would be 649,425 pounds (324.71 tons). For the second lifting bracket, the total pressure was 3,700 psi x 173.18 square inches = 640,766 pounds (320.38 tons). The lifting brackets were qualified for the 205 ton rated load of the transporter. So each lifting bracket would carry 102.5 tons (205,000 pounds). A load test of 300% on the bracket would require a load of 307.5 tons (615,000 pounds). The test load was held for ten minutes and a visual examination was performed and documented in Procedure HSP-335, Exhibit 3.3 "Pre and Post Load Test NDE Inspection Data." Dye penetrant exams were also performed on the connecting pins and top pin.

No indications were observed on the tested components. The transporter was rated for 205 tons. The transporter was not considered a special lifting device by Holtec. The transporter was load tested per Procedure 36304-05 to approximately 267 tons (130% of the rated capacity), held for ten minutes and visually examined afterwards with no indications.

Documents (a) Holtec Procedure HSP-335 HI-STORM Lifting Bracket Load and Functional Test, Reviewed: Revision 2 (b) Holtec Document Package DP 0323-019 "Document Package for Lift Brackets for Comanche Peak," Revision 0 (c) Morris Material Handling (MMH)

Procedure #36304-05 Holtec Vertical Cask Transporter Factory Acceptance Test Procedure Device Serial # CN-36309, Revision 5 (d) E-mail from Frayne Ronkowski, Holtec, Inc. to Craig Montgomery, Luminant Power entitled "NRC Action Item,"

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providing hydraulic test cylinder effective areas, dated August 17, 2011 (e) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Transporter Inspection - Quarterly Reference: ANSI N14.6, Sect 6.3.7 Revision 1993 Requirement: Special lifting devices shall be visually inspected by maintenance or other non-operating personnel at intervals not to exceed three months in length for indications of damage or deformation.

Observation: The licensee had established a daily/monthly inspection schedule for the transporter and transporter components in Procedures DCS-109 and DCS-111. Procedure DCS-109 provided instructions on performing general inspections of the transporter on a monthly basis, or daily if the transporter was being used for a loading campaign. These inspections included the various components of the transporter structure and the hydraulic and engine components. Procedure DCS-111 included daily/monthly inspection requirements related to several components including the lift slings and the lifting brackets. Section 8.7 "Daily/Monthly HI-STORM Lifting Bracket Inspection" included inspecting all welds for cracks, pits, damage, or deformation. All accessible areas of the lifting brackets were visually inspected for signs of wear, cracks, corrosion, loose fasteners, etc. Any discrepancies required initiation of a condition report.

Documents (a) Procedure DCS-109 "Vertical Cask Transporter Maintenance," Revision 0 (b)

Reviewed: Procedure DCS-111 "Inspection and Testing of Dry Cask Storage Lifting Devices,"

Revision 2 (c) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Transporter Lift Bracket Inspection Prior to Use Reference: ANSI N14.6, Sect 6.3.6 Revision 1993 Requirement: Special Lifting Devices shall be visually inspected by operating personnel for indications of damage prior to each use.

Observation: The requirement to inspect the lifting brackets on the transporter in accordance with ANSI N14.6 requirements had been incorporated into Procedure DCS-201, Steps 6.6 and 8.4.18. Step 6.6 required completion of the inspection of the lift brackets in accordance with Procedure DCS-111. Step 8.4.18 required the inspection of the lifting brackets, prior to use of the transporter, to include a visual exam for physical/structural deformation, abnormal wear, corrosion, loose fasteners, and weld cracking. Procedure DSC-111, Section 8.7 "Daily/Monthly HI-STORM Lifting Bracket Inspection" provided detail instructions for conducting the inspection including a figure showing the key elements of the lifting bracket. The inspection included looking for structural damage, wear, or deformation.

Documents (a) Procedure DCS-111 "Inspection and Testing of Dry Cask Storage Lifting Devices,"

Reviewed: Revision 2 (b) Procedure DCS-201 "Transporting Loaded and Unloaded HI-STORM,"

Revision 2 (c) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Page 128 of 149

Category: Special Lifting Device Topic: Yoke Annual Testing Reference: ANSI N14.6, Sect 7.3.1; Sect 6.3.1 Revision 1993 Requirement: Annually, not to exceed 14 months, the yoke shall be subjected to a test load equal to 300% of the maximum service load if a single component failure on the yoke could result in an uncontrolled lowering of the load. If the design for handling the load incorporates a single-failure proof concept, then each path in the dual-load-path device shall be tested to 150% of the load instead of the 300%. After sustaining the load for a period of not less than 10 minutes, critical areas, including major load bearing welds, shall be subject to visual inspection for defects and all components shall be inspected for permanent deformation. In cases where surface cleanliness and conditions permit, the load testing may be omitted and dimensional testing, visual inspection and nondestructive testing of major load-carrying welds and critical areas shall suffice.

Observation: Annual inspection requirements for the yoke and yoke extension were included in Procedure DCS-112. The initial 300% load test of the yoke and yoke extension was completed on January 7, 2011. Procedure DCS-112, Sections 8.2 through 8.6 included instructions for performing the annual inspections for the lift yoke lifting pin assemblies, lifting sling shackles and pins, yellow lift arm assembly, blue lift arm assembly, and the lift yoke strongback load bearing surfaces. The annual inspection requirement incorporated the ANSI N14.6 option for conducting the testing by disassembling the yoke and yoke extension components and performing visual exams, dimensional measurement of the different parts, and by performing liquid penetrant or magnetic particle exams of major load carrying welds and critical areas. Any issues required issuance of a condition report. This procedure was scheduled for completion prior to each campaign or annually if the yoke and yoke extension will be used for long than a one year period.

Documents (a) Procedure DCS-112 Inspection and Testing of Dry Cask Storage Lifting Devices, Reviewed: Revision 3 (b) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Yoke Initial Acceptance Testing Reference: ANSI N14.6, Sect 7.3.1; Sect 6.2.1; Sect 6.5 Revision 1993 Requirement: Prior to initial use, the yoke shall be subjected to a test load equal to 300% of the maximum service load if a single component failure on the yoke could result in an uncontrolled lowering of the load. If the design for handling the load incorporates a single-failure proof concept, then each path in the dual-load-path device shall be tested to 150% of the load instead of the 300%. After sustaining the load for a period of not less than 10 minutes, critical areas, including load bearing welds, shall be subject to nondestructive testing using liquid penetrant or magnetic particle examination.

Observation: Both the lift yoke and lift yoke extension were test to 300% of the design load. Holtec Document No. HI-2104639 calculated the maximum weight of the HI-TRACK transfer cask used at Comanche Peak to be 244,646 pounds (122.32 tons). The HI-TRACK transfer cask was designed to 250,000 pounds (125 tons). The test load applied to the lift yoke and lift yoke extension per Holtec Documents DOC-1027-702-307R0 and DOC-1027-1117-308R0 was 779, 400 pounds (389.7 tons). After the load was applied, a Page 129 of 149

visual test and liquid penetrant was performed and confirmed that no deformation, distortion or cracking had occurred. The lift yoke extension was tested on December 13, 2010. The lift yoke was tested on January 7, 2011.

Documents (a) Holtec Document No. HI-2104639 "Cask Handling Weight and Cask Handling Reviewed: Dimensions for Comanche Peak," Revision 0 (b) Holtec Document DOC-1027-702-307 R0 "Documentation Package for Lift Yoke for Comanche Peak," dated April 21, 2011; (c) Holtec Document DOC-1027-1117-308 R0 "Documentation Package for Lift Yoke Extension for Comanche Peak," dated April 21, 2011 (d) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Yoke Inspection Prior to Use Reference: ANSI N14.6, Sect 6.3.6 Revision 1993 Requirement: The yoke shall be visually inspected by operating personnel for indications of damage prior to each use.

Observation: The yoke and yoke extension was visually inspected prior to each use. Procedure DCS-203, Step 8.3.3 required inspection of the yoke and yoke extension prior to lifting the HI-TRAC transfer cask loaded with an empty canister from the dry cask pit to the spent fuel pool wet cask pit. The inspection items included inspection for physical/structural deformation, abnormal wear, corrosion, loose fasteners, and weld cracking. After the transfer cask and canister were placed in the wet cask pit lower level, the canister was loaded with spent fuel and the loaded canister moved to the wet cask pit upper shelf.

After completion of this lift, the yoke extension was removed and the yoke connected to the crane hook. An inspection was performed of the yoke per Step 8.6.3. The lid was then attached to the yoke and placed on the loaded canister. After successful positioning of the lid, the yoke was used to lift the loaded transfer cask and canister to the dry cask pit. The same inspection criteria in Step 8.3.3 was listed in Step 8.6.3. Procedure DCS-205 provided instructions for lifting the HI-TRAC transfer cask with the loaded and sealed canister from the dry cask pit onto the top of the HI-STORM storage cask for downloading the canister. Step 8.4.2 required inspection of the lift yoke prior to use.

The yoke inspection criteria matched the criteria in Procedure DCS-203, Steps 8.3.3 and 8.6.3.

Documents (a) Procedure DCS-203 MPC Handling and Fuel Loading Operations, Revision 3 (b)

Reviewed: Procedure DCS-205 Stack-up Transfer of Loaded MPC, Revision 2 (c) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Yoke Records of Annual Testing Reference: ANSI N14.6, Sect 6.3.8 Revision 1993 Requirement: Each yoke shall be tagged or the record system updated after annual testing, or both, indicating the expiration date of the test.

Observation: The yoke and yoke extension record of annual inspection was recorded in Procedure DCS-112. Attachment 10.1.13 "HI-TRAC Lift Yoke Recertification Inspection," and Page 130 of 149

Attachment 10.1.14 "HI-TRAC Lift Yoke Extension Recertification Inspection,"

provided forms to document that each of the required inspections had been completed for the yoke and yoke extension. Procedure DCS-112 was scheduled for completion prior to each campaign or annually if the yoke and yoke extension will be used for long than a one year period.

Documents (a) Procedure DCS-112 Inspection and Testing of Dry Cask Storage Lifting Devices, Reviewed: Revision 3 (b) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Special Lifting Device Topic: Yoke Stress Design -Dual-Load-Path Reference: ANSI N14.6, Sect 4.2.1.1; Sect 4.1.3; Sect 7.2.3 Revision 1993 Requirement: For yokes that are single failure proof by having dual-load-path attachments, the load bearing members of the yoke shall be capable of lifting three (3) times the combined weight of the shipping container plus the weight of the intervening components of the special lifting device, without generating a combined shear stress or maximum tensile stress at any point in the device in excess of the corresponding minimum tensile yield strength of the material of construction. They shall also be capable of lifting five (5)

times the weight without exceeding the ultimate tensile strength of the materials. The dual load-path attachment points on the yoke shall be designed such that each load path will be able to support a static load of three (3) times the weight of the critical load, including intervening components of the lifting device.

Observation: The design of the lift yoke and lift yoke extension incorporated the stress design factors specified by ANSI N14.6. Holtec Documents HI-2104602 and HI-2104603 for the lift yoke and lift yoke extension required the bearing stresses in the equipment to be maintained below 90% of the material yield strength under three times the lifted load.

The calculations showed that the devices would be capable of lifting five times the weight without exceeding the ultimate tensile strength of the materials used. Both the lift yoke and lift yoke extension were load tested to 300% (389.7 tons) on January 7, 2011 and December 13, 2010, respectively, in accordance with Holtec documents DOC-1027-702-307R0 and DOC-1027-1117-308R0.

Documents (a) Holtec Procedure HI-2104602 Structural Analysis of HI-TRAC 125 Ton Lift Yoke Reviewed: for Comanche Peak, Revision 2 (b) Holtec Procedure HI-2104603 Structural Analysis of HI-TRAC 125 Ton Lift Yoke Extension for Comanche Peak, Revision 1 (c) Holtec Document DOC-1027-702-307 R0 "Documentation Package for Lift Yoke for Comanche Peak," dated April 21, 2011; (d) Holtec Document DOC-1027-1117-308 R0

"Documentation Package for Lift Yoke Extension for Comanche Peak," dated April 21, 2011 (e) American National Standard Institute (ANSI) N14.6 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More, Revision 1993 Category: Storage Operations Topic: Cask Spacing Reference: FSAR 1014, Sect 1.4 Revision 9 Requirement: From a thermal standpoint, regardless of the size of the ISFSI, the casks should be arrayed in such a manner that the tributary area for each cask is a minimum of 225 square feet.

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Observation: The cask spacing for Comanche Peaks ISFSI was 271.25 square feet, well above the 225 square feet minimum requirement. The Comanche Peak layout was a 14 x 6 square array (84 total casks). The center to center distance was 17 6 x 15 6 which was a thermal tributary area of 271 square feet. Procedure DCS-201, Attachment 10.1.9 provided a figure showing where each cask was required to be placed in the 14 x 6 array on the ISFSI pad. Prior to placement of the canister on the pad, the cask location was outlined on the pad to ensure correct placement was met.

Documents (a) Procedure DCS-201 Transporting Loaded and Unloaded HI-STORM, Revision 2 Reviewed:

Category: Storage Operations Topic: Heat Transfer Validation Test Reference: CoC 1014, License Condition 9 Amendment 7 Requirement: The air mass flow rate through the cask system will be determined by direct measurements of air velocity in the overpack cooling passages for the first HI-STORM 100 cask system placed into service by any user with a heat load greater than 20 kW.

The velocity will be measured in the annulus formed between the canister shell and the overpack inner shell. An analysis shall be performed that demonstrates the measurements validate the analytical methods and thermal performance predicted by the licensing basis thermal models in Chapter 4 of the FSAR. Letter reports summarizing the results of each thermal validation test shall be submitted to the NRC in accordance with 72.4. Cask users may reference validation test reports submitted to the NRC by other cask users.

Observation: A heat transfer validation test was completed by Southern Company and was applicable to Comanche Peak. The study had been submitted to Holtec for review but had not been submitted to the NRC as of September 1, 2011. Holtec Final Safety Analysis Report (FSAR), Section 8.1.7 "Placement of HI-STORM into Storage," Step 23 provided additional information related to this test. The first cask planned for loading at Comanche Peak was below 20 kW. The second cask will be above 20 kW, and as such will either rely upon the Southern Company test report, if submitted to the NRC, or Comanche Peak will be required to perform their own heat transfer validation test.

Documents (a) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-Reviewed: STORM 100 Cask System," Revision 9 Category: Storage Operations Topic: Overpack Vent Screen Inspections Reference: FSAR 1014, Table 9.2.1 Revision 9 Requirement: The overpack [storage cask] vent screens shall be visually examined for damage monthly Observation: The monthly storage cask vent screen visual inspection for damage, holes, etc was not required for the Comanche Peak ISFSI. The Holtec Final Safety Analysis Report (FSAR), Section 2.3.3.2 Instrumentation, stated that in lieu of performing the periodic inspection of the HI-STORM storage cask vent screens, temperature elements may be installed in two of the storage cask exit vents to continuously monitor the air temperature. Comanche Peak used the temperature monitoring system. The temperature monitoring system readout was required to be checked daily in accordance with Shift Surveillance OPT-102A (Mode 1 thru 6).

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Documents (a) Shift Surveillance OPT-102A-1 Mode 1 and 2 Shiftly Surveillances, Revision 37 Reviewed: (b) Shift Surveillance OPT-102A-3 Mode 3 Shiftly Surveillances, Revision 25 (c) Shift Surveillance OPT-102A-4 Mode 4 Shiftly Surveillances, Revision 22 (d) Shift Surveillance OPT-102A-5 Mode 5 Shiftly Surveillances, Revision 22 (e) Shift Surveillance OPT-102A-6 Mode 6 Shiftly Surveillances, Revision 24 (f) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Storage Operations Topic: Storage Cask Temperature Monitoring Reference: CoC 1014, Tech Spec A.3.1.2 Amendment 7 Requirement: Verify all storage cask inlet and outlet air ducts are free of blockage from solid debris or floodwater every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, OR for storage casks with installed temperature monitoring equipment, verify that the difference between the average storage cask air outlet temperature and ISFSI ambient temperature is less than or equal to 155 degrees F every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for storage casks containing PWR canisters and less than or equal to 137 degree F for storage casks containing BWR canisters.

Observation: Inspection of the storage cask air temperature monitoring system was performed daily.

Shift Surveillance OPT-102A, Forms 1, 3, 4, 5, and 6 (which represent the different modes, i.e. Mode 1, Mode 2, etc.) listed the required surveillance of Overpack Inlet/Outlet or Temperature Differential per Certificate of Compliance SR 3.1.2," as a required surveillance to be performed on the day shift. The acceptance criteria was to verify all storage cask inlets and outlets were free of blockage from solid debris/floodwater or for storage casks with installed temperature monitoring equipment to verify the difference between average storage cask air outlet temperature and ISFSI ambient temperature was less than or equal to 155 degree F.

Documents (a) Shift Surveillance OPT-102A-1 Mode 1 and 2 Shiftly Surveillances, Revision 37 Reviewed: (b) Shift Surveillance OPT-102A-3 Mode 3 Shiftly Surveillances, Revision 25 (c) Shift Surveillance OPT-102A-4 Mode 4 Shiftly Surveillances, Revision 22 (d) Shift Surveillance OPT-102A-5 Mode 5 Shiftly Surveillances, Revision 22 (e) Shift Surveillance OPT-102A-6 Mode 6 Shiftly Surveillances, Revision 24 Category: Unloading Operations Topic: Canister Gas Sampling Reference: FSAR 1014, Section 8.3.3.7 Revision 9 Requirement: During unloading of a cask, gas sampling is performed to assess the condition of the fuel assembly cladding. The gas sample bottle is connected to the vent port RVOA and the RVOA body and sample bottle are evacuated. The vent port cap is then slowly opened using the RVOA, and the gas sample is obtained.

Observation: If a canister was required to be unloaded, instructions for taking a gas sample of the canister prior to cutting the lid off was provided in Procedure DCS-207, Section 8.4 MPC Gas Sampling, and Attachment 10.1.16 MPC Gas Sampling System Configuration. The MPC-32 canister has approximately 6,484 liters of free volume according to the Holtec Final Safety Analysis Report (FSAR), Table 4.4.8 "Summary of MPC Free Volume Calculations." A 1000 cubic centimeter (cc) [1 liter] gas sample bottle was planned for use by Comanche Peak. The sample bottle would be evacuated Page 133 of 149

using a vacuum pump, then connected to the vent removable valve operating assembly (RVOA) in accordance with Procedure DCS-207, Steps 8.4.7 through 8.4.12.

Radiological personnel were notified in Step 8.4.2 prior to the collection of the sample.

A caution in Section 8.4 warned personnel that high radiation levels could be encountered during sample collection. After the valve line-ups were performed on the gas sampling rig, Step 8.4.16 provided for notification of chemistry personnel to take the sample. Procedure DCS-507, Section 6.7 "Gamma Count Survey of MPC Gas Sample" was used for sample collection and analysis of the gas sample by chemistry personnel.

Procedure RPI-627 was used by the radiation protection personnel for their role in monitoring radiation levels and assisting with gas sample collection.

The licensee performed calculations to determine what the estimated dose rate from the sample bottle would be using the MegaShield 3.0 code. For a 1,000 cc sample bottle, the dose rate could be 36 R/hr at 1 inch and 2 R/hr at 1 foot. These values were incorporated into Procedure RPI-627, Attachment 2 "Abnormal Operations," Section 1.6 "MPC Gas Sampling." Section 1.6 provided instructions and cautions for taking the gas sample.

The dose rate of the sample being taken should be limited, preferably to 100 mR/hr.

Continuous monitoring of both beta and gamma dose rates should be performed during sampling. Based on the sample counting results, an evaluation of the condition of the fuel would be made and the canister unloading procedures modified, if necessary, if the fuel was damaged. ALARA job planning for unloading operations required an evaluation of the adequacy of contamination controls, airborne monitoring, skin monitoring, and provisions for radiation surveys. An evaluation was required of the gas in the sample bottle to determine if it could be released to the atmosphere, released through a building HEPA filtering system, or directed to the waste gas system.

Documents (a) Procedure DCS-207 "Unloading a Loaded MPC, Revision 2 (b) Procedure DCS-507 Reviewed: Chemistry Sampling of the Multi-Purpose Canister (MPC)," Revision 0, PCN #1 (c)

Procedure RPI-627 "Job Coverage for Dry Run Fuel Storage," Revision 1 (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Unloading Operations Topic: Canister Reflooding Reference: FSAR 1014, Section 8.3.3.8 Revision 9 Requirement: Reflood the canister slowly with a pressure of less than 90 psi through the drain port until bubbling from the vent line has terminated.

Observation: Reflooding of the canister was performed through the drain port at a pressure of less than 90 psig. Procedure DCS-207, Section 8.5 Reflood System Staging and Section 8.6 MPC Water Reflood described the reflooding process. Attachment 10.1.17 Multipurpose Pump System Set-Up for MPC Reflood provided a drawing of the various components of the systems used for reflooding. Section 8.5 of the procedure provided the initial valve line-ups for the reflooding . Section 8.6 included the opening of the valves and starting the pumps which provided water through the drain port into the canister. Water was injected at less than or equal to 7 gpm per Step 8.6.6 and a pressure of less than 90 psig was maintained. Step 8.6.16 stated that water injection should continue until a solid water stream was observed through the sight glass. Step 8.6.18 directed that water recirculation should continue as directed by the cask loading Page 134 of 149

supervisor. The note immediately above Step 8.6.18 allowed water recirculation to be discontinued once the temperature of the water had stabilized and was changing less than a few degrees every thirty minutes. Step 8.6.25 recorded the exit temperature of the water and determined a new time-to-boil value for the canister.

Documents (a) Procedure DSC 207 "Unloading a Loaded MPC," Revision 2 (b) Holtec Report HI-Reviewed: 2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Unloading Operations Topic: Canister Shell Cooling - High Burnup Fuel Reference: CoC 1014, Tech Spec A.3.1.4 Amendment 7 Requirement: Whenever a canister containing one or more spent fuel assemblies with average burnup values greater than 45 GWD/MTU is inside the transfer cask or the canister heat load exceeds 28.74 kW, the canister shell must be cooled using the Supplemental Cooling System (SCS). The SCS must be placed in operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of completion of canister drying operations during loading, or within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of transferring the canister into the transfer cask during unloading. Once steady state operation is achieved, the SCS may be disabled for up to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to facilitate operational evolutions.

Observation: The use of the supplemental cooling system (SCS) was included in Procedure DCS-204, Section 8.16 Supplemental Cooling System Operations. The first loading campaign at Comanche Peak will not involve high burn-up fuel and as such, the supplemental cooling system will not be used. Procedure DCS-204 had included the operations of the supplemental cooling system, even though it would not be initially used. Step 6.17 of the procedure referenced Form NUC-212-4 Cask Acceptability Report as the source to determine if supplemental cooling was required for the canister being loaded. Page 1 of Form NUC-212-4, stated Supplemental cooling system requirements are applicable when maximum burn-up is greater than 42,750 or total heat greater than 28.74 kW. The statement referenced Technical Specification A.3.1.4 and provided for either a True or a False conclusion, as determined by the TARPIT software. The 42,750 MWD/MTU was used to account for a 5% error in the determination of the burn-up such that the value would be below the 45,000 MWD/MTU requirement in the technical specification. This was the information that was used to complete Procedure DCS-204, Step 6.17.4 to identify that supplemental cooling was required. Validation Test Report NUC006 had verified that the TARPIT software would correctly identify when supplemental cooling was required based on several data sets entered with values above and below the limits.

If the supplemental cooling system was required, Step 6.20.1 required the initiation of Procedure DCS-102 DCS Ancillary Equipment Pre-Use Preparations. Section 8.16 of Procedure DCS-204 provided the steps for operating the system. Several attachments in Procedure DCS-204 related to the supplemental cooling system including Attachment 10.1.16 Initial Set-Up and Functional Test of the Supplemental Cooling System, Attachment 10.1.17 Supplemental Cooling System Schematic, and Attachment 10.1.18 Supplemental Cooling System Operability Surveillance. The supplemental cooling system was demonstrated as part of the dry run activities on May 31, 2011.

Section 8.16 of Procedure DCS-204 included a reference to Technical Specification A.3.1.4 of Certificate of Compliance 1014. The Caution at the first of Section 8.16 Page 135 of 149

stated that the supplemental cooling system must be started within four hours of completing canister drying operations during loading. Attachment 10.1.18 provided instructions to verify the supplemental cooling system was operating in accordance with Technical Specification A.3.1.4.1. This attachment included in Section 5.0 Precautions, Limitations, and Notes, referenced the requirement to start the supplemental cooling system within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of completing canister drying if the canister contained one or more fuel assemblies with an average burn-up greater than 45,000 MWD/MTU or the canister heat load exceeds 28.74 kW. Attachment 10.1.18 provided a note in Section 5.1.1 that once steady state operations was reached, the supplemental cooling system could be disabled for up to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Steady state operations was defined in Step 5.1.2 as the air cooler inlet temperature change was less than 2 degree F every 15 minutes and the water flow in the HI-TRAC annulus was continuous. Steady state operations was confirmed every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

For the unloading process, Procedure DCS-206 provided instructions for transporting the canister from the ISFSI back to the fuel building. This procedure included instructions for the use of the supplemental cooling system in Section 2.1 and Section 8.7. The instructions included the criteria to use the supplemental cooling system if the average burn-up of one or more fuel assemblies exceeds 45,000 MWD/MTU or the cask heat load exceeds 28.74 kW. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requirement to start cooling after placing the canister in the HI-TRAC transfer cask and the 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> limit on temporarily disabling the system after steady state operations was reached were both included in the procedure.

Procedure DCS-207, which provides instructions for unloading the canister after it has been returned to the fuel building, also provided instructions related to the supplemental cooling system. Section 5.28 SCS Operations included the requirement to start the supplemental cooling system operations within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of placing the canister in the HI-TRAC transfer cask if the canister contains one or more fuel assemblies with an average burn-up greater than 45,000 MWD/MTU or the decay heat load exceeded 28.74 kW Documents (a) Procedure DCS-204 "MPC Closure Operations (Sealing, Drying, and Backfilling),"

Reviewed: Revision 2 (b) Procedure DCS-206 "Transporting and Transferring a Loaded MPC for Unloading," Revision 1 (c) Procedure DCS-207 "Unloading a Loaded MPC," Revision 2 (d) Procedure NUC-212 Spent Fuel Limits for Dry Cask Operations, Revision 1 (e)

Document NUC006 Validation Test Plan for TARPIT, Revision 0 Category: Unloading Operations Topic: Cavity Reflooding Reference: CoC 1014, Tech Spec A.3.1.3 Amendment 7 Requirement: Prior to and during reflooding of the canister to unload the spent fuel, the canister cavity pressure shall be maintained less than 100 psig. If cavity pressure exceeds this level, reflooding is to be stopped immediately until the pressure is within limits and the vent port is checked to verify it is not closed or blocked.

Observation: Procedure DCS-207, Section 8.6 MPC Water Reflooding incorporated the 100 psig limit in a caution statement at the first of the section. Based on the gauge uncertainty and conservatism, an administrative limit of 90 psig was established as the limit. Step 8.6.6 of the procedure assigned an individual to continuously monitor the reflooding operation to maintain the pressure on the vent port pressure gauge to less than 90 psig Page 136 of 149

and to maintain the reflood rate at less than or equal to 7 gpm on the drain port flow rate meter. Step 8.6.7 stated that if the pressure increased above 90 psig, to immediately stop the reflooding operation until the pressure dropped and to ensure the vent port was not closed or blocked.

Documents (a) Procedure DSC 207 "Unloading a Loaded MPC," Revision 2 Reviewed:

Category: Unloading Operations Topic: Hydrogen Monitoring Reference: FSAR 1014, Section 3.4.1 Revision 9 Requirement: To preclude the potential for hydrogen ignition during lid cutting, operating procedures require monitoring for combustible gas and purging the space beneath the canister lid with an inert gas.

Observation: Procedure DCS-207, Step 8.7.19 required hydrogen monitoring on the vent port prior to cutting the lid off. A caution statement in Step 8.7.19 required the discontinuance of cutting operations if any combustible gas readings exceeded 50% of the lower explosive limit (LEL), which would be 2% hydrogen, and to continue inert gas purging until the hydrogen level dropped below 25% of the LEL (1% hydrogen).

Documents (a) Procedure DCS 207 " Unloading a Loaded MPC," Revision 2 Reviewed:

Category: Welding Topic: Combustible Gas Monitoring Reference: CoC 1014, Appendix B, Section 3.8 Amendment 7 Requirement: During canister lid-to-shell welding operations, combustible gas monitoring of the space under the lid is required to ensure that there is no combustible gas mixture present in the welding area.

Observation: Combustible gas monitoring was required during the lid to shell welding activities performed on the canister lid. Procedure HSP-504, Steps 6.3.5 thru 6.3.7 required the use of two independent hydrogen detectors. One hand held hydrogen monitor (Hy-Alerta-500) was required to be used every ten minutes during the initial root weld per Step 6.3.7 and values recorded in Attachment 9.18 "Combustible Gas Monitoring and Argon Purge Log". The second detector, an in-line hydrogen monitor (Hy-Optima 1700)

was required to run continuously during all welding passes, from root to final pass per Step 6.3.7.4. Step 6.3.5 discussed the required hydrogen purge of the hydrogen monitor prior to use. The Hy-Optima 1700 model in-line hydrogen monitor required a one hour purge in a mixture of 5% hydrogen and 95% nitrogen prior to use. Procedure HSP-504, Step 4.9 required welding to stop immediately if the monitored hydrogen gas concentration exceeded 2% hydrogen, which is 50% the lower explosive limit for hydrogen.

Documents (a) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC and Reviewed: MPC Lid," Revision 6

.

Page 137 of 149

Category: Welding Topic: GTAW Essential Variables Reference: ASME Section IX, Part QW-256 Code Year 1995 Requirement: The welding procedure specification for gas tungsten arc welding (GTAW) shall describe the following essential variables: (1) base metal thickness range, qualified; (2)

base metal P number; (3) filler metal F number; (4) filler metal A number; (5) filler metal; (6) filler metal product form (flux, metal, powder); (7) maximum weld deposit thickness; (8) minimum preheat temperature; (9) post-weld heat treatment (PWHT)

conditions; (10) post-weld heat treatment (PWHT) thickness; (11) shielding gas mixture; (12) trailing shielding gas mixture and flow rate; and (13) technique change from closed to out chamber Observation: Welding Procedure Specifications WPS-13, WPS-41, WPS-47L, WPS-246LW contained the gas tungsten arc welding (GTAW) essential variables identified in ASME Section IX, Part QW-256 applicable to the canister used at Comanche Peak. These variables were reviewed with the welders during the welding dry run on May 2-5, 2011.

Documents (a) Holtec Welding Procedure Specifications WPS-13 "Manual GTAW," Revision 0 (b)

Reviewed: Holtec Welding Procedure Specifications WPS-41 "Manual GTAW," Revision 0 (c)

Holtec Welding Procedure Specifications WPS-47L "Manual GTAW," Revision 0 (d)

Holtec Welding Procedure Specifications WPS-246LW "Machine GTAW Hotwire,"

Revision 1 Category: Welding Topic: GTAW Non Essential Variables (1-12)

Reference: ASME Section IX, Part QW-256 Code Year 1995 Requirement: The welding procedure specification for gas tungsten arc welding (GTAW) must describe the following non-essential variables: (1) joint groove design; (2) backing; (3)

root spacing; (4) retainers; (5) filler metal size; (6) consumable inserts; (7) filler metal AWS classification number; (8) welding positions; (9) welding progression; (10)

trailing shielding gas composition; (11) gas flow rate; and (12) backing flow; Observation: Welding Procedure Specifications WPS-13, WPS-41, WPS-47L, and WPS-246LW contained the gas tungsten arc welding (GTAW) non-essential variables identified in ASME Section IX, Part QW-256 applicable to the canister used at Comanche Peak.

These variables were reviewed with the welders during the welding dry run on May 2-5, 2011.

Documents (a) Holtec Welding Procedure Specifications WPS-13 "Manual GTAW," Revision 0 (b)

Reviewed: Holtec Welding Procedure Specifications WPS-41 "Manual GTAW," Revision 0 (c)

Holtec Welding Procedure Specifications WPS-47L "Manual GTAW," Revision 0 (d)

Holtec Welding Procedure Specifications WPS-246LW "Machine GTAW Hotwire,"

Revision 1 Category: Welding Topic: GTAW Non Essential Variables (13-26)

Reference: ASME Section IX, Part QW-256 Code Year 1995 Requirement: The welding procedure specification for Gas Tungsten Arc Welding (GTAW) must also describe the following non-essential variables: (13) electrical pulsing; (14) electrical current or polarity; (15) amperage and voltage range; (16) tungsten size; (17) string or Page 138 of 149

weave bead; (18) orifice or gas cup size; (19) method of initial and interpass cleaning; (20) method of back gouging; (21) oscillation width; (22) multiple or single pass per side; (23) multiple or single electrodes; (24) electrode spacing; (25) travel mode and speed; and (26) peening.

Observation: Welding Procedure Specifications WPS-13, WPS-41, WPS-47L, and WPS-246LW contained the gas tungsten arc welding (GTAW) non-essential variables identified in ASME Section IX, Part QW-256 applicable to the canister used at Comanche Peak.

These variables were reviewed with the welders during the welding dry run on May 2-5, 2011.

Documents (a) Holtec Welding Procedure Specifications WPS-13 "Manual GTAW," Revision 0 (b)

Reviewed: Holtec Welding Procedure Specifications WPS-41 "Manual GTAW," Revision 0 (c)

Holtec Welding Procedure Specifications WPS-47L "Manual GTAW," Revision 0 (d)

Holtec Welding Procedure Specifications WPS-246LW "Machine GTAW Hotwire,"

Revision 1 Category: Welding Topic: GTAW Supplementary Essential Variables Reference: ASME Section IX, Part QW-256 Code Year 1995 Requirement: The welding procedure specification for gas tungsten arc welding (GTAW) must describe the following supplementary essential variables, when required: (1) base metal group number; (2) base metal thickness range; (3) AWS classification; (4) welding positions; (5) maximum interpass temperature; (6) PWHT conditions; (7) maximum heat input; (8) current type and polarity; (9) multiple or single pass per side; and (10)

multiple or single electrodes.

Observation: Welding Procedure Specifications WPS-13, WPS-41, WPS-47L, AND WPS-246LW contained the gas tungsten arc welding (GTAW) supplementary essential variables identified in ASME Section IX, part QW-256 applicable to the canister used at Comanche Peak. These variables were reviewed with the welders during the welding dry run on May 2-5, 2011.

Documents (a) Holtec Welding Procedure Specifications WPS-13 "Manual GTAW," Revision 0 (b)

Reviewed: Holtec Welding Procedure Specifications WPS-41 "Manual GTAW," Revision 0 (c)

Holtec Welding Procedure Specifications WPS-47L "Manual GTAW," Revision 0 (d)

Holtec Welding Procedure Specifications WPS-246LW "Machine GTAW Hotwire,"

Revision 1 Category: Welding Topic: Material Specifications Reference: 10 CFR 72.154 Published 2011 Requirement: The licensee shall establish measures to ensure that purchased material, equipment, and services conform to procurement documents. These measures must include provisions for source evaluation and selection, objective evidence of quality furnished by the contractor/subcontractor, inspection at the contractor/subcontractor source and examination of product on delivery. Records shall be available for the life of the ISFSI.

The effectiveness of the control of quality by contractors/subcontractors shall be assessed at intervals consistent with the importance, complexity and quantity of the Page 139 of 149

product or service.

Observation: Records were reviewed for the weld wire and filler material and were found to conform to the procurement documents. The licensee's quality assurance program related to materials specifications had been adequately applied to the purchase. The Holtec Final Safety Analysis Report (FSAR), Section 3.3.1.4 "Weld Material" required the minimum tensile strength of the weld wire and filler metal to be equal to or greater than the tensile strength of the base metal listed in the ASME Code. Purchase Order # S07076056S2 purchased three different filler rods and one weld wire spool. All four were E308L metal type with a minimum tensile strength of 75 ksi. The base metal of the canister was stainless steel of miscellaneous types: 304, 304L, 316, 316L. All base metals types had a minimum tensile strength of 75 ksi.

Documents (a) Purchase Order # S07076056S2, dated May 17, 2011 (b) Holtec Report HI-2002444 Reviewed: "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System,"

Revision 9 Category: Welding Topic: Minimum Delta Ferrite Content Reference: ASME Section III, Article NB-2433; Reg Guide 1.31 Code Year 1995 Requirement: A delta ferrite determination must be made for A-No.8 consumable inserts, bare electrode, rod, or wire filler metal. Exceptions: 1) A-No.8 metal used for weld metal cladding; 2) SFA-5.4 and SFA-5.9 metal; 3) Type 16-8-2 metal. The minimum acceptable delta ferrite content is 5 FN and it must be stated in the certification records.

Observation: All of the weld filler metals purchased and used at Comanche Peak conformed to a delta ferrite number greater than 5. Four different types of filler metals were used in the welding operations. The 1/16" diameter Arcaloy ER308L filler metal had a ferrite number of 8, which was documented on Certified Materials Test Report (CMTR) Order

1. 772007. The 1/8" diameter Lincoln ER308L filler metal had a ferrite number of 9, which was documented on CMTR Customer PO #112140644. The 0.045 MIG 308LSI spool filler metal had a ferrite number of 10, which was documented on CMTR Customer PO # 112130773. The 3/32" Arcaloy ER308L filler metal had a ferrite number of 10, which was documented on CMTR Order # 704081.

Documents (a) Certified Material Test Report, Order # 772007 from ESAB Group Inc., dated March Reviewed: 9, 2011 (b) Certified Material Test Report, Customer PO # 112140644 from The Lincoln Electric Company, dated November 15, 2010 (c) Certified Material Test Report, Customer PO # 112130773 from The Lincoln Electric Company, dated September 16, 2010 (d) Certified Material Test Report, Order # 704081 from ESAB Group Inc., dated August 10, 2010 Category: Welding Topic: NDE Inspection Documentation Reference: FSAR 1014, Section 9.1.1.4 Revision 9 Requirement: Weld inspections shall be detailed in a weld inspection plan which shall identify the weld and the examination requirements, the sequence of examination, and the acceptance criteria.

Observation: The required information was incorporated into Procedures HSP-506 for penetrant Page 140 of 149

examinations and Procedure HSP-507 for visual examination. Procedure HSP-506, Section 1.0 "Scope" stated that the penetrant examination procedure was applicable to the closure welds for the canister and base materials. The shapes of the welds covered by the procedure included groove welds, fillet welds, and base metal repair welds consisting of an acceptable weld joint. The base metal was stainless steel. The procedure provided the examination requirements and sequence of examination including a list of acceptable penetrant materials; weld cleaning requirements; penetrant and developer application methods, sequence, and limitations; requirements for conducting the examinations; acceptance criteria; post cleaning requirements; and requirements for documentation.

Procedure HSP-507 provided instructions for visual examinations of the welds. Section 1.0 "Scope" stated that the procedure provided visual weld inspection criteria for the Holtec canister lid welding. Section 6.0 "Procedure" stated that visual weld examinations could be done either directly or by remote visual examination. The procedure included instructions to visually examine tack welds if they become part of the finished weld. Cleanliness requirements, acceptance criteria for the welds, and documentation requirements were specified. Step 6.4 required the quality control inspector to verify the location, configuration, size, and length of the weld in accordance with the approved Holtec fabrication drawing. Direct visual examinations required the distance between the eye and the surface being examined to not exceed 24" and a minimum of 30 degree. Magnifying lenses were allowed. For remote viewing, telescopes, borescopes, fiber optics, cameras, and other suitable instruments could be used. A flashlight or other auxiliary lighting was required to maintain a minimum of 100-foot candles at the weld location under examination.

Documents (a) Holtec Procedure HSP-506 "Liquid Penetrant Examination for MPC Closure Reviewed: Welding," Revision 2 (b) Holtec Procedure HSP-507 "Visual Weld Examination for MPC Field Closure Welding," Revision 2 Category: Welding Topic: Procedure Qualification Record (PQR)

Reference: ASME Section IX, Part QW-200.2 Code Year 1995 Requirement: Each manufacturer or contractor shall prepare a Procedure Qualification Record (PQR)

for each procedure. The completed PQR shall document all essential and, when required, all supplementary essential variables of QW-250 through QW-280 for each welding process used during the welding of the test coupon. Non essential variables may be documented at the contractor's option. The PQR shall be certified accurate by the manufacturer or contractor.

Observation: The procedure qualification records (PQR-4A, PQR-4B, PQR-4C, PQR-1C, PQR-1062, and PQR-1146) for Weld Procedure Specifications (WPS) 47LW and 246LW listed the proper essential, supplementary, and non-essential variables. The ASME code for gas metal arc welding (GTAW) was contained in ASME Code Section IX, Part QW-256 and QW-256.1 "Welding Variables Procedure Specifications (WPS) for Gas Tungsten Arc Welding." Holtec certified the procedure qualification records as accurate on May 4, 2011. During the period from 1983 to 2001, the listed PQRs were prepared, welded, tested and certified in accordance with ASME Section IX, Part QW-200.2 by US Tool and Die. Holtec acquired US Tool and Die in 2006, and as such, became the current Page 141 of 149

owners of the procedure qualification records. During the welding dry run on May 2, 2011, the procedure qualification records were found to not be in compliance with this ASME code requirement because they had not been certified by Holtec (i.e. the current owner). The issue of US Tool and Die ownership of the procedure qualification records versus Holtec's ownership was documented in Condition Report CR-2011-5573. Holtec took immediate action to resolve this issue and presented documentation dated May 4, 2011, showing that the current Holtec welding engineer had reviewed and accepted the procedure qualification records being used at Comanche Peak.

Documents (a) Procedure Qualification Record PQR-4A, "Manual GTAW on P-No.8 Stainless Reviewed: Steel," Revision 1 (b) Procedure Qualification Record PQR-4B "Manual GTAW on P-No.8 Stainless Steel," Revision 1 (c) Procedure Qualification Record PQR-4C "Manual GTAW on P-No.8 Stainless Steel," Revision 1 (d) Procedure Qualification Record PQR-1C "Manual GTAW on P-No.8 Stainless Steel," Revision 1 (e) Procedure Qualification Record PQR-1062 "Manual GTAW on P-No.8 Stainless Steel," Revision 2 (f) Procedure Qualification Record PQR-1146 "Machine Hot Wire GTAW on P-No. 8 Stainless Steel,"

Revision 1 (f) Condition Report CR-2011-5573 "NRC Observations During Welding Dry Runs," initiated May 3, 2011 Category: Welding Topic: Procedure Qualification Tests Reference: ASME Section III, Article NB-4331 Code Year 1995 Requirement: All welding procedure qualification tests shall be in accordance with the requirements of Section IX. ASME Section IX Article II QW-202.2 (b) requires partial penetration groove welds to be qualified in accordance with the requirements of QW-451. QW-202.2 (c) states that welding procedure specification (WPS) qualification for fillet welds may be made on groove-weld test coupons using test specimens specified in (b) above.

Observation: The procedure qualification record (PQR) test coupons, which qualified Welding Procedure Specifications WPS-47LW and WPS-246LW, all satisfactorily passed the required tests per Table QW-451.1 "Groove -Weld Tension Tests and Transverse-Bend Tests." Holtec Procedure HSP-504, Steps 6.1.2 and 6.1.3 identified Welding Procedure Specifications WPS-47LW as the process for performing manual welding on the canister lid and Welding Procedure Specifications WPS-246LW as the process for performing the machine automated welding. The welds associated with the canister lid welding were all partial penetrant welds except for one fillet weld for the closure ring to shell weld. The partial penetrant welds and the fillet weld were qualified using the required tests per Table QW-451.1 of ASME Section IX Article II. The Procedure Qualification Record PQR-1C was perform on test coupons that were 1/4 inch thick. Table QW-451.1 required two tension tests, two face bend tests, and two root bend tests for procedure qualification records with a metal thickness of 1/16 inch but less than 3/8 inch.

Procedure Qualification Record PQR-1C had acceptable results on all six tests. The Procedure Qualification Records PQR-4A , 4B, and 4C were performed on test coupons that were 0.015 inch, 0.025 inch, and 0.035 inch thick, respectively. Table QW-451.1 required two tension tests, two face bend tests, and two root bend tests for procedure qualification record with metal thickness less that 1/16 inch. Each procedure qualification record (PQR-4A, PQR-4B, and PQR-4C) had acceptable results on all six tests. Procedure Qualification Record PQR-1062 and PQR-1162 were performed on test coupons that were 3/4 inch thick. Table QW-451.1 required two tension tests and four Page 142 of 149

side bend tests for procedure qualification records with a metal thickness of 3/4 inch to less than 1-1/2 inch. Each procedure qualification record (PQR-1062 and PQR-1162)

had acceptable results on all six tests. Weld Procedure Specification WPS-47LW was qualified by Procedure Qualification Records PQRs-1C, 4A, 4B, and 1062. Weld Procedure Specification WPS-246LW was qualified by Procedure Qualification Record PQR-1162.

Documents (a) Procedure Qualification Record PQR-4A, "Manual GTAW on P-No.8 Stainless Reviewed: Steel," Revision 1 (b) Procedure Qualification Record PQR-4B "Manual GTAW on P-No.8 Stainless Steel," Revision 1 (c) Procedure Qualification Record PQR-4C "Manual GTAW on P-No.8 Stainless Steel," Revision 1 (d) Procedure Qualification Record PQR-1C "Manual GTAW on P-No.8 Stainless Steel," Revision 1 (e) Procedure Qualification Record PQR-1062 "Manual GTAW on P-No.8 Stainless Steel," Revision 2 (f) Procedure Qualification Record PQR-1146 "Machine Hot Wire GTAW on P-No. 8 Stainless Steel,"

Revision 1 (g) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on MPC and MPC Lid," Revision 6 Category: Welding Topic: Tack Welds Reference: ASME Section III, Article NB-4231.1 Code Year 1995 Requirement: Tack welds used to secure alignment shall either be removed completely when they have served their purpose, or their stopping and starting ends shall be properly prepared by grinding or other suitable means so that they may be satisfactorily incorporated into the final weld. When tack welds are to become part of the finished weld, they shall be visually examined and defective tack welds shall be removed.

Observation: Weld Procedure Specification WPS-47LW contained the essential welding instructions for setup of the welding equipment to perform the tack welding. Grinding of the start and stop points of the tack welds for inclusion into the root weld and visual examination of the tack welds was required for all tack welds by Procedure HSP-504. Step 6.3.13.3 through Step 6.3.13.5 included the tack welding of the shims, grinding the ends of the tack welds, and performing a visual inspection of the tack welds. Step 6.3.18 provided instructions for tack welding the lid to the shell. This included grinding the start and stop points of all tack welds for inclusion into the root weld. Step 6.3.19 required a visual inspection of the tack welds. The vent port cover and drain port cover tack welds were specified in Steps 6.7.3 and 6.8.3 respectively and required four tack welds at 90 degree from each other. Grinding of the start and stop points of the tack welds were required for inclusion into the root weld. Steps 6.7.4 and 6.8.4 required visual examinations of the tack welds. Step 6.11.4 provided instructions for tack welding the closure rings and included the requirement to grind the start and stop points on all tack welds for inclusion into the root weld. Step 6.11.5 required a visual examination of the tack welds on the closure rings. Tack welds could be applied either manually of by the automatic welding machine. Proper welding and examination of tack welds was demonstrated during the welding dry run May 2-5, 2011.

Documents (a) Holtec Weld Procedure Specification WPS-47LW "Manual GTAW for MPC Closure Reviewed: Welds - Tacks," Revision 0 (b) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC," Revision 6 Page 143 of 149

Category: Welding Topic: Vent and Drain Port Cover Plate Weld PT Reference: CoC 1014, Appendix B, Table 3-1 Amendment 7 Requirement: A liquid penetrant (PT) examination is required on the root (if more than one weld pass is required) and the final pass on the vent and drain port cover plate welds. The PT examination shall be performed in accordance with NB-5245.

Observation: Liquid penetrant examinations of the root weld and final weld on the vent and port cover plates was required in Procedure HSP-504, Section 6.0 "Procedure, Instructions and Definitions." A liquid penetrant examination was required by Step 6.7.7 for the root pass of the lid vent port cover plate. Step 6.7.10 required a liquid penetrant examination after the final pass on the vent port cover plate. Step 6.8.7 required a liquid penetrant examination after the root pass on the drain port cover plate. Step 6.8.10 required a liquid penetrant examination after the final pass on the drain port cover plate. The weld for the vent and drain port cover plates was a partial penetration groove weld, per Table 7.1.2 of the Holtec Final Safety Analysis Report (FSAR). ASME Section III, Division I -

NB-5245 "Partial Penetrant Welded Joints" stated that partial penetrant weld joints shall be examined progressively using either magnetic particle or liquid penetrant and that the surface of the finished welded joint be examined by either method. This requirement was met in Procedure HSP-504, Section 6.0 using liquid penetrant examinations.

Requirement NB-5245 also required that the increments of examination shall be the lesser of 1/2 of the maximum welded joint dimensions measured parallel to the center line of the connection or 1/2". Procedure HSP-504, Step 4.17 required liquid penetrant examination for each approximately 3/8" of weld depth. This was not necessary on the port cover plates since only a root pass and final weld were required to complete the welding of the covers.

Documents (a) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC,"

Reviewed: Revision 6 (b) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR)

for the Hi-STORM 100 Cask System," Revision 9 Category: Welding Topic: Weld Grinding and Machining Reference: FSAR 1014, Section 9.1.1.8 Revision 9 Requirement: Grinding and machining operations on the canister confinement boundary shall be controlled through written and approved procedures and quality assurance oversight to ensure grinding and machining operations do not reduce base metal wall thicknesses of the confinement boundary beyond that allowed per the design drawings.

Observation: Caution related to grinding into the base metal was included in Procedure HSP-504.

Step 4.13 stated "Whenever the welder is grinding to address workmanship of the weld, care shall be taken to not contact the base metal. If such an incident occurs, the welder is to notify the Site Welding Supervision and the Holtec QC immediately so that the area can be properly inspected for minimum wall thickness and addressed accordingly." Step 4.14 stated "Visual inspections by Holtec QC of all welds shall include an inspection for any evidence of inadvertent grinding or arc strikes into the base metal and appropriate follow-up inspections for base metal minimum wall thickness." Step 4.16 stated "All grinding activities on the canister shall be controlled via Holtec Procedure HSP-510."

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Documents (a) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC and Reviewed: MPC Lid," Revision 6 (b) Holtec Procedure HSP-510 "Grinding Control Procedure for Site MPC Welding Operations," Revision 0 Category: Welding Topic: Weld Repairs - Base Metal Defects Reference: ASME Section III, Article NB-4132 Code Year 1995 Requirement: Weld repairs exceeding in depth the lesser of 3/8 inch (10 mm) or 10 percent of the section thickness, shall be documented on a report which shall include a chart which shows the location and size of the prepared cavity, the welding material identification, the welding procedure, the heat treatment, and the examination results of the weld repair.

Observation: Procedure HSP-513 incorporated the weld repair requirements specified in ASME Section III, Article NB-4132. Procedure HSP-513, Step 5.2.5 required repairs exceeding in depth, the lesser of 3/8 inch, or 10% of the section thickness shall include a map, documented in Exhibit 9.1 "Repair Excavation Map," which showed the location and size of the prepared cavity. Section 8.0 of the procedure required that a field condition report or field non-conformance report be written for any repairs that meet the criteria of Step 5.2.5.

Documents (a) Holtec Procedure HSP-513 "Base Metal Repair Procedure for MPC Field Closure Reviewed: Welding," Revision 1 Category: Welding Topic: Weld Repairs - Surface Defects Reference: ASME Section III, Article NB-4452; NB-2538.c Code Year 1995 Requirement: Surface defects may be removed by grinding or machining without weldout provided the minimum section thickness is maintained, the depression is blended and liquid penetrant testing or magnetic particle testing is performed to ensure the defect is removed. Areas ground to remove oxide scale or other mechanically caused impressions for appearance or to facilitate proper ultrasonic testing need not be examined by the magnetic particle or liquid penetrant test method.

Observation: The weld repair requirements for surface defects were incorporated into the welding procedure consistent with the requirements of ASME Section III, Article NB-4452, NB-2538.c. Procedure HSP-508, Section 5.0 Rework and Repair of Weld Metal Defects stated in Steps 5.1.1.1 thru 5.1.1.3 that the minimum thickness of the weld was to be maintained, the depressions were to be blended, and liquid penetrant examination performed after blending to ensure the defect had been removed.

Documents (a) Holtec Procedure HSP-508 Repair of Deposited Weld Material for MPC Field Reviewed: Closure Welding, Revision 1 Category: Welding Topic: Weld Types for Canister Lid Reference: FSAR 1014, Table 7.1.2 Revision 9 Requirement: The canister closure welds on the canister lid shall be the following types: (a) canister lid to shell - partial penetration groove (b) vent and drain port cover plates - partial penetration groove (c) closure ring to shell - fillet (d) closure ring to closure ring radial -

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partial penetration groove (e) closure ring to lid - partial penetration groove.

Observation: Holtec Fabrication Drawings 8216 had incorporated the correct weld types as listed in the Holtec Final Safety Analysis Report (FSAR), Table 7.1.2 for the canister lid closure welds. Procedure HSP-504, Attachment 9.5 "MPC Lid Welding Production Work Routing Plan (PWRP)" referenced Drawing 8216 as the applicable drawing for performing the welding on the canister lid. The closure ring to shell weld was a fillet weld. All other welds were partial penetration groove welds. The closure ring to closure ring radial and the vent and drain ports cover plates were manually welded. All other welds were performed using the automatic gas tungsten arc welding machine. Welding personnel performing the weld demonstration May 2-5, 2011 were familiar with the welding requirements in the fabrication drawings and knew the correct welds to perform on the canister lid. During the welding demonstration, an error was found in Procedure HSP-504. Attachment 9.15 "Weld Map of MPC Closure Ring" listed double groove field weld symbols instead of single groove field welds. This error was documented in Condition Report CR-2011-5573 and the drawing corrected in the next revision.

Documents (a) Holtec Procedure HSP-504 "Procedure to Perform Closure Welds on the MPC and Reviewed: MPC Lid," Revision 7 (b) Holtec Fabrication Drawing 8216 "MPC Field Welding Details for MPC-24, 32, 24E, and 68," Revision 0 (c) Condition Report CR-2011-5573

"NRC Observations During Welding Dry Runs," initiated May 3, 2011 (d) Holtec Report HI-2002444 "Holtec Final Safety Analysis Report (FSAR) for the Hi-STORM 100 Cask System," Revision 9 Category: Welding Topic: Welder Performance Qualification Test Reference: ASME Section IX, Part QW-301.2 Code Year 1995 Requirement: The welder performance qualification test shall be welded in accordance with a qualified welding procedure specification (WPS), unless preheat or post weld heat treatment is specified.

Observation: The welder performance qualification (WPQ) test records for the Holtec welders that participated in the May 2-5, 2011 welding dry run demonstrations and were scheduled to participate in the first canister loading, were reviewed. The records demonstrated that the welders had met the qualification testing requirements for manual welding on the canister lid. Manual welding on the canister lid involved the closure ring seam weld and the vent and drain port cover plates. Procedure HSP-504, Attachment 9.8 "Vent Port Cover Plate Welding Production Work Routing Plan," Attachment 9.10 "Drain Port Cover Plate Welding Production Work Routing Plan," Attachment 9.12 "Port Cover Plate Plug Welding Production Work Routing Plan," and Attachment 9.14 "MPC Closure Ring Welding Production Work Routing Plan" all specified welding per manual Welding Procedure Specification WPS-47LW. No preheat or post weld heat treatment was specified in Procedure HSP-504. Review of the welder performance qualification test records for the Holtec welders found that none of the five welders had qualified directly to Welding Procedure Specification WPS-47LW. Procedure HQP 9.2, Step 6.1.2 stated that a welder or welding operator who successfully passed a welding test in accordance with a weld procedure specification is qualified to use that process with the same weld procedure specification and any other weld procedure specification using the same process within the limits of the essential variables described in ASME Code Section IX.

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Documentation was provided showing that the welders had qualified to the essential variables for gas tungsten arc welding as listed in Welding Procedure Specification WPS-13, WPS-47, WPS-93 and WPS-93B. These welding procedure specifications contained equivalent essential variables as specified in Welding Procedure Qualification WPS-47LW. During the review of the records, it was noted that several of the welder's qualification records were approved by their former employer, U. S. Tool and Die, and not Holtec Manufacturing as required by ASME Section IX, QW-301.2. Holtec purchased U.S. Tool and Die in 2006. On May 4, 2011, the Holtec welding engineer reviewed and accepted the U.S. Tool and Die welder performance qualification test records as Holtec Manufacturing records.

Documents (a) Holtec Procedure HQP-9.2 "Welder Qualification Requirements," Revision 6 (b)

Reviewed: Holtec Welding Procedure Specification WPS-47LW "Manual GTAW on MPC Closure Welds," Revision 0 (c) Holtec Welding Procedure Specification 93B "Gas Tungsten Arc Welding," Revision 10 (d) Holtec Welding Procedure Specification 47 "Gas Tungsten Arc Welding," Revision 15 (e) System One Radiographic Inspection Reports (f) Holtec Procedure HSP-504 Procedure to Perform Closure Welds on the MPC and MPC Lid, Revision 6 (g) US Tool & Die Welder Performance Qualification (WPQ) Test Record for John Ciesielski on WPS-13, Revision 0, dated December 22, 1992 and August 1, 1989 (h) US Tool & Die Welder Performance Qualification (WPQ) Test Record for John Ciesielski on WPS-47, Revision 0, dated May 7, 1993 (i) Holtec Welder Performance Qualification (WPQ) Test Record for Tim Ciesielski on WPS-13, Revision 0, dated January 14, 1987 (j) Holtec Welder Performance Qualification (WPQ) Test Record for Tim Ciesielski on WPS-93, Revision 2, dated August 21, 1995 (k) Holtec Welder Performance Qualification (WPQ) Test Record for Keith Eggar on WPS-47, Revision 4, dated July 23, 1999 (l) Holtec Welder Performance Qualification (WPQ) Test Record for Keith Eggar on WPS-93B, Revision 8, dated January 30, 2005 (m) Holtec Welder Performance Qualification (WPQ) Test Record for Tom Hagner on WPS-47, Revision 14, dated July 13, 2004 (n) Holtec Welder Performance Qualification (WPQ) Test Record for Tom Hagner on WPS-93B, Revision 8, dated April 11, 2005 (o) US Tool &

Die Welder Performance Qualification (WPQ) Test Record for Joseph McMahon on WPS-47, Revision 12, dated June 26, 2001 (p) Procedure Qualification Record (PQR)

No. 1C for Weld Procedure Specification WPS-13, Revision 0, dated February 21, 1983 (q) Procedure Qualification Record (PQR) No. 4A for Weld Procedure Specification WPS-41, Revision 0, dated February 21, 1983 (r) Procedure Qualification Record (PQR)

No. 4B for Weld Procedure Specification WPS-41, Revision 0, dated February 21, 1983 (s) Procedure Qualification Record (PQR) No. 4C for Weld Procedure Specification WPS-41, Revision 0, dated February 21, 1983 (t) Procedure Qualification Record (PQR)

No. 1062 for Weld Procedure Specification WPS-47, dated March 31, 2000 (u)

Procedure Qualification Record (PQR) No. 1146 for Weld Procedure Specification WPS-237, dated October 5, 2001 Category: Welding Topic: Welding Operator Performance Qualification Reference: ASME Section IX, Parts QW-301.4, 361.2, 452.1, 6 Code Year 1995 Requirement: The record of welding operator performance qualification (WOPQ) tests shall include the essential variables listed in QW-360, the type of test and test results, and the ranges qualified in accordance with QW-452. The essential variables for machine welding are:

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(1) welding process; (2) direct or remote visual control; (3) automatic arc voltage control (GTAW); (4) automatic joint tracking; (5) position qualified; (6) consumable inserts; (7) backing; and (8) single or multiple passes per side. Two side bend tests are required for groove weld test coupons 3/4 inch thick or greater. Groove weld tests qualify fillet welds.

Observation: All welders that participated in the May 2-5, 2011 welding dry run and were scheduled to perform welding on the first canister were qualified on the robotic welding machine to perform welding in accordance with Procedure HSP-504. Welding operator performance qualification (WOPQ) test records documented that all five welders had qualified to the essential variables of the gas tungsten arc welding (GTAW) process listed in Welding Procedure Specification WPS-246LW for both remote welding and direct welding.

Procedure DCS-504, Attachment 9.5 "MPC Lid Welding Production Work Routing Plan (PWRP)" listed Welding Procedure Specification WPS-246LW as the applicable welding process for the automatic lid welding activities. The welders were qualified by radiography examinations of the welds.

Documents (a) Weld Procedure Specification WPS-246LW "Machine GTAW - Hot Wire on MPC Reviewed: Closure Welds," Revision 0 (b) System One Radiographic Inspection Reports (c) Holtec Procedure HSP-504 Procedure to Perform Closure Welds on the MPC and MPC Lid, Revision 6 (d) Holtec Welder Operator Performance Qualification (WOPQ) Test Records for John Ciesielski on WPS-246, Revision 2, dated April 6, 2011 [WQT 6032 and 6033] and April 18, 2011 [WQT 6052 and 6053] (e) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for John Ciesielski on WPS-246LW, Revision 0, dated April 18, 2011 [WQT 6053, 6071, and 6072] (f) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Tim Ciesielski on WPS-246LW, Revision 0, dated April 18, 2011 [WQT 6073 and 6074] (g) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Tim Ciesielski on WPS-246, Revision 2, dated April 6, 2011 [WQT 6030 and 6031] (h) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Keith Eggar on WPS-246, Revision 2, dated April 6, 2011 [WQT 6034 and 6035] and April 18, 2011 [WQT 6048 and 6049]

(i) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Keith Eggar on WPS-246LW, Revision 0, dated April 18, 2011 [WQT 6069 and 6070] (j)

Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Joe McMahon on WPS-246LW, Revision 0, dated April 18, 2011 [WQT 6075 and 6076] (k)

Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Joe McMahon on WPS-246, Revision 2, dated April 6, 2011 [WQT 6028 and 6029] and April 18, 2011 [WQT 6044 and 6045] (l) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Tom Hagner on WPS-246, Revision 2, dated April 18, 2011 [WQT 6046, 6047, and 6066] (m) Holtec Welder Operator Performance Qualification (WOPQ) Test Record for Tom Hagner on WPS-246LW, Revision 0, dated April 18, 2011 [WQT 6067 and 6068]

Category: Welding Topic: Welding Procedure Specification (WPS)

Reference: ASME Section IX, Part QW-200.1 Code Year 1995 Requirement: Each manufacturer or contractor shall prepare written Welding Procedure Specifications for making production welds to code requirements. Welding Procedure Specifications shall include the essential, non-essential, and (when required) supplementary essential Page 148 of 149

variables for each welding process. The variables are listed in QW-250 through QW-280 and are defined in Article IV, Welding Data.

Observation: Weld Procedure Specifications WPS-13, WPS-41, WPS-47, WPS-47LW, WPS-93B, and WPS-246LW were reviewed against the requirements in ASME Section IX, Part QW-250 through QW-280 and found to contain the essential variables as required for welding the canister lid. The applicable sections of the ASME code for gas metal arc welding (GTAW) were in Part QW-256 and 256.1 "Welding Variables Procedure Specifications (WPS) for Gas Tungsten Arc Welding." Parts QW-256 and 256.1 listed the essential, supplementary, and non-essential variables for such topics as joints, base metal, filler metals, preheat, etc. The Holtec weld procedure specifications had incorporated the variables applicable to the canister welding at Comanche Peak.

Documents (a) Holtec Weld Procedure Specification WPS-13 "Manual GTAW," Revision 0 (b)

Reviewed: Holtec Weld Procedure Specification WPS-41 "Manual GTAW," Revision 0 (c) Holtec Weld Procedure Specification WPS-47LW "Manual GTAW," Revision 0 (d) Holtec Weld Procedure Specification WPS-47 "Gas Tungsten Arc Welding," Revision 15 (e)

Holtec Weld Procedure Specification WPS-246LW "Machine GTAW Hotwire,"

Revision 0 (f) Holtec Weld Procedure Specification WPS-93B "Gas Tungsten Arc Welding," Revision 10 Page 149 of 149