Text

IR 05000206-03-010; 05000361-03-010; 05000362-03-010; 07200041-03-001; on 08/4/2003 - 10/01/2003; Southern California Edison Co., San Onofre Nuclear Generating Station; Units 1, 2, 3. ISFSI Report. No Violations. SONGS ISFSI Inspection 72-4
	ML040070255
Person / Time
Site:	San Onofre
Issue date:	01/06/2004
From:	Spitzberg D; NRC/RGN-IV/DNMS/FCDB
To:	Ray H; Southern California Edison Co
References
	-RFPFR IR-03-001, IR-03-010
	Download: ML040070255 (70)
	v • d • e

SONGS ISFSI INSPECTION 72-41/2003-01 (INSPECTOR NOTES)

Category: Drying/Helium Topic: Helium Backfill Pressure Requirement Reference: CoC 1029, Tech Spec 3.1.2; FSAR, Sect 3.4.1 Requirement After vacuum drying, the canister is backfilled with helium providing a non-corrosive environment. The canister helium backfill pressure shall be 1.5 + or - 1.5 psig (stable for 30 minutes after filling).

Finding: A helium backfill pressure limit of 1.5 psig +/- 1.0 psig was incorporated into Procedure SO23-X-9, Step 6.12.11. This was more conservative than the Technical Specification 3.1.2 requirement. Step 6.12.13 recorded the final helium pressure after 30 minutes with the canister isolated. Step 6.12.13 also included an independent verification and sign-off by quality control. For the first canister loaded, the final helium pressure was 0.8 psi This was achieved on September 29, 2003. In addition to the official pressure gauge on the helium backfill system, the licensee had installed two highly sensitive Heise pressure gauges, model HQS-2, on the system for real time confirmation of the system pressure gauge reading as a quality assurance check. The two Heise gauges both read 1.4 psig at the end of the 30 minute test.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Drying/Helium Topic: Helium Leak Rate Test on Inner Lid Reference: CoC 1029, Tech Spec 3.1.3; FSAR 1.2.2.2 & 7.1.1 Requirement Leak testing of the canister inner top cover plate to shell weld, vent and siphon cover plate and vent/siphon block to shell weld is performed by pulling a vacuum between the inner and outer top cover plates through a test port in the outer top cover plate and monitoring for helium in accordance with the requirements of ANSI N14.5. The canister's helium leak rate of the inner top cover plate and vent and siphon port cover welds shall be less than or equal to 10(-7) standard cubic centimeters/second (std-cc/sec). The test is performed after the root pass weld of the outer lid.

Finding: The helium leak test of the inner lid and siphon/vent ports was conducted using Procedure SO1-XII-9.103, Section 6.6. Instructions were provided for conducting the leak test of the inner lid after the outer lid was welded in place. Step 6.6.4.1 specified an acceptance criteria of 3.4 x 10(-8) std-cc/sec. This met the Technical Specification 3. requirement for the leak rate of less than 1 x 10(-7) std-cc/sec. Section 6.6 described the leak testing of the inner cover using a hood technique to satisfy ANSI N14.5 "Leakage Tests on Packages for Shipment." The leak test was performed by pulling a vacuum between the inner and outer top cover plates through a test port in the outer top cover plate.

Documents (a) Procedure SO1-XII-9.103 "Helium Leak Detection of Advance NUHOMS 24PT1 Reviewed: Dry Shielded Canisters after Fuel Loading and Final Weld Closure," Rev 0 (b) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Page 1 of 70

Category: Drying/Helium Topic: Helium Leak Test Personnel/Equipment Reference: FSAR 1029, Sect 9.1.3 Requirement Personnel performing the helium leak tests must be qualified in accordance with SNT-TC-1A "Personnel Qualification and Certification in Nondestructive Testing."

Finding: Procedure SO1-XII-9.103, Step 3.2 required personnel performing leak testing to be certified in accordance with Procedure SO123-XII-2.16. Procedure SO123-XII-2.16 required leak test personnel to be qualified in accordance with SNT-TC-1A. For the first canister loading, the helium leak testing was performed by Leak Test Specialists, In Personnel assigned to the SONGS site were interviewed and were qualified in accordance with SNT-TC-1A.

Documents (a) Procedure SO1-XII-9.103 "Helium Leak Detection of Advance NUHOMS 24PT1 Reviewed: Dry Shielded Canisters after Fuel Loading and Final Weld Closure," Rev 0 (b)

Procedure SO123-XII-2.16 "Qualifications and Certification of Nondestructive Examination Personnel" Category: Drying/Helium Topic: Helium Purity Reference: CoC 1029, Tech Spec 3.1.3 Requirement The helium leak test of the inner top cover plate weld and vent/siphon port cover welds shall be performed using helium gas greater or equal to 99% purity.

Finding: Procedure SO1-XII-9.103 established the criteria for performing the helium leak tests for the inner top cover plate and the siphon/vent port covers. Section 6.2 described the equipment and material for the tests and specified in Step 6.2.6 that helium gas with a purity of 99.995% or greater was to be used for the tests. The helium gas bottles for the ISFSI project were stored in the warehouse. A review of the bottles found them to be properly tagged with the correct color coded tag in compliance with the quality assurance program requirements. The tag listed the purchase order number for the helium bottles. During the loading of the first canister, the helium bottles available in the Unit 3 fuel building were all labeled as high purity with the correct material code as specified in the note above Step 6.10.2 of Procedure SO23-X-9 for the 99.995% pure helium.

Documents (a) Procedure SO1-XII-9.103 "Helium Leak Detection of Advanced Nuhoms 24PT1 Dry Reviewed: Shielded Canisters after Fuel Loading and Final Weld Closure," Rev 0 (b) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (c) Procedure SO123-XII-7.12 "Source Verification," Rev 3 (d) Procedure SO123-MS-1 "Material Support Program," Rev 5 (e)

Topical Quality Assurance Manual (TQAM) Section 4B "Handling, Storage and Shipping," Rev 18 Category: Drying/Helium Topic: Time Limit for Vacuum Drying Reference: CoC 1029, Tech Spec 3.1.1 Requirement Vacuum drying of the canister shall be achieved within the following time durations after the start of bulk water removal (blowdown): 1) No limit for 12 kW or less, 2) 71 hours8.217593e-4 days 0.0197 hours 1.173942e-4 weeks 2.70155e-5 months for more than 12 kW but less than or equal to 13 kW and 3) 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months for more than 13 kW but less than or equal to 14 k .

Page 2 of 70

Finding: The licensee had incorporated the time limits for vacuum drying after blowdown listed in Technical Specification 3.1.1 into Procedure SO23-X-9. Specifically, Step 6.1.1 of procedure SO23-X-9, listed the time limits for canisters up to 14 kW. The heat load for canisters loaded at SONGS was limited by Technical Specification 2.1(c) to 14 kW except for the mixed-oxide (MOX) assemblies, which were limited to 13.706 kW. For assemblies of 12 kW or less, there was no time limit requirement for how long the spent fuel could be in the canister during vacuum drying. All five canisters to be loaded in the first loading campaign to remove the Unit 1 spent fuel from the Unit 3 spent fuel pool were below 12 kW. For the first canister, the heat load was 9.352 kW. Procedure SO23-X-9, Step 6.8.15 documented the start of the blowdown process. For the first canister loaded, the time recorded was September 14, 2003 at 11:22 a.m. On September 15, 2003, initial vacuum drying was started. The vacuum drying process for the SONGS casks was complicated by the design of the fuel assemblies which included dash-pots that held water after the canister was initially drained. The licensee estimated that the dash-pots associated with the 24 fuel assemblies would hold approximately 4.3 gallons of water that could not be drained. During the first few days of vacuum drying, excess moisture was successfully being removed from the cask by the vacuum drying system and was being collected in a bottle at the end of the vacuum drying system exhaust lin After 7 days, approximately 19.8 gallons had been collected. This exceeded the initial estimate of the amount of water left in the dash-pots. The licensee determined that the design of the stainless steel spacers being used with the Unit 1 spent fuel may have also held water. Vacuum drying was stopped and an additional 18.6 gallons of water was drained from the canister. Vacuum drying was re-initiated. On September 28, 2003, the licensee succeeded in passing the 2.8 torr/30 minute dryness test. Upon completion of the vacuum drying test, the licensee backfilled the canister with high purity helium and performed the inner lid leak test. This was completed on September 28, 2003. The canister was then vacuum dried again to the 2.8 torr limit for 30 minutes, which was successfully completed the following day. During the second vacuum drying process, an additional 1/2 cup of water was drained from the canister. Later, white crystals, thought to be boron, were observed in a clear section of the drain line that connected the canister to the vacuum drying system. The licensee initiated a work order to modify the spacers to prevent water from remaining in the spacer cavity.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Drying/Helium Topic: Time Limit for Vacuum Drying Not Met Reference: CoC 1029, Tech Spec 3.1.1 Requirement If the required vacuum limit cannot be established within the time limits specified in Technical Specification 3.1.1, then within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months establish a helium pressure of at least 1 atmosphere and no greater than 20 psig in the canister or flood the canister with water submerging all fuel assemblies.

Finding: Procedure SO23-X-9, Step 6.1 incorporated the requirement of Technical Specification 3.1.1 for the vacuum drying time limits and the actions to take if the limits were not me Step 6.1.1 specified the Technical Specification 3.1.1 actions required to be completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to establish a helium environment in the canister or re-flood the canister. Step 6.1.1 referenced Attachment 6 "DSC Helium Fill" and Attachment 7

"Flooding the DSC with Spent Fuel Pool Water" as the two actions available if the time Page 3 of 70

limits were not met. Step 6.8.17 of Procedure SO23-X-9 required the time limit for the canister being loaded to be determined based on the canister's heat load. The licensee then predicted the date and time that the Limiting Condition of Operation for the technical specification would be entered and recorded this information in Step 6.8.18 of the procedure.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Drying/Helium Topic: Vacuum Drying Pressure Limit - Initial Drying Reference: CoC 1029, Tech Spec 3.1.1; FSAR, Sect 8.1.1.3 (21)

Requirement After the inner top cover is welded in place, the cavity pressure should be reduced in steps to approximately 100 torr, 50 torr, 25 torr, 15 torr, 10 torr, 5 torr and 3 torr. This staged drawdown will prevent ice blockage of the evacuation path. After pump down to each level, the pump is valved off and the cavity pressure monitored. The cavity pressure will rise as water and other volatiles in the cavity evaporate. When the cavity pressure stabilizes, the pump is valved in to complete the vacuum drying proces Vacuum drying is complete when the pressure stabilizes for a minimum of 30 minutes at 3 torr or less.

Finding: Procedure SO23-X-9, Section 6.9 described the process for the "initial" vacuum drying operation and required the canister pressure to be reduced, after the inner cover was welded in-place, in steps of 100 torr, 50 torr, 25 torr, 15 torr, 10 torr, 5 torr, and 2.8 tor After each of the stepped levels were reached, the canister was isolated and the pressure monitored. If the pressure rose above the previous stepped limit within 5 minutes, the canister was vacuum dried again until the 5 minute criteria was met. At the end of the stepped vacuum drying process, the procedure required the canister pressure to be reduced to less than 1 torr before the final pressure test was performed. Vacuum drying was successfully completed during the final pressure test when the pressure remained below 2.8 torr for a minimum of 30 minutes. The value of 2.8 torr was used as a quality assurance measure to account for instrument error and to provide an extra margin for compliance instead of the 3 torr limit specified in Technical Specification 3.1.1. For the first canister loaded, the final pressure at the end of the 30 minutes was 2.75 torr for the

"initial" drying of the canister. This was achieved on September 28, 2003.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Drying/Helium Topic: Vacuum Drying Pressure Limit for Final Drying Reference: CoC 1029, Tech Spec 3.1.1; FSAR, Sect 8.1.1.3 (27)

Requirement After the helium leak test on the inner lid welds, re-evacuate the canister. The cavity pressure should be reduced in steps to approximately 10 torr, 5 torr and 3 torr. After pumping down to each level, the pump is valved off and the cavity pressure is monitored. When the cavity pressure stabilizes, the pump is valved in to continue the vacuum drying process. Vacuum drying is complete when the pressure stabilizes for a minimum of 30 minutes at 3 torr or less.

Finding: Procedure SO23-X-9, Section 6.11 described the process for the "final" vacuum drying operation and required vacuum drying to be performed in steps of 10 torr, 5 torr and torr. The final vacuum drying was performed after completion of the initial vacuum Page 4 of 70

drying and the helium backfill and helium test of the inner lid welds. After each of the stepped levels were reached, the canister was isolated and the pressure monitored. If the pressure rose above the previous stepped limit within 5 minutes, the canister was vacuum dried again until the 5 minute criteria was met. At the end of the stepped vacuum drying process, the canister pressure was reduced to less than 1 torr before the final pressure test was performed. Vacuum drying was successfully completed during the final pressure test when the pressure remained below 2.8 torr for a minimum of 30 minutes. For the first canister loaded, the final pressure at the end of the 30 minutes was 2.73 torr for the "final" drying of the canister. This was achieved on September 29, 2003. In addition to the official pressure gauge on the vacuum drying system, the licensee had installed two highly sensitive Heise pressure gauges on the system as real time confirmation of the system pressure gauge reading. The two Heise gauges both read 2.50 torr at the end of the 30 minute final test.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Emergency Planning Topic: Emergency Drills Reference: 10 CFR Part 50, App E, Sect F.1 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.

Finding: The licensee had not conducted an emergency drill involving ISFSI operations. The licensee planned to conduct an emergency drill in September 2003 and a graded site exercise in October 2003. Neither scenario involved the ISFSI. However, the events that could occur at the ISFSI would be similar to the types of radiation emergencies included in the drill and exercise program being implemented for the reactor facilities and would involve the same emergency response personnel.

Documents None Reviewed:

Category: Emergency Planning Topic: Emergency Plan Reference: 10 CFR 72.32(c)

Requirement For an ISFSI that is located on the site of a nuclear power plant licensed for operation, the emergency plan required by 10CFR50.47 shall be deemed to satisfy the requirements of this section.

Finding: The licensee had incorporated the ISFSI into the SONGS site-wide emergency plan. By letter dated July 25, 2003 a revision to the SONGS emergency plan was issued incorporating the ISFSI. The site-wide emergency plan complied with 10 CFR 50.47 for the operating reactor facilities at the San Onofre site. The licensee had submitted the revised emergency plan to the NRC by letter dated July 25, 2003 in accordance with 10 CFR 50.54(q). Procedure SO123-VIII-1 was updated to include ISFSI related emergency action levels. The emergency action levels included loss of canister integrity, fire within the ISFSI and security events. These incidents would be classified as Unusual Events. Action Request #031001626-01 was issued to incorporate a tornado touching down in the ISFSI as an emergency action level in Tab E1 of Procedure SO123-VIII-1. The licensee used Regulatory Guide 1.101, "Emergency Planning and Page 5 of 70

Preparedness for Nuclear Power Reactors" and the Nuclear Energy Institute document NEI 99-01 "Methodology for Development of Emergency Action Levels," as guidance in developing the emergency action levels and the classification determination. The licensee had also developed security contingency procedures for security incidents and abnormal operating procedures for a dropped cask and for natural disasters.

Documents (a) "Emergency Plan," Rev 14 (b) Procedure SO123-VIII-1, "Recognition and Reviewed: Classification of Emergencies," Rev 20 (b) Action Request #031001626-01 (c) Letter to NRC from Southern California Edison entitled "Docket Nos. 50-206, 50-361 and 50-362 Revisions to Emergency Plan at SONGS," dated July 25, 2003 (NRC Adams Document ML032100742)

Category: Emergency Planning Topic: Emergency Plan Changes Reference: 10 CFR 72.44(f)

Requirement Within six months of any changes made to the emergency plan, the licensee shall submit a report containing a description of the changes to the appropriate regional office and HQ.

Finding: The ISFSI emergency planning program was incorporated into the site-wide reactor emergency planning program developed to comply with 10 CFR Part 50. The site-wide emergency planning program required notification to the NRC within 30 days of any change to the emergency plan in accordance with 10 CFR 50.54(q). The 30-day notification requirement was also specified in Step 6.2.4 of Procedure SO123-VIII-0.100.

Documents Procedure SO123-VIII-0.100 "Maintenance and Control of Emergency Planning Reviewed: Documents," Rev 5 Category: Emergency Planning Topic: Offsite Emergency Support Reference: 10 CFR 72.122(g)

Requirement Structures systems and components important to safety must be designed for emergencies. The design 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.

Finding: The design of the ISFSI pad provided for accessibility to onsite and offsite emergency response personnel and equipment. The ISFSI was located in the Unit 1 industrial are There were two access gates leading into the ISFSI security-controlled area. The access points were wide enough for emergency response vehicle entry. In case of fire or medical emergency, the licensee's onsite fire brigade and emergency medical technicians would be the first responders to the event. The licensee's emergency response teams would be augmented, as necessary, by the Camp Pendleton fire department or local, offsite emergency medical teams. The licensee provided training to offsite response organizations. Training was provided to the Camp Pendleton fire department and emergency medical technician personnel during October 2002. Training consisted of four hours of classroom training and four hours of site tour/mini-drill. The next training was scheduled for October 2003 and was expected to include ISFSI-specific training. At the June 2003 Interjurisdictional Planning Committee Meeting, the licensee briefed the committee on proposed changes to the emergency action levels. The committee Page 6 of 70

consisted of local city, county, state and federal representatives. The committee approved the proposed changes to the emergency action level procedure to incorporate the ISFSI.

Documents "Interjurisdictional Planning Committee Monthly Meeting Minutes," dated June 4, 2003 Reviewed:

Category: Fire Protection Topic: Combustible Materials at the ISFSI Reference: FSAR 1029, Sect 11.2.4.1 Requirement Combustible materials will not normally be stored at the ISFSI. However, a hypothetical fire accident is evaluated for the advanced NUHOMS system based on a fuel fire. The source of fuel is postulated to be from a ruptured fuel tank of the transfer cask transporter tow vehicle. The bounding capacity of the fuel tank is 300 gallons and the bounding hypothetical fire is an engulfing fire around the transfer cask.

Finding: The licensee had established controls to limit combustible material stored at the ISFSI pad and during movement of spent fuel to the ISFSI pad to well below the 300 gallon limit. Procedure SO23-X-9, Step 6.15.1 required notification of the site fire department and completion of Attachment 5 "Hazardous Materials Checklist" prior to transporting the canister from the fuel building area to the ISFSI. Attachment 5 included verification that combustible and flammable liquids in quantities exceeding 56 gallons were not in nearby buildings or within 10 feet of the transportation route, except for materials stored in flammable liquid storage lockers. The prime mover, which pulled the trailer containing the canister, held 50 gallons of diesel fuel. Procedure SO1-XXVIII-6. provided controls for storing transient combustibles near the ISFSI. The licensee had painted the ISFSI pad to clearly delineate the ISFSI combustible control zone. Action Request #020701550 identified plans to install eight dry chemical fire extinguishers around the ISFSI. The fire extinguishers would provide manual fire fighting capability for any transient combustible fire event. Security Procedure SO123-IV-5.3.7 provided administrative controls for movement of vehicles transporting 300 or more gallons of flammable liquid near the ISFSI protected area.

Documents (a) Procedure SO123-IV-5.3.7 "Vehicle Access Controls, Search and Inspection," Rev 0 Reviewed: (b) Procedure SO1-XXVIII-6.1.2 "Control of Work and Storage Areas," Rev 3 (c)

Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (d) Action Request

020701550 dated July 30, 2002 Category: Fire Protection Topic: External Explosion Reference: FSAR 1029, Sect 2.3.6 & 3.1.2.2.4.2 Requirement Externally initiated explosions are considered to be bounded by the design basis tornado generated missile load analysis. The licensee will evaluate the site specific external hazards to the ISFSI to demonstrate that externally initiated explosions are bounded by the design basis tornado generated missile load analysis, as described in FSAR Section 2.2.1.

Finding: The licensee evaluated a site specific explosion that could affect the ISFSI and determined that an over-pressure of 7 pounds/square inch (psi) could result. This exceeded the analyzed value of 3 psi pressure drop for the tornado analyzed in the FSAR. The licensee performed Calculation C-296-01.02 "ISFSI Pad 'Other Events'

Page 7 of 70

Hazards Analysis" to determined that the 7 psi value was acceptable. Per FSAR Table 3.1-7 "24PT1-DSC External Pressure Loads," the canister was qualified to an external pressure of 22 psi.

Documents Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Reviewed:

Category: Fire Protection Topic: Fire Protection Plan Reference: CoC 1029, Tech Spec 4.4.3.6; 10 CFR 50.48(a)(1)

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. The potential for fires and explosions, affecting the ISFSI, must be addressed based on site-specific considerations.

Finding: The licensee evaluated the hazards from fires and explosions that could affect the ISFSI in the Updated Fire Hazards Analysis and the 10 CFR 72.212 Evaluation. The analysis included consideration of the 20-foot combustible control zone and the 300-gallon diesel fuel limit. Administrative controls were established to limit the amount of combustible material near the ISFSI. Section 6 "SONGS ISFSI and Unit 1 Industrial Area," of the Updated Fire Hazards Analysis divided the Unit 1 area into three subsections: the ISFSI area, the common site facilities area and the industrial area. The fire protection equipment available in the Unit 1 area readily available to the ISFSI included portable extinguishers, wet pipe sprinkler systems and fire hydrants. The licensee had a full time fire department consisting of professionally trained personnel. A minimum of five certified fire fighters were on-duty per shift. A mutual aid agreement was established with the Camp Pendleton Marine Base to provide additional fire fighting capability.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) Updated Fire Reviewed: Hazards Analysis (draft).

Category: Fuel Verification Topic: Center of Gravity Reference: CoC 1029, Tech Spec 2.1(b)/(d); FSAR, Sect 2.1.1 Requirement The canister may include two empty fuel slots and/or multiple dummy fuel assemblie The empty slots must be located at symmetrical locations about the 0-180 degree and the 90-270 degree axes. The canister's center of gravity, under all loading conditions (including two empty slots and/or dummy fuel assemblies) must be maintained within 0.1 inches of its design basis (symmetrical loading) radial position to ensure that deviations in the center of gravity are minimal. This includes consideration of the center of gravity when control components are loaded in the canister. Dummy fuel assemblies are unirradiated stainless steel encased structures that approximate the weight and center of gravity of a fuel assembly.

Finding: The licensee had completed Calculation No. N-1020-164 to determine the deviation from the canister's radial center of gravity for each of the loading patterns for the first five canisters and verified that no canister had a deviation that exceeded the 0.1 inch limi The calculation provided a description of the procedure used to calculate the overall canister center of gravity based on the weights and locations of the specific fuel assemblies and non-fuel assembly hardware to be stored. Table 8-11 "Radial Center of Gravity Calculation for Canister Loading Pattern Number 1" of Calculation No. N-1020-Page 8 of 70

164 included the four failed fuel cans. These were identified for loading in the four outermost guide sleeves in accordance with Technical Specification 2.1(a). The deviation of the center of gravity for this canister from the radial center of gravity was 0.0042 inches. Canister Loading Pattern Number 5 identified only twenty-two assemblies for loading. Since each canister was designed to hold twenty-four assemblies, this would leave two empty slots. The two empty canister locations were included in Calculation No. N-1020-164, Table 8-15 "Radial Center of Gravity Calculation for Canister Loading Pattern Number 5." The two empty slots were located symmetrically about the 0-180 degree and 90-270 degree axes as required by Technical Specification 2.1(d). The deviation of the center of gravity for this canister from the radial center of gravity was 0.0 inches. For the first five canisters, the calculated deviation from the radial center of gravity ranged from 0.0 inches to 0.0042 inches, all within the 0.1 inch limit.

Documents Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Patterns Reviewed: for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 Category: Fuel Verification Topic: Damaged Fuel Pins Reference: CoC 1029, Tech Spec 2.1(d); FSAR, Sect 2.1.1 Requirement No more than 14 fuel pins in each assembly may exhibit damage.

Finding: The criteria for no more than 14 fuel pins exhibiting damage per assembly was specified in the canister loading criteria, Section 1.2 of Calculation No. N-1020-164. The visual inspection of the Unit 1 fuel described in the memorandum dated October 29, 2001, indicated that there were four Unit 1 assemblies in the Unit 3 spent fuel pool with damaged pins. Each of these assemblies had only one damaged pin.

Documents (a) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Reviewed: Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 (b) SCE Memorandum to file "Visual Inspection of Unit 1 Fuel in the Units 1, 2 &

3 Spent Fuel Pools," dated October 29, 2001 Category: Fuel Verification Topic: Damaged Fuel/Fuel Configuration Reference: CoC 1029, Tech Spec 2.1(d)

Requirement Three configurations for the fuel to be stored in the canister are allowed by the technical specifications. These include: a) 24 intact assemblies, b) up to 4 damaged stainless steel assemblies with the balance intact stainless steel assemblies, c) one mixed oxide (MOX)

damaged assembly with the balance intact stainless steel assemblies.

Finding: The licensee had documented a loading pattern in Calculation No. N-1020-164 for the first five canisters that complied with the requirements in Technical Specification 2.1(d)

for fuel configuration. Four damaged stainless steel assemblies, all placed in the same canister, would be loaded into one of the canisters. The remaining spent fuel assemblies planned for loading in the first five canisters were intact stainless steel assemblies.

Documents (a) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Reviewed: Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 (b) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1

.

Page 9 of 70

Category: Fuel Verification Topic: Failed Fuel Can Cover Reference: FSAR 1029, Sect 8.1.1.1 (6)

Requirement The failed fuel can cover is to be removed from the failed fuel can for installation after fuel loading.

Finding: Procedure SO23-X-9, Step 6.2.37 removed the failed fuel can cover after the failed fuel can was installed in the canister and prior to placement in the spent fuel pool. Step 6. required installation of the failed fuel can cover after the failed fuel assembly was placed into the can and prior to the canister being removed from the spent fuel pool.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Fuel Verification Topic: Failed Fuel Can Inspection Reference: FSAR 1029, Sect 8.1.1.1 (6)

Requirement Failed fuel cans should be placed into the canister's basket prior to lowering the transfer cask/canister into the spent fuel pool. The failed fuel cans shall be inspected prior to placement in the canister. This inspection shall ensure cleanliness and determine if any damage has occurred to the failed fuel cans since receipt inspection.

Finding: Procedure SO23-X-9, Step 6.2.35 required the licensee to verify that the failed fuel can was not damaged and was Class B clean prior to installation in the canister. The failed fuel can was placed in the canister per Step 6.2.36 in the location specified by Attachment 2 "DSC Fuel Loading Pattern," of Procedure SO23-X-9 prior to placement in the spent fuel pool. Class B cleanliness was defined in Procedure SO123-I-1.38 as having no evidence of surface contamination under visual inspection without magnification and with optimum lighting conditions on the surface being inspected.

Documents (a) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Reviewed: Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 (b) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (c) Procedure SO123-I-1.38 "Internal Cleanliness of Components," Rev 2 Category: Fuel Verification Topic: Failed Fuel Cans Reference: CoC 1029, Tech Spec 2.1(a); FSAR, Sect 2.1.1 Requirement Damaged fuel assemblies shall be placed in screened confinement cans (failed fuel cans)

and shall be stored in outermost guidesleeves located at the 45, 135, 225, and 315 degree azimuth locations. Damaged fuel may include assemblies with known or suspected cladding defects greater than hairline cracks or pinhole leaks, up to and including broken rods, portions of broken rods and rods with missing sections.

Finding: The licensee identified twenty-seven Unit 1 damaged fuel assemblies for storage in failed fuel cans. Procedure SO23-X-8 provided guidance consistent with the Transnuclear FSAR, Section 2.1.1 and NRC's Interim Staff Guidance (ISG)-1 "Damaged Fuel," for classifying spent fuel as damaged. The licensee used a computer program called Cask Works, for identifying acceptable canister loading patterns. Cask Works limited placement of the damaged assemblies to the outermost guide sleeves at the 45, 135, 225, and 315 degree azimuth locations consistent with the requirements in Page 10 of 70

Technical Specification 2.1(a). A total of twenty-seven fuel assemblies (four in the Unit 3 pool, twenty-one in the Unit 1 pool, and two in the Unit 2 pool) had suspected defects greater than pinhole leaks or hairline cracks on exterior fuel pins and/or suspected gross cladding defects on interior fuel pins. For the first loading campaign, which will remove all Unit 1 spent fuel from the Unit 3 spent fuel pool, one canister will contain all four of the damaged Unit 1 fuel assemblies.

Documents (a) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Reviewed: Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 (b) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (c) Procedure SO23-X-8 "Fuel Bundle Visual Examination in the Spent Fuel Pool," Rev 5 Category: Fuel Verification Topic: Fuel Dimensions and Weights Reference: CoC 1029, Tech Spec 2.1(e); Table 2-2 Requirement Table 2-2 of the technical specifications provides the fuel dimensions and weights for the assemblies that can be stored in the canisters.

Finding: The licensee performed an evaluation of the Unit 1 spent fuel stored in the Unit 3 spent fuel pool in Calculation No. N-1020-164. Included in this calculation was Table 1-2 which specified the required dimensions and weights for the spent fuel allowed for storage in the ISFSI. A comparison of the Table 1-2 requirements against the SONGS Unit 1 spent fuel in the Unit 3 pool was documented as Table 4-2 in the calculation. A review of Table 1-2 and Table 4-2 verified that the dimensions and weights evaluated by the licensee to confirm acceptability of the spent fuel were the same as the values listed in Table 2-2 of the technical specifications in Certificate of Compliance #1029.

Documents (a) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Reviewed: Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 Category: Fuel Verification Topic: Fuel Spacers and Failed Fuel Cans Reference: FSAR 1029, Sect 1.2.1.1 Requirement Stainless steel fuel spacers are used to center the short 14x14 Westinghouse fuel in the 24PTI-DSC canister. Stainless steel screened confinement cans (failed fuel cans), placed inside the 24PTI-DSC guide sleeves, are provided for storage of damaged fuel assemblies.

Finding: Provisions had been incorporated into procedures for the installation of stainless steel fuel spacers and the use of failed fuel cans. Procedure SO23-X-9, Step 6.2.52 required installation of the DSC bottom spacers per Procedure SO1-XXVIII-5.89. Step 6.2.53 required verification that the spacers were installed in each canister location which will be receiving a fuel assembly. Procedure SO1-XXVIII-5.89 included steps for ensuring the cleanliness of the spacers, installing the spacers and verifying that the spacers were in each canister location where a fuel assembly will be stored. Procedure SO23-X-9, Step 6.2.36 required installation of failed fuel cans in accordance with the loading pattern provided in Attachment 2 of the procedure. Attachment 2 was a form that would be completed for each canister showing which fuel assembly would be placed in each of the canister locations. Attachment 2 required final canister loading patterns to be Page 11 of 70

verified against Calculation No. N-1020-164 and signed off by both the Fuels Service Engineer and the Supervisor, Fuels Services.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) SO1-XXVII-5.89 Reviewed: "DSC Bottom Spacer Installation," Rev 0 (c) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 Category: Fuel Verification Topic: Fuel Types Allowed for Storage Reference: CoC 1029, Tech Spec 2.1(b); Table 2-1; Table 2-4 Requirement The fuel types which can be stored in the 24PT1-DSC canister used at SONGS are limited to Westinghouse 14x14 assemblies consisting of UO(2) with stainless steel cladding or mixed oxide fuel (MOX) with Zircalloy cladding. Table 2-1 and Table 2-4 of the technical specifications provides acceptable fuel specifications.

Finding: Calculation No. N-1020-164 provided general information on all the Unit 1 spent fuel and provided specific information on the 118 Unit 1 spent fuel assemblies stored in the Unit 3 spent fuel pool to confirm compliance of these 118 assemblies against the Technical Specification 2.1 requirements for storage in the NUHOMS 24PT1-DSC canister. Removal of the Unit 1 spent fuel from the Unit 3 spent fuel pool will be the first loading campaign for SONGS and will require five canisters. The SONGS Unit 1 spent fuel was stored in four locations: 270 assemblies at the General Electric spent fuel storage pool located in Morris, Illinois; 207 assemblies in the SONGS Unit 1 spent fuel pool; 70 assemblies in the SONGS Unit 2 spent fuel pool; and 118 assemblies in the SONGS Unit 3 spent fuel pool. The Unit 1 spent fuel was stainless steel-clad UO(2) fuel with the exception of four zircalloy-clad MOX assemblies stored in the Unit 1 poo Four of the 207 assemblies in the Unit 1 pool were intact zircalloy-clad MOX assemblies while the rest of the assemblies in all three pools were stainless steel-clad UO(2). Four Unit 1 assemblies in the Unit 3 pool, twenty-one in the Unit 1 pool and two in the Unit 2 pool were classified as damaged for a total of twenty-seven damaged assemblies. Seven fuel assemblies in the Unit 1 pool will not meet the cooling time criteria for loading until June 1, 2004. For determining burnup of the assemblies, flux traces from fission chambers in the core were input into the computer code INCORE, which generated core power distribution, peaking and assembly burnup rates. Another computer code, SCENIC, used assembly burnup rates from INCORE along with each assembly's core location to determine individual assembly burnups. A conservative 10% uncertainty factor was added to all burnup values. This calculation, outlined in "Fuel Assembly Database for SNM Accountability," was performed according to Procedure SO123-XXIV-7.15. The maximum burnup for any of the 118 assemblies was 37.6 GWd/MT Cooling times ranged from 13.1 to 23.3 years. Initial enrichments range from 3.983 to 4.006 wt-%. The licensee also planned to store two canisters containing greater than class C (GTCC) waste from Unit 1 at the ISFSI.

Documents (a) Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Reviewed: Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 (b) Procedure SO123-XXIV-7.15 "Preparation and Verification of Design Calculations," Rev 3 (c) "Fuel Assembly Database for SNM Accountability," Rev 0 Page 12 of 70

Category: Fuel Verification Topic: Material Balance, Inventory, and Records Reference: 10 CFR 72.72(a)

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).

Finding: Procedure SO123-X-1.7 identified "Item Control Areas (ICAs)" for tracking the inventory of Special Nuclear Material (SNM) at SONGS. The procedure established the Unit 1 ISFSI as an Item Control Area. The transfer of SNM from any of the spent fuel pools to the Unit 1 ISFSI generated a Form 741. Form 741 identified each assembly transferred and its inventory of SNM.

Documents (a) Procedure SO123-X-1.7 "Special Nuclear Material Accountability," Rev 7 (b) DOE Reviewed: NRC Form 741 "Nuclear Material Transaction Report: Transfer of Nuclear Material from SONGS Unit 3 Spent Fuel Pool to SONGS Unit 1 ISFSI," dated July 28, 2003 Category: Fuel Verification Topic: Maximum Heat Load Limit per Canister Reference: CoC 1029, Tech Spec 2.1(c)

Requirement The maximum heat load designed for the canister is 14 kW when loaded with all stainless steel clad (SC) fuel assemblies and 13.706 kW when loaded with any mixed oxide (MOX) fuel assemblies. For a single fuel assembly, the limits are 0.583 kW per stainless steel clad assembly and 0.294 kW per MOX fuel assemblies. (See also FSAR 1029 Sect 1.2.1.1 & 1.2.3)

Finding: Heat loads for each canister were calculated in Section 8.4 "Canister Loading Patterns" of Calculation No. N-1020-164. The five canisters planned for loading during the first campaign had total decay heat loads of 8.836, 9.353, 9.355, 9.310 and 8.337 kW. All were stainless steel clad and were below the 14 kW limit of Technical Specification 2.1(c). The maximum individual assembly decay heat load in each canister was 0.445, 0.414, 0.405, 0.436, and 0.450 kW, respectively. The canister loading pattern was developed using the computer program Cask Works, which used the ORIGEN-S isotope depletion and decay computer code to determine assembly decay heat and source ter The loading patterns assigned to each canister were relatively uniform with respect to decay heat, with the hottest assemblies stored on the outside of the pattern when there were larger decay heat differences between the spent fuel assemblies.

Documents Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Patterns Reviewed: for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 Category: Fuel Verification Topic: Neutron/Gamma Source Terms for Fuel Reference: CoC 1029, Tech Spec 2.1(f); Table 2-3 Requirement The limits for the neutron and gamma source terms for the fuel assemblies is provided in Table 2-3 of the technical specifications.

Finding: Calculation No. N-1020-164 included calculations of the neutron and gamma source term for each of the fuel assemblies for the first five canisters and verified that the source terms were less than the limits specified in Table 2-3 of the technical specifications. The hottest assembly (0.450 kW) had a gamma source term of 2.29 E+15 Page 13 of 70

gamma/sec/assembly and a neutron source term of 8.41 E+7 neutron/sec/assembly. This was below the 3.43 E+15 gamma/sec/assembly and 2.84 E+8 neutron/sec/assembly specified for stainless steel assemblies in Table 2-3 of the technical specifications. Two neutron source assemblies (NSAs) will be stored with the fuel assemblies in the first loading campaign. Since the antimony/beryllium (SbBe) neutron source assemblies had decayed to less than 1 E-18 of their original source strength in the 10 years since shutdown, the neutron sources were not considered in the analysis. The two neutron source assemblies were in the Unit 3 spent fuel pool and were included in the loading pattern for the first five canisters. The two neutron source assemblies will be placed in different canisters.

Documents Calculation No. N-1020-164 "Dry Cask Storage 24PT1-DSC Canister Loading Patterns Reviewed: for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 Category: Fuel Verification Topic: Verification of Fuel Prior to Loading Reference: FSAR 1029, Sect 8.1.1.2 (5)

Requirement Prior to insertion of a spent fuel assembly into the canister, the identity of the assembly shall be verified by two individuals using an underwater video camera. Read and record the fuel assembly identification number from the fuel assembly and check this identification against the canister loading plan.

Finding: Procedure SO23-X-9, Step 6.4.1 required two individuals to verify the serial number of the spent fuel assemblies in the fuel racks prior to removal for placement in the caniste Step 6.4.2 specified that the spent fuel was to be loaded into the canister in accordance with Procedure SO23-X-7.2 "Nuclear Fuel Movement - Spent Fuel Pool." Step 6. required the top of the spent fuel assemblies to be scanned with a camera and video recorder. Step 6.4.4 required visual verification of the proper loading of the canister by serial number and verification that the assembly was properly seated in relationship to the top of the canister. Step 6.4.5 provided for an independent review of the video tape against the loading pattern in Attachment 2 "DSC Fuel Loading Pattern," of Procedure SO23-X-9 to confirm the spent fuel had been properly loaded into the caniste Movement of the first fuel assembly for loading into a canister was begun at 2:45 pm on September 12, 2003. The spent fuel loading pattern specified in Attachment 2 of Procedure SO23-X-9 was consistent with Figure 8-2 "Canister Loading Pattern Number 2," of Calculation No. N-1020-164. The loading of the 24 fuel assemblies into the canister was completed at 9:05 p.m. on September 12, 2003 for a total of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 20 minutes.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) Calculation No. N-1020-Reviewed: 164 "Dry Cask Storage 24PT1-DSC Canister Loading Patterns for Unit 1 Fuel Assemblies Stored in the Unit 3 Spent Fuel Pool," dated July 18, 2003 Category: Fuel Verification Topic: Visual Examination of Fuel Assemblies Reference: CoC 1029, Tech Spec 2.1(d); FSAR, Sect 8.1.1.1 (1)

Requirement Prior to placement in dry storage, the candidate fuel assemblies are to be visually examined to insure that no known or suspected gross cladding breaches exist. Pinholes and hairline cracks are acceptable. Visual verification of fuel integrity may also be Page 14 of 70

accomplished in conjunction with existing plant records.

Finding: A review of the SONGS fuel inspection program was performed by the NRC during an inspection on April 16-19, 2001 and documented in NRC Inspection Report 50-206/01-06; 72-41/01-01 dated May 11, 2001. The SONGS program for visual examination of the spent fuel was acceptable. On October 29, 2001, SONGS issued a report entitled

"Visual Inspection of Unit 1 Fuel in the Units 1, 2 & 3 Spent Fuel Pools." The visual inspection was performed in accordance with the guidance in NRC Interim Staff Guidance (ISG) - 1, "Damaged Fuel." The licensee's visual inspection consisted of a camera inspection of exterior rods on all four faces of each fuel assembly, and a camera inspection of interior rows, using a back-lighting technique, looking for gross blockage/rod failure. The licensee identified twenty-seven Unit 1 damaged fuel assemblies. Of these, twenty-one were in the Unit 1 spent fuel pool, two were in the Unit 2 spent fuel pool and four were in the Unit 3 spent fuel pool.

Documents (a) SCE Memorandum to file "Visual Inspection of Unit 1 Fuel in the Units 1, 2 & 3 Reviewed: Spent Fuel Pools," dated October 29, 2001 (b) Procedure SO23-X-7.2 "Nuclear Fuel Movement - Spent Fuel Pool," Rev 7 (c) NRC Inspection Report 50-206/01-06; 72-41/01-01 dated May 11, 2001 (NRC Adams Document ML011340078)

Category: General License Topic: Annual Collective Dose During Loading Operations Reference: FSAR 1029, Sect 10.3 Requirement Assumed annual occupancy times, including the anticipated maximum total hours per year for any individual and total person-hours per year for all personnel for each radiation area during normal operation and anticipated operational occurrences will be evaluated by the licensee in a 10 CFR 72.212 evaluation to address the site specific ISFSI layout, inspection and maintenance requirements. The estimated annual collective person rem doses associated with loading operations shall be addressed by the licensee in a 10CFR 72.212 evaluation.

Finding: Calculation No. DCS-001, Attachment 4 provided dose assessment estimates for loading operations and annual ISFSI activities. The licensee conservatively estimated a combined dose of 3117 millirem to load each canister. The licensee also estimated the personnel dose for annual operational activities. The estimated dose to perform the daily visual inspections inside the ISFSI area was calculated to be 172 millirem/year. The estimated dose to perform weekly radiation protection surveys at the ISFSI was calculated to be 49 millirem/year. The dose estimates included site-specific dose rates and estimated occupancy times.

Documents Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Reviewed:

Category: General License Topic: Environmental Extremes Reference: CoC 1029, Tech Spec 4.4.3; FSAR, Sect 2.2.5 Requirement The following parameters and analysis shall be verified by the user for applicability at their specific site: a) tornados/maximum winds of 290 mph rotational and 70 mph translational, b) flood levels of 50 feet and water velocity of 15 feet/second, c) one hundred year roof snow load of 110 pounds/square foot, d) normal ambient temperatures of 0 degree F to 104 degree F, and e) off-normal ambient temperature range of -40 Page 15 of 70

degree F without solar insolation to 117 degree F with full solar insolation. Other site-specific events, such as lightning, tsunamis, hurricanes, and seiches should also be evaluated.

Finding: The licensee evaluated the SONGS site specific environmental conditions in Section X of Calculation No. DSC-001 against the parameters established in the NUHOMS Final Safety Analysis Report (FSAR) and Certificate of Compliance #1029 technical specifications and found the ISFSI site to be bounded by the design parameters established by the NUHOMS FSAR. This included: tornados/high winds, flood, seismic events, snow/ice loading, tsunami, lightning, burial under debris, normal and abnormal temperatures, and fires/explosions. The data used for the analysis was from the SONGS reactor facility Updated Final Safety Analysis Report (UFSAR) and supporting calculations. The maximum tornado wind speed for the SONGS site was 260 mph which consisted of a 220 mph rotational speed and a 40 mph translational speed. These values were below the NUHOMS design criteria of 290 mph rotational and 70 mph translational. The tornado pressure differential for a tornado at SONGS was pounds/square inch (psi). The NUHOMS design basis was 3.0 psi. The tornado generated missiles analyzed by the NUHOMS FSAR were also bounded by the SONGS UFSAR tornado generated missiles. For flooding, the SONGS site maximum probable flood was 0.1 feet with a flow rate of 0.34 feet/second (fps). The NUHOMS design criteria was a flood height of 50 feet and a flow rate of 15 fps. For seismic criteria, the ISFSI design requirement specified in the NUHOMS FSAR was 1.5 g horizontal and g vertical. An NRC inspection of the ISFSI pad conducted on April 22-25, 2002 compared the seismic criteria for the SONGS site to the design basis for the NUHOMS cask system. The results of that inspection were documented in NRC Inspection Report 50-206/2002-07;72-41/2002-01 issued on May 21, 2002. The evaluation found that the SONGS site was bounded by the seismic criteria used for the design of the NUHOMS casks. Snow and ice considerations were identified in Calculation No. DSC-001 as not being applicable, since the likelihood of snow and ice at the SONGS site was insignificant due to the climate in southern California. The licensee performed a site specific analysis for a tsunami at the SONGS site. A tsunami was a possibility due to the coastal location of the ISFSI. The licensee completed Calculation No. C-296-01.02

"ISFSI Pad 'Other Events' Hazards Analysis" which evaluated the potential for the overturning of the concrete storage modules, flooding and debris. The results of the calculation demonstrated that the concrete modules would not be overturned and would not be pushed off the ISFSI pad due to a tsunami. The flooding height of a tsunami was lower than the design basis flood already analyzed for and the problem with debris blocking the cooling vents was already addressed by procedural controls. Lightning was addressed by installing a lightning protection system at the ISFSI. Burial of the ISFSI under debris was addressed by building the ISFSI far enough away from the bluffs that a collapse of the bluff would not affect the ISFSI. Temperature considerations for the ISFSI were evaluated. Both normal and extreme temperatures at the SONGS site were bounded by the NUHOMS design criteria of 0 degree F to 104 degree F normal temperatures and -40 degree F to 117 degree F extremes. For fires and explosions, the licensee evaluated the ISFSI as well as the transportation route from the nuclear plants to the ISFSI. Administrative procedures were established for time periods when the canister was being moved to the ISFSI pad to control explosive and combustible material around the transport route. The ISFSI did not have combustible material in it's construction. Controls were established to limit the amount of flammable material Page 16 of 70

brought onto the pad. SONGS also had 24-hour onsite fire response capability.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) NRC Inspection Reviewed: Report 50-206/2002-07;72-41/2002-01 dated May 21, 2002 (NRC Adams Document ML021410532)

Category: General License Topic: Evaluation of Effluents/Direct Radiation Reference: 10 CFR 72.212(b)(2)(i)(C); 10 CFR 72.104 Requirement The general licensee shall perform a written evaluation prior to use that establishes that the requirements of 10CFR72.104 "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.

Finding: An evaluation of the radiological conditions created by the ISFSI were included in Section VIII of Calculation No. DSC-001 and confirmed that the radiological limits specified in 10 CFR 72.104 would not be exceeded. The evaluation included direct radiation, contamination and liquid/gases effluents. The licensee determined that storage of spent fuel in the ISFSI would not result in any effluent discharges and that contamination potential was minimal. Accordingly, there would be no contribution to thyroid or organ doses resulting from use of the ISFSI. The licensee calculated that the maximum does at the ISFSI controlled area boundary due to direct radiation shine would be less than 1 millirem per year.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) Transnuclear Reviewed: Calculation SCE-23.0505 "San Onofre ISFSI Dose Rate Calculation," Rev A Category: General License Topic: Initial Compliance Evaluation Against CoC Reference: 10 CFR 72.212(b)(2)(i)(A)

Requirement A general licensee shall perform written evaluations, prior to use, that establish that the conditions set forth in the Certificate of Compliance (CoC) have been met.

Finding: The licensee performed an extensive evaluation documented as Attachment 1 to Calculation No. DSC-001 to verify compliance of the SONGS programs against Certificate of Compliance #72-1029. The licensee identified a number of changes and additions to the SONGS programs and procedures including the development of new procedures related to the ISFSI operations, the requirement to conduct a pre-operational exercise, the establishment of technical specifications and surveillance requirements for the ISFSI, documentation that the spent fuel planned for storage at the ISFSI met the certificate of compliance requirements and incorporation of applicable ASME code requirements into the SONGS programs.

Documents Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Reviewed:

.

Page 17 of 70

Category: General License Topic: Initial Compliance Evaluation Against FSAR Reference: 10 CFR 72.212(b)(3)

Requirement The general licensee shall review the Final Safety Analysis Report referenced in the 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 paragraph 72.212(b)(2).

Finding: The licensee documented the review of the NUHOMS Final Safety Analysis Report (FSAR) and the NRC's Safety Evaluation Report (SER) as Attachments 2 and 3 of Calculation No. DSC-001 and identified specific procedures and calculations where the requirements identified in the FSAR and SER had been evaluated to confirm compliance. This evaluation included a review of the various postulated accidents that could occur during cask handling and storage to confirm that the events had been addressed in site procedures or programs. The review included the seismic and tornado issues, which were evaluated in Section XIII of Calculation No. DSC-001 and are discussed in more detail in the above section entitled "Environmental Extremes."

Documents Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Reviewed:

Category: General License Topic: Initial Evaluation Against Part 50 License Reference: 10 CFR 72.212(b)(4)

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 Part 50.59(c)(2). Results of this determination must be documented in the evaluation made in paragraph 72.212(b)(2).

Finding: The licensee documented a review of the activities related to the ISFSI against the reactor facility technical specifications in Section XI of Calculation No. DSC-001 and determined that a license amendment to their Part 50 license was not require Engineering Change Package (ECP) 020701186-33 "Implementation of ISFSI," which included a 10 CFR 50.59 evaluation, assessed the impact of the ISFSI fuel transfer process on the Unit 2 & 3 operating plants. This evaluation included the transfer cask path from the fuel buildings to the ISFSI. No impact on the operating reactor's Part 50 license was identified requiring a license amendment from the NRC. The licensee had also identified several plant modifications required for the ISFSI including the pad construction, security systems for the pad, the temperature monitoring system for the concrete modules, lightning protection system for the ISFSI, Unit 3 fuel handling building modifications and the Unit 3 cask handling crane upgrade. The 10 CFR 50.59 screenings/evaluations performed on these plant modifications also determined that a license amendment and NRC approval was not required.

Documents Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Reviewed:

Category: General License Topic: Program Review - RP, EP, QA, and Training Reference: 10 CFR 72.212(b)(6)

Requirement The general licensee shall review the reactor emergency plan, quality assurance program, Page 18 of 70

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.

Finding: The licensee documented the required reviews of the reactor programs in Section XIII of Calculation No. DSC-001. A number of changes to existing site programs were identified and implemented including adding new emergency action levels to the emergency plan implementing procedures, applying the reactor's Part 50 quality assurance program to the ISFSI such as adding the ISFSI "important to safety" items to the SONGS Unit 1 Q-List, establishing an ISFSI training program, and adding the ISFSI to the site radiation program including developing ALARA goals, establishing radiation work permits for ISFSI work and scheduling radiological surveillances for the ISFSI.

Documents Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Reviewed:

Category: General License Topic: Revisions to 72.212 Analysis Reference: 10 CFR 72.212(b)(2)(ii)

Requirement The general licensee shall evaluate any changes to the written evaluations required by 72.212(b)(2) using the requirements of 72.48(c). A copy of this record shall be retained until spent fuel is no longer stored under the general license issued under 72.210.

Finding: The licensee had not clearly established in their procedures that a change to Calculation No. DCS-001 required a 10 CFR 72.48 evaluation. Action Request #030800197 was issued by the licensee on August 5, 2003 to require changes to Calculation No. DCS-001 to be performed under Procedure SO123-XV-44 "10 CFR 50.59 and 72.48 Program."

The Action Request would change Procedure SO123-XXIV-7.15 "Preparation and Verification of Design Calculations," to include the requirement that all 10 CFR 72.212 evaluation changes be evaluated against 10 CFR 72.48 using Procedure SO123-XV-4 Retention of required records related to the ISFSI were retained for the life of the SCE Corporation in accordance with requirements issued in a memo dated March 25, 2002.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) Internal Memo to Reviewed: CDM-SONGS from J. T. Reilly, Director, SONGS 1 Decommissioning entitled

"Retention of SONGS 1 Decommissioning and SONGS 1,2,3 ISFSI Documents," dated March 25, 2002 Category: General License Topic: Seismic Restraints on Transfer Cask Reference: CoC 1029, Tech Spec 4.4.3.8 Requirement Seismic restraints shall be provided to prevent overturning of a loaded transfer cask during a seismic event if the site can experience a horizontal acceleration of 0.40 g or greater. The determination of horizontal acceleration acting at the center of gravity of the loaded transfer cask must be based on a peak horizontal ground acceleration at the site.

Finding: The licensee identified in Calculation No. C-296-03.04 that the horizontal acceleration from a seismic event could be 1.2 g at the Unit 3 cask washdown area based on the SONGS design basis earthquake peak horizontal ground acceleration of 0.67 g. This seismic acceleration exceeded the 0.4 g requirement in the technical specification. The licensee designed and installed a seismic restraint system in the Unit 3 cask washdown Page 19 of 70

area for attachment to a loaded transfer cask while in the cask washdown area. A 10 CFR 50.59 evaluation was performed for the modification to install the seismic restraint system.

Documents Calculation No. C-296-03.04 "Unit 2 & 3 110-Ton Cask Seismic Constraints" Reviewed:

Category: Heavy Loads Topic: Heavy Loads Requirements Reference: CoC 1029, Condition 5 Requirement Each lift of the canister and transfer 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 safety review (under 10 CFR 50.59 or 10 CFR 72.48) is required to show operational compliance with existing plant specific heavy loads requirements.

Finding: The licensee revised several procedures and completed a 10 CFR 50.59 safety review in Action Request #020101437-63 to incorporate the lifting and movement of the NUHOMS canisters and transfer cask into the existing SONGS heavy loads progra Selected procedures were reviewed during this inspection related to cask handling, rigging and crane maintenance. The procedures had incorporated specific information related to the ISFSI cask lifting and transfer activities. The 10 CFR 50.59 safety review provided a comprehensive evaluation of the Unit 3 crane upgrade to verify requirements for the new crane trolley were incorporated into the Unit 3 programs and procedure Discussions with crane operators during the pre-operational testing activities and during the first cask loading operations at Unit 3 found the individuals to be very knowledgeable of the crane's operating capabilities and safety restrictions.

Documents (a) Action Request #020101437-63 "10 CFR 50.59 Screening of Upgrade of Unit 3 Cask Reviewed: Handling Crane" (b) Procedure SO3-I-3.32 "Unit 3 Cask Handling Crane Checkout and Operation," Rev 1 (c) Procedure SO123-I-7.24 "Rigging Manual," Rev 12 (d) Procedure SO123-I-1.13 "NUREG-0612 Cranes, Rigging and Lifting Control," Rev 9 (e) Procedure SO123-I-7.14 "Maintenance and Inspection of Cranes," Rev 7 Category: Heavy Loads Topic: Inspection After Cask Drop Reference: CoC 1029, Tech Spec 5.3.2 Requirement The canister will be inspected for damage after any transfer cask drop of 15 inches or greater.

Finding: The requirement to inspect the canister for damage after a drop of 15" or greater was specified in Procedure SO23-X-9, Step 6.1.9. Specific provisions for the inspection of the canister for damage following a 15 " drop was incorporated into Procedure SO123-X-9.3, Section 6.3. Procedure SO123-X-9.3 required the canister to be inspected for damage after any transfer cask drop of 15 inches or greater. The procedure required the damage to be evaluated against the system's design requirements. Required rework, repair or retesting would be performed to restore the system to its original design requirements. The licensee's corrective action program would be used to document and track any inspection and recovery activities.

Documents (a) Procedure SO123-X-9.3 "Dropped Cask," Rev 0 (b) Procedure SO23-X-9 "Dry Cask Reviewed: Storage Loading," Rev 1 Page 20 of 70

Category: Heavy Loads Topic: Maximum Lift Height of Cask/Canister Reference: CoC 1029, Tech Spec 5.3.1(a)/(b)/(c)

Requirement The maximum lift height of the transfer cask/canister inside the fuel handling building shall be 80 inches if the ambient temperature is below 0 degree F but higher than -80 degrees F. No lift height restriction is imposed on the cask/canister during loading operations if the ambient temperature is higher than 0 degree F. The maximum lift height and handling height for all transfer operations shall be 80 inches and the ambient temperature must be greater than 0 degree F. Transfer operations begin when the transfer cask is placed on the transfer trailer following loading operations and ends when the canister is placed into the concrete storage module on the storage pad at the ISFSI.

Finding: The 80" lifting limit specified in the technical specification for the transfer cask when temperatures were below 0 degree F was incorporated into Procedure SO23-X-9, Step 6.1.6. Section 6.15 of the procedure also provided a caution related to the 80" lifting height. Step 6.15.11 required that the ambient air temperature of the fuel handling building be measured and recorded prior to lifting the transfer cask. The mild climate of Southern California made it highly unlikely that a limitation on the lifting height of the transfer cask while in the fuel handling building will ever be imposed. The licensee planned the transfer route from Unit 3 to the ISFSI such that the height of the cask would never be over the 80" limit during the movement. With the transfer cask on the transport trailer, the maximum drop height would be 58". When inserting the canister into the advanced horizontal storage module, the maximum height of the canister was approximately 68".

Documents (a) Calculation No. C-296-04.01 "Dry Cask Storage Route," Rev 0 (b) Procedure SO23-Reviewed: X-9 "Dry Cask Storage Loading," Rev 1 Category: Heavy Loads Topic: Maximum Weight of Cask/Canister Reference: FSAR 1029, Table 3.2-1; FSAR 1004, Sect 8.1.1.9 Requirement The maximum weight for the cask during cask loading activities must be verified as less than the capacity of any crane that will be moving the cask. Maximum weight will occur when the transfer cask is being removed from the spent fuel pool while containing a loaded canister full of water.

Finding: The maximum estimated weight of a cask loaded with Unit 1 spent fuel and water at Unit 3 was 105 tons. The capacity of the Unit 3 crane was 125 tons. For cask operations involving the heavier Unit 2/3 Combustion Engineering 16x16 fuel, the maximum estimated weight was 109 tons. Calculation SO1-207-1-C11 documented the weights expected for the various cask movement activities. This included lifting a loaded cask out of the spent fuel pool and moving the cask to the washdown area. This portion of the movement involved the maximum weight. Calculation SO1-207-1-C11 appropriately accounted for all the components that would contribute to the maximum weight, including the added water for neutron shielding at Unit 3. The weight of the crane hook and cabling was accounted for during the design and certification of the crane. At the time of the inspection, Revision 8 to Calculation SO1-207-1-C11 estimated the maximum weight that would be moved by the Unit 3 crane was approximately 10 tons. Since Revision 8, more accurate estimates and actual measurements had been performed which resulted in a new estimated weight of approximately 105.2 tons. This Page 21 of 70

estimate was less than the 125 ton rated capacity of the Unit 3 crane. Calculation SO23-207-16-C5 documented the weight analysis for a transfer cask filled with the Unit 2/3 Combustion Engineering 16x16 fuel. The weight estimate at the time of the inspection for a fully loaded cask and water was approximately 109 tons which was still less than the 125 ton rated capacity of the Unit 3 crane. The Unit 1 turbine gantry crane will be used as the cask handling crane at the Unit 1 spent fuel pool. This crane will be rated for only 105 tons. In order to reduce the weight of the loaded cask, neutron shielding water will not be used. The current estimated weight of the loaded transfer cask at Unit 1 was approximately 104.1 tons or 99.1 percent of the crane's planned rated capacity. Before actual fuel movement, certified weight tests will be performed on the measurable components to ensure that the weight rating capacity of the Unit 1 crane is not exceede The yoke to be used for lifting the loaded casks was labeled as weighing 6415 pounds with a 125 ton load capacity.

Documents (a) Calculation No. SO1-207-1-C11 "DSC/AHSM/OS197 Cask Component Weights, Reviewed: Mass Properties and Lift Weight Calculation and Evaluation," Rev 11 (b) Calculation No. SO23-207-16-C5 "Advanced NUHOMS 24PT4 Weight Calculation," Rev 1 (c)

Draft Updated Weight Calculations for Transfer Cask Containing Units 1/2/3 Spent Fuel (d) Ederer Inc. Drawing No. C-41901, "Lower Block Assembly 125 Ton Crane," Rev B Category: Heavy Loads Topic: Safe Loads Path Reference: NUREG 0612, Sect 5.1.1; FSAR 1029, Sect Requirement Safe loads 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 the spent fuel pool, or to impact safe shutdown equipment.

Finding: The safe load path for the movement of the transfer cask was adequately defined and analyzed to minimize the potential for the heavy load to impact the spent fuel pool or safe shutdown equipment. Procedure SO3-I-3.32 documented the safe load path for movement of the loaded cask from the Unit 3 spent fuel pool, over the fuel handling building hatch, to the washdown area. The crane was prevented from deviating beyond the boundaries of the safe load path through the use of limit switches and the physical boundaries of the crane rails. Prior to using the crane, the limit switches were tested for proper operation. The safe load path that was identified would not place the transfer cask directly over the irradiated fuel in the spent fuel pool or over any safe shutdown equipment. The licensee physically marked the floor of the fuel handling building with green tape to identify the safe load path. Procedure SO3-I-3.32 stated that all heavy loads with a documented load path were required to follow the identified path and that any deviation from the path required written authorization from Maintenance Engineering and the Maintenance Manager. Calculation C-296-04.01 documented the analysis of the path for transporting the transfer cask from Unit 3 to the ISFSI. Starting from the outside of the Unit 3 fuel handling building, the prime mover would transport the cask at approximately 3 mph along the east road just west of the switch yard. The cask would pass the Unit 2 containment, the Unit 2 reserve auxiliary transformers and a hazardous waste storage area before reaching the security hold down area at the Unit 2/3 protected area exit. The cask would then proceed past the security processing facility and enter the Unit 1 industrial area where the ISFSI was located. Calculation C-296-04.01 also included an evaluation of the major underground components that the transfer Page 22 of 70

trailer and cask would drive over. The two most significant systems were portions of the firewater main piping and the emergency diesel generator fuel tanks. The calculation demonstrated that the extra weight from the loaded transfer trailer was within the design limits of the two systems with an approximate margin of 25 percent. The NRC inspectors walked the transfer route with the licensee to verify that the calculation accurately accounted for all systems along the safe load path.

Documents (a) Calculation No. C-296-04.01 "Dry Cask Storage Route," Rev 0 (b) Procedure SO3-I-Reviewed: 3.32 "Unit 3 Cask Handling Crane Checkout and Operation," Rev 1 (c) Procedure SO123-I-1.13 "NUREG 0612 Cranes, Rigging and Lifting Controls," Rev 9 (d)

Procedure SO1-I-7.27 "Turbine Gantry Crane Checkout and Operation," Rev 9 (e)

Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Category: Heavy Loads Topic: Single Failure Proof Crane Reference: NUREG 0554; NUREG 0612, Sect 5.1.6, App C Requirement SONGS will use new crane trolleys for movement of the transfer cask/canister in the fuel buildings. The cranes are being considered single failure proof. In order for the cranes to be considered single failure proof, a number of evaluations and tests must be performed in accordance with the manufactures specifications, NUREG 0554 "Single Failure Proof Cranes for Nuclear Power Plants" and NUREG 0612 "Control of Heavy Loads at Nuclear Power Plants."

Finding: The licensee had upgraded their heavy lift capability to support the fuel movement activities by purchasing two Ederer X-SAM single failure proof Seismic Category I crane trolleys. One crane was purchased to upgrade the Unit 3 facility. A second crane would initially be used at Unit 1, then moved to Unit 2 after the Unit 1 spent fuel pool was emptied. The licensee completed a 10 CFR 50.59 safety review for the crane trolley replacement. The main hoist capacity of the Unit 3 crane was not up-rated and remained at 125 tons. The auxiliary hoist remained at 10 tons. There was no change in the Unit 3 crane's capability to perform the design and accident mitigation functions described in the SONGS Part 50 Final Safety Analysis Report. For Unit 1 however, the load capacity of the crane was increased from 100 tons to 105 tons. The licensee submitted a license amendment to the NRC on July 25, 2003 for modifications to the Unit 1 turbine building and turbine building gantry crane. The licensee also performed structural calculations of the Unit 1 building and support structure associated with the crane trolley and identified modifications to the facility to strengthen the support structure. During this inspection, the NRC reviewed the crane design, along with the licensees testing and maintenance program, against selected guidance in NUREG-0554 "Single Failure Proof Cranes for Nuclear Power Plants" and NUREG-0612 "Control of Heavy Loads at Nuclear Power Plants". Procedure 251-F2704 was also reviewed and portions of the Unit 3 crane test observed. The final results of the licensee's testing of the crane were reviewed. No issues were identified.

Documents (a) Ederer Single Failure Proof Trolley Project Specification SO123-209-01-M331, Reviewed: Procedure 251-F2704 "Site Pre-Operational Test for X-SAM System," Rev 2 (b) Ederer Single Failure Proof Trolley Project Specification SO123-209-01-M316, Procedure 250-F2704 "Field Acceptance, Inspection & Testing, Rev 2 (c) NUREG-1774 "A Survey of Crane Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002,"

dated July 2003 (d) NUREG-0554 "Single-Failure-Proof Cranes for Nuclear Power Page 23 of 70

Plants," dated May 1979 (e) NUREG-0612 "Control of Heavy Loads at Nuclear Power Plants" dated July 1980 (f) Transnuclear West Calculation Number SCE-01.0215

"Design and Evaluation of the OS197-3 Cask Lifting Yoke" (g) NRC Letter from Robert L. Baer, Chief, Light Water Reactors Branch No.2, Division of Project Management, to Mr. C. William Clark Jr., Manager of Engineering, Ederer Incorporated "Review and Acceptance of Topical Report EDR-1, Ederers Nuclear Safety Related eXtra-Safety And Monitoring (X-SAM) Cranes, Rev 1," dated January 2, 1980.

Category: Heavy Loads Topic: Sling Safety Margins Reference: ANSI N 14.6 Requirement For the slings that will lower the top shield plug into place on the loaded canister, verify that the slings to be used meet the required safety margins specified in the site heavy loads program.

Finding: The slings purchased for lowering the top shield plug into place on the loaded canister were purchased with certifications for a rating of 2 times the estimated weight of the top shield plug. This exceeded the ANSI N 14.6 requirement for the slings to have a rated capacity of 1.5 times the weight of the object being lifted. The slings were properly labeled, listing both their load capacity and the due date of the next visual inspection, and were in good physical condition. Procedure SO123-I-7.10 provided the licensee's criteria for inspecting the slings. Procedure SO123-I-7.24 provided the rigging requirements for use of the slings.

Documents (a) ANSI N 14.6, "Special Lifting Devices for Shipping Canisters Weighing 10,000 Reviewed: Pounds or More" (b) Procedure SO123-I-7.10 "Periodic Inspection and Testing of Rigging and Accessories," Rev 7 (c) Procedure SO123-I-7.24 "Rigging Manual" Category: Heavy Loads Topic: Transfer Cask Trunnion Tests/Inspections Reference: FSAR 1029, Sect 9.2.1.1; FSAR 71-9302, Sect 8.2.1 Requirement FSAR 1029, Section 9.2.1.1 establishes a number of routine and annual inspection requirements for the OS-197 transfer cask. FSAR 71-9302 for the MP-197 has additional requirements for load tests and inspections of the transfer cask trunnion SONGS is using the OS-197 which should have similar load test requirements for the trunnions as described in Section 8.2.1 of the MP-197 FSAR.

Finding: Documentation provided by Equipos Nucleares, S.A. (ENSA), the manufacturer of the fuel transfer cask, listed the design and testing criteria required by Southern California Edison for the transfer cask trunnions and provided the results of the tests. The trunnions were load tested to 375 Tons (187.5 Tons per trunnion) and subsequently penetrant tested as required by the design specifications. The testing criteria was consistent with the requirements in Safety Analysis Report MP-197, Section 8.2.1. The licensee had also incorporated the periodic trunnion penetrant testing requirements required by Section 9.2.1.1 of Safety Analysis Report NUH-003 into Procedure SO23-X-9.

Documents (a) Safety Analysis Report NUH-003 for Docket #72-1004, "Safety Analysis Report for Reviewed: the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," Rev 4A (b) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (c) Safety Analysis Report MP-197 (Docket #71-9302) (d) ENSA Equipos Nucleares, S.A. Do Page 24 of 70

OFWS CS 020 (SO1-207-1-M2659-1) "Load Test of Upper Trunnions" Category: Heavy Loads Topic: Transfer Trailer Reference: FSAR 1029, Sect 11.2.5.1.1 Requirement Once the transfer cask is loaded onto the transport trailer and secured, it is pulled to the ISFSI site by a tractor vehicle. The hold down mechanisms that secure the transfer cask to the transport trailer remain in place at all times during the canister transport. As a result, there is no reasonable way during the operations for a cask drop accident to occur.

Finding: The licensee had incorporated pre-operational instructions, testing and annual maintenance for the transfer trailer into Procedure SO123-X-9.2 consistent with the trailer vendor's operational manual. The licensee performed portions of the trailer inspection as a demonstration for the NRC inspectors. The positioning of the transfer cask and the hold down mechanisms on the trailer were examined. A tour of the transport path was completed to verify that the transport path was sufficiently level and no obstacles were present that could result in the unbalancing of the trailer or result in a condition where the transfer cask could become dislodged from the trailer. Due to the manner in which the transfer cask trunnions were secured to the transfer trailer, the cask weight, and the overall trailer design, it was determined that a cask drop accident during transport to the ISFSI was not a credible event. The weight capacity of the trailer was confirmed against the weight of a loaded transfer cask to verify that the trailer capacity to move the spent fuel was sufficient.

Documents (a) Procedure SO123-X-9.2 "Prime Mover and Transfer Trailer Operation," Rev 0 (b)

Reviewed: Procedure SO1-207-M301 "Greiner Operational Manual, Transfer Trailer," Rev 0 (c)

Docket 72-1004 "Safety Analysis Report NUH-003 for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel," Rev 4A Category: Pad/Storage Modules Topic: Air Inlet Cleaning Reference: FSAR 1029, Sect 3.1.1.2 Requirement A removable block is provided below the air inlet to provide a means for cleanout of the inlet in case of accidental blockage.

Finding: The design of the concrete storage modules for the NUHOMS system, referred to as the advanced horizontal storage modules (AHSMs), made it relatively simple to remove the bolts on the screen and two bolts attaching the block below the air inlet to obtain access to clean the air inlet. With the storage modules on the ISFSI pad in various stages of construction, access was readily available to the area below the air inlet to examine how the removable block was placed and attached to the storage module. A streaming path for radiation that could present high radiation levels around the storage module when the block was temporarily moved was not likely due to the design of the storage module.

Documents None Reviewed:

.

Page 25 of 70

Category: Pad/Storage Modules Topic: Air Inlet/Outlet Inspection Prior to Loading Reference: FSAR 1029, Sect 8.1.1.6 (3)

Requirement Prior to canister insertion into the concrete storage module, the air inlet and outlet as well as the screens should be inspected to ensure that they are clear of debris and/or damaged.

Finding: Procedure S023-X-9, Steps 6.16.7, 6.16.8 and .6.16.9 required verification that the storage module cavity and air inlet and outlet vents were free of debris and that the air inlets, outlets and heat shield were not damaged prior to inserting the canister loaded with spent fuel.

Documents Procedure S023-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Pad/Storage Modules Topic: Berm Construction Reference: CoC 1029, Tech Spec 4.4.3.7; 10 CFR 72.104(a)

Requirement In cases where engineering features, such as the berm around the ISFSI pad and the shield walls, are used to ensure that the requirements of 10 CFR 72.104(a) are met, such features are to be considered important to safety and must be evaluated to determine the applicable quality assurance category. 10 CFR 72.104(a) establishes radiological exposure limits to the public from the ISFSI.

Finding: SONGS had not established a berm around the ISFSI. The thickness of the concrete used with the concrete storage modules was calculated to be adequate to reduce the radiation levels from the stored spent fuel to below the limits in 10 CFR 72.104(a) for the public.

Documents None Reviewed:

Category: Pad/Storage Modules Topic: Built to ACI-318/349 Standards Reference: CoC 1029, Tech Spec 4.3.1: FSAR, Sect 2.5.2 Requirement The concrete storage module for the NUHOMS system, referred to as the Advanced Horizontal Storage Module (AHSM), is considered "important to safety" since it provides physical protection and shielding for the canister during storage. The reinforced concrete in the storage module is designed in accordance with American Concrete Institute Standard (ACI)-349-97 and built to ACI-318. The level of testing, inspection and documentation provided during construction and maintenance is in accordance with the quality assurance requirements defined in 10CFR 72, Subpart G.

Finding: Construction of the advanced horizontal storage modules (AHSMs) was inspected by the NRC's Spent Fuel Project Office on November 18-21, 2002 and documented in Inspection Report 72-1029/2002-201 dated December 26, 2002. The inspection concluded that the concrete components for the storage modules being constructed at the Kie-Con facility in Antioch, California for the SONGS ISFSI were being fabricated in accordance with the requirements of 10 CFR Parts 21 and 72, the applicable Certificate of Compliance and Safety Analysis Report and Transnuclear's NRC-approved quality assurance program. Selected requirements specified by ACI-318 were inspected and had been adequately implemented during construction of the storage modules. At the SONGS site, the licensee was properly storing the various storage module component Page 26 of 70

Adequate documentation of the receipt inspections of the components was being generated in accordance with the SONGS quality assurance program. Selected quality assurance paperwork was reviewed to confirm that the licensee was appropriately documenting and tracking receipt inspection issues. Items noted on the receipt inspections that required rework had been referred back to the vendor.

Documents (a) Receiving Inspection Data Reports (RIDR) RSO-1167-03-00, RSO-0567-03-00, RSO-Reviewed: 0427-03-00 and RSO-1409-03-00 (b) NRC Inspection Report 72-1029/2002-201 dated December 26, 2002 (NRC Adams Document ML023610073)

Category: Pad/Storage Modules Topic: Horizontal Storage Module-to-Module Connection Reference: FSAR 1029, Sect 1.2.1.2 Requirement The concrete storage modules are "tied" to adjacent storage modules but are not tied to the reinforced concrete basemat.

Finding: The licensee was in the process of constructing the first twelve concrete storage modules at the ISFSI. Three storage modules were substantially completed. A review of the construction packages and inspection of the construction underway verified that the concrete storage modules were being tied to each other, but not to the basemat.

Documents (a) Construction Work Orders 03030021000, 03030022000 and 03030647000 for the Reviewed: "Assembly and Installation of Advanced Horizontal Storage Module (AHSM) 001" (b)

Construction Work Orders 03032104000, 03032105000 and 03032106000 for the

"Assembly and Installation of Advanced Horizontal Storage Module (AHSM) 002" Category: Pad/Storage Modules Topic: Seismic Design for the Pad Reference: CoC 1029, Tech Spec 4.2.2; FSAR, Sect 2.2.3 Requirement The licensee is to perform site-specific analysis considering the effects of soil-structure interaction. Amplified seismic spectra at the location of the concrete storage module center of gravity is to be developed based on the SSI responses. The storage pad location shall have no potential for liquefaction at the site-specific Safe Shutdown Earthquake (SSE) level earthquake.

Finding: On April 9, 2002, the NRC observed the concrete placement of the first ISFSI basema This inspection was documented in an NRC inspection report dated May 21, 2002. The inspection included observation of the pouring of the pad and a review of the technical design of the pad. The licensee had performed an extensive analysis of the seismic conditions that could effect the ISFSI pad including the sub-soil conditions at the location of the pad. The site specific seismic conditions used for the ISFSI analysis were detailed in the reactor facility's Updated Final Safety Analysis Report. Analysis showed the seismic requirements for the ISFSI were met at the SONGS site and that no potential for liquefaction of the soil existed during a safe shutdown earthquake.

Documents NRC Inspection Report 50-206/02-07;72-41/02-01 dated May 21, 2002 (NRC Adams Reviewed: Document ML021410532)

.

Page 27 of 70

Category: Pad/Storage Modules Topic: Shield Walls Reference: FSAR 1029, Sect 3.1.1.2 Requirement The top shield block of the concrete storage module is attached to the base unit by eight steel rods in the vertical direction. Three-foot thick shield walls are installed behind each storage module (single row array only) and at the ends of each row to provide additional shielding and missile protection.

Finding: Twelve concrete storage modules were in various stages of completion at the ISFSI. The first four were substantially complete with the top shield blocks properly connected and the three foot shielded wall installed on the back of the storage module and at the beginning of the row.

Documents None Reviewed:

Category: Pad/Storage Modules Topic: Shield Walls at End of Array Reference: FSAR 1029, Sect 1.2.1.2 Requirement Separate shield walls at the end of a storage module row in conjunction with the storage module wall, provide a minimum thickness of four feet for shielding.

Finding: The first row of concrete storage modules was under construction. The wall on the end of the row was measured. The concrete wall of the storage module was 1 foot thic Attached was a concrete shield wall that was an additional 3 feet thick. The measurements were consistent with the requirements in the Final Safety Analysis Report and the concrete module design drawings.

Documents (a) Drawings SCI-01-2000 "Nuhoms Advanced Horizontal Storage Module General Reviewed: Arrangement," Rev 0 (b) Drawings SCI-01-2001 "Nuhoms Advanced Horizontal Storage Module Main Assembly," Rev 0 (c) Drawings SCI-01-2003 "Nuhoms Advanced Horizontal Storage Module Roof," Rev 0 (d) Drawings SCI-01-2004 "Nuhoms Advanced Horizontal Storage Module Walls," Rev 0 Category: Pad/Storage Modules Topic: Storage Module Array Configuration Reference: FSAR 1029, Sect 1.4 & 3.1.1.2 Requirement The required array size for the advanced horizontal storage modules (AHSMs) to meet the high seismic design criteria is a minimum of three storage modules in a single row array. Expansion can be accomplished as necessary by the licensee provided the criteria of 10CFR 72.104 and 10CFR 72.106 are met.

Finding: All rows of advanced horizontal storage modules (AHSMs) planned for the SONGS ISFSI exceeded the three concrete storage module minimum. The first row of storage modules was being constructed with twelve storage modules in various stages of completion. The first row will consist of 19 storage modules, 17 for spent fuel and two for Greater Than Class C Waste (CTCC). Plans were to expand the ISFSI pad to three rows of storage modules once room was available after Unit 1 decommissioning. The licensee had performed the necessary analysis concerning compliance with the dose rate limits specified in 10 CFR 72.104. This is discussed in the Inspector Notes section entitled Radiological: Dose to Individuals Beyond the Controlled Area. Analysis confirming compliance with the dose rate limits specified in 10 CFR 72.106 are Page 28 of 70

discussed in the Inspector Notes section entitled Radiological: Controlled Area Radiological Doses.

Documents None Reviewed:

Category: Pad/Storage Modules Topic: Storage Module Configuration Reference: CoC 1029, Tech Spec 4.4.1 Requirement Each group of concrete storage modules not tied together must be separated from other groups by a minimum of 20 feet to accommodate possible sliding during a seismic even The distance between any storage module and the edge of the ISFSI pad shall be no less than 10 feet.

Finding: At the time of the inspection, there was only one group of concrete storage modules in the process of being installed. Twelve modules were in various stages of being complete, with three on one end complete. The 10 foot criteria between the module group and the edge of the ISFSI pad was verified. The 20 foot criteria was not demonstrated since the placement of the second group of concrete storage modules had not occurred at the time of the inspection. Licensee personnel were aware of the 20 foot requirement in the technical specification.

Documents None Reviewed:

Category: Pad/Storage Modules Topic: Thermal Monitoring Program - Concrete Temperature Reference: CoC 1029, Tech Spec 5.2.5(a)

Requirement A program shall be established to monitor the performance of each concrete storage module. The temperature measurements will be a direct measurement of the storage module's concrete temperature, or other means that would identify and allow for the correction of off-normal thermal conditions that could lead to exceeding the concrete and fuel clad temperature criteria. A temperature measurement of the thermal performance for each storage module will be taken on a daily basis.

Finding: Engineering Change Package (ECP) No. 010901111-4 described the temperature monitoring systems to be installed in the concrete storage modules. Procedure SO23-X-9, Step 6.18.66 documented notification of the Unit 1 Operations Shift Supervisor when a canister was inserted into the concrete storage module and the temperature monitoring program required by Procedure SO1-12.1-9 should be initiated. Procedure SO1-12.1-9 provided for daily temperature monitoring.

Documents (a) Engineering Change Package (ECP) No. 010901111-4 "Temperature Monitoring Reviewed: System, ISFSI," Rev 0 (b) Procedure S0123-XXIV-10.1 "Preparation, Review Approval, Issuance Implementation and Closure of Engineering Change Packages (ECPs) and Engineering Change Notices (ECNs)," Rev 5-1 (c) Procedure SO1-12.1-9 "ISFSI Daily Survey" (d) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Category: Pad/Storage Modules Topic: Thermal Monitoring Program - Vent Inspection Reference: CoC 1029, Tech Spec 5.2.5(c); FSAR, Sect 9.2.1.5 Requirement Site personnel will conduct a daily visual inspection of the concrete storage module air Page 29 of 70

vents to ensure the air vents are not blocked for more than 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months and that blockage will not exist for periods longer than assumed in the safety analysis.

Finding: The licensee had incorporated the requirement for the visual inspection of the concrete storage module vents on a daily basis into Procedures S01-9-18 and S01-12.1-0.

Documents (a) Procedure S01-9-18 "Advanced Horizontal Storage Modules (AHSMs) Monitoring Reviewed: and System Operation," Rev 0 (b) Procedure S01-12.1-0 "Inspection of Advanced Horizontal Storage Modules (AHSMs)," Rev 0 Category: Pad/Storage Modules Topic: Thermal Monitoring Program - Vent Temperature Reference: CoC 1029, Tech Spec 5.2.5(b)

Requirement Following initial canister transfer to the concrete storage module, the air temperature difference between ambient temperature and the roof vent temperature will be measured 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the canister insertion into the storage module and again 7 days after insertion. If the air temperature differential is greater than 100 degree F, the air inlets and outlets should be checked for blockage.

Finding: Procedure S023-X-9, Steps 6.18.68 through 6.18.70 provided instructions to check the ventilation temperature at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at 7 days. Steps 6.18.69 and 6.18.71 required removal of any blockage of the air inlets or outlets and initiation of a condition report if the temperature exceeded 100 degree F.

Documents Procedure S023-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Pad/Storage Modules Topic: Thickness of Concrete Storage Module Roof Reference: FSAR 1029, Sect 1.2.1.2 Requirement The nominal thickness of the concrete storage module roof is five feet for biological shielding.

Finding: The concrete storage modules were being constructed at the licensee's ISFSI with twelve in various stages of completion. The roof on four of the storage modules was examined and verified to be five feet in thickness.

Documents None Reviewed:

Category: Pre-Operational Tests Topic: Cask/Canister Fit Test Reference: FSAR 1029, Sect 9.2.1.2(A)

Requirement The canister will be loaded into the transfer cask to verify fit and adequacy of the transfer cask/canister annulus seal.

Finding: The licensee successfully demonstrated the placement of a canister into the transfer cask on June 26-28, 2003 and August 18-19, 2003 during the pre-operational demonstrations observed by the NRC. The adequacy of the annulus seal between the canister and the transfer cask was demonstrated to the NRC on September 12, 2003 as part of the preparations for placement of the cask/canister into the spent fuel pool for fuel loading.

Documents None Reviewed:

Page 30 of 70

Category: Pre-Operational Tests Topic: Insertion of Canister into Storage Module Reference: FSAR 1029, Sect 9.2.1.2(D)

Requirement The transfer trailer is aligned and docked with the concrete storage module. The hydraulic ram is used to insert the canister loaded with mock-up fuel assemblies into the storage module and then to retrieve it. Transfer of the canister to the storage module will verify that support skid positioning system and the hydraulic ram system operate safety for both insertion and retrieval.

Finding: On June 27-28, 2003 and August 18-19, 2003 the licensee demonstrated the insertion of a weighted canister from the transfer cask into a concrete storage module. The demonstrations were performed in accordance with Procedure SO23-X-9. The licensee also demonstrated that the transport trailer, after being backed away from the storage module, could be realigned with the storage module and the canister retrieved back into the transportation cask. During a practice test at the ISFSI, the ram tilt cylinder support pillow block bearings failed while tilting the hydraulic ram. The hydraulic ram is used to insert, and if necessary, to remove the canister from the storage module. The licensee issued Action Request #030800608 to replace the blocks and to evaluate the cause of the failure. Modifications were made which strengthened the blocks. The modifications provided for a successful demonstration of the hydraulic ram during the August 19, 2003 loading/unloading demonstration.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Pre-Operational Tests Topic: Lifting Yoke/Load Path Test Reference: FSAR 1029, Sect 9.2.1.2(B)

Requirement Functional testing is performed with the transfer cask and lifting yoke. These tests are to ensure that the transfer cask can be safely lifted from the plant's cask receiving area to the cask washdown area. The cask is also partially lowered into the spent fuel pool and positioned in the cask loading area to verify clearances and travel path.

Finding: The licensee performed numerous cask movement activities to verify the operability of the crane, travel paths, clearances for the yoke and the adequacy of the extender for lowering the cask into the lower portion of the spent fuel pool. During June 27-28, 2003, August 18-19, 2003 and immediately prior to initiation of canister loading on September 10-11, 2003, the NRC observed the cask handling crane load tests, loading of a weighted canister into the transfer cask, movement of a weighted canister/transfer cask into the cask handling area of the spent fuel pool, raising and moving a weighted canister/transfer cask from the cask handling area to the decontamination area, moving a weighted canister/transfer cask into the fuel handling bay and lowering of a weighted canister/transfer cask onto the transfer trailer. The demonstrations were performed safely and without any significant heavy loads or rigging problems. Comprehensive pre-job briefings were performed prior to the dry-runs and included safe work practices, radiological controls, communications, acceptance criteria and personnel responsibilities.

Documents None Reviewed:

.

Page 31 of 70

Category: Pre-Operational Tests Topic: Pre-Operational Testing Requirement Reference: CoC 1029, Condition 8; FSAR, Sect 1.2.2.1 Requirement A dry run training exercise of the loading, closure, handling, unloading and transfer of the Advanced NUHOMS system shall be conducted prior to the first use of the syste The training exercise shall not be conducted with spent fuel in the canister. The dry run may be performed in an alternate step sequence from the actual procedural guidelines in Chapter 8 of the SAR. The dry run shall include: a) fuel loading, b) canister sealing, drying and backfilling operations, c) transfer cask downending and transport to the ISFSI, d) canister transfer to the concrete storage module, e) canister retrieval from an storage module, f) flooding a canister, g) opening a canister.

Finding: The licensee completed all required pre-operational testing requirements during several demonstrations between June 26, 2003 and September 12, 2003 that were observed by the NRC. Certificate of Compliance #1029, Condition 8 listed seven specific demonstrations that were required. Loading Condition 8.a required a demonstration of fuel loading. This demonstration confirmed that the fuel handling machine was capable of reaching all guide sleeve positions in the canister and would be able to successfully align the fuel assemblies with the correct guide sleeve position to allow insertion of the assembly. On September 12, 2003, SONGS used a dummy fuel assembly and successfully inserted the assembly into four locations in a canister. The locations represented the furthest extreme guide sleeve positions in the canister that the fuel handling machine was required to reach. The first insertion was followed by the uncoupling of the handling machine from the dummy assembly and a demonstration that the fuel handler could return to the assembly, regrapple and remove the dummy assembly, should unloading the canister be necessary. Loading Condition 8.b required demonstration of canister sealing, drying and backfilling with helium. This was observed by the NRC during the week of August 4-8, 2003. This included the welding of a lid on a truncated canister and the nondestructive testing of the weld. The operability of the vacuum drying and helium backfill equipment and the ability to connect the equipment to the canister were demonstrated. Loading Condition required the downending of the transfer cask onto the transport trailer and movement of the cask to the ISFSI. This was demonstrated with a transfer cask containing a weighted canister similar in weight to an actually loaded canister. The transport route demonstration was from Unit 3 to the ISFSI and was performed on June 27-28, 2003 and again on August 18, 2003. Loading Condition 8.d and Unloading Condition 8.a required demonstration of the insertion of a canister into a concrete storage module and retrieval back into the transfer cask. This was demonstrated on August 19, 2003 and included the insertion of a weighted canister into the concrete storage module, the disconnecting and removal of the transfer cask and transport trailer from their position at the storage module, then repositioning the transport trailer back to the storage module and the removal of the canister from the storage module. Unloading Condition 8.b required demonstration of the flooding of a loaded canister should this be required to prepare the canister for removing the lid and placing the spent fuel back into the spent fuel poo This capability was demonstrated through a review of the licensee's Procedure SO23-X-9.1 which described the process of unloading a canister. The licensee walked through the steps that would be performed and demonstrated that the necessary equipment and connectors were available. The siphon port would be connected to the spent fuel pool discharge line as a source of water for flooding the canister. The vent port would be Page 32 of 70

used as a discharge line back to the spent fuel pool. Step 6.6.15 of Procedure SO23-X-9.1 required the canister atmosphere to be sampled before flooding. A special precaution was included in the procedure above Step 6.6.19 to control the entry of water into the canister due to the potential for the water to be flashed to steam. The procedures and equipment for flooding a loaded canister were found to be acceptable. Unloading Condition 8.c required the capability to open the canister. SONGS had made arrangement with Tri-Tool Inc. for use of their equipment to cut a canister open and had developed a written purchase order for the equipment and support from Tri-Tool to be issued if needed. A video tape of the cutting open of a canister using the Tri-Tool equipment was observed by the NRC. Tri-Tool, Inc had extensive experience cutting large piping and contaminated components at nuclear plants and had been selected at other NRC licensed ISFSIs to provide equipment and support for opening loaded canisters.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) Procedure SO23-X-9.1 Reviewed: "Dry Cask Storage Unloading," Rev 0 (c) Purchase Order 8X073101 to Tri-Tool Inc for

"Equipment and Support to Cut Open a Loaded Canister" Category: Pre-Operational Tests Topic: Transfer Route Verification Reference: FSAR 1029, Sect 9.2.1.2(C)

Requirement The transfer cask is placed on the trailer, which is moved to the ISFSI along the predetermined route and aligned with a concrete storage module. Compatibility of the transfer trailer with the transfer cask, verification of the transfer route to the ISFSI and maneuverability within the confines of the ISFSI are verified.

Finding: On June 27-28, 2003 and August 18-19, 2003 the licensee demonstrated the movement of the transfer cask containing a weighted canister from the Unit 3 facility to the ISFS The transport route used for the demonstration was the same route that would be used for movement of the actual casks. The demonstration included maneuvering the transport trailer and transfer cask at the ISFSI to align with the concrete storage module for insertion of the canister. The licensee had developed Calculation C-296-04.01 to evaluate the proposed route for transporting the spent fuel from Units 2 and 3 to the ISFSI. The calculation included an evaluation of the dimensional clearances and underground utilities and structures along the transport route. The calculation concluded that the proposed transfer route was acceptable and would not cause damage to underground utilities and structures. The calculation recommended that plates be installed over the rail spur switches. The NRC confirmed that the plate(s), with a minimum 1-inch thickness, were installed over the switches in accordance with the calculation recommendations. During the preoperational inspection, an NRC inspector conducted a detailed assessment and walkdown of the proposed route with the Fuel Movement Project Manager. The route clearances were found to be acceptable to allow for prime mover/transfer trailer passage. Gates were wide enough to accommodate the prime mover and transfer trailer. The transfer route took the cask near the paint shop and a flammable liquid storage area. These facilities had been analyzed by the licensee to verify that the flammable liquids allowed for storage in the area were below the 300 gallon limit in FSAR Section 11.2.4.1. For Unit 1, the route was short since the Unit 1 spent fuel pool was near the ISFSI pad. The route for moving the Unit 1 spent fuel to the ISFSI was reviewed with the licensee. No issues were identified with any of the planned Page 33 of 70

transport routes.

Documents Calculation No. C-296-04.01 "Dry Cask Storage Route," Rev 0 Reviewed:

Category: Pre-Operational Tests Topic: Weighted Canister for Dry Run Reference: FSAR 1029, Sect 1.2.2.1 Requirement A dry run using a loaded canister with mock-up fuel assemblies will be performed prior to loading the first canister to demonstrate the adequacy of training, familiarity of system components and operational procedures. Mock-up fuel assemblies shall provide a representation of the maximum fuel assembly cross sectional envelope and provide a reasonable approximation of fuel assembly length and weight. The licensee shall determine the quantity of mock-up fuel assemblies for the dry run to demonstrate that the loading and unloading processes are sound and the operations personnel are adequately trained. (Note: Table 3.2-1 of the FSAR provides the weight of the canister)

Finding: The licensee performed a number of heavy lift and cask movement activities with a canister loaded with mock fuel assemblies which approximated the weights listed in Table 3.2-1 of the Final Safety Analysis Report. The loaded canister was used for testing of the Unit 3 fuel building crane and the transport trailer. The NRC observed the lowering of the weighted cask from the Unit 3 fuel building onto the transport trailer, moving the canister to the ISFSI, then inserting the canister into the concrete storage module on June 27-28, 2003 and on August 18-19, 2003.

Documents (a) Ederer Single Failure Proof Trolley Project Specification SO123-209-01-M331, Reviewed: Procedure 251-F2704 "Site Pre-Operational Test for X-SAM System," Rev 2 (b) Ederer Single Failure Proof Trolley Project Specification SO123-209-01-M316, Procedure 250-F2704 "Field Acceptance, Inspection & Testing," Rev 2 (c) SO123-I-7.24 "Rigging Manual" (d) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Category: Procedures & Tech Specs Topic: Annulus Seal Reference: FSAR 1029, Sect 2.3.2.1 & 3.4.1 Requirement The transfer cask/canister annulus opening at the top of the cask must be sealed using an inflatable seal to prevent pool water from entering the annulus. The annulus is filled with uncontaminated, demineralized water prior to placing the transfer cask/canister in the fuel pool.

Finding: Step 6.2.57 of Procedure SO23-X-9 required filling the annulus with nuclear service water. Steps 6.2.61 and 6.2.62 inserted the annulus seal and pressurized it to 20 to 25 pounds/square inch. These activities were performed prior to the cask being placed in the spent fuel pool.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Procedures & Tech Specs Topic: Cask Surface Wash After Withdrawal from Pool Reference: FSAR 1029, Sect 8.1.1.2 (13)

Requirement As the cask is raised from the pool, the exposed portions of the cask shall be sprayed with demineralized wate Page 34 of 70

Finding: Steps 6.5.19 and 6.5.20 of Procedure SO23-X-9 required spraying the transfer cask, yoke and yoke extension with nuclear service water when being lifted out of the spent fuel pool.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Procedures & Tech Specs Topic: Cask Surface Wash Prior to Insertion into Pool Reference: FSAR 1029, Sect 8.1.1.2 (2)

Requirement As the cask is lowered into the pool, the exterior surface of the cask must be sprayed with demineralized water to minimize surface adhesion of contaminated particles.

Finding: Step 6.3.15 of Procedure SO23-X-9 contained the requirement to spray the transfer cask with nuclear service water as it was being lowered into the spent fuel pool.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Procedures & Tech Specs Topic: Lifting Yoke Wash Reference: FSAR 1029, Sect 8.1.1.2 (4)

Requirement After disengaging the lifting yoke from the transfer cask (positioned in the transfer pit),

the lifting yoke has to be sprayed with demineralized water if it is raised out of the fuel pool.

Finding: Step 6.3.27 of Procedure SO23-X-9 required the yoke and yoke extension to be sprayed with nuclear service water while being lifted from the spent fuel pool after disengagement from the transfer cask.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Procedures & Tech Specs Topic: Operating Procedures Reference: CoC 1029, Condition 2 Requirement Written operating procedures shall be prepared for cask handling, loading, movement, surveillance and maintenance. The user's site-specific written operating procedures shall be consistent with the technical basis described on Chapter 8 of the Safety Analysis Report.

Finding: Procedures for cask handling, loading, movement, surveillance and maintenance had been developed by the licensee. Most of these activities were described in Procedure SO23-X-9. Additional procedures developed by the licensee related to these activities are listed in these inspector notes under the category "Procedures and Tech Specs" and topic "Written Procedures Required." Selected portions of the technical basis requirements in Section 8 of the Final Safety Analysis Report were reviewed and found to be consistent with the procedures.

Documents (a) List entitled "Station Program & Procedures Generated or Modified to Support Reviewed: ISFSI," dated August 5, 2003 (b) Procedure SO1-207-1-M135 "Advanced NUHOMS System Final Safety Analysis Report," Rev 3 (c) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Page 35 of 70

Category: Procedures & Tech Specs Topic: Programs Required for ISFSI Operations Reference: CoC 1029, Tech Spec 5.2 Requirement The licensee shall implement the programs listed in Technical Specification Programs listed include a safety review program, training program, radiological environmental monitoring program, radiation protection program and advanced thermal monitoring program for the concrete storage modules.

Finding: The licensee had established the following procedures to meet the requirements in Technical Specification 5.2: (a) Procedure SO123-XV-44.2 "10 CFR 72.48 Program Implementation Guidelines," Rev. 0 covered safety reviews (b) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 covered training, (c) Procedure SO123-IX-1.3 "Environmental TLD Exchange," Rev 11 covered radiological environmental monitoring, (d) Procedure SO123-VII-20.6 "External Occupational Exposure Monitoring," Rev 4 covered radiation protection, and (e) Procedure SO123-X-9.6 "Excessive AHSM Temperature," Rev 0 (draft) covered AHSM (concrete storage module) thermal monitoring.

Documents List entitled "Station Program & Procedures Generated or Modified to Support ISFSI,"

Reviewed: dated August 5, 2003 Category: Procedures & Tech Specs Topic: Stuck Fuel Assembly During Loading Reference: None Requirement During the loading of the fuel assemblies into the canister, it is possible that a fuel assembly could become stuck. Removing the fuel assembly requires establishing a lifting limit to prevent damage to the assembly.

Finding: There were no procedural requirements for dealing with a stuck fuel assembly during the loading of the SONGS Unit 1 fuel into the canisters. The Unit 1 fuel assemblies had an outside dimension of 7.76 inches while the canister basket slot had an inside dimension of 8.86 inches. With this tolerance and the observation, made during the visual inspection of the fuel, that there was no significant bowing of any of the Unit 1 fuel assemblies, the licensee did not consider a stuck fuel assembly as a credible event.

Documents None Reviewed:

Category: Procedures & Tech Specs Topic: Top Fuel Spacers Reference: CoC 1029, Tech Spec 4.2.5 Requirement Top fuel spacers are required to be located above each intact fuel assembly stored in the canister (the failed fuel can design includes an integral top fuel spacer and therefore does not require a top fuel spacer).

Finding: Steps 6.4.7 and 6.4.8 of Procedure SO23-X-9 required upper spacer assemblies to be installed and verified as properly seated below the top of the guide sleeves before the shield plug was installed in step 6.5.14.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Page 36 of 70

Category: Procedures & Tech Specs Topic: Top Shield Plug Inspection Reference: FSAR 1029, Sect 8.1.1.2 (9) & (12)

Requirement Visually verify that the top shield plug is properly seated on the canister. Raise the transfer cask to the pool surface. Prior to raising the top of the cask above the water surface, stop vertical movement. Inspect the top shield plug to re-verify that it is properly seated onto the canister.

Finding: Procedure SO23-X-9, Step 6.5.14 required the shield plug to be visually inspected to verify proper seating during placement into the canister after the spent fuel had been loaded. As the canister was being removed from the spent fuel pool, Step 6.5.21 required visual verification of proper seating of the shield plug. This was performed immediately after the canister was stopped just below the surface of the spent fuel pool, a radiation reading was performed and health physics personnel approved the continued raising of the canister from the pool.

Documents Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 Reviewed:

Category: Procedures & Tech Specs Topic: Water Level Adjustments for Spent Fuel Pool Reference: FSAR 1029, Sect 8.1.1.1 (16)

Requirement Prior to the cask being placed into the fuel pool, the water level in the pool should be adjusted as necessary to accommodate the cask/canister volume. If the water placed in the canister's cavity was obtained from the fuel pool, a level adjustment may not be necessary.

Finding: Adjustment to the spent fuel pool water level during lowering of the transfer cask/canister into the pool was not necessary. The water used to fill the canister prior to lowering into the cask loading pit was taken from the spent fuel pool. An evaluation of the need to adjust the water level of the spent fuel pool during loading activities had been completed in response to Action Request #020501126. The evaluation determined that the water level changes in the spent fuel pool during cask loading activities would not result in the pool water level being out of the normal band for the pool.

Documents (a) Action Request #020501126, (b) Procedure SO23-X-9 "Dry Cask Storage Loading,"

Reviewed: Rev 1 Category: Procedures & Tech Specs Topic: Written Procedures Required Reference: CoC 1029, Tech Spec 5.1; 10 CFR 72.212(b)(9)

Requirement Written procedures shall be established, implemented, and maintained covering the 14 topical areas listed in Technical Specification 5.1. Topics include ISFSI operations, health physics. maintenance, quality assurance, records, security, emergency operations, etc.

Finding: The topical areas listed in Technical Specification 5.1 were covered by the following procedures or program documents: (a) Procedure SO123-OP-1 "Operations Division Program," Rev 5 and the Topical Quality Assurance Manual, Chapter 1-B

"Organization" covered organization and management, (b) Procedure SO1-9-18

"Advanced Horizontal Storage Modules (AHSM) Monitoring & System Operation," Rev 0 (draft) covered routine ISFSI operations, (c) Procedure SO1-9-18 "Advanced Page 37 of 70

Horizontal Storage Modules (AHSM) Monitoring & System Operation," Rev 0 (draft)

covered alarms and annunciators, (d) Procedure SO123-VIII-10 "Emergency Coordinator Duties," Rev 16 covered emergency operations, (e) Procedure SO123-XV-44.2 "10 CFR 72.48 Program Implementation Guidelines," Rev 0 covered design control and facility change/modification, (f) Procedure SO1-12.1-9 "Inspection of Advanced Horizontal Storage Modules (AHSM)," Rev 0 (draft) covered control of surveillances and tests, (g)

Procedure SO123-XV-44 "10 CFR 50.59 and 72.48 Program," Rev 5 covered control of special processes, (h) Procedure SO123-I-7.102 "Dry Fuel Storage Special Lifting Devices," Rev 0 (draft) covered maintenance, (i) Procedure SO123-VII-20.6 "External Occupational Exposure Monitoring," Rev 4 covered health physics, including ALARA practices, (j) Procedure SO123-X-1.7 "Special Nuclear Material Accountability," Rev 7 covered special nuclear material accountability, (k) the Topical Quality Assurance Manual and Procedure SO123-XII-18.1 "Audit Program," Rev 8 covered quality assurance, inspections and audits, (l) Procedure SO123-IV-6.3.7 "ISFSI Security Test, Inspection and Surveillance Program," Rev 0 (draft) covered physical security and safeguards, (m) Procedure SO123-VI-1 "Review/Approval Process for Orders, Procedures and Instructions," Rev 21 covered records management, (n) Procedure SO123-0-14 "Notification and Reporting of Significant Events," Rev 10 covered reporting, (o) Procedure SO123-XV-44.2 "10 CFR 72.48 Program Implementation Guidelines," Rev 0 covered safety review, (p) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 covered training, (q) Procedure SO123-IX- "Environmental TLD Exchange," Rev 11 covered radiological environmental monitoring, (r) Procedure SO123-VII-20.6 "External Occupational Exposure Monitoring," Rev 4 covered radiation protection, and (s) Procedure SO123-X- "Excessive AHSM Temperature," Rev 0 (draft) covered AHSM Thermal Monitoring.

Documents (a) List entitled "Station Program & Procedures Generated or Modified to Support Reviewed: ISFSI," dated August 5, 2003 (b) "Topical Quality Assurance Manual," Rev 18 (c)

Procedure SO123-OP-1 "Operations Division Program," Rev 5 (d) Procedure SO123-X-1.7 "Special Nuclear Material Accountability," Rev 7 (e) Procedure SO123-0-14

"Notification and Reporting of Significant Events," Rev 10 Category: QA Topic: Approved QA Program Reference: 10 CFR 72.140(d)

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.

Finding: Southern California Edison notified the NRC by letter dated September 27, 2002 of their plans to apply the NRC approved Part 50, Appendix B Quality Assurance Program for the reactor facilities (Dockets 50-206, 50-361 and 50-362) to the SONGS ISFSI. The Part 50 Quality Assurance Program had been originally submitted to the NRC on October 27, 1976 and amended on March 24, 1977. NRC had approved the Part 50 Quality Assurance Program on June 28, 197 Page 38 of 70

Documents SCE letter to NRC entitled "Quality Assurance Program for ISFSI," dated September 27, Reviewed: 2002 (NRC Adams Document ML022740187)

Category: QA Topic: Canister Construction Onsite Reference: CoC 1029, Condition 4 Requirement Activities in the areas of design, purchase, fabrication, assembly, inspection, testing, operations, maintenance, repair, modifications of structures, systems and components and decommissioning that are important to safety shall be conducted in accordance with a Commission approved quality assurance program.

Finding: The NRC's Spent Fuel Project Office performed an inspection on March 3-6, 2003 of the quality assurance activities related to the fabrication of the canisters at the Southern California Edison facility near the SONGS site. Southern California Edison was constructing their own canisters under the oversight of Transnuclear, Inc. The fabrication activities were being performed under the Southern California Edison quality assurance program. Overall, fabrication activities by Southern California Edison and the oversight activities by Transnuclear, Inc were assessed to be good. Specific areas inspected included overall quality assurance controls including audits and corrective action processes, design controls including translation into fabrication drawings, material procurement, and fabrication activities including welding and nondestructive examinations. No significant adverse findings were noted and no cited or non-cited violations were identified.

Documents NRC Inspection Report 72-1029/2003201 "Transnuclear Inc," dated March 6, 2003 Reviewed: (NRC Adams Document ML030940163)

Category: QA Topic: Control of Measuring and Test Equipment Reference: 10 CFR 72.164 Requirement The licensee shall establish measures to ensure that tools, gauges, instruments and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specific periods to maintain accuracy within necessary limits.

Finding: The licensee was implementing a program for measuring and test equipment for dry cask storage activities under the reactor facility Part 50 quality assurance program. Procedure SO23-X-9 was reviewed to identify test equipment that would be used to verify compliance with specific technical specification requirements or important to safety design aspects of the cask design. The helium leak detectors (Mass Spectrometer Model 959 and 959D) used to perform the nondestructive testing on the canister lid welds was selected for a detail review. Based on the requirements in Vendor Manual N , "Operations Manual/Varian Vacuum Technologies - Model 959 and 959D Mass Spectrometer," the licensee developed a calibration procedure for this detector as Attachment 1, "Startup and Calibration Procedure for Varian Model 959," to Procedure SO1-XII-9.103. The calibration procedure and range of accuracy of the detector was reviewed against the vendor manual. Calibration records for the helium standards used for the helium leak detection equipment were verified as current. A second instrument selected for review was the pressure gauge on the vacuum drying skid used to verify the vacuum pressure during drying of the canister. This was Equipment ID #

S1232208MW002 calibrated under Maintenance Order #03080092000. The required Page 39 of 70

readings and tolerances were specified in the calibration sheet and included calibration at the 3 torr range specified in Technical Specification 3.1.1. The calibration as-found and as-left values were 3.03 torr. In addition to the pressure gauge on the vacuum drying skid, SONGS used two Heise PTE-1 vacuum gauges attached to the skid to provide a real time verification of the accuracy of the reading on the skid mounted gauge. The vacuum gauges used Heise Model HQS-2 pressure modules. The calibration sheets for the pressure modules were reviewed. The calibration was performed on the 0.025%

pressure modules (HQS-24216 and HQS-24217) from 1 psia to 10 psia. Converting this to torr resulted in a calibration range from 51.7 torr to 517 torr. These calibration points were high compared to the 3 torr acceptance value in the technical specification even though the gauges could read well below the 3 torr limit. It was also noticed that the Edison ESI calibration report for pressure module HQS-24216 had extremely accurate readings to four significant digits of accuracy between the nominal value and the as-found value on all 10 calibration points. The calibration records from the manufacture, Dresser Inc., were requested for the factory calibration performed. The Dresser calibration certificate report dated August 6, 2003 confirmed the same calibration values, supporting the results of the calibration check performed by Edison ESI.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) "Topical Quality Reviewed: Assurance Manual (TQAM)," Change 03-02 (c) Vendor Manual No. 699909750

"Operations Manual/Varian Vacuum Technologies - Model 959 and 959D Mass Spectrometer" (d) Procedure SO1-XII-9.103 "Helium Leak Detection of Advance NUHOMS 23PT1 Dry Shielded Canisters After Fuel Loading and Final Weld Closure" (e) Attachment 1 to Procedure SO1-XII-9.103 "Startup and Calibration Procedure for Varian Model 959" (f) Edison ESI "Calibration Report for Heise HQS-2 Pressure Modules," dated September 8, 2003 for Serial Nos. HQS-24216, HQS-24217, HQS-24211 and HQS-24212 (g) Maintenance Order 03080092000 "Calibration of Miscellaneous ISFSI Instrumentation," completed September 8, 2003 for the Vacuum Drying Skid Torr Gauge (h) Heise Operating Manual "PTE-1 Handheld Calibrator with HQS-2 Pressure Module," Rev 1 (i) Dresser, Inc "Calibration Certification Report for HQS-2 Serial No 24216," dated August 6, 2003 Category: QA Topic: Corrective Actions Reference: 10 CFR 72.172 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.

Finding: The licensee's corrective action reports, audit finding reports, vendor nonconformance reports and supplier deviation requests were reviewed to verify that the licensee was applying the SONGS corrective action program appropriately to various issues identified during construction of the cask components. Items reviewed were found to have been properly evaluated with adequate corrective actions identified to address the concerns raised. The licensee had a program for tracking non-conforming items identified by the vendors, in addition to a process to relay SONGS identified non-conforming items back Page 40 of 70

to the vendor for resolution. Audits performed by the licensee of vendors included a review of non-conformance items that had been identified by Southern California Edison and referred back to the vendor. The review included the adequacy of the corrective actions taken by the vendor for these issues.

Documents (a) Procedure SO123-XV-50 "Corrective Action Process," Rev 5 (b) Procedure SO123-Reviewed: XXXII-2.27 "Supplier Deviation Requests," Rev 1 (c) Transnuclear Nonconformance Report No. F-03.042 "Batching Tolerances Were Exceeded for Aggregates and Cement in Some of the Batches," dated May 12, 2003 (d) Transnuclear Nonconformance Report No. F-03.073 "During Setting of Roof to Base on AHSM002 (SCE ID) Grout and Roof Alignment Conflicts Will Not Let the Roof Bolt at the South-East Location to Mate With the Richmond Anchor Due to a Slight Vertical Offset," dated July 24, 2003 (e)

Transnuclear Nonconformance Report No. F-03.019 "Dimensional Inspection of the DSC Basket Assembly Support Rods Did Not Meet the Drawing Requirements," dated May 5, 2003 (f) SCE Supplier Deviation Request No. 48501-03095 "Revise SO1-207-01, Section 3.11.3.1, 5th and 7th Paragraphs to Reflect the Updated Crane Capacity of 105 tons instead of the Currently Specified 100 Tons," dated May 14, 2003 Category: QA Topic: Design Change Controls Reference: 10 CFR 72.146(c)

Requirement The licensee shall subject design changes, including field changes, to design control measures commensurate with those applied to the original design.

Finding: The licensee was implementing a design change control process using the reactor facility Part 50 quality assurance program and Procedures SO123-XXIV-10.1 and SO123-XV-51. Selected engineering change packages were reviewed to verify that the licensee was applying the design change program appropriately. The engineering change packages included the reason for the change, provided an assessment of the effect of the change on the original design and provided a determination that the original design analysis was still valid. Engineering evaluations were found to be thorough and changes needed to drawings and procedures were identified and completed. The licensee also had a program for tracking the design changes.

Documents (a) Procedure SO123-XXIV-10.1 "Preparation, Review, Approval, Issuance, Reviewed: Implementation, and Closure of Engineering Change Packages (ECPs) and Engineering Change Notices (ECNs)," Rev 5 (b) Procedure SO123-XV-51 "Identifying and Assessing Impact to Site Programs and Procedures," Rev 4 (c) "Topical Quality Assurance Manual," Rev 18 (d) Procedure SO123-XXIV-37.30.41 "Specification/Mini-Specifications," Rev 3 Category: QA Topic: Handling, Storage and Shipping Control Reference: 10 CFR 72.166 Requirement The licensee shall establish measures to control, in accordance with work and inspection instructions, the handling, storage, shipping, cleaning and preservation of material and equipment to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere and specific moisture content and temperature levels must be specified and provide Page 41 of 70

Finding: The licensee had adequately established measures to control the handling, storage, shipping, cleaning, and preservation of material and equipment used for the fabrication and maintenance of the ISFSI consistent with the requirements specified in Procedures SO123-XI-3.2, SO123-I-1.33 and SO123-XI-3.3. A tour was conducted of the storage locations for ISFSI equipment in the warehouse, the fuel handling building and the ISFSI facility. The warehouse had a separate area dedicated for the storage of ISFSI related equipment and components. All equipment was properly stored and in good physical condition. Appropriate measures were being implemented to ensure that the canisters, lids and other hardware were stored to preclude physical damage or material deterioration.

Documents (a) Procedure SO123-XI-3.2 "Storage of Quality Affecting Items," Rev 3-4 (b) Procedure Reviewed: SO123-I-1.33 "Storage of Quality-Affecting Items during Staging Period," Rev 2 (c)

Procedure SO123-XI-3.3 "Packaging and Preservation Requirements for Storage and Shipment of Quality-Affecting Items," Rev 3-2 Category: QA Topic: Important to Safety Items Reference: FSAR 1029, Table 2.5-1; Sect 13.2 Requirement Components "Important to Safety" are listed in Table 2.5-1. During the design process, items that are considered important to safety are further categorized using a graded quality approach. When the graded quality approach is used, a list shall be developed for each important to safety item which includes an assigned quality category consistent with the item's importance to safety. Quality categories are determined based on guidance provided in NUREG/CR-6407. The categories are A, B and C. For "safety related" items the quality assurance program is applied as described for Category A items.

Finding: The components identified as "important to safety" in the FSAR were incorporated into the SONGS quality assurance program. Drawing M-37560 was updated on April 12, 2002, with ECN A12371 to incorporate the important to safety components identified in the Final Safety Analysis Report into the Unit 1 Q-List. All of the components identified in the Final Safety Analysis Report were treated as "safety related" under the quality assurance program and Classification Category A as described in NUREG/CR-6407.

Documents (a) Drawing M-37560 "San Onofre Nuclear Generating Station Unit 1 Q-List," Rev 8 (b)

Reviewed: Engineering Change Notice (ECN) A12371 to Drawing M-37560, dated April 12, 2002 (c) Topical Report SCE-1-A "Quality Assurance Program," Amendment 22 (d) "Topical Quality Assurance Manual," Rev 18 Category: QA Topic: Inspection and Testing Status Reference: 10 CFR 72.168(a)

Requirement The licensee shall establish measures to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable means, the status of inspections and tests performed upon individual items. These measures must provide for the identification of items which have satisfactorily passed required inspections or tests where necessary to preclude inadvertent bypassing of the inspection or test.

Finding: The licensees procurement program provided appropriate guidance for determining the Page 42 of 70

inspection and testing requirements for procured products related to the dry cask storage project. Interviews of licensee personnel, review of selected purchase orders and associated design specifications, and verification that appropriate inspection and testing requirements were developed for selected components and equipment was complete Selected components examined were found to be appropriately tagged and stored. A program was in place to track the inspection and testing requirements for these components.

Documents (a) Material Support Order SO123-MS-1 "Material Support Program," Rev 5 (b)

Reviewed: Procedure SO123-XII-7.12 "Source Verification," Rev 3 Category: QA Topic: Nonconforming Material and Parts Reference: 10 CFR 72.170 Requirement The licensee shall establish measures to control materials, parts or components that do not conform to their requirements in order to prevent their inadvertent use or installation. These measures must include procedures for identification, documentation, segregation, disposition and notification to affected organizations. Nonconforming items must be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.

Finding: The licensee implemented the ISFSI nonconformance material and parts program using the reactor facility Part 50 nonconformance procedures. Any material or equipment received in the warehouse that was identified as nonconforming was segregated and appropriately tagged. An action request was generated in accordance with Procedure SO123-XV-5 to ensure appropriate tracking and resolution of the nonconformanc Nonconforming material and equipment related to the ISFSI was verified by the NRC as being properly segregated and tagged in the warehouse to prevent inadvertent use or installation. Selected action requests associated with the concrete storage modules were reviewed. The nonconformance conditions were limited to minor surface defects and spalling of the concrete. The identification, documentation and disposition of the nonconforming conditions were reviewed and determined to be in accordance with the licensee's program.

Documents (a) Procedure SO123-XV-5 "Nonconforming Material, Parts or Components," Rev 16 Reviewed: (b) "Topical Quality Assurance Manual," Rev 18 (c) Procedure SO123-XI-3.2 "Storage of Quality Affecting Items," Rev 4 (d) Procedure SO123-XII-20.4 "Receiving Inspection," Rev 7-2 (e) Procedure SO123-XXXII-3 "Warehouse Nonconformances,"

Rev 1-4 Category: QA Topic: Operating Status Reference: 10 CFR 72.168(b)

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.

Finding: The licensee identified the operating status of ISFSI related equipment using the existing reactor facility Part 50 work authorization program. Attachment 3 of Procedure SO123-XX-5 contained examples of the type of work that did not require formal processing Page 43 of 70

through the work authorization program. This work was limited to minor repairs or maintenance that had essentially no impact on the operating status of safety-related equipment. The vacuum drying system and the helium leak detection system were selected for inspection and found to be properly tagged and labeled for operational us The vacuum drying hose connections and the helium supply lines were properly tagged, labeled, color coded and fitted with unique connectors.

Documents (a) Procedure SO123-XX-4 "SONGS Work Control," Rev 7 (b) Procedure SO123-XX-5 Reviewed: "Work Authorizations," Rev 12-2 Category: QA Topic: Procurement Controls for Material Reference: 10 CFR 72.154(a)/(b)/(c)

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 ISFS The effectiveness of the control of quality by contractors/subcontractors shall be assessed at intervals consistent with the importance, complexity and quantity of the product or service.

Finding: The licensees reactor facility Part 50 program for controlling the procurement of materials was applied to the dry cask storage program. The licensees quality assurance personnel monitored all stages of the procurement process. This included reviewing purchase orders to ensure that all the correct specifications were being applied and performing receipt inspections to verify that the product met the specifications in the purchase orders. Audits of vendors were performed to verify that the vendor adhered to the manufacturing requirements set forth in the purchase order. The licensee and the vendors were documenting problems and errors identified during the manufacturing and shipping process and appropriately tracking the review, disposition and closure of these issues. Due to a number of problems at one of the vendors performing concrete storage module construction, the licensee assigned two quality assurance personnel to the vendor's facility. This level of responsiveness by the licensee to correct quality affecting issues resulted in a significant performance improvement by the vendor. For the construction of the canisters, Southern California Edison was in a unique relationship with the cask designer, Transnuclear, Inc. The canisters were being built at the SONGS site by Southern California Edison workers. This required Southern California Edison to function as a "vendor" to Transnuclear, Inc. As a result of this relationship, Southern California Edison established a separation and independence between the SONGS employees that were subcontracted to Transnuclear and the employees involved in inspecting and accepting the procured and manufactured products. This separation of the two organizations was reviewed and found to be appropriate.

Documents (a) Material Support Order SO123-MS-1 "Material Support Program," Rev 5 (b)

Reviewed: Procedure SO123-XII-7.12 "Source Verification," Rev 3 (c) Procedure SO123-XII-18.19

"Supplier Audits," Rev 6 (d) Procedure SO123-XXIV-37.30.41 "Specifications/Mini-Specifications," Rev 3 (e) Procedure SO123-XXIV-1.1 "Document Review and Approval Control," Rev 8 Page 44 of 70

Category: QA Topic: QA Audits Reference: 10 CFR 72.176 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.

Finding: The licensee was implementing a comprehensive audit program of dry cask storage activities under the reactor facility Part 50 quality assurance program. Due to the complexity of the dry cask storage project, dedicated quality assurance personnel were assigned to work exclusively on this effort. This was most clearly noted in the area of procurement and receipt inspection. The licensee had also assigned quality assurance personnel to one of the vendor facilities to monitor that facilitys day-to-day activities and to verify the vendor's adherence to the SONGS quality assurance program requirements. Review of selected vendor audits performed by the licensee and third party audits of on-site activities performed by the licensee found the audit program to be effectively implemented. Deficiencies identified in these audits were well characterized. The licensee had a program for tracking, addressing and closing audit items.

Documents (a) Procedure SO123-XII-18.19 "Supplier Audits," Rev 6 (b) Audit No. SCES-002-03 Reviewed: "Design Engineering," dated July 23, 2003 (c) Procedure SO123-XII-2.19 "Qualification and Certification of Auditing Personnel" (d) Evaluation of Supplier Audits/Surveys Performed by Contractors, Consultants Utilities or Licensees (e) Procedure SO123-XI-8

"Supplier Evaluation and Qualification," (f) Topical Quality Assurance Manual, Chapter 1-E "Audits," Rev 18 Category: QA Topic: Test Controls Reference: 10 CFR 72.162 Requirement The licensee shall establish a test program to ensure that all testing, required to demonstrate that the structures, systems, and components will perform satisfactorily in service, is identified and performed in accordance with written test procedures. The test procedure must include provisions to ensure that all prerequisites for the given tests are met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions. The licensee shall document and evaluate the test results to ensure that test requirements have been satisfied.

Finding: The Topical Quality Assurance Manual had been used by the licensee to develop testing, handling and storage criteria for the components and equipment in the dry cask storage program. Procedure SO23-X-9 was reviewed against the licensees 10CFR Part 50 quality assurance program, as described in Topical Report SCE-1-A, to determine if the various test requirements for the dry cask storage loading met quality assurance program requirements. As part of this review, vendor recommendations and requirements were evaluated concerning equipment testing, handling and storage to verify that requirements had been incorporated into the licensee's programs.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) Topical Report SCE-1-Reviewed: A "Quality Assurance Program," Amendment 22 (c) "Topical Quality Assurance Manual (TQAM)," Change 03-02 Page 45 of 70

Category: QA Topic: Visual Examination of Canisters/Storage Modules Reference: FSAR 1029, Sect 9.1.1 Requirement Upon arrival at the site, the canisters and concrete storage modules are inspected to ensure that they have not been damaged during shipment. Conditions which are not in conformance with the drawings and specifications will be repaired or evaluated, in accordance with 10 CFR 72.48, for the effect of the condition on the safety function of the component.

Finding: The licensee had implemented a program to visually inspect the canisters and the concrete storage modules upon their arrival at the site. Procedure SO123-XII-2 outlined the scope of the visual examinations and included provisions for documenting and tracking any non-conformance conditions that were identified. Receipt records for the concrete storage modules currently on-site that had been receipt inspected by the licensee were reviewed. Damage that was noted was limited to minor spalling of the concrete and surface anomalies. All damaged conditions identified by the receipt inspections were documented in the licensees corrective action system and tracked to resolution. Repairs to the modules were made in accordance with Procedure SO1-207-1-M226. At the time of the inspection, the licensee had not received any canisters.

Documents (a) Procedure SO123-XII-20.4 "Receiving Inspection," Temporary Change Notice 7-2 Reviewed: (b) Receiving Inspection Data Reports (RIDR) RSO-1364-03-00, RSO-1061-03-00, RSO-1419-03-00, RSO-1365-03-00, RSO-0427-03-00, and RSO-0623-03-00 (c) Procedure SO1-207-1-M226 "Field Erection of NUHOMS Precast AHSMs," Rev 1 Category: Radiological Topic: ALARA Dose Estimates Reference: FSAR 1029, Sect 10.3.1.1 Requirement The estimated occupational exposures to ISFSI personnel during loading, transfer, and storage of the canister are expressed in Table 10.3-1 of the FSAR.

Finding: Table 10.3-1 of the Final Safety Analysis Report showed the estimated occupational exposures to ISFSI personnel during loading and transfer operations. The design occupational dose for each canister was estimated to be 3118 person-millirems. The highest dose activity was expected to be non-destructive testing. Based in part on actual experience at other sites, the licensee estimated that occupational doses would be much lower than design doses. As an example, the licensee estimated the occupational doses for the first canister would be 598 person-millirems. The lowered estimate was based in part on a realistic estimate of the amount of time required to conduct non-destructive examination activities, including dye penetrant testing.

Documents (a) Docket 72-1029 "Transnuclear Advanced NUHOMS Final Safety Analysis Report,"

Reviewed: Rev 0 (b) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 Category: Radiological Topic: Cask Radiation Levels Prior to Drying Reference: FSAR 1029, Sect 8.1.1.3 (1)

Requirement After the cask is moved into the decon area, the radiation levels along the perimeter of the cask must be checked and temporary shielding may be installed as necessary to minimize personnel exposur Page 46 of 70

Finding: The Health Physics Work Control Plan (WCP)03-003, Section 9.B required monitoring the cask radiation levels during removal of the cask from the spent fuel pool and placement in the washdown area. Once the cask was in the washdown area, radiation levels were measured around the perimeter of the cask and shielding was used, as necessary, to reduce worker exposures. The licensee performed the required survey of the first loaded cask on September 13, 2003 and documented the survey as Survey N . The canister was full of water and was in the cask washdown area. The highest beta/gamma contact reading was 40 mrem/hr approximately mid-way up the cask. The highest neutron dose rate was 1.8 mrem/hr. The highest removable contamination level measured prior to decon was 5,000 disintegrations per minute (dpm)

beta/gamma. No alpha contamination was detected.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: Contamination Survey of Canister Reference: CoC 1029, Tech Spec 5.2.4(c), FSAR 8.1.1.3 (5)

Requirement Following placement of each loaded transfer cask into the cask decontamination area prior to transfer to the ISFSI, the canister smearable surface contamination levels on the outer surface of the canister shall be less than 2,200 dmp/100 square cm from beta and gamma emitting sources and less than 220 dpm/100 square cm from alpha emitting sources. Allow water from the annulus to drain out until the water level is approximately twelve inches below the top edge of the canister. Take swipes around the outer surface of the shell (and top cover plate) and check for smearable contamination.

Finding: Procedure SO23-X-9, Step 6.6.3 drained approximately 12 inches from the annulu Step 6.6.5 directed that the exposed outer canister surface and inner transfer cask surface be surveyed. Step 6.6.6 specified the contamination limits listed in Technical Specification 5.2.4(c) of 2,200 dpm/100 square cm beta/gamma and 220 dpm/100 square cm alpha. If the contamination limits were exceeded, Step 6.6.6 directed that the area be decontaminated by hand and if unsuccessful with the hand method, to proceed to procedure SO23-X-9.7. Procedure SO23-X-9.7 Section 6.2 specified decontaminating the annulus by flushing with nuclear service water. The licensee recognized that if the annulus seal had leaked while the canister was in the spent fuel pool, that the water in the annulus would be contaminated and hand decon would not be successful unless the contamination was limited to only the area where the annulus seal had been locate During the contamination survey of the first canister, the licensee cleaned the top of the canister and annulus area before conducting the survey. The purpose of the survey required by the technical specifications was to obtain a reasonable level of confidence that the canister was not contaminated above the shipping limits specified in 49 CFR 173.443. Since attempting to perform a full contamination survey of the canister by taking smears in the annulus with all the water removed could result in personnel exposures from the high radiation levels that were associated with the annulus area, the NRC allows the contamination survey in the accessible areas of the top of the annulus to be considered representative of the entire canister contamination levels. By cleaning this area prior to the survey, the top of the canister would no longer be representative of the contamination on the entire canister. The licensee recovered the cloths used for cleaning the top of the canister and confirmed that the contamination levels were below the Page 47 of 70

technical specification limits. The licensee revised Work Control Plan 03-003 to require the survey prior to cleaning the canister lid area.

Documents (a) SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) Procedure SO23-X-9.7 "Transfer Reviewed: Cask/Dry Shielded Canister Contamination," (Draft) (c) Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Fuel From Units 2/3," Rev 4 Category: Radiological Topic: Controlled Area Radiological Doses Reference: 10 CFR 72.106(a)/(b)/(c)

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 total effective dose equivalent (TEDE), or sum of the deep dose equivalent (DDE) + committed dose equivalent (CDE) of 50 rem, or 15 rem to lens of eye, or 50 rem skin/extremities. 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.

Finding: The licensee had defined the controlled area for the ISFSI in Calculation DSC-001 as the SONGS Part 50 exclusion area. The minimum distance from the ISFSI to the exclusion area boundary was approximately 318 meters. The license had analyzed the worst case accident for the ISFSI and estimated that the dose to an individual would be less than 22 millirem at 100 meters. This value was significantly less than the 5000 millirem limit specified in 10 CFR 72.106. The exclusion area was traversed by the Burlington Northern & Santa Fe railroad, Interstate Highway #5, old U.S. Highway #101, a navigable portion of the Pacific Ocean and a public access sections of the California Parks and Recreation State Beach.Section VIII of DCS-001 described the licensee's authorization to control activities within the exclusion area in accordance with the authority acquired by grant of easement from the United States of America made by the Secretary of the Navy as amended on September 18, 1975. This easement included the authority to remove personnel from the exclusion area. The licensee had established provisions for implementing protective actions in the exclusion area through the site emergency plan and mutual aid agreements with offsite authorities.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) Docket 72-1029 Reviewed: "Transnuclear Advanced NUHOMS Final Safety Analysis Report," Rev 0 Category: Radiological Topic: Dose Rates for Storage Modules and Transfer Cask Reference: FSAR 1029, Sect 5.1 Requirement The maximum, minimum and average calculated dose rates on the surface of the concrete storage module are shown in Table 5.1-2. The dose rates on and around the transfer cask (top, bottom and sides) during fuel loading and transfer operations are shown in Tables 5.1-3 thru 5.1-5.

Finding: The Health Physics Work Control Plan (WCP)03-003 included the tables provided in Section 5.1 of the FSAR. Training on the work control plan provided the dose rate information to the radiation staff. Pre-job briefings were conducted for the workers which included a review of expected radiation levels for the various job task Page 48 of 70

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: Dose to Individuals Beyond Controlled Area Reference: FSAR 1029, Sect 10.2.3 Requirement For a single concrete storage module containing design basis fuel, a minimum distance of approximately 80 meters is necessary to meet the 10 CFR 72.104 annual limits (25 mrem whole body, 75 mrem thyroid and 25 mrem to a critical organ) for a member of the public beyond the controlled area. For a 2 x 10 array of storage modules without site specific shielding, a distance of approximately 200 meters is required.

Finding: Using the Monte Carlo computer code MCNP, the licensee confirmed that the dose limits in 10 CFR 72.104 would not be exceeded at the ISFSI controlled area boundary which is approximately 318 meters at it's nearest location. In the Advanced NUHOMS System FSAR, the computer code MCNP was used to calculate dose rates at locations around a single storage module and around a 2 x 10 array of storage modules. Using the same computer code, a calculation based on site specific parameters at the SONGS ISFSI was conducted. As documented in the licensee's Calculation No. DCS-001, storage of spent fuel at the ISFSI would not result in any effluent discharges and there would be no contribution to the thyroid or organ doses. All doses to the public would be from external radiation. The calculation considered seven different phases of the SONGS ISFSI construction where each phase included an increasing number of storage modules on the ISFSI pad. Phase I consisted of Unit 1 fuel only at the ISFSI. According to the licensee's Phase I computer code calculation, the maximum annual dose from the ISFSI at the controlled area boundary (the same boundary as the 10 CFR Part 50 exclusion area boundary) was determined to be 0.62 millirems. This calculated value was well below the annual whole body limit of 25 millirem specified in 10 CFR 72.104.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) Transnuclear Reviewed: Calculation No. SCE-23.0505 "San Onofre ISFSI Dose Rate Calculation," Rev A Category: Radiological Topic: Highest Potential Dose Streaming Paths Reference: FSAR 1029, Sect 10.3.1.1 Requirement The areas of highest operational dose (potential streaming paths) are the front of a loaded concrete storage module at the air inlet vent, at the transfer cask side surface containing a dry canister during outer cover plate welding and transfer operations and at the transfer cask/canister annulus. Operating procedures and personnel training minimize personnel exposure in these areas.

Finding: Training, pre-job briefings, Work Control Plan (WCP)03-003 and a review of the ALARA post job reports and lessons learned from the cask loading activities at Rancho Seco were the primary means of providing radiological precautions to workers of expected high dose rates. The work control plan included numerous statements concerning the potential for high dose rates during certain activities. These included raising the loaded cask out of the spent fuel pool, performing work near the cask while in the cask washdown area, working around the annulus after the water is lowered in the annulus, manual welding if the automatic welder malfunctions, canister insertion into the concrete storage module, entry into an empty storage modules when the adjacent module Page 49 of 70

was loaded and work activities immediately around the concrete storage module after the canister is loaded. The work control plan provided estimated dose rate information, including highest operational doses during selected loading evolutions. Health Physics technicians demonstrated the use of the work control plan during dry run activitie During fuel loading of the first canister, the health physics staff implemented the work control plan and used the regular health physics program procedures for conducting routine radiological surveys and posting controlled area boundaries. Pre-job briefings were used to provide up-to-date information to workers.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: ISFSI Radiation Safety and ALARA Reference: FSAR 1029, Sect 10.1.1 Requirement The licensee's existing radiation safety and ALARA policies for the plant should be applied to the ISFSI.

Finding: Loading, transfer and storage activities for the ISFSI were performed under the existing SONGS Nuclear Organization Directive D-013 "Radiation Protection," that implemented the corporate management support for maintaining occupational and public doses ALARA. Work Control Plan 03-003, used for the ISFSI work, had incorporated ALARA concepts including requirements for job planning, pre-job briefings, specialized shielding and use of radiation work permits to control access to restricted areas. The licensee had implemented controls requiring specific briefings and review of radiological work permits for different phases of the ISFSI work activities before the computerized system would allow issuance of dosimetry for entry into the radiologically controlled area. Health physics personnel were on-duty during all phases of cask loading and movement to ensure compliance with ALARA requirements. Observation by the NRC during the loading of the first canister verified that health physics personnel were very active in conducting radiation surveys and ensuring that personnel who were not actively performing work remained in low background areas.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: Neutron Energies for Dosimetry Reference: FSAR 1029, Sect 5.2.2 Requirement The neutron source term consists primarily of spontaneous fission neutrons largely from Cm-244 with (alpha, O-18) source terms of lesser importance, both causing secondary fission neutrons. The overall spectrum is well represented by the Cm-244 fission spectrum.

Finding: The licensee had developed a plan for measuring neutron doses to account for the energy spectrum expected from the casks. Occupational neutron dose measurements would consist of a combination of neutron dosimeters and measurement of neutron exposures using portable measuring equipment. Individual Task Assignment 03-007 identified the methods available to estimate neutron dose. The ratio method was selected as the preferred method. The ratio method used a ratio value between the theoretical gamma Page 50 of 70

dose and the neutron doses. Using this ratio, the licensee could estimate the occupational neutron dose based on time spent in an area with known gamma dose rate The licensee developed theoretical ratios based on information provided in the FSA The ratios varied from 0.002 to 1.5, depending on the work in progress and worker location relative to the cask. The licensee planned to confirm the theoretical ratios during loading of the first cask. In accordance with Individual Task Assignment 03-008, the licensee evaluated the current calibration process for the ASP-1 rem ball survey meter. The survey meter was checked using a plutonium-beryllium source which released higher energy neutrons than was expected to be emitted from the Unit 1 fue The licensee determined that no change in their current calibration process was necessary, in part, because the survey meter would tend to overestimate actual dose rates. Per Individual Task Assignment 03-009, the licensee determined a dose conversion factor for neutron dosimeters issued to site workers. A neutron spectrum factor of 15 was chosen. The licensee had reviewed an extensive study conducted at a Virginia nuclear plant which used a conversion factor of 20. The factor of 15 planned for use by SONGS was more conservative.

Documents (a) Individual Task Assignment 03-007 "Determine Method to Estimate Neutron Dose Reviewed: and Control Total Dose During Unit 1 Fuel Handling," dated April 3, 2003 (b) Individual Task Assignment 03-008 "Evaluate Applicability of Current ASP-1 Rem Ball Calibration Process for Unit 1 Fuel Handling Survey Purposes," dated May 6, 2003 (c)

Individual Task Assignment 03-009 "Using Anticipated Unit 1 Fuel Neutron Energy Spectrum, Determine Dose Conversion Factor for Neutron Badge," (undated)

Category: Radiological Topic: Radiological Environmental Monitoring Program Reference: CoC 1029, Tech Spec 5.2.3, 10 CFR 72.104(a)

Requirement A radiological environmental monitoring program will be implemented to ensure that the annual dose equivalent to an individual located outside the ISFSI controlled area does not exceed the annual dose limits specified in 10 CFR 72.104(a).

Finding: The licensee had incorporated the ISFSI Radiological Environmental Monitoring Program into the existing Part 50 Radiological Environmental Monitoring Program. The licensee had augmented the number of direct radiation monitoring devices at the SONGS site to evaluate the radiological environmental impact of the ISFSI. Step 6.1.3.6 of Procedure SO123-IX-1.10 noted that 21 additional thermoluminescent dosimeters (TLDs) had been installed to evaluate the ISFSI pre-operational radiological environment. The additional TLDs had been placed in service during the fourth quarter of 2001. The licensee had collected data for seven quarters prior to the operation of the ISFSI. During the fourth quarter 2003 the licensee had removed the Unit 1 reactor vessel from the Unit 1 containment. Some of the TLD locations had measured increased direct radiation readings as a result of the storage of the reactor vessel near the ISFSI are Excluding the fourth quarter 2003 readings, the TLD quarterly readings ranged from 16.15 to 36.84 millirem. The licensee will report the results of the ISFSI Radiological Environmental Monitoring Program to the NRC as part of the annual report issued for the Part 50 Radiological Environmental Monitoring Program.

Documents (a) Procedure SO123-IX-1.10 "Review, Analysis and Reporting of Radiological Reviewed: Environmental Monitoring Program (REMP) Data," Rev 4 (b) Action Request

030800320-01 "Per NRC ISFSI Pre-Operational Readiness Inspection Report 50-Page 51 of 70

206/2003-10; 72-41/2003-01, Generate TCN or Rev to SO123-IX-1.10" Category: Radiological Topic: Radiological Surveys During Draining the Canister Reference: FSAR 1029, Sect 8.1.1.3 (18)

Requirement Allow compressed gas to force the water from the canister cavity through the siphon port. (Note: This action drains the canister. Should crud be in the canister water from the spent fuel, draining the water from the canister may result in crud moving through the drain line.)

Finding: The Health Physics Work Control Plan (WCP)03-003, Section 9.C required monitoring of the drain line during the cask blowdown to verify that no crud had become dislodged from the spent fuel and passed into the drain line. The work control plan provided contingency actions in case elevated dose rates were detected on the drain line. These included notifying the Fuels Engineer, shielding the line and flushing the line. After the water was drained from the canister, radiation levels increased to as high as 100 mrem/hr beta/gamma and 15 mrem/hr neutron on the side of the canister. This was an increase from the 25 to 40 mrem/hr beta/gamma and less than 1 mrem/hr neutron levels that existed prior to draining the canister.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: Temporary Shielding of Inner Top Cover Plate Reference: FSAR 1029, Sect 8.1.1.3 (7)

Requirement Radiation levels along the surface of the inner top cover plate of the canister should be checked and temporary shielding may be installed as necessary to minimize personnel exposure.

Finding: The Health Physics Work Control Plan (WCP) #03-003 had incorporated instructions to install temporary shielding on the inner top cover plate after the canister was removed from the spent fuel pool. Section 9.B of the work control plan instructed health physics technicians to place flame retardant shielding "snakes" around the annulus gap to help minimize exposure to the welders. The licensee performed the required survey of the first loaded cask on September 14, 2003 and documented the survey as Survey N . Dose rates on the lid were typically below 1 mrem/hr with small areas up to 5 mrem/hr around the annulus gap where the snake had been inserted.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: Top Shield Plug Radiation Monitoring Reference: FSAR 1029, Sect 8.1.1.2 (15)

Requirement Radiation levels shall be checked at the center of the top shield plug and around the perimeter of the cask after the cask is removed from the fuel pool.

Finding: The Health Physics Work Control Plan (WCP)03-003, Section 9.B included instructions to perform a dose rate survey at the center of the shield plug and around the perimeter of the transfer cask as it was being withdrawn from the spent fuel pool. Section 9.B had Page 52 of 70

specified limits of 200 mrem/hr for the gamma dose on the lid and 10 mrem/hr for the neutron dose in the general area occupied by workers. The licensee performed the required survey of the first loaded cask on September 13, 2003 and documented the survey as Survey No. 030913-003. The dose rate on the shield plug was 1.5 mrem/hr beta/gamma and 0.3 mrem/hr neutron. On the sides of the cask as it was being removed from the spent fuel pool, radiation levels were typically between 25 to 40 mrem/hr beta/gamma with neutron levels around 1 mrem/hr. General area radiation levels around the perimeter of the cask during removal from the spent fuel pool were below 1 mrem/hr.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev 2 Category: Radiological Topic: Vacuum Drying Discharge Monitoring Reference: FSAR 1029, Sect 10.1.2 & 2.3.2.2 Requirement Monitoring of the vacuum drying system discharge and diversion to the gaseous radwaste system or other appropriate filtration systems shall be implemented.

Finding: The Health Physics Work Control Plan (WCP)03-003, Section 9.D discussed monitoring of the vacuum drying system exhaust line during drying operations. The exhaust line from the vacuum drying system was routed into a large poly bottle to collect any moisture. A high efficiency particulate filter (HEPA) vacuum cleaner suction line was positioned at the mouth of the poly bottle to draw exhaust air from the poly bottle and into the HEPA filter. The vacuum cleaner discharged into the fuel building. Air monitoring in the fuel building was conducted anytime the vacuum system was operating by an air monitor located near the work area. The air monitoring system was set for a low alarm at 0.3 derived air concentrations (DAC) and a high alarm at five DAC. The fuel building ventilation system was also monitored for radiation and would isolate and automatically switch to 100% recirculation upon detection of high radiation levels. The fuel building ventilation system included a post-accident clean-up system with prefilters, HEPA filters and an activated charcoal filter. Discharge air from the fuel building ventilation system went into the plant exhaust vent stack. Work Control Plan 03-003 noted that Krypton-85 may be released during the drying process and could increase the general background, which would be detected on a nearby frisker and the air monitor.

Documents Health Physics Work Control Plan (WCP)03-003 "Loading and Transporting of Unit 1 Reviewed: Fuel From Units 2/3," Rev. 2 Category: Records Topic: Cask Records (#1)

Reference: 10 CFR 72.212(b)(8)

Requirement The licensee shall accurately maintain the records provided by the cask supplier for each cask that shows, in addition to the information provided by the cask vendor, the following: (a) the name and address of the cask vendor, (b) the listing of the spent fuel stored in the cask and (c) any maintenance performed on the cask. This record must include sufficient information to furnish documentary evidence that any testing and maintenance of the cask has been conducted under an NRC approved QA plan.

Finding: The licensee had developed a document control program to maintain cask records under their reactor Part 50 quality assurance program. Vendor information provided for each Page 53 of 70

canister was reviewed as part of the receipt inspection process in accordance with Procedure SO123-XII-20.4. Receipt inspection records were then forwarded to the document control group. The list of spent fuel placed in each canister was documented in Attachment 2, "DCS Fuel Loading Pattern," to Procedure S023-X-9. Step 7.2 of Procedure SO23-X-9 stated that a completed copy of this procedure (with Attachment 2)

will be forwarded to Corporate Document Management (CDM-SONGS). Maintenance records were included in document control records as maintenance activities were competed. The maintenance orders were tracked by equipment identification number.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) Procedure SO123-XII-Reviewed: 20.4 "Receiving Inspection," Rev 7 (c) Topical Report SCE-1-A "Quality Assurance Program," Change Notice 53 (d) Action Request #030800329 dated August 7, 2003 Category: Records Topic: Cask Records (#2)

Reference: 10 CFR 72.234(d)(2) & (d)(3)

Requirement A list of records required for each cask is provided in 72.234(d)(2). The certificate holder is required by 72.234(d)(3) to provide an original of these records to the user.

Finding: The licensee had established provisions for receiving and storing the required records listed in 10 CFR 72.234(d)(2). Construction of the canisters was initiated by maintenance work orders. The maintenance work orders were scanned and stored in the licensee's optical disk storage system and were retained for the life of the corporatio Upon completion of construction, the cask was transferred to the licensee with a set of original records. Once accepted by the licensee's quality control group, the documentation package was transferred to document control for scanning into the optical disk storage system. The canister fabrication group also maintained a copy of the documentation package.

Documents (a) Record Processing Agreement (RPA)03-003 "Dry Shielded Canister Fabrication Reviewed: Maintenance Orders," Rev 1 (b) Procedure SO1-XXVII-5.7.1 "ASME Section III Fabrication Maintenance Order Preparation and Processing," Rev 1 Category: Records Topic: Maintaining a Copy of the CoC and Documents Reference: 10 CFR 72.212(b)(7)

Requirement The general licensee shall maintain a copy of the CoC and documents referenced in the certificate.

Finding: The licensee had current copies of the Certificate of Compliance, Technical Specifications, Final Safety Analysis Report and the NRC's Safety Evaluation Repor These documents were added to the licensee's optical scanner record retention syste The licensee also had a reference library and on-line subscription service for access to industry standards.

Documents (a) Docket 72-1029 "Transnuclear Advanced NUHOMS Final Safety Analysis Report,"

Reviewed: Rev 0 (b) "Certificate of Compliance #1029 for the NUHOMS 24PT1 Cask System,"

Amendment 0 (c) Appendix A to Certificate of Compliance #1029 "Technical Specifications for the Advanced NUHOMS System Operating Controls and Limits," Rev 0 (d) Appendix B to Certificate of Compliance #1029 "Bases for Technical Specifications for the Advanced NUHOMS System," Rev 0 Page 54 of 70

Category: Records Topic: Maintenance Records Reference: FSAR 1029, Sect 9.2.1.4 Requirement The licensee shall maintain the maintenance records for the Advanced NUHOMS System components provided by Transnuclear Inc.

Finding: The licensee established record processing agreements for maintenance and construction work orders in RPA 03-0020, RPA 03-0021 and RPA 03-0022. These agreements required the maintenance records to be scanned and stored in the licensee's optical disk storage system. The SONGS records center had received 42 completed construction work orders on August 5, 2003 for optical scanning. These work orders were for ISFSI construction/installation activities. As of August 5, 2003, there were no maintenance and repair records in possession of document control.

Documents (a) Record Processing Agreement (RPA) 03-0020 "ISFSI Spent Fuel Transfer Reviewed: Maintenance Orders," Rev 0 (b) RPA 03-0021 "ISFSI Spent Fuel Transfer Maintenance Orders," Rev 0 (c) RPA 03-0022 "ISFSI Spent Fuel Transfer Maintenance Orders," Rev 0 (d) RPA 03-0023 "ISFSI Spent Fuel Transfer Maintenance Orders," Rev 0 Category: Records Topic: Notice of Initial Loading Reference: 10 CFR 72.212(b)(1)(i)

Requirement The general licensee shall notify the NRC at least 90 days prior to first storage of spent fuel.

Finding: The licensee complied with the requirement to notify the NRC at least 90 days prior to first storage of spent fuel at the ISFSI by letter dated May 5, 2003. Loading of the first canister at SONGS began 131 days later on September 12, 2003.

Documents Southern California Edison's letter to the NRC entitled "Docket Nos. 50-206, 50-361, 50-Reviewed: 362 and 72-41 San Onfre Nuclear Generating Station, Units 1, 2, and 3" dated May 5, 2003 (NRC Adams Document ML031270243)

Category: Records Topic: QA Records Reference: 10 CFR 72.174 Requirement The licensee shall maintain sufficient records to furnish evidence of activities affecting quality. The records must include the following: design records, records of use, and the results of reviews, inspections, tests, audits, monitoring of work performance, and materials analysis. The records must include closely related data such as qualifications of personnel, procedures, and equipment. Inspection and test records must identify the inspector/data recorder, type of observation, results, acceptability, and actions taken concerning deficiencies. Records must be maintained until termination of the license.

Finding: The licensee had implemented a records control and retention program for the ISFSI using the reactor Part 50 quality assurance program. The Topical Quality Assurance Manual, Section 17.2.6 "Document Control" and Section 7.2.17 "Quality Assurance Records" had been updated to incorporate the ISFSI records requirements. Procedures SO123-VI-28, SO123-VI-29 and SO123-XXIV-37.8.26 provided instructions for implementing the records program requirements. The types of records specified in 10 Page 55 of 70

CFR 72.174 had been incorporated into the licensee's records program. The Quality Assurance Manual listed the record retention time interval as "the duration of the license or the Certificate of Compliance." The Director, SONGS 1 Decommissioning issued a letter to the records management group on March 25, 2002 which specified that ISFSI records shall be maintained for the life of the SCE Corporation. SONGS had established the optical disk system as part of their permanent records system and had informed the NRC by letter dated August 6, 1996 of the use of the system for record retention. The system incorporated redundancy by using two optical disks. One disk was used for on-line retrieval and the second disk was used as a backup disk.

Documents (a) Topical Report SCE-1-A "Quality Assurance Program," Amendment 22 (b) Topical Reviewed: Quality Assurance Manual, Chapter 1-D," Document Management," Rev 18 (c)

Procedure SO123-VI-29 "Quality Assurance Record Retention Assessment, Processing and Destruction," Rev 5 (d) Procedure SO123-VI-28 "Document Preparation, Transmittal and Processing," Rev 4 (e) Internal Memo to CDM-SONGS from J. Reilly, Director, SONGS 1 Decommissioning entitled "Retention of SONGS 1 Decommissioning and SONGS 1,2,3 ISFSI Documents," dated March 25, 2002 (f)

Southern California Edison's letter to the NRC entitled "Optical Storage of Plant Records," dated August 6, 1996 (Not available in Adams) (g) SO123-XXIV-37.8.26

"Processing of Supplier Documents" Category: Records Topic: Record Retention for 72.212 Analysis Reference: 10 CFR 72.212(b)(2)(i)(C)

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.

Finding: The licensee had scanned the 10 CFR 72.212 analysis (Calculation No. DSC-001) into their optical records retention system. The licensee planned to retain the 10 CFR 72.212 analysis for the life of the corporation.

Documents None Reviewed:

Category: Records Topic: Registration of Casks with NRC Reference: 10 CFR 72.212(b)(1)(ii)

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.

Finding: Calculation DCS-001,Section V specified the 30 day registration requirement of 10 CFR 72.212(b)(1)(ii). This requirement was incorporated into Attachment 7, "10 CFR Reporting Requirements," of Procedure SO123-XV-3.3.

Documents (a) Calculation No. DCS-001 "10 CFR 72.212 Evaluation," Rev 1 (b) Procedure SO-123-Reviewed: XV-3.3 "NRC Reporting Requirements," Rev 9 Category: Safety Reviews Topic: Changes, Tests, and Experiments Reference: 10 CFR 72.48(c)(1)

Requirement A licensee can make changes to their facility or storage cask design if certain criteria are Page 56 of 70

met as listed in 10 CFR 72.48.

Finding: The licensee was performing safety screenings and evaluations for the ISFSI using Procedure SO123-XV-44. Modifications to the site and the ISFSI that were reviewed under Procedure SO123-XV-44 included the temperature monitoring system at the storage module, ISFSI lightning protection, ISFSI pad and module construction, ISFSI security systems, ISFSI security closed circuit TV, Unit 3 fuel handling building modifications, Unit 3 cask handling crane upgrade, calculation to analyze blast effects on the ISFSI to determine security measures and performing a calculation to determine canister fuel loading patterns for storing Unit 1 fuel from the Unit 3 pool. In addition to reviews performed by the licensee, numerous issues had been forwarded back to the cask vendor, Transnuclear, Inc. for review against the design aspects of the NUHOMS FSAR. No issues were identified during the NRC review of selected safety screenings.

Documents (a) Procedure SO123-XV-44 "10 CFR 50.59 and 72.48," Rev 6 (b) 10 CFR 50.59/10 Reviewed: CFR 72.48 Screening Forms for Action Request #'s: AR 030701011-21 "ISFSI Temperature Monitoring," AR 030400078-89 "Canister Loading Patterns," AR 020701186-84 "ISFSI Blast Effects" and AR 030701011-22 "ISFSI Lightning Protection" Category: Security Topic: Back-up Power Supply Reference: 10 CFR 75.55(e)(1) & (f)(4)

Requirement Onsite secondary power supply systems for alarm annunciator equipment and non-portable communications equipment as required in 10 CFR 75.55(f) must be located within vital areas. Non-portable communications equipment controlled by the licensee and required by this section shall remain operable from independent power sources in the event of the loss of normal power.

Finding: The licensee had installed an uninterruptible power supply in the event of an off-site power failure. The uninterruptible power supply was located inside the ISFSI protected area. The uninterruptible power supply could provide power for a specific period of time and would indicate an alarm when there was a loss of primary power. If the uninterruptible power supply failed, an audible alarm would annunciate in the continuously manned alarm station. Should any of the primary batteries fail, the licensee had additional backup batteries available on site.

Documents "Physical Security Plan," Rev 76 Reviewed:

Category: Security Topic: Compensatory Measures Reference: 10 CFR 73.55(g)(1)

Requirement The licensee shall develop and employ compensatory measures including equipment, additional security personnel and specific procedures to assure that the effectiveness of the security system is not reduced by failure or other contingencies affecting the operation of the security related equipment or structures.

Finding: The licensee had established clear compensatory measure requirements to be implemented in the event of failure or degradation of ISFSI security equipment or systems including provisions for the deployment of compensatory measures and the effectiveness of those measures. Compensatory measures were described in the Physical Security Plan, Chapters 4 and 6 and Procedures SO123-IV-6.8.5 and SO123-IV-6. Page 57 of 70

Documents (a) "Physical Security Plan," Rev 76 (b) Procedure SO123-IV-6.8.5 "Security Reviewed: Compensatory Measures," Rev 0/EC 0-5 (c) Procedure SO123-IV-6.8.6 "ISFSI Security Operations," Rev 0 Category: Security Topic: Coordination with Local Law Enforcement Agencies Reference: 10 CFR 73.55(h)(2) & (h)(4)

Requirement The licensee shall establish and document liaison with local law enforcement authorities. Upon detection of abnormal presence or activity of persons or vehicles within an isolation zone, a protected area, material access area, or a vital area; or upon evidence or indication of intrusion into a protected area, the licensee security organization shall inform local law enforcement agencies of the threat and request assistance.

Finding: The licensee had established liaison with local law enforcement agencies and had a current Memorandum of Understanding on file. The law enforcement agencies with which an assistance agreement had been made included the Federal Bureau of Investigation. The licensee was scheduled to conduct their annual Site Orientation Briefing with the FBI in early August 2003, which included the ISFS security physical protection requirements and general layout of the area.

Documents (a) "Physical Security Plan," Rev 76 (b) "Safeguards Contingency Plan," Rev 30 (c)

Reviewed: Procedure SO123-IV-7.1 "Security Communications System," Rev 6 (d) E-mail to Rick Beatty "FBI Orientation Briefing Visit," dated August 4, 2003 Category: Security Topic: Intrusion Detection System Reference: 10 CFR 73.55(e)(1)

Requirement All alarms required must annunciate in a continuously manned central alarm station located within the protected area and in at least one other continuously manned station not necessarily onsite.

Finding: All alarms generated from the ISFSI were transmitted to the power reactor's central alarm station and the secondary alarm station located within the power reactor protected area and within bullet resistant structures. An additional secondary alarm station operator had been assigned specifically for the ISFSI. The ISFSI secondary alarm station operator was responsible for determining the status of all ISFSI alarms, requesting a security officer to respond to the ISFSI and contacting local law enforcement agencies if assistance was necessary. The ISFSI intrusion detection system was highly effective. All intrusion detection alarms for the ISFSI protected area annunciated audibly and visually in the central alarm station and secondary alarm station. When a security alarm was detected, an audible alarm sounded and an electronic record created. An indication of the type and location of the alarm appeared on an alarm monitor at both alarm stations. A visual display of the zone in alarm status was automatically displayed on monitors at each alarm station. During the inspection, the licensee performed functional tests of several alarm zones selected by the NRC inspector. All attempts to surreptitiously penetrate into the protected area were detected and all alarms annunciated in the continuously manned alarmed station as require Page 58 of 70

Documents "Physical Security Plan," Rev 76 Reviewed:

Category: Security Topic: ISFSI Security after Loading Operations Reference: FSAR 1029, Sect 8.1.1.6 (23)

Requirement After completion of the concrete storage module loading, the ISFSI access gate should be closed and locked, and activation of ISFSI security measures should be performed.

Finding: Security Procedure SO123-IV-6.8.6, Section 6.9 required the security officer at the vehicle gate to confirm that the ISFSI protected area was unoccupied and the concrete storage module loading was complete. That officer would then request the security alarm station to close the vehicle gate and would lock the gate. The alarm station would then activate the intrusion detection zone covering the vehicle access gate. The officer at that location was required to conduct a test of the zone to ensure it was functioning.

Documents Procedure SO123-IV-6.8.6 "ISFSI Security Operations," Rev 0 Reviewed:

Category: Security Topic: ISFSI Security Fence Reference: FSAR 1029, Sect 1.4 Requirement Arrays of concrete storage modules are arranged within the ISFSI site on a concrete basemat with the entire area enclosed by a security fence.

Finding: The entire ISFSI area was enclosed by a security fence. The ISFSI protected area fence design and installation was excellent. The fence was a nominally 8-foot high fence constructed of heavy welded wire fabric topped by several strands of barbed wire. There were two motor-operated rolling vehicle gates. The protected area barrier provided penetration resistance to both forced and surreptitious entries adequate to ensure delay of a potential adversary and effectively limited access to only authorized personnel.

Documents "Physical Security Plan," Rev 76 Reviewed:

Category: Security Topic: ISFSI Security Surveillance Reference: FSAR 1029, Sect 8.1.1.7 (1)

Requirement Routine security surveillance shall be performed in accordance with the licensee's ISFSI security plan.

Finding: The Physical Security Plan, Section 6.8 "Security Patrols," provided information on the licensee's security patrol program. A video assessment system monitored by operators in the alarm stations provided the remote capability to routinely observe the ISFSI protected area barrier and intrusion detection zones. The video system was capable of performing checks of all observable areas within the ISFSI protected area for the presence of unauthorized persons, vehicles or materials. Random security patrols of the exterior of the protected area perimeter to check the integrity of security barriers, security equipment hardware and degraded perimeter lighting, etc., were required. If the video assessment system was inoperable, or incapable of observing the ISFSI protected area (e.g. inclement weather conditions), security patrols would be performed in accordance with security procedures. During daylight hours on August 6, 2003, an NRC security inspector conducted an observation of the exterior and interior isolation zones and exterior areas within the ISFSI protected area using monitors located in the Page 59 of 70

secondary alarm station. The positioning of the cameras provided adequate assessment of all intrusion detection zones. In addition, during the hours of darkness on August 6, 2003, the NRC inspector conducted an evaluation of the assessment aids around and within the ISFSI protected area. The NRC inspector observed alarms in the secondary alarm station while several security personnel walked around the ISFSI isolation zone The licensee also had a video capture capability. The video capture system enhanced the licensee's ability to immediately identify the cause of the alarms. Alarm station operators were well trained on the use of the video capture system to assess alarms.

Documents (a) "Physical Security Plan," Rev 76 (b) Procedure SO123-IV-6.8.6 "ISFSI Security Reviewed: Operations," Rev 0 Category: Security Topic: Lock and Key Controls Reference: 10 CFR 73.55(d)(8)

Requirement All keys, locks, combinations and related access control devices used to control access to the protected areas must be controlled to reduce the probability of compromise.

Finding: The licensee had an excellent lock and key control program. The issuance and control of ISFSI keys was administered through Procedures SO123-IV-4.4 and SO123-IV-4.5 and was the responsibility of the Security Shift Commander. The Shift Commander or designee supervised maintenance of key issue logs and ensured a continuous accounting of all locks and keys. Accountability records were checked at each shift change and the results were recorded in appropriate logs. All locks and keys were inventoried at least once every 12 months by a qualified security individual. Keys to the ISFSI protected area were retained in locked cabinets in designated secured areas.

Documents (a) "Physical Security Plan," Rev 76 (b) Procedure SO123-IV-4.4 "Security Lock and Reviewed: Key Issue, Control and Accountability," Rev 8 (c) Procedure SO123-IV-4.5 "Security Lock and Key Repair/Replacement Program," Rev 2/EC 2-1 Category: Security Topic: Observation of Protected Area Reference: 10 CFR 72.212(b)(5)(iv)

Requirement The observational capability required by 10 CFR 73.55(h)(6) as applied to a new protected area may be provided by a guard or watchman on patrol in lieu of closed circuit television.

Finding: Observation of the isolation zones and the physical barrier at the ISFSI protected area was provided by closed circuit television. Fixed closed circuit television cameras were installed along the boundary as the primary means of alarm assessment by both the central and secondary alarm station operators. When an individual(s) attempting to gain unauthorized access to the ISFSI protected area generated an alarm, the closed circuit television system automatically displayed images associated with the alarmed zoned on dedicated monitors. This system was observed in operation by the NRC inspector during a night time test. Discussions with the secondary alarm station operator during the test found the operator to be very knowledgeable of his duties and responsibilities.

Documents (a) "Physical Security Plan," Rev 76 (b) ISFSI Protected Area Perimeter CCTV Testing Reviewed: conducted August 6, 2003 Page 60 of 70

Category: Security Topic: Personnel Search Prior to ISFSI Entry Reference: 10 CFR 72.212(b)(5)(iii)

Requirement For the purpose of this general license, searches required by 10CFR73.55(d)(1) before admission to a new protected area may be performed by physical pat-down searches of persons in-lieu of firearms and explosives detection equipment.

Finding: Procedures SO123-IV-5.3.6 and SO123 IV-5.3.7 required all individuals, vehicles and hand carried packages entering the ISFSI protected area to be properly authorized and visually searched for explosives prior to entry. Access was controlled by a member of the security force to ensure only authorized personnel, vehicles and materials entered on the basis of a need to access and were approved.

Documents (a) "Physical Security Plan," Rev 76 (b) Procedure SO123-IV-5.3.6 "ISFSI Personnel Reviewed: Access Controls, Search/Inspection and Emergency Personnel Access," Rev 0 (c)

Procedure SO123 IV-5.3.7 "ISFSI Vehicle Access Controls, Search/Inspection and Emergency Personnel Access," Rev 0 Category: Security Topic: Physical Security Organization & Program Reference: 10 CFR 72.212(b)(5)(i)

Requirement The physical security organization and program for the (reactor) facility must be modified as necessary to assure that activities conducted under this general license do not decrease the effectiveness of the protection of vital equipment in accordance with 10 CFR 73.55 Finding: The security organization was sufficiently staffed and equipped with weapons, radios, and other types of personal security items to implement an effective response to a security event at the ISFSI without decreasing the effectiveness of the protection of vital equipment at the reactor facilities. The licensee had integrated security for the ISFSI into the existing site security program. The ISFSI was a separate protected area located outside the power block protected area separated from any vital areas. The ISFSI physical security program included the following features: protected area barriers, intrusion detection system, isolation zones, closed circuit television cameras, lighting, and search requirements for individuals, packages and vehicles entering the protected area. The security organization had designated and trained security officers specifically assigned to the ISFSI facility.

Documents "Physical Security Plan," Rev 76 Reviewed:

Category: Security Topic: Protected Area Reference: 10 CFR 72.212(b)(5)(ii)

Requirement Storage of spent fuel must be within a protected area, in accordance with 10 CFR 73.55(c), but need not be within a separate vital area. Existing protected areas may be expanded or new protected areas added for the purpose of storage of spent fuel.

Finding: The ISFSI was a stand-alone protected area located outside the power block protected area and was separated from any vital areas. Two vehicle gates were provided. The gates were constructed to provide penetration resistance equal to, or greater than, the protected area fence. The design of the ISFSI protected area perimeter ensured that Page 61 of 70

assessment aids had a clear unobstructed view of all intrusion detection alarm zone The design features meet the requirements of 10 CFR 73.55. The size of the storage casks coupled with the use of trained personnel, physical barriers, access controls, intrusion detection, communications, contingency response and local law enforcement support made the ISFSI an unlikely target for radiological sabotage.

Documents "Physical Security Plan," Rev 76 Reviewed:

Category: Security Topic: Protected Area Lighting Reference: 10 CFR 73.55(c)(5)

Requirement Isolation zones and all exterior areas within the protected area shall be provided with illumination sufficient for the monitoring and observation requirements of paragraph (c)(3), (c)(4) and (h)(4) of 10 CFR 73.55, but not less than 0.2 foot-candles measured horizontally at ground level.

Finding: A tour of the ISFSI area was performed to observe the ISFSI protected area lightin The tour was conducted in the early morning at 4:00 am on August 6, 2003. ISFSI lighting was excellent. The ISFSI isolation zones and all exterior areas were illuminated to allow (1) effective monitoring and observation of activity on each side of the protected area barrier within the isolation zone; (2) detection of penetration or attempted penetration of a protected area barrier; (3) detection of unauthorized persons, vehicles and materials within the protected area; and (4) determination of whether a threat existed and the extent of that threat. The illumination of the ISFSI met the 0.2 foot-candles requirement measured horizontally at ground level.

Documents (a) "Physical Security Plan," Rev 76 (b) Procedure SO123-IV-6.8.2 "Protected Area Reviewed: Lighting," Rev 1 (c) "ISFSI Protected Area Lighting Survey," dated August 4, 2003 Category: Security Topic: Security Guard Force Training Reference: 10 CFR 73.55(b)(4)(i) & (b)(4)(ii)

Requirement The licensee may not permit an individual to act as a guard or armed response person, or other member of the security organization unless the individual has been trained, equipped and qualified to perform each assigned security job duty in accordance with 10 CFR 73 Appendix B. Each licensee shall establish, maintain, and follow an NRC-approved training and qualifications plan.

Finding: SONGS had modified the Security Force Training and Qualification Plan to incorporate training and qualification program requirements that the security personnel may be required to perform at the ISFSI. A review of training records verified that all ISFSI training for qualified security officers would be completed by August 2003.

Documents (a) "Physical Security Plan," Rev 76 (b) "Training and Qualification Plan," Rev 26 (c)

Reviewed: "SONGS 1, 2, and 3 Security Force Training and Qualification Plan (T&Q)," Rev 0 (d)

Procedure SO123-IV-3.3.1 "Security Training and Qualification Program," Rev 0 (e)

ISFSI T&Q Individual Attendance and Training Certification Records (f) SONGS T2000 Training History by Attainment - ISFSI T&Q Page 62 of 70

Category: Security Topic: Security Plan Review Reference: 10 CFR 73.55(g)(4)(i) & (g)(4)(i)(B)

Requirement The licensee shall review implementation of the security program by individuals who have no direct responsibility for the security program, as necessary, based on an assessment by the licensee against performance indicators and as soon as reasonably practicable after a change occurs in personnel, procedures, equipment or facilities that potentially could adversely affect security.

Finding: The licensee had conducted Surveillance #07108103 to verify proposed changes to encompass the ISFSI into the security plans were being implemented in accordance with security plan requirements. The SONGS audit program requirements included periodic review of the application and adequacy of the SONGS physical security program to assess station compliance with security plan and procedure requirements. Audits were performed at intervals not to exceed 12 months, or as necessary, based on assessment by the Nuclear Oversight and Assessment (NOA) group against performance indicators, and as soon as reasonably practicable after a change occurred in personnel, procedures, equipment, or facilities that potentially could adversely affect security, but no longer than 12 months after the change. The results of all audits were summarized and reported in writing to the Vice President, Nuclear Generation and to site managemen Recommendations for correction and improvement were also documented and reported.

Documents Nuclear Oversight and Assessment (NOA) Surveillance Report #07108103 "SONGS Reviewed: ISFSI" dated July 8, 2003 Category: Security Topic: Tamper Seal on Storage Modules Reference: FSAR 1029, Sect 2.3.5.1 Requirement A security tamper seal on the concrete storage module door may be included after insertion of a loaded canister. This may be, but is not limited to, one of the following: 1)

tack welding the storage module access door, 2) tack welding two or more closure bolts on the storage module access door, 3) tamper seals.

Finding: The licensee had considered the installation of a tamper seal but decided that given the security of the ISFSI and the difficulty associated with removing the outer cover door, (i.e. a crane and a lifting rig would be required) that a security seal was not required.

Documents E-mail dated August 5, 2003: Subject: Security Seal Reviewed:

Category: Security Topic: Transport of Canister to ISFSI Reference: FSAR 1029, Sect 11.2.5.1.1 Requirement Once the transfer cask is loaded onto the transport skid/trailer and secured, it is pulled to the ISFSI by a tractor vehicle. A predetermined route is chosen to minimize the potential hazards that could occur during transport.

Finding: The licensee had established security plans and contingency procedures for the protection of the spent fuel canister during movement between the Unit 3 fuel building, the Unit 1 fuel building and the ISFSI. The security procedures restricted vehicle traffic near the canister movement route by controlling access on plant roadways at specific locations with armed security personnel. Provisions were established by security to Page 63 of 70

ensure there was no degradation of protection of the spent fuel during the time the cask left the reactor protected area until it entered the ISFSI protected area. Security escorts and a "moving" protected area were used to limit access to the canister during the transfer. A walk-down of the transportation route confirmed that adequate security provisions had been identified for transporting the canister to the ISFSI.

Documents (a) Procedure SO123-IV-6.8.6 "ISFSI Security Operations," Rev 0 (b) Security Force Reviewed: Supplemental Instructions SE123-686 "Unit 1 Spent Fuel Cask Transfer Operations from Unit 3 to ISFSI Protected Area," Rev 0 Category: Training Topic: Approved Training Program Reference: CoC 1029, Tech Spec 5.2.2; 10 CFR 72.44(b)(4)

Requirement The licensee shall have a training program in effect that covers the training and certification of personnel that meet the requirements of subpart I before the licensee receives spent fuel at the ISFSI. Training modules shall require a comprehensive program for the operations and maintenance of the Advanced NUHOMS system and the ISFSI. The training modules shall include the elements listed in Technical Specification 5.2.2.

Finding: Selected topics from Technical Specification 5.2.2 were verified as being incorporated into the licensee's training modules, lesson plans and personnel qualification standard ISFSI training modules ISFS01 through ISFS04 and several ISFSI related personnel qualification standards were reviewed. Personnel qualification standards were training modules that involved demonstration of skills associated with a particular task. These included: (1) Training Module ISFQ03 for startup, operation, and shutdown of the vacuum dryer per Procedure SO23-X-9, (2) Training Module ISFQ04 for operation of the transfer trailer per Procedure SO123-X-9.2, (3) Training Module ISFQ05 for operation of the prime mover per Procedure SO123-X-9.2, (4) Training Module ISFQ06 for alignment of the transfer cask with the storage module per Procedure SO23-X-9, and (5) Training Module ISFQ07 for operation of the hydraulic ram system. Topics from the Technical Specification 5.2.2 list that were associated with activities already being performed under the Part 50 license were addressed in the reactor facility training programs. For example, rigging and handling of loads was addressed in the mechanical maintenance training program under Lesson Plan MT7400, "Basic Rigging," Revision 3.

Documents (a) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 (b)

Reviewed: "ISFSI Personnel Training Status Matrix Printout," dated August 4, 2003 (c) Training Module ISFS01 "NUHOMS 72-1029 Certificate of Compliance," Rev 0 (d) Training Module ISFS02 "NUHOMS Cask/DSC Preparations," Rev 0 (e) Training Module ISFS03 "Draining and Drying Operations," Rev 0 (f) Training Module ISFS04 "Transfer Trailer," Rev 0 Category: Training Topic: Certification of Personnel Reference: 10 CFR 72.190 Requirement Operations of equipment and controls that have been identified as important to safety in the Safety Analysis Report and in the license must be limited to trained and certified personnel or be under the direct visual supervision of an individual with training and Page 64 of 70

certification in the operation. Supervisory personnel who personally direct the operation of equipment and controls that are important to safety must also be certified in such operations.

Finding: The certification program for personnel involved in ISFSI operations was described in procedure SO-123-XXI-1.11.27. Attachment 1 of the procedure contained a task qualification list and the training requirements to obtain qualification for each task. The following tasks were addressed in Attachment 1: ISFSI General Worker, ISFSI Crane Operator, ISFSI Prime Mover Operator, ISFSI Transfer Trailer Operator, ISFSI Alignment Operator, ISFSI Draining and Drying Operator, ISFSI Hydraulic Ram Operator, ISFSI Health Physicist, ISFSI Fuel Team Leader and ISFSI Fuel Movement Supervisor. Some ISFSI-related procedures stated specific personnel qualifications in the procedure. For example, SO3-1-3.3.2, "Unit 3 Cask Handling Crane Checkout and Operation," Revision 1 required that the crane be operated only by a NUREG 0612 qualified crane operator. A review of the training matrix status dated August 4, 2003 showed that the majority of personnel responsible for ISFSI operations had completed all required applicable training. The remaining individuals were in the process of completing required training. The training certification records for approximately one third of the personnel listed on the training status matrix were reviewed and determined to be complete. Verification that the individuals assigned to tasks were trained and qualified was the responsibility of the supervisor to confirm during the pre-job preparations and briefings.

Documents (a) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 (b)

Reviewed: "ISFSI Personnel Training Status Matrix Printout," dated August 4, 2003 Category: Training Topic: Re-Training Program Reference: FSAR 1029, Sect 9.3.2 Requirement Re-training is generally consistent with the re-training requirements in effect at the plant for personnel involved in fuel handling operations.

Finding: Procedure SO123-XXI-1.11.27, Section 6.6 described the requirement for a continuing training program. Step 6.6.1.1 stated that the ISFSI continuing training curriculum shall be determined by the Maintenance Training Manager based on new industry events and the time since last training was performed. The licensee had not established a re-training schedule at the time of the NRC pre-operational inspection and was still in the process of completing all initial training requirements.

Documents Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 Reviewed:

Category: Training Topic: Training for Health Physics Personnel Reference: FSAR 1029, Sect 9.3.1.3 Requirement Generalized training should be provided to health physics personnel in the applicable regulations and standards and the engineering principles of passive cooling, radiological shielding and structural characteristics of the ISFSI. Specific training shall be provided in the radiological shielding design of the system, particularly the canister's top shield plug, the transfer cask and the concrete storage module.

Finding: Procedure SO123-XXI-1.11.27, Attachment 2 listed the training qualifications for ISFSI Page 65 of 70

health physics personnel. The specific training provided to the health physics personnel included ISFS01"NUHOMS 72-1029 Certificate of Compliance" and ISFS02

"NUHOMS Cask/DSC Preparations." The lesson plans and student handouts covered the required specific training listed in the Final Safety Analysis Report. Selected individual health physics personnel training records were reviewed in the training status matrix dated August 4, 2003. All selected personnel had completed training.

Documents (a) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 (b)

Reviewed: "ISFSI Personnel Training Status Matrix Printout," dated August 4, 2003 (c) Training Module ISFS01 "NUHOMS 72-1029 Certificate of Compliance Overview, " Rev 0 (d)

Training Module ISFS02 "NUHOMS Cask/DSC Preparation Operations," Rev 0 Category: Training Topic: Training for Maintenance Personnel Reference: FSAR 1029, Sect 9.3.1.2 Requirement Generalized training should be provided to plant maintenance personnel in the applicable regulations and standards and the engineering principles of passive cooling, radiological shielding and structural characteristics of the ISFSI. Specific training shall be provided for use of the vacuum drying system; the automated welding equipment; operation of the transfer trailer; alignment of the transfer cask skid with the concrete storage module; assembly of the hydraulic ram system; and normal and off-normal operation of the hydraulic ram system. Specific training shall also be provided for cleaning the storage module air inlet and outlet vents.

Finding: Procedure SO123-XXI-1.11.27, Attachment 2 listed the training qualifications for various tasks that would be performed by maintenance personnel and included all the required training listed in the Final Safety Analysis Report.

Documents (a) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 (b)

Reviewed: "ISFSI Personnel Training Status Matrix Printout," dated August 4, 2003 Category: Training Topic: Training for Operations Personnel Reference: FSAR 1029, Sect 9.3.1.1 Requirement Generalized training should be provided to plant operations personnel in the applicable regulations and standards and the engineering principles of passive cooling, radiological shielding and structural characteristics of the ISFSI. Detailed operator training shall be provided for canister preparation and handling, fuel loading, transfer cask preparation and handling, and transfer trailer loading.

Finding: Operations personnel were provided training on ISFSI operations as part of the license operator continuing training program. Training was provided under Lesson Plan 2RP453. The lesson plan included the subjects required by the Final Safety Analysis Report. The training records for one of the three ISFSI Fuel Team Leader positions was reviewed and was current.

Documents (a) Lesson Plan No. 2RP453 "Independent Spent Fuel Storage Installation" (b) "ISFSI Reviewed: Personnel Training Status Matrix Printout," dated August 4, 2003 Page 66 of 70

Category: Training Topic: Training Program Records Reference: FSAR 1029, Sect 9.3.3 Requirement The licensee's plant training organization is responsible for training programs and for maintaining up-to-date records on the status of personnel training.

Finding: Procedure SO123-XXI-1.11.27 identified the requirements for establishing and maintaining the training records. The licensee had established a computerized training matrix that could provide the current status of training for each individual by name.

Documents (a) Procedure SO123-XXI-1.11.27 "ISFSI Training Program Description," Rev 0 (b)

Reviewed: "ISFSI Personnel Training Status Matrix Printout," dated August 4, 2003 Category: Welding/NDE Topic: Canister Fabrication/Inspection of Welds Reference: FSAR 1029, Sect 2.3.2.1 Requirement The canister's cylindrical shell is fabricated from rolled ASME stainless steel plate which is joined with full penetration welds that are 100% inspected by non-destructive examination.

Finding: Canister fabrication was performed in accordance with the licensee procedures and fully satisfied ASME specifications. The canisters were fabricated in a shop at the SONGS site. The fabrication involved the cold rolling of flat stock material. All weld seams received multiple nondestructive examinations both during and after complete fabrication. The nondestructive examination records for Canister 02010375001 were reviewed and found to adequately document the required examinations. The examinations performed included dye penetrant, radiography and visual.

Documents (a) Procedure SO1-XXVIII-5.4 "Closure Welding of DSC Vessel," Rev 0 (b) Procedure Reviewed: SO123-V-7.20.13 "ASME Spent Fuel Dry Storage Canister Closure Welding Standard,"

Rev 0 (c) Procedure SO123-XII-9.9 "Welding Inspection," Rev 4 Category: Welding/NDE Topic: Dye Penetrant Exam of Welds Reference: FSAR 1029, Sect 8.1.1.3 (13) & 8.1.1.4 (1, 3)

Requirement Weld examination of the inner top cover plate weld, prefabricated covers over the vent and siphon ports and the outer top cover plate weld root pass should be performed using a dye penetrant.

Finding: Procedures SO123-XII-9.301 and SO123-XII-9.303 provided a description of the dye penetrant exam process. During the pre-operational inspections, the licensee successfully demonstrated the dye penetrant testing techniques on a truncated caniste During the welding on the first canister, the NRC observed the dye penetrant exams performed on September 14, 2003 of the welds on the canister lids and the vent and siphon port covers. The examinations fully met the requirement for examination of the root weld and the final weld surface. The examinations were performed by qualified personnel using certified materials and techniques.

Documents (a) Procedure SO123-XII-9.301 "Liquid Penetrant Examination," Rev 2 (b) Procedure Reviewed: SO123-XII-9.303 "Liquid Penetrant Examination - High Temperature," Rev 0 Page 67 of 70

Category: Welding/NDE Topic: Exceptions to the ASME Code Reference: CoC 1029, Tech Spec 4.3.4 Requirement Proposed alternatives to the ASME code, other than the exceptions mentioned in Tech Specs 4.3.4 for the canister, may be used when authorized by the Director, Nuclear Materials Safety and Safeguards, or designee. The licensee should demonstrate that the proposed alternatives would provide acceptable level of quality and safety or, comply with the specified requirements of ASME Code,Section III, 1992 Edition. Requests for exceptions to the ASME codes should be submitted in accordance with 10 CFR 72.4.

Finding: There were no identified exceptions to the ASME Code related to the canister lid welding program.

Documents None Reviewed:

Category: Welding/NDE Topic: Hydrogen Monitoring During Welding Reference: FSAR 1029, Sect 8.1.1.3 (11 & 12)

Requirement Continuous hydrogen monitoring during the welding of the inner cover plate is required (including tack welding of the lid). Insert a piece of tygon tube through the vent port such that it terminates just below the shield plug. Connect the tube to a hydrogen monitor to allow continuous monitoring of the hydrogen atmosphere in the canister during welding. Hydrogen concentration shall not exceed a safety limit of 2.4% during welding of the inner top cover plate to the canister's shell.

Finding: Procedure SO1-XXVIII-5.4, Step 6.1.15 required continuous monitoring of the space under the inner cover plate during welding to assure hydrogen levels remained below 2.4%. The lower explosive limit (LEL) for hydrogen is 4.8%. If the 2.4% value was exceeded (50% LEL), purging of the space was allowed. On September 14, 2003 tack welding was started on the first canister. A USI hydrogen detector was used by the licensee with a 0-100% LEL read-out. The alarm point was set at 50% LEL. During tack welding, the hydrogen detector was reading 6-7% LEL. This would equate to approximately 0.3% hydrogen present. After 7 to 8 tack weld segments, the hydrogen detector failed due to low battery. Moisture was noted in the line between the detector and the canister, however, the user manual did not identify high humidity as a potential cause for failure of the detector, though this has been a problem at other sites with hydrogen detectors. A replacement detector was installed and welding continued. The new hydrogen detector readings ranged from 5% LEL to 8.4% LEL with typical readings around 6-7% LEL. This was consistent with the readings observed on the first hydrogen detector. A water trap was added to the line to reduce the moisture getting to the detector. During a break in the welding, the licensee performed a field calibration check of the hydrogen detector's zero calibration setting using a bottle gas calibrator as provided for in the user manual. The detector required a slight adjustment. After re-calibration, the detector typically read 5-6% LEL. After the tack welds were completed, the welding of the root pass was started. The hydrogen detector reading was 3% LEL.

Documents Procedure SO1-XXVIII-5.4 "Closure Welding of DSC Vessel," Rev 0 Reviewed:

.

Page 68 of 70

Category: Welding/NDE Topic: Lid Weld per Subsection NB Reference: FSAR 1029, Sect 1.2.1.1 Requirement The top closure confinement and the top plug penetrations (siphon and vent ports) are welded compliant to requirements of the ASME Code, Subsection NB.

Finding: The welds were performed in accordance with Subsection NB of Section III of the ASME Code. The welding was performed using an automated welding system. A review of the welding procedure and observation of the welding demonstrations during the pre-operational inspection were performed. A very high quality weld was produce During the actual loading of the first canister, welding personnel were familiar with their assigned tasks for set-up and operation of the automatic welding equipment. The welding personnel were cognizant of the radiological conditions around the canister and practiced good ALARA techniques by staying back away from the canister as much as practical. Welds produced on the first canister were of high quality.

Documents (a) Procedure SO1-XXVIII-5.4 "Closure Welding of DSC Vessel," Rev 0 (b) Procedure Reviewed: SO123-V-7.20.13 "ASME Spent Fuel Dry Storage Canister Closure Welding Standard,"

Rev 0 Category: Welding/NDE Topic: Lowering Water Level Prior to Welding Reference: FSAR 1029, Sect 8.1.1.3 (8)

Requirement Connect the vacuum drying system to the canister and use the liquid pump to drain approximately 60 gallons to the fuel pool prior to welding the inner cover plate.

Finding: Procedure SO23-X-9, Steps 6.6.8 through 6.6.14 provided for the draining of approximately 60 gallons of water from the canister using the vacuum drying system liquid pump. The water was drained to the spent fuel pool. During the draining of the 60 gallons from the first loaded canister prior to welding the inner cover plate, the radiation levels remained unchanged at 1-2 mrem/hr gamma and less than 1 mrem/hr neutron on the lid.

Documents (a) Procedure SO23-X-9 "Dry Cask Storage Loading," Rev 1 (b) Procedure SO1-XXVIII-Reviewed: 5.4 "Closure Welding of DSC Vessel," Rev 0 Category: Welding/NDE Topic: NDE Personnel Qualifications Reference: FSAR 1029, Sect 9.1.2 Requirement Non-destructive examination (NDE) personnel are qualified in accordance with SNT-TC-1A.

Finding: During the pre-operational inspection, the qualifications documents for the licensees nondestructive examination personnel assigned to the weld inspection program for the ISFSI project were reviewed. All personnel fully satisfied the SNT-TC-1A requirements.

Documents Recommended Practice No. SNT-TC-1A, Society for Nondestructive Testing (SNT) -

Reviewed: Technical Council (TC) - Document "Personnel Qualification and Certification in Nondestructive Testing," 1996 Page 69 of 70

Category: Welding/NDE Topic: NDE Procedures Reference: FSAR 1029, Sect 9.1.3 Requirement Non-destructive examination (NDE) requirements for welds are specified on drawing Weld related NDE is performed in accordance with written and approved procedures.

Finding: A review of the licensees approved drawings and procedures verified that the nondestructive examination requirements for the welds were fully specified in the drawings. The licensee had developed two procedures for performing the nondestructive examinations of the welds, one for normal temperature ranges and one for high temperature ranges. The high temperature nondestructive examination process was needed for the canisters with high heat loads.

Documents (a) Procedure SO123-XII-9.301 "Liquid Penetrant Examination," Rev 2 (b) Procedure Reviewed: SO123-XII-9.303 "Liquid Penetrant Examination - High Temperature," Rev 0 Category: Welding/NDE Topic: Qualifications of Welds and Welders Reference: FSAR 1029, Sect 2.3.2.1 Requirement Pressure boundary weld and welders are qualified in accordance with Section IX of the ASME B&PV Code and inspected according to the appropriate articles of Section III, Division 1, Subsection NB.

Finding: The qualifications for the welders were reviewed and found to meet the requirements of Section IX of the ASME Boiler and Pressure Vessel Code.

DocumentsSection IX of the ASME Boiler and Pressure Vessel Code Reviewed:

Category: Welding/NDE Topic: Welding Consumables Purchased Under QA Reference: FSAR 1029, Sect 13.2(A)

Requirement Welding consumables shall be procured as a Category A item if the intended use is unknown. If purchased for a specific Category B or C application, the material must be identified and its use restricted to fabrication of the same level.

Finding: Specification SO1-207-22 for welding consumables was reviewed. Welding consumables were purchased under the proper quality assurance category in accordance with the licensee approved procedures, consistent with ASME Code requirements.

Documents Specification SO1-207-22 "Material Specification for Unit 1 DSC Vessel GTAW Filler Reviewed: Metal," Rev 0 Page 70 of 70

IR 05000206/2003010

Text

Navigation menu

v t e Inspection Report - San Onofre - 2003010
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
2003 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 22(0) 10(0) 7(0) 11(0) Security Operator Exams 301(0) Emergency Planning Other