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Public Watchdogs - NRC 2.206 Petition Exhibits 1-38, Part 6
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Maureen Brown, (626) 302-2255 SCE Selects Robust Underground System to Store San Onofre Used Nuclear Fuel ROSEMEAD, Calif., Dec. 11, 2014 Southern California Edison (SCE) has selected Holtec International to expand the San Onofre nuclear plants storage of used nuclear fuel in a robust underground facility.

The contract with Holtec represents a major step in the decommissioning of the nuclear plant. It sets the stage to transfer San Onofres used fuel from steel-lined concrete storage pools to steel-and-concrete-encased canisters, with a goal of completing the work by mid-2019.

After reviewing leading designs with the San Onofre Community Engagement Panel, we concluded this underground design is best suited to safely and securely store used nuclear fuel at San Onofre until the federal government removes the fuel from site, as required, said Chris Thompson, SCE vice president of Decommissioning. Our decision to move expeditiously to transfer the fuel also reflects feedback from community leaders who prefer dry storage of used nuclear fuel.

Thompson noted the robust Holtec design exceeds California earthquake requirements and protects against hazards such as water, fire or tsunamis.

I especially want to thank the Community Engagement Panel for its thoughtful questions and enormous time commitment during SCEs evaluation, said Thompson, noting that SCE ultimately focused on cask designs licensed by the Nuclear Regulatory Commission for both storage and transport of used nuclear fuel.

While dry storage of nuclear fuel is a proven technology used for almost three decades in the United States, Thompson said SCE will go beyond industry practices by partnering with the Electric Power Research Institute to develop new inspection techniques to monitor cask integrity.

Holtecs HI-STORM UMAX underground storage system features corrosion-resistant, stainless-steel fuel canisters topped with a 24,000-pound steel and concrete lid. The canisters will be encased in a concrete monolith. Holtec is a global supplier and has nuclear fuel storage systems at two other California locations, Humboldt Bay and Diablo Canyon. More information is available in this fact sheet.

Thompson said engineering work begins immediately, followed by fabrication of canisters. Completion of the dry storage project facilitates major dismantlement work SCE plans to complete within 20 years.

SCE announced in June 2013 that it would retire San Onofre Units 2 and 3, and begin preparations to decommission the facility. SCE has established core principles of safety, stewardship and engagement to guide decommissioning. For more information about SCE, visit www.songscommunity.com.

About Southern California Edison An Edison International (NYSE:EIX) company, Southern California Edison is one of the nations largest electric utilities, serving a population of nearly 14 million via 4.9 million customer accounts in a 50,000-square-mile service area within Central, Coastal and Southern California.

San Onofre Nuclear Waste Problems Tom English, Ph.D., Samuel Lawrence Foundation Subrata Chakraborty, Ph.D., UCSD, Dept. of Chemistry and Biochemistry Rear Admiral Len Hering Sr. USN (ret)

January 2019 INTRODUCTION In August 2018, a near-accident during the loading of nuclear waste into dry storage triggered a federal investigation and brought new urgency to the debate of how best to store some of the most dangerous waste known to humankind - spent nuclear fuel. The San Onofre Nuclear Generating Station (S.O.N.G.S.) closed in 2012 after a number of serious failures. Since then, Southern California Edison and its contractor, Holtec International, built a concrete storage vault to hold 3.6 million pounds of nuclear waste in dry storage. That vault is footsteps from the rising Pacific Ocean. In our brief report, we explore the fatal flaws of this location and recommend moving the storage facility to a technically defensible storage facility at a significantly higher elevation with distance from the ocean. We address the inadequacy of the equipment used to move and contain the nuclear waste material. We explore the gouging that occurs when stainless steel canisters are lowered into the storage vault and how gouging compromises the integrity of the containers. Finally, we examine management practices at San Onofre and an apparent lack of supervision, training and protocols. The examination of the perils of S.O.N.G.S. Independent Spent Fuel Storage Installations poor location, poor technology and poor management, presents an urgent situation for regulators to: order Edison to permanently stop the loading of canisters into dry storage, require Edison to store the waste in canisters that may be inspected, and secure an independent analysis and risk assessment of canister loading procedure.

RATIONALE Most serious of the issues facing the interim storage of nuclear waste at S.O.N.G.S. include the gouging damage to fully-loaded steel canisters upon downloading into the storage vault. These 54-ton thin-walled steel canisters are loaded with nuclear waste in wet storage - spent fuel pools - and are transported to the on-site concrete storage vault, adjacent to the reactor domes. With the Brinell hardness scale calculations our team demonstrates the depth and width of canister gouges upon downloading into the storage system. The current downloading procedure and on-site storage configuration provides the factors necessary to create gouges in the external steel walls of the canisters: operators have no visibility of the canister during downloading and precise adjustments to canister orientation cannot be made. These gouges remain undetected and unrepaired due to the lack of thorough inspection and monitoring at

San Onofre Nuclear Waste Problems the San Onofre Independent Spent Fuel Storage Installations (ISFSIs). The preliminary findings are found in this report.

1. POOR LOCATION Today, two separate Independent Spent Fuel Storage Installations (ISFSIs) exist at San Onofre.

The newest, built by Holtec, is located about 100 feet from the Pacific Ocean on the 85-acre grounds of S.O.N.G.S. The property is part of Marine Corps Base Camp Pendleton and is owned by the Department of the Navy. Two of the nations busiest transportation corridors --

Interstate 5 and the Los Angeles-San Diego-San Luis Obispo Rail Line -- flank the site. The ISFSIs are clearly visible in Google Earth images and in numerous published photographs. The high accessibility and visibility of the site leaves it extremely vulnerable to an act of malfeasance.

Figure 1. Independent Spent Fuel Storage Installations and Storage Vault.

Forces of nature, exacerbated by sea-level rise, carry further risks. Frequent high humidity and coastal fog make the metal at the site susceptible to short-term corrosion and stress-induced corrosion cracking. Also located at this site is a second, older ISFSI, which contains 51 thin-walled steel canisters that are up to 15 years old.

Numerous reports show that mean high tide level is about 18 inches below the base of the newer, oceanfront ISFSI, which was designed by Holtec. Since this is the mean height, the sea level frequently exceeds this height. Hence, it is likely the present ground water table will leach into the storage vault and result in at least damp storage. Further sea level rise due to climate change will make this problem far worse.

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San Onofre Nuclear Waste Problems Dr. James Hansen, who managed NASAs climate change program for about 25 years, predicts sea levels could rise up to 10 feet during the next 50 years. At San Onofre, this would cause the bottom seven feet of the Holtec nuclear storage canisters to be submerged in seawater, unintentionally resulting in wet storage. This would invite a crisis similar to that of Fukushima, where spent fuel was exposed to moisture.

A second estimate appears in a comprehensive report by the Working Group of the California Ocean Protection Council Science Advisory Team. Published in 2017, the report shows 75%

likelihood sea levels will rise by two feet by 2100. Either of these scenarios envisions that a major portion of the nuclear storage canisters as San Onofre would be submerged in seawater.

The combination of the effects of sea-level rise and ground water inundation at the current location would change the Holtec ISFSI to wet storage site, for which it was not designed.

Hence, little if anything would be accomplished by moving the waste from the spent-fuel pool to the dry storage ISFSI. The dangers would not be decreased. If anything, the inability to adequately measure and mitigate the impacts of corrosion on the underground nuclear canisters would lead to a significant increase in risk.

All of this can be avoided. If the nuclear waste at the two ISFSIs is transferred into thick-walled casks and then moved to a technically defensible storage facility at higher ground, the problems of ocean water and ground water intrusion can be avoided. As an added benefit, the waste would be easier to secure from an act of malfeasance.

2. POOR TECHNOLOGY In California, the storage tanks at gas stations must be double-walled; painful experience has shown that single-walled containers can leak gasoline into the groundwater system. With a double-walled fuel tank, if a leak occurs it can be detected and the storage container can be repaired or replaced before any gasoline is released. At San Onofre, we certainly should expect that some kind of leak prevention system would be in place to contain extremely toxic high-level radioactive waste. Additionally, the canisters should be able to be monitored and inspected. The thin-walled canisters at the San Onofre ISFSIs cannot be adequately monitored or inspected. Regulators and Holtec officials have stated that the canisters cannot be inspected from the inside or the outside for cracks or other degradation and that, even if damage could be identified, it would be impossible to fix.

To illustrate the importance of adequate monitoring, we analyze a scenario in which one vent of a canister clogs. We refer to a Holtec non-proprietary safety analysis report 1 that calculates a temperature rise to about 90% of the maximum permissible limit (MPL) in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This infers that within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the system will exceed the MPL rating and lead to a meltdown 2.

1 Table 4.I.9, page 1050, Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System.

USNRC Docket No.: 72-1014, Holtec Report No.: HI-2002444.

2 S. Alyokhina, Thermal analysis of certain accident conditions of dry spent nuclear fuel storage, Nuclear Engineering and Technology 50 (2018) 717-723.

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San Onofre Nuclear Waste Problems Through our own statistical analysis, 3 we prove that if the probability of clogging one of the vents during an event is 1%, then the chance that one of the 146 total vents (two vents on each of 73 canisters) will clog in such an event is 78%. This chance reduces to 53% if we reduce the probability of occurrence to .5% from 1%. Tsunamis followed by clogging are dependent events and thus the combined chance of such an event is about 11% during a 30-year period. The sea level rise, the rise of tide levels and the associated rise in the coastal aquifer are all interlinked, as discussed previously. These climate-related phenomena could cause serious damage to the ISFSIs. Therefore, close monitoring and the use of proven thick-walled cask technology for all nuclear waste storage containers is not only necessary but urgent. A mishap could imperil the lives and livelihoods of more than 8 million people who live within 50 miles of the ISFSIs.

2.1 NEAR MISS EVENT David Fritch, an industrial safety inspector turned whistleblower, remembers August 3, 2018, as a bad day. Fritch worked at San Onofre during a loading failure that left a fully-loaded 54-ton canister of high-level radioactive waste stuck on the lip of a guide ring. Above the 17-foot-tall canister, the slings that attached it to the behemoth loading rig had gone slack.

The canister was, hanging by about a quarter inch, Fritch told attendees of the community engagement panel on August 9. Its a bad day. That happened, and you havent heard about it, and thats not right. What we have is a canister that could have fallen 18 feet.

Subsequent investigations revealed that the operators and managers could not see Canister No.

29 as it was being loaded into the storage cavity and became stuck for nearly an hour.

Since the near-accident, regulators have halted further loading of canisters into the seaside storage vault and researchers have explored what could have happened if Canister No. 29 had fallen.

Our own research explores the basic physics of a fully-loaded 54-ton canister in free fall to extrapolate the upper energy involved in the initial impact.

For example, the falling canister could hit the steel-lined concrete floor of the nuclear waste storage facility with explosive energy greater than that of several large sticks of dynamite. The resultant damage to the canister could cause a large radiation release.

At point of contact at the bottom of the storage cavity, damage to the concrete and metal structure could ruin the cooling system. The damage to the concrete would equal that of a fully-loaded 18-wheeler truck, with a gross weight of 80,000 pounds, crashing into reinforced concrete at 23 miles per hour. Our preliminary calculations show the combination of the weight and velocity of the dropped canister exceeds the ISFSIs design criteria for tornado missiles, by a factor of 4. Future experiments should include drop tests of the actual canisters with non-3 Chakraborty and English, 2019, ES&H Risk Estimation from Interim Storage of SNF at the Beach: The San Onofre NPP, WM2019 Conference, March 3-7, 2019, Phoenix, Arizona, USA (under review).

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San Onofre Nuclear Waste Problems radioactive loads that simulate the weight of the spent fuel assemblies and fuel baskets to determine what would happen to the actual canisters.

Southern California Edison is set to move 73 canisters into the seaside storage vault and, at the time of publication, has moved 29. Each nuclear storage canister contains 37 spent fuel assemblies, which generate enormous amounts of heat. The systems are cooled by a simple air duct system, which could have been blocked by the damage caused by the canisters fall. If that had happened, great quantities of water would have been needed to cool the reaction and prevent or control a meltdown. The enveloping water would instantly become radioactive steam, as we saw at Fukushima. In the heavily-populated area surrounding San Onofre, however, radioactive steam could prompt the evacuation of millions of people. Whats more, since both the canister and the surrounding structure could be badly damaged, there would be no available way to pull the damaged canister from the storage cavity.

Nuclear Regulatory Commission (NRC) computer simulations show what happens when a nuclear storage canister with slightly thinner walls 4 drops from 19 feet. In the test, a canister falls from a transfer cask onto a storage pedestal. The canister failure rate was 28%. Similar calculations must be performed at San Onofre to determine if that storage system has a similar probability of canister failure. At 28%, that is more than a one-in-four chance of catastrophic failure. Would you fly on an airplane with those odds? Our analysis alone should place the NRC, policymakers and Edison on alert. A more substantial analysis must be completed to examine the potential damage that can be caused by a falling, fully-loaded 54-ton nuclear storage canister.

Continued loading of the nuclear waste into canisters threatens the lives and livelihood of more than 8 million people. Software and computer resources are available by which estimates can be made of the impacts of a dropped canister on both the reinforced concrete and the canister walls. The NRC-approved Holtec technical specifications state that a canister drop of more than 11 inches requires the contents of the canister to be inspected for damage. This specification assumed the canister was in a transfer cask. The impact of an un-casked canister was never analyzed because Holtec and the NRC assumed it could never happen, citing triple-redundancy of the fuel transfer system. But a subsequent NRC inspection revealed that on August 3rd, all three components of this system simultaneously failed. Only the accidental snag of a quarter-inch of the 54-ton canister on the lip of the guide ring prevented a catastrophe.

Our research suggests the entire storage system may need to be redesigned to reduce the probability of canister failure to levels that are acceptable in such a highly-populated area.

4 Pg. 4-24 Table 12, NUREG-1864 - A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a Nuclear Power Plant, March 2007, A. Malliakos, NRC Project Manager Page 5 of 11

San Onofre Nuclear Waste Problems RESULTS 2.2 GOUGES IN DROPPED CANISTER In their 2007 report, the NRCs analysts did not consider the impact of gouges on the strength of canister walls. There was no need, the analysts and a Holtec official said, as gouges were not important to the system under examination. We disagree. A detailed analysis of gouging is necessary to properly evaluate the damage to Canister No. 29 during the botched loading and to every other canister loaded into the ISFSI.

We established preliminary results of such an analysis using the Brinell hardness scale approach to estimate the depth and width of expected gouges in 316 stainless steel, of which the Holtec canisters at San Onofre is made.

While the canister is stuck, the guide ring gouges the bottom of the canister.

As the canister drops it is gouged on two sides by a combination of the guide ring, the storage cavity wall and the inner diameter of the transfer cask. This gouging absorbs some of the kinetic energy of the canister.

When the canister smashes into the bottom of the cavity, the kinetic energy and momentum from the fall will be dissipated by damage to:

the ISFSI;

the canister; and

the contents of the canister.

The formation process of gouges will exert a force on the canister. This is the force, P, shown in Figure 2.

Figure 2. Brinell hardness scale calculation. Credit: The Samuel Lawrence Foundation.

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San Onofre Nuclear Waste Problems In Figure 3, the width of a gouge is shown in relationship to the canisters weight. The expected range of gouge widths is shown in Figure 3. A variety of indenter widths are used as a surrogate for the gouging. The gouging widths range from 2 mm to 16 mm. This is highly significant, since the thickness of the nuclear canisters is 5/8, which is close to 16 mm. We recommend that tests be performed on actual canisters to experimentally determine the accuracy of these predictions.

Canister Gouge Width vs. % Canister Weight 16 14 Canister Gouge Width (mm) 12 16 mm 10 10 mm 8

8 mm 6

6 mm 4

3 mm 2 2 mm Canister thickness = 5/8 = 16 mm 0

0 10 20 30 40 50 60 70 80 90 100

% Canister Weight Figure 3. Calculated penetration width of gouge as a function of load for different intender diameter.

The hardness number in Brinell scale for stainless steel 316 (BHN) is 217 kgf/mm2. Saturated zone is eliminated.

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San Onofre Nuclear Waste Problems The expected range of gouge depths is shown in Figure 4. A variety of indenter depths are used as a surrogate for the gouging. The gouging depths expected to be found range from 1 mm to 4.5 mm. This is highly significant, since 4.5 mm is 28% of the thickness of the nuclear storage canister.

Canister Gouge Depth vs. % Canister Weight 5

10 mm 4

Canister Gouge Depth (mm) 8 mm 3

6 mm 2

3 mm 1

Canister thickness = 5/8" = 16 mm 0

0 20 40 60 80 100

% Canister Weight Figure 4. Calculated penetration depth of gouge as a function of load for different intender diameter. The hardness number in Brinell scale for stainless steel 316 (BHN) is 217 kgf/mm2.

2.3 GOUGES DURING ROUTINE LOADING Extensive gouging will also occur during routine loading of the nuclear storage canister into the storage cavity. By moving the Vertical Cask Transporter, shown in Figure 5, crude adjustments can be made to the alignment of the canister as it is lowered into the storage cavity. The bulky, tank-like machine travels on steel treads, like those found on earth-moving or military equipment. The transporter is not equipped to make the fine adjustments required to insert the nuclear storage canister into the narrow spacing of the storage cavity without banging the canister against the guide ring. This banging gouges the canister and causes the canister to move side-to-side, similar to a pendulum. An Edison official has referred to this process as jiggling. This jiggling process continues for 15 to 30 minutes as the canister is lowered to the bottom of the storage cavity. Each jiggle causes the type of gouging shown in Figure 3 and Page 8 of 11

San Onofre Nuclear Waste Problems Figure 4. We expect that this routine loading process produces a multitude of gouges that significantly damage the canister walls, rendering them unsuitable for storage of nuclear waste.

Figure 5. Vertical Cask Transporter during downloading and alignment of a canister.

Credit: San Onofre Special Inspection Webinar Presentation (NRC).

We strongly recommend that a sampling of the canisters previously lowered into the storage vault be removed and inspected so the extent of gouging can be experimentally determined.

We expect the damage will be so severe that the current ISFSI will need to be replaced.

3. POOR MANAGEMENT During the late 1970s and early 1980s, Rear Admiral Len Hering, USN (ret) served as a Nuclear Weapons Safety Officer, Handling Officer and Surety Officer. Admiral Hering provides the following assessment of management practices at the S.O.N.G.S. ISFSI.

When it comes to the handling and movement of nuclear material, you would expect that only those specifically qualified and trained for such an important task would be deployed to ensure the safe movement of that material. In the Department of Defense (DOD), strict requirements are in place to make sure this very dangerous material is properly handled, transported and stowed.

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San Onofre Nuclear Waste Problems The DOD and Navy programs were created and built to make certain nuclear material was secure, safely handled and accounted for. Every person who has any contact with nuclear material is required to have a security clearance. A two-person rule is in effect at all times.

Personnel at all levels perform countless hours of training, obtain certifications of qualification, and complete rigorous inspection and training events to both prove and assure their proficiency in performing the job they are assigned. All of this is all done before anyone is permitted to even gaze upon a real weapon.

Handling gear and all aspects of the evolution are vigilantly maintained, inspected, weight-tested and inspected again. Cranes and dollies or hoist equipment are tested, placed under extreme loading conditions and prepared for specific tasks. Nothing goes untested. Nothing.

We leave nothing to chance and we never hypothetically presume. If it isnt tested and proven, it isnt done with the actual material in question.

Ashore, and specifically at S.O.N.G.S, I find that virtually none of the protocols that should be expected for the safe handling of this dangerous material are present. I find that personnel and companies are being hired virtually off the street, no specific qualification standards are present or for that matter even required, training is not specific to the risks of the material involved, and there is no fully-qualified and certified team assembled for this highly-critical operation. They have not been required to conduct dry runs to ensure handling teams are proficient and, more importantly, they have never trained specifically to be ready to execute emergency procedures should the unexpected occur. The manuals are not on site, nor are they being followed to step a team through the evolution of moving the nuclear waste. Team leaders have no specific handling qualifications or training. Even the industrial safety inspectors are not specifically nuclear-certified but are general industrial specialists. No manuals are available for procedural review and, by their own admission, the required number of safety officials are often absent during movement of the nuclear storage canisters. In the Navy, if a near-accident such as the one at S.O.N.G.S is uncovered, the Commanding Officer, Weapons Officer -- and anyone else with a significant position on the team -- are relieved. The ship is then ordered to stand-down while a team of experts off-loads its cargo.

The widely reported incident in which a 54-ton, thin-walled container nearly fell 18 feet while it was being lowered into its silo rocked me to the core. What made things worse was narrative in a follow-up report that stated the canister was left suspended for nearly an hour, held up by a mere guide ring installed in the silo, cables slack and operators clueless. There is no doubt that this incident occurred because those on-scene were completely unqualified, unprepared, untrained and incompetent. This very dangerous operation was being performed as if this crew were moving a simple stack of wood around a construction site when, in actuality, the crew was conducting one of the most dangerous operations in the industrial sector. No one was relieved, fired or held accountable. The investigation being conducted is flawed in that those responsible for this deplorable safety environment are the same people who will feed findings to the investigation.

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San Onofre Nuclear Waste Problems The handling of nuclear waste at San Onofre and other sites across our country should scare every single American. We have a regulatory agency that has failed to make sure the most basic safety precautions are being applied to one of the most dangerous industrial evolutions of our time. The number of waivers being issued where safety is of concern is staggering.

In the DOD, the reason why there were and continue to be no significant accidents with the handling of nuclear material is because there are no waivers and there are no quick wins.

Workers are fully qualified, inspected and certified to handle this very dangerous material. In this case, there is no room for error. One mistake is too many. It is my professional opinion that we need to hit the reset button before a disaster of unparalleled portion occurs.

CONCLUSION The nuclear waste at San Onofre requires a much better storage configuration and must be moved to a technically defensible storage facility to reduce threats. From a security standpoint, the waste should be moved further away from major transportation corridors. The thin-walled nuclear waste storage canisters are at risk of failure due to gouging when downloaded into the seaside storage vault. Once lowered into the storage system, the canisters cannot be thoroughly inspected, monitored or repaired. A near-accident on August 3rd demonstrated that safety protocols are lacking, and that further study is needed to understand the consequences of dropping a fully-loaded 54-ton canister of nuclear waste. The incident revealed that the loading equipment is imprecise and revealed a pattern of mismanagement in canister loading procedure. A complete analysis of canister loading procedure and comprehensive risk assessment must be conducted by an independent party with absolute transparency. If an accident, natural disaster, negligence, or an act of terrorism were to cause a large-scale release of radiation, the health and safety of 8.4 million people within a 50-mile radius would be put at risk. To secure the nuclear waste properly, we recommend a permanent stop to the loading of nuclear storage canisters into the seaside storage vault, placing spent fuel into reliable canisters that can be monitored, inspected and repaired, and moving these canisters to an acceptable storage facility at a significantly higher elevation.

ACKNOWLEDGEMENTS We thank UCSD Departments of Chemistry and Biochemistry and The Samuel Lawrence Foundation. For more information visit www.samuellawrencefoundation.org/nuclear-energy.

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Case 3:19-cv-01635-JM-MSB Document 1-29 Filed 08/29/19 PageID.886 Page 1 of 6 EXHIBIT 28

Case 3:19-cv-01635-JM-MSB Document 1-29 Filed 08/29/19 PageID.887 Page 2 of 6 Case 3:19-cv-01635-JM-MSB Document 1-29 Filed 08/29/19 PageID.888 Page 3 of 6 Case 3:19-cv-01635-JM-MSB Document 1-29 Filed 08/29/19 PageID.889 Page 4 of 6 Case 3:19-cv-01635-JM-MSB Document 1-29 Filed 08/29/19 PageID.890 Page 5 of 6 Case 3:19-cv-01635-JM-MSB Document 1-29 Filed 08/29/19 PageID.891 Page 6 of 6

NRC FORM 651 U.S. NUCLEAR REGULATORY COMMISSION (10-2004) 10 CFR 72 CERTIFICATE OF COMPLIANCE FOR SPENT FUEL STORAGE CASKS Page 1 of 4 The U.S. Nuclear Regulatory Commission is issuing this Certificate of Compliance pursuant to Title 10 of the Code of Federal Regulations, Part 72, "Licensing Requirements for Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste" (10 CFR Part 72). This certificate is issued in accordance with 10 CFR 72.238, certifying that the storage design and contents described below meet the applicable safety standards set forth in 10 CFR Part 72, Subpart L, and on the basis of the Final Safety Analysis Report (FSAR) of the cask design. This certificate is conditional upon fulfilling the requirements of 10 CFR Part 72, as applicable, and the conditions specified below.

Certificate No. Effective Expiration Date Docket No. Amendment No. Amendment Effective Date Package Identification No.

Date 1040 TBD TBD 72-1040 0 USA/72-1040 Issued To: (Name/Address)

Holtec International Holtec Center 555 Lincoln Drive West Marlton, NJ 08053 Safety Analysis Report Title Holtec International Final Safety Analysis Report for the HI-STORM UMAX Canister Storage System This certificate is conditioned upon fulfilling the requirements of 10 CFR Part 72, as applicable, the attached Appendix A (Technical Specifications) and Appendix B (Approved Contents and Design Features), and the conditions specified below:

APPROVED SPENT FUEL STORAGE CASK Model No.: HI-STORM UMAX Canister Storage System DESCRIPTION:

The HI-STORM UMAX Canister Storage System consists of the following components: (1) interchangeable multi-purpose canisters (MPCs), which contain the fuel; (2) underground Vertical Ventilated Modules (VVMs),

which contains the MPCs during storage; and (3) a transfer cask (HI-TRAC VW), which contains the MPC during loading, unloading and transfer operations. The MPC stores up to 37 pressurized water reactor fuel assemblies or up to 89 boiling water reactor fuel assemblies.

The HI-STORM UMAX Canister Storage System is certified as described in the UMAX Final Safety Analysis Report (FSAR) supplemented by the information on the MPCs and transfer cask in the HI-STORM FW FSAR, and in the U. S. Nuclear Regulatory Commissions (NRC) Safety Evaluation Report (SER) accompanying the Certificate of Compliance (CoC).

The MPC is the confinement system for the stored fuel. It is a welded, cylindrical canister with a honeycombed fuel basket, a baseplate, a lid, a closure ring, and the canister shell. All MPC components that may come into contact with spent fuel pool water or the ambient environment are made entirely of stainless steel or passivated aluminum/aluminum alloys. The canister shell, baseplate, lid, vent and drain port cover plates, and closure ring are the main confinement boundary components. All confinement boundary components are made entirely of stainless steel. The honeycombed basket provides criticality control.

NRC FORM 651 U.S. NUCLEAR REGULATORY COMMISSION (3-1999) 10 CFR 72 CERTIFICATE OF COMPLIANCE Certificate No. 1040 FOR SPENT FUEL STORAGE CASKS Amendment No. 0 Supplemental Sheet Page 2 of 4 DESCRIPTION (continued)

There are two types of MPCs permitted for storage in HI-STORM UMAX VVM: the MPC-37 and MPC-89. The number suffix indicates the maximum number of fuel assemblies permitted to be loaded in the MPC. Both MPC models have the same external diameter.

The HI-TRAC VW transfer cask provides shielding and structural protection of the MPC during loading, unloading, and movement of the MPC from the cask loading area to the VVM. The transfer cask is a multi-walled (carbon steel/lead/carbon steel) cylindrical vessel with a neutron shield jacket attached to the exterior and a retractable bottom lid used during transfer operations.

The HI-STORM UMAX VVM utilizes a storage design identified as an air-cooled vault or caisson. The HI-STORM UMAX VVM relies on vertical ventilation instead of conduction through the fill material around the VVM, as it is essentially a below-grade storage cavity. Air inlets and an air outlet allow air to circulate naturally through the cavity to cool the MPC inside. The subterranean steel structure is seal welded to prevent ingress of any groundwater in the MPC storage cavity from the surrounding subgrade, and it is mounted on a stiff foundation. The surrounding subgrade and a top surface pad provide significant radiation shielding. A loaded MPC is stored within the HI-STORM UMAX VVM in a vertical orientation.

CONDITIONS

1. OPERATING PROCEDURES Written operating procedures shall be prepared for handling, loading, movement, surveillance, and maintenance. The users site-specific written operating procedures shall be consistent with the technical basis described in Chapter 9 of the FSAR.

2. ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Written acceptance tests and a maintenance program shall be prepared consistent with the technical basis described in Chapter 10 of the FSAR. At completion of welding the MPC shell to baseplate, an MPC confinement weld helium leak test shall be performed using a helium mass spectrometer. This test shall include the base metals of the MPC shell and baseplate. A helium leakage test shall also be performed on the base metal of the fabricated MPC lid. The confinement boundary welds leakage rate test shall be performed in accordance with ANSI N14.5 to leaktight criterion. If a leakage rate exceeding the acceptance criteria is detected, then the area of leakage shall be determined and the area repaired per ASME Code Section III, Subsection NB, Article NB-4450 requirements. Re-testing shall be performed until the leakage rate acceptance criterion is met.

3. QUALITY ASSURANCE Activities in the areas of design, purchase, fabrication, assembly, inspection, testing, operation, maintenance, repair, modification of structures, systems and components, and decommissioning that are important-to-safety shall be conducted in accordance with a Commission-approved quality assurance program which satisfies the applicable requirements of 10 CFR Part 72, Subpart G, and which is established, maintained, and executed with regard to the storage system

NRC FORM 651 U.S. NUCLEAR REGULATORY COMMISSION (3-1999) 10 CFR 72 CERTIFICATE OF COMPLIANCE Certificate No. 1040 FOR SPENT FUEL STORAGE CASKS Amendment No. 0 Supplemental Sheet Page 3 of 4

4. HEAVY LOADS REQUIREMENTS Each lift of an MPC or a HI-TRAC VW transfer cask must be made in accordance to the existing heavy loads requirements and procedures of the licensed facility at which the lift is made. A plant-specific review of the heavy load handling procedures (under 10 CFR 50.59 or 10 CFR 72.48, as applicable) is required to show operational compliance with existing plant specific heavy loads requirements. Lifting operations outside of structures governed by 10 CFR Part 50 must be in accordance with Section 5.2 of Appendix A.

5. APPROVED CONTENTS Contents of the HI-STORM UMAX Canister Storage System must meet the fuel specifications given in Appendix B to this certificate.

6. DESIGN FEATURES Features or characteristics for the site or system must be in accordance with Appendix B to this certificate.

7. CHANGES TO THE CERTIFICATE OF COMPLIANCE The holder of this certificate who desires to make changes to the certificate, which includes Appendix A (Technical Specifications) and Appendix B (Approved Contents and Design Features), shall submit an application for amendment of the certificate.

8. PRE-OPERATIONAL TESTING AND TRAINING EXERCISE A dry run training exercise of the loading, closure, handling, unloading, and transfer of the HI-STORM UMAX Canister Storage System shall be conducted by the licensee prior to the first use of the system to load spent fuel assemblies. The training exercise shall not be conducted with spent fuel in the MPC. The dry run may be performed in an alternate step sequence from the actual procedures, but all steps must be performed. The dry run shall include, but is not limited to the following:

a. Moving the MPC and the transfer cask into the spent fuel pool or cask loading pool.

b. Preparation of the HI-STORM UMAX Canister Storage System for fuel loading.

c. Selection and verification of specific fuel assemblies to ensure type conformance.

d. Loading specific assemblies and placing assemblies into the MPC (using a dummy fuel assembly),

including appropriate independent verification.

e. Remote installation of the MPC lid and removal of the MPC and transfer cask from the spent fuel pool or cask loading pool.

f. MPC welding, NDE inspections, pressure testing, draining, moisture removal (by vacuum drying or forced helium dehydration, as applicable), and helium backfilling. (A mockup may be used for this dry-run exercise.)

g. Transfer of the MPC from the transfer cask to the VVM.

NRC FORM 651 U.S. NUCLEAR REGULATORY COMMISSION (3-1999) 10 CFR 72 CERTIFICATE OF COMPLIANCE Certificate No. 1040 FOR SPENT FUEL STORAGE CASKS Amendment No. 0 Supplemental Sheet Page 4 of 4

h. HI-STORM UMAX Canister Storage System unloading, including flooding MPC cavity and removing MPC lid welds. (A mockup may be used for this dry-run exercise.)

Any of the above steps can be omitted if the site has already successfully loaded a Holtec MPC System.

9. AUTHORIZATION The HI-STORM UMAX Canister Storage System, which is authorized by this certificate, is hereby approved for general use by holders of 10 CFR Part 50 licenses for nuclear reactors at reactor sites under the general license issued pursuant to 10 CFR 72.210, subject to the conditions specified by 10 CFR 72.212, this certificate, and the attached Appendices A and B. The HI-STORM UMAX Canister Storage System may be fabricated and used in accordance with any approved amendment to CoC No. 1040 listed in 10 CFR 72.214.

Each of the licensed HI-STORM UMAX Canister Storage System components (i.e., the MPC, overpack, and transfer cask), if fabricated in accordance with any of the approved CoC Amendments, may be used with one another provided an assessment is performed by the CoC holder that demonstrates design compatibility.

FOR THE U. S. NUCLEAR REGULATORY COMMISSION DRAFT Michele M. Sampson, Chief Licensing Branch Division of Spent Fuel Storage and Transportation Office of Nuclear Material Safety and Safeguards Washington, DC 20555 Dated TBD Attachments:

1. Appendix A

2. Appendix B

CERTIFICATE OF COMPLIANCE NO. 1040 APPENDIX A TECHNICAL SPECIFICATIONS FOR THE HI-STORM UMAX CANISTER STORAGE SYSTEM

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1-1 1.1 Definitions ............................................................................................ 1.1-1 1.2 Logical Connectors .............................................................................. 1.2-1 1.3 Completion Times ................................................................................ 1.3-1 1.4 Frequency ............................................................................................ 1.4-1 2.0 NOT USED 2.0-1 3.0 LIMITING CONDITIONS FOR OPERATION (LCO) APPLICABILITY ............ 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ............................. 3.0-2 3.1 SFSC INTEGRITY ............................................................................ 3.1.1-1 3.1.1 Multi-Purpose Canister (MPC) ............................................... 3.1.1-1 3.1.2 SFSC Heat Removal System ................................................. 3.1.2-1 3.1.3 MPC Cavity Reflooding .......................................................... 3.1.3-1 3.2 SFSC RADIATION PROTECTION ................................................... 3.2.1-1 3.2.1 TRANSFER CASK Surface Contamination ............................ 3.2.1-1 3.3 SFSC CRITICALITY CONTROL ....................................................... 3.3.1-1 3.3.1 Boron Concentration .............................................................. 3.3.1-1 Table 3-1 MPC Cavity Drying Limits .................................................................... 3.4-1 Table 3-2 MPC Helium Backfill Limits .................................................................. 3.4-4 4.0 NOT USED 4.0-1 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS ...................................... 5.0-1 5.1 Radioactive Effluent Control Program .................................................. 5.0-1 5.2 Transport Evaluation Program ............................................................. 5.0-2 5.3 Radiation Protection Program .............................................................. 5.0-3 Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.1-i

Definitions 1.1 1.0 USE AND APPLICATION

NOTE-----------------------------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

1.1 Definitions Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

AMBIENT TEMPERATURE AMBIENT TEMPERATURE for Short Term Operations (operations involving use of the HI-TRAC, a Lifting device, and/or an on-site transport device) is defined as the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months average of the local temperature as forecast by the National Weather Service.

DAMAGED FUEL ASSEMBLY DAMAGED FUEL ASSEMBLIES are fuel assemblies with known or suspected cladding defects, as determined by a review of records, greater than pinhole leaks or hairline cracks, empty fuel rod locations that are not filled with dummy fuel rods, missing structural components such as grid spacers, whose structural integrity has been impaired such that geometric rearrangement of fuel or gross failure of the cladding is expected based on engineering evaluations, or that cannot be handled by normal means. Fuel assemblies that cannot be handled by normal means due to fuel cladding damage are considered FUEL DEBRIS.

DAMAGED FUEL CONTAINER DFCs are specially designed enclosures for (DFC) DAMAGED FUEL ASSEMBLIES or FUEL DEBRIS which permit gaseous and liquid media to escape while minimizing dispersal of gross particulates. DFCs authorized for use in the HI-STORM UMAX System are as follows:

1. Holtec Generic BWR design

2. Holtec Generic PWR design Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.1-1

Definitions 1.1 1.1 Definitions Term Definition FUEL DEBRIS FUEL DEBRIS is ruptured fuel rods, severed rods, loose fuel pellets, containers or structures that are supporting these loose fuel assembly parts, or fuel assemblies with known or suspected defects which cannot be handled by normal means due to fuel cladding damage.

FUEL BUILDING The FUEL BUILDING is the site-specific power plant facility, governed by the regulations of 10 CFR Part 50, where the loaded OVERPACK or TRANSFER CASK is transferred to or from the transporter.

Spent nuclear fuel rod with a cladding defect that GROSSLY BREACHED could lead to the release of fuel particulate greater SPENT FUEL ROD than the average size fuel fragment for that particular assembly. A gross cladding breach may be confirmed by visual examination, through a review of reactor operating records indicating the presence of heavy metal isotopes, or other acceptable inspection means.

LOADING OPERATIONS LOADING OPERATIONS include all licensed activities on a TRANSFER CASK while it is being loaded with fuel assemblies. LOADING OPERATIONS begin when the first fuel assembly is placed in the MPC and end when the TRANSFER CASK is suspended from or secured on the transporter. LOADING OPERATIONS does not include MPC TRANSFER.

MULTI-PURPOSE CANISTER MPCs are the sealed spent nuclear fuel canisters (MPC) which consist of a honeycombed fuel basket contained in a cylindrical canister shell which is welded to a baseplate, lid with welded port cover plates, and closure ring. The MPC provides the confinement boundary for the contained radioactive materials.

MPC TRANSFER MPC TRANSFER begins when the MPC is lifted off the TRANSFER CASK bottom lid and ends when the MPC is supported from beneath by the OVERPACK (or the reverse).

NON-FUEL HARDWARE NON-FUEL HARDWARE is defined as Burnable Poison Rod Assemblies (BPRAs), Thimble Plug Devices (TPDs), Control Rod Assemblies (CRAs),

Axial Power Shaping Rods (APSRs), Wet Annular Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.1-2

Definitions 1.1 1.1 Definitions Term Definition Burnable Absorbers (WABAs), Rod Cluster Control Assemblies (RCCAs), Control Element Assemblies (CEAs), Neutron Source Assemblies (NSAs), water displacement guide tube plugs, orifice rod assemblies, instrument tube tie rods (ITTRs),

vibration suppressor inserts, and components of these devices such as individual rods.

OVERPACK For the HI-STORM UMAX, the term OVERPACK is synonyms with the term VVM defined below.

PLANAR-AVERAGE INITIAL PLANAR AVERAGE INITIAL ENRICHMENT is the ENRICHMENT average of the distributed fuel rod initial enrichments within a given axial plane of the assembly lattice.

REPAIRED/RECONSITUTED Spent nuclear fuel assembly which contains dummy FUEL ASSEMBLY fuel rods that displaces an amount of water greater than or equal to the original fuel rods and/or which contains structural repairs so it can be handled by normal means.

SPENT FUEL STORAGE SFSCs are containers approved for the storage of CASKS (SFSCs) spent fuel assemblies at the ISFSI. The HI-STORM UMAX SFSC System consists of the OVERPACK and its integral MPC.

STORAGE OPERATIONS STORAGE OPERATIONS include all licensed activities that are performed at the ISFSI while an SFSC containing spent fuel is situated within the ISFSI perimeter. STORAGE OPERATIONS does not include MPC TRANSFER.

TRANSFER CASK TRANSFER CASKs are containers designed to contain the MPC during and after loading of spent fuel assemblies, and prior to and during unloading and to transfer the MPC to or from the OVERPACK.

TRANSPORT OPERATIONS TRANSPORT OPERATIONS include all licensed activities performed on a TRANSFER CASK loaded with one or more fuel assemblies when it is being moved after LOADING OPERATIONS or before UNLOADING OPERATIONS. TRANSPORT OPERATIONS begin when the TRANSFER CASK is first suspended from or secured on the transporter and end when the TRANSFER CASK is at its destination and no longer secured on or suspended from the transporter. TRANSPORT OPERATIONS includes Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.1-3

Definitions 1.1 1.1 Definitions Term Definition MPC TRANSFER.

VERTICAL VENTILATED The VVM is a subterranean type overpack which MODULE (VVM) receives and contains the sealed MPC for interim storage at the ISFSI. The VVM supports the MPC in a vertical orientation and provide gamma and neutron shielding and also provides air flow through cooling passages to promote heat transfer from the MPC to the environs.

UNDAMAGED FUEL UNDAMAGED FUEL ASSEMBLIES are: a) fuel ASSEMBLY assemblies without known or suspected cladding defects greater than pinhole leaks or hairline cracks and which can be handled by normal means; or b) a BWR fuel assembly with an intact channel, a maximum planar average initial of 3.3 wt% U-235, without known or suspected GROSSLY BREACHED SPENT FUEL RODS, and which can be handled by normal means. An UNDAMAGED FUEL ASSEMBLY may be a REPAIRED/RECONSTITUTED FUEL ASSEMBLY.

UNLOADING OPERATIONS UNLOADING OPERATIONS include all licensed activities on an SFSC to be unloaded of the contained fuel assemblies. UNLOADING OPERATIONS begin when the TRANSFER CASK is no longer suspended from or secured on the transporter and end when the last fuel assembly is removed from the SFSC.

UNLOADING OPERATIONS does not include MPC TRANSFER.

ZR ZR means any zirconium-based fuel cladding or fuel channel material authorized for use in a commercial nuclear power plant reactor.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.1-4

Definitions 1.1 PURPOSE The purpose of this section is to explain the meaning of logical connectors.

Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.

BACKGROUND Several levels of logic may be used to state Required Actions.

These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentions of the logical connectors.

When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.1-5

Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors EXAMPLES The following examples illustrate the use of logical connectors.

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 VERIFY . . .

AND A.2 Restore . . .

In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.2-1

Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES EXAMPLE 1.2-2 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Stop . . .

OR A.2.1 Verify . . .

AND A.2.2.1 Reduce . . .

OR A.2.2.2 Perform . . .

OR A.3 Remove . . .

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three ACTIONS may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.2-2

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

BACKGROUND Limiting Conditions for Operation (LCOs) specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the HI-STORM UMAX System is in a specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time.

An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the HI-STORM UMAX System is not within the LCO Applicability.

Once a Condition has been entered, subsequent subsystems, components, or variables expressed in the Condition, discovered to be not within limits, will not result in separate entry into the Condition unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.3-1

Completion Times 1.3 1.3 Completion Times (continued)

EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Perform Action B.1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action and associated AND Completion Time not met. B.2 Perform Action B.2 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered.

The Required Actions of Condition B are to complete action B.1 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND complete action B.2 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A total of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed for completing action B.1 and a total of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (not 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ) is allowed for completing action B.2 from the time that Condition B was entered. If action B.1 is completed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , the time allowed for completing action B.2 is the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months because the total time allowed for completing action B.2 is 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.3-2

Completion Times 1.3 1.3 Completion Times (continued)

EXAMPLES EXAMPLE 1.3-2 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One system A.1 Restore system to 7 days not within limit. within limit.

B. Required B.1 Complete action 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action and B.1.

associated Completion AND Time not met.

B.2 Complete action 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months B.2.

When a system is determined not to meet the LCO, Condition A is entered. If the system is not restored within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the system is restored after Condition B is entered, Conditions A and B are exited, and therefore, the Required Actions of Condition B may be terminated.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.3-3

Completion Times 1.3 1.3 Completion Times (continued)

EXAMPLES EXAMPLE 1.3-3 (continued)

ACTIONS

NOTE------------------------------------------

Separate Condition entry is allowed for each component.

CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Restore 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months compliance with LCO.

B. Required B.1 Complete action 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and B.1.

associated Completion AND Time not met.

B.2 Complete action 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.2.

The Note above the ACTIONS table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.

The Note allows Condition A to be entered separately for each component, and Completion Times tracked on a per component basis. When a component is determined to not meet the LCO, Condition A is entered and its Completion Time starts. If subsequent components are determined to not meet the LCO, Condition A is entered for each component and separate Completion Times start and are tracked for each component.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.3-4

Completion Times 1.3 1.3 Completion Times (continued)

IMMEDIATE When "Immediately" is used as a Completion Time, the Required COMPLETION Action should be pursued without delay and in a controlled manner.

TIME Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.3-5

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated Limiting Condition for Operation (LCO). An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.4-1

Frequency 1.4 1.4 Frequency (continued)

EXAMPLES The following examples illustrate the various ways that Frequencies are specified.

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify pressure within limit 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an extension of the time interval to 1.25 times the interval specified in the Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment or variables are outside specified limits, or the facility is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the facility is in a condition specified in the Applicability of the LCO, the LCO is not met in accordance with SR 3.0.1.

If the interval as specified by SR 3.0.2 is exceeded while the facility is not in a condition specified in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the specified condition. Failure to do so would result in a violation of SR 3.0.4 (continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.4-2

Frequency 1.4 1.4 Frequency (continued)

EXAMPLES (continued) EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to starting activity AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time the example activity is to be performed, the Surveillance must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to starting the activity.

The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2.

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If the specified activity is canceled or not performed, the measurement of both intervals stops. New intervals start upon preparing to restart the specified activity.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 1.4-3

2.0 2.0 This section is intentionally left blank Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 2.0-1

LCO Applicability 3.0 3.0 LIMITING CONDITIONS FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during specified conditions in the Applicability, except as provided in LCO 3.0.2.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.

LCO 3.0.3 Not applicable.

LCO 3.0.4 When an LCO is not met, entry into a specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in specified conditions in the Applicability that are required to comply with ACTIONS or that are related to the unloading of an SFSC.

LCO 3.0.5 Equipment removed from service or not in service in compliance with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate it meets the LCO or that other equipment meets the LCO. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.0-1

LCO Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as once, the above interval extension does not apply. If a Completion Time requires periodic performance on a once per... basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.0-2

LCO Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.3 When the Surveillance is performed within the delay period and the (continued) Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

SR 3.0.4 Entry into a specified condition in the Applicability of an LCO shall not be made unless the LCO's Surveillances have been met within their specified Frequency. This provision shall not prevent entry into specified conditions in the Applicability that are required to comply with Actions or that are related to the unloading of an SFSC.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.0-3

Multi-Purpose Canister (MPC) 3.1.1 3.1 SFSC INTEGRITY 3.1.1 Multi-Purpose Canister (MPC)

LCO 3.1.1 The MPC shall be dry and helium filled.

Table 3-1 provides decay heat and burnup limits for forced helium dehydration (FHD) and vacuum drying.

APPLICABILITY: Prior to TRANSPORT OPERATIONS ACTIONS

NOTES---------------------------------------------------------

Separate Condition entry is allowed for each MPC.

COMPLETION CONDITION REQUIRED ACTION TIME 7 days A. MPC cavity vacuum A.1 Perform an engineering drying pressure or evaluation to determine the demoisturizer exit gas quantity of moisture left in temperature limit not the MPC.

met. AND 30 days A.2 Develop and initiate corrective actions necessary to return the MPC to compliance with Table 3-1.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.1.1-1

Multi-Purpose Canister (MPC) 3.1.1 ACTIONS (continued)

B. MPC helium backfill limit B.1 Perform an engineering 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months not met. evaluation to determine the impact of helium differential.

AND B.2.1 Develop and initiate 14 days corrective actions necessary to return the MPC to an analyzed condition by adding helium to or removing helium from the MPC.

OR B.2.2 Develop and initiate corrective actions necessary to demonstrate through analysis, using the models and methods from the HI-STORM UMAX FSAR, that all limits for MPC components and contents will be met.

C. MPC helium leak rate C.1 Perform an engineering 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limit for vent and drain evaluation to determine the port cover plate welds impact of increased helium not met. leak rate on heat removal capability and offsite dose.

AND C.2 Develop and initiate 7 days corrective actions necessary to return the MPC to compliance with SR 3.1.1.3.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.1.1-2

Multi-Purpose Canister (MPC) 3.1.1 D. Required Actions and D.1 Remove all fuel assemblies 30 days associated Completion from the SFSC.

Times not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify that the MPC cavity has been dried in Once, prior to accordance with the applicable limits in Table TRANSPORT 3-1. OPERATIONS SR 3.1.1.2 Verify MPC helium backfill quantity is within the Once, prior to limit specified in Table 3-2 for the applicable MPC TRANSPORT model. Re-performance of this surveillance is not OPERATIONS required upon successful completion of Action B.2.2.

SR 3.1.1.3 Verify that the helium leak rate through the MPC Once, prior to vent and drain port cover plates (confinement TRANSPORT welds and the base metal)meets the leaktight OPERATIONS criteria of ANSI N14.5-1997.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.1.1-3

SFSC Heat Removal System 3.1.2 3.1 SFSC INTEGRITY 3.1.2 SFSC Heat Removal System LCO 3.1.2 The SFSC Heat Removal System shall be operable

NOTE--------------------------------------------------

The SFSC Heat Removal System is operable when 50% or more of the inlet vent duct areas are unblocked and available for flow or when air temperature requirements are met.

APPLICABILITY: During STORAGE OPERATIONS.

ACTIONS

NOTE--------------------------------------------------

Separate Condition entry is allowed for each SFSC.

COMPLETION CONDITION REQUIRED ACTION TIME A. SFSC Heat Removal A.1 Remove blockage. N/A System operable, but partially (<50%) blocked.

B. SFSC Heat Removal B.1 Restore SFSC Heat 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months System inoperable. Removal System to operable status.

C. Required Action B.1 and C.1 Measure SFSC dose rates Immediately and associated Completion in accordance with the once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Time not met. Radiation Protection thereafter Program.

AND C.2.1 Restore SFSC Heat 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Removal System to operable status.

OR C.2.2 Transfer the MPC into a 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TRANSFER CASK.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.1.2-1

SFSC Heat Removal System 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2 Verify all VVM inlets and outlets duct screen are 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months free of blockage from solid debris or floodwater.

OR For VVMs with installed temperature monitoring 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months equipment, verify that the difference between the average VVM air outlet duct temperature and ISFSI ambient temperature is 80oF for VVMs containing MPC-37s and 85oF for VVMs containing MPC-89s.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.1.2-2

Fuel Cool-Down 3.1.3 3.1 SFSC INTEGRITY 3.1.3 MPC Cavity Reflooding LCO 3.1.3 The MPC cavity pressure shall be < 100 psig

NOTE--------------------------------------------------------

The LCO is only applicable to wet UNLOADING OPERATIONS.

APPLICABILITY: UNLOADING OPERATIONS prior to and during re-flooding.

ACTIONS

NOTE--------------------------------------------------------

Separate Condition entry is allowed for each MPC.

COMPLETION CONDITION REQUIRED ACTION TIME A. MPC cavity pressure A.1 Stop re-flooding operations Immediately not within limit. until MPC cavity pressure is within limit.

AND A.2 Ensure MPC vent port is not Immediately closed or blocked.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Ensure via analysis or direct measurement that Once, prior to MPC cavity pressure is within limit. MPC re-flooding operations.

OR Once every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter when using direct measurement.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.1.3-1

TRANSFER CASK Surface Contamination 3.2.1 3.2 SFSC RADIATION PROTECTION.

3.2.1 TRANSFER CASK Surface Contamination.

LCO 3.2.1 Removable contamination on the exterior surfaces of the TRANSFER CASK and accessible portions of the MPC shall each not exceed:

a. 1000 dpm/100 cm2 from beta and gamma sources

b. 20 dpm/100 cm2 from alpha sources.

APPLICABILITY: During TRANSPORT OPERATIONS.

ACTIONS

NOTE--------------------------------------------------------

Separate Condition entry is allowed for each TRANSFER CASK.

COMPLETION CONDITION REQUIRED ACTION TIME A. TRANSFER CASK or A.1 Restore removable surface 7 days MPC removable surface contamination to within contamination limits not limits.

met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify that the removable contamination on the Once, prior to exterior surfaces of the TRANSFER CASK and TRANSPORT accessible portions of the MPC containing fuel is OPERATIONS within limits.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.2.1-1

Boron Concentration 3.3.1 3.3 SFSC CRITICALITY CONTROL 3.3.1 Boron Concentration LCO 3.3.1 The concentration of boron in the water in the MPC shall meet the following limits for the applicable MPC model and the most limiting fuel assembly array/class to be stored in the MPC:

MPC-37: Minimum soluble boron concentration as required by the table below.

One or more Damaged Fuel All Undamaged Fuel Assemblies Assemblies or Fuel Debris Array/Class Maximum Initial Maximum Initial Maximum Initial Maximum Initial Enrichment Enrichment 5.0 Enrichment Enrichment 5.0 4.0 wt% 235U wt% 235U 4.0 wt% 235U wt% 235U (ppmb) (ppmb) (ppmb) (ppmb)

All 14x14 and 16x16A 1000 1500 1300 1800 All 15x15 and 17x17 1500 2000 1800 2300 For maximum initial enrichments between 4.0 wt% and 5.0 wt% 235U, the minimum soluble boron concentration may be determined by linear interpolation between the minimum soluble boron concentrations at 4.0 wt% and 5.0 wt%.

APPLICABILITY: During PWR fuel LOADING OPERATIONS with fuel and water in the MPC AND During PWR fuel UNLOADING OPERATIONS with fuel and water in the MPC.

ACTIONS

NOTE----------------------------------------------------

Separate Condition entry is allowed for each MPC.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.3.1-1

Boron Concentration 3.3.1 COMPLETION CONDITION REQUIRED ACTION TIME A. Boron concentration not A.1 Suspend LOADING Immediately within limit. OPERATIONS or UNLOADING OPERATIONS.

AND A.2 Suspend positive reactivity Immediately additions.

AND A.3 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE------------------------------------ Once, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to This surveillance is only required to be performed if the MPC is entering the submerged in water or if water is to be added to, or recirculated through the MPC. Applicability of

this LCO.

SR 3.3.1.1 Verify boron concentration is within the AND applicable limit using two independent measurements. Once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months thereafter.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.3.1-2

3.4 Tables Table 3-1 MPC Cavity Drying Limits MPC Type Cell Heat Load Limits Method of Moisture Fuel Burnup (Note 5) Removal (MWD/MTU) (Note 6) (Notes 1 and 2)

MPC-37 Figure 2.3-1, 2.3-2, or 2.3-3 of (Short Fuel) Appendix B MPC-37 Figure 2.3-1, 2.3-2, or 2.3-4 of All Assemblies (Standard Fuel) Appendix B VDS (Notes 3 and 4)

< 45,000 or FHD (Note 4)

MPC-37 Figure 2.3-5, 2.3-6, or 2.3-7 of (Long Fuel) Appendix B MPC-89 Figure 2.3-10 of Appendix B MPC-37 One or more (Short, Standard Figure 2.3-12 of Appendix B VDS (Notes 3 and 4) assemblies and Long Fuel) or FHD (Note 4)

> 45,000 MPC-89 Figure 2.3-13 of Appendix B MPC-37 Figure 2.3-1, 2.3-2, or 2.3-3 of (Short Fuel) Appendix B MPC-37 Figure 2.3-1, 2.3-2, or 2.3-4 of One or more (Standard Fuel) Appendix B assemblies FHD (Note 4)

> 45,000 MPC-37 Figure 2.3-5, 2.3-6, or 2.3-7 of (Long Fuel) Appendix B MPC-89 Figure 2.3-10 of Appendix B Notes:

1. VDS means a vacuum drying system. The acceptance criterion when using a VDS is the MPC cavity pressure shall be 3 torr for 30 minutes while the MPC is isolated from the vacuum pump.

2. FHD means a forced helium dehydration system. The acceptance criterion when using an FHD system is the gas temperature exiting the demoisturizer shall be 21oF for 30 minutes or the gas dew point exiting the MPC shall be 22.9oF for 30 minutes.

3. Vacuum drying of the MPC must be performed with the annular gap between the MPC and the TRANSFER CASK filled with water.

4. Heat load limits are set for each cell; see Appendix B Section 2.3.

5. The fuel assembly lengths loaded in MPC-37 are catalogued as short, standard and long fuel based on the active fuel lengths specified in Appendix B Table 2.1-4.

6. For additional aggregate heat load limits for storage, see Appendix B Table 2.3-1 Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.4-1

3.4 Tables Table 3-2 MPC Helium Backfill Limits 1 Helium Backfill Helium Backfill MPC Type Pressure Pressure Range Option (psig) 1 41.0 and 44.2 MPC-37 2 41.0 and 44.5 3 39.0 and 46.0 1 42.0 and 45.2 MPC-89 2 39.0 and 46.0 Note: For Permissible Aggregate Heat Load Limit for each helium backfill pressure option see Appendix B, Table 2.3-1.

1 Helium used for backfill of MPC shall have a purity of 99.995%. Pressure range is at a reference temperature of 70oF Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 3.4-2

4.0 4.0 This section is intentionally left blank Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 4.0-1

Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS The following programs shall be established, implemented and maintained.

5.1 Radioactive Effluent Control Program This program implements the requirements of 10 CFR 72.44(d).

a. The HI-STORM UMAX Canister Storage System does not create any radioactive materials or have any radioactive waste treatment systems.

Therefore, specific operating procedures for the control of radioactive effluents are not required. Specification 3.1.1, Multi-Purpose Canister (MPC), provides assurance that there are not radioactive effluents from the SFSC.

b. This program includes an environmental monitoring program. Each general license user may incorporate SFSC operations into their environmental monitoring programs for 10 CFR Part 50 operations.

c. An annual report shall be submitted pursuant to 10 CFR 72.44(d)(3).

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 5.0-1

Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS (continued) 5.2 Transport Evaluation Program

a. For lifting of the loaded MPC or TRANSFER CASK using equipment which is integral to a structure governed by 10 CFR Part 50 regulations, 10 CFR 50 requirements apply.

b. This program is not applicable when the TRANSFER CASK is in the FUEL BUILDING or is being handled by equipment providing support from underneath (i.e., on a rail car, heavy haul trailer, air pads, etc...).

c. The TRANSFER CASK when loaded with spent fuel, may be lifted to and carried at any height necessary during TRANSPORT OPERATIONS and MPC TRANSFER, provided the lifting equipment is designed in accordance with items 1, 2, and 3 below.

1. The metal body and any vertical columns of the lifting equipment shall be designed to comply with stress limits of ASME Section III, Subsection NF, Class 3 for linear structures. All vertical compression loaded primary members shall satisfy the buckling criteria of ASME Section III, Subsection NF.

2. The horizontal cross beam and any lifting attachments used to connect the load to the lifting equipment shall be designed, fabricated, operated, tested, inspected, and maintained in accordance with applicable sections and guidance of NUREG-0612, Section 5.1. This includes applicable stress limits from ANSI N14.6.

3. The lifting equipment shall have redundant drop protection features which prevent uncontrolled lowering of the load.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 5.0-2

Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS (continued) 5.3 Radiation Protection Program 5.3.1 Each cask user shall ensure that the Part 50 radiation protection program appropriately addresses dry storage cask loading and unloading, as well as ISFSI operations, including transport of the loaded TRANSFER CASK outside of facilities governed by 10 CFR Part 50. The radiation protection program shall include appropriate controls for direct radiation and contamination, ensuring compliance with applicable regulations, and implementing actions to maintain personnel occupational exposures As Low As Reasonably Achievable (ALARA). The actions and criteria to be included in the program are provided below.

5.3.2 As part of its evaluation pursuant to 10 CFR 72.212(b)(2)(i)(C), the licensee shall perform an analysis to confirm that the dose limits of 10 CFR 72.104(a) will be satisfied under the actual site conditions and ISFSI configuration, considering the planned number of casks to be deployed and the cask contents.

5.3.3 Based on the analysis performed pursuant to Section 5.3.2, the licensee shall establish individual cask surface dose rate limits for the TRANSFER CASK and the VVM to be used at the site. Total (neutron plus gamma) dose rate limits shall be established at the following locations:

a. The top of the VVM.

b. The side of the TRANSFER CASK

c. The outlet vents on the VVM 5.3.4 Notwithstanding the limits established in Section 5.3.3, the average of the measured dose rates on a loaded VVM or TRANSFER CASK shall not exceed the following values:

a. 30 mrem/hr (gamma + neutron) on the top of the closure lid of the VVM

b. 3500 mrem/hr (gamma + neutron) on the side of the TRANSFER CASK 5.3.5 The licensee shall measure the TRANSFER CASK and VVM surface neutron and gamma dose rates as described in Section 5.3.8 for comparison against the limits established in Section 5.3.3 or Section 5.3.4, whichever are lower.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 5.0-3

Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS (continued) 5.3 Radiation Protection Program (continued) 5.3.6 If the measured surface dose rates exceed the lower of the two limits established in Section 5.3.3 or Section 5.3.4, the licensee shall:

a. Administratively verify that the correct contents were loaded in the correct fuel storage cell locations.

b. Perform a written evaluation to verify whether a VVM at the ISFSI containing the as-loaded MPC will cause the dose limits of 10 CFR 72.104 to be exceeded.

c. Perform a written evaluation within 30 days to determine why the surface dose rate limits were exceeded.

5.3.7 If the evaluation performed pursuant to Section 5.3.6 shows that the dose limits of 10 CFR 72.104 will be exceeded, the MPC shall not be placed into a VVM or the MPC shall be removed from the VVM until appropriate corrective action is taken to ensure the dose limits are not exceeded.

5.3.8 TRANSFER CASK and VVM surface dose rates shall be measured at approximately the following locations:

a. A minimum of four (4) dose rate measurements shall be taken on the top of the VVM. These measurements shall be taken approximately 90 degrees apart around the circumference of the lid, approximately 18 inches radially inward from the edge of the lid.

b. A minimum of four (4) dose rate measurements shall be taken adjacent to the outlet vent duct screen of the VVM, approximately 90 degrees apart.

c. A minimum of four (4) dose rate measurements shall be taken on the side of the TRANSFER CASK approximately at the cask mid-height plane. The measurement locations shall be approximately 90 degrees apart around the circumference of the cask. Dose rates shall be measured between the radial ribs of the water jacket.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 5.0-4

Programs 5.0 5.0 ADMINISTRATIVE CONTROLS AND PROGRAMS (continued) 5.3 Radiation Protection Program (continued) 5.3.9 The Radiation Protection Space (RPS) is the prismatic subgrade buffer zone surrounding the VVMs in a loaded ISFSI. The RPS boundary is indicated in the Licensing Drawings in Section 1.5 of the system FSAR. The RPS boundary shall not be encroached upon during any site construction activity. The jurisdictional boundary of the RPS extends down from the top of the ISFSI pad to the elevation of the Bottom surface of the Support Foundation Pad. The ISFSI design shall ensure that there is no significant loss of shielding in the RPS due to a credible accident or an extreme environment event during construction activity involving excavation adjacent to the RPS boundary.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix A 5.0-5

CERTIFICATE OF COMPLIANCE NO. 1040 APPENDIX B APPROVED CONTENTS AND DESIGN FEATURES FOR THE HI-STORM UMAX CANISTER STORAGE SYSTEM

TABLE OF CONTENTS 1.0 DEFINITIONS ........................................................................................................ 1-1 2.0 APPROVED CONTENTS ...................................................................................... 2-1 2.1 Fuel Specifications and loading conditions ........................................................ 2-1 2.2 Violations ........................................................................................................... 2-1 2.3 Decay Heat Limits ........................................................................................... 2-15 Table 2.1-1 Fuel Assembly Limits .......................................................................... 2-2 Table 2.1-2 PWR Fuel Assembly Characteristics .................................................. 2-6 Table 2.1-3 BWR Fuel Assembly Characteristics .................................................. 2-9 Table 2.1-4 Classification of Fuel Assembly for MPC-37 in the HI-STORM UMAX System ...................................................................................................... 2-14 Table 2.3-1 Permissible Heat Load for long term storage .................................... 2-16 Figure 2.3-1 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 1 for Long-term Storage for Short and Standard Fuel ........................................ 2-19 Figure 2.3-2 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 2 for Long-Term Storage for Short and Standard Fuel ....................................... 2-20 Figure 2.3-3 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 3 for Long-term Storage for Short Fuel .............................................................. 2-21 Figure 2.3-4 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 3 for Long-term Storage for Standard Fuel ........................................................ 2-22 Figure 2.3-5 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 1 for Long-term Storage for Long Fuel ............................................................... 2-23 Figure 2.3-6 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 2 for Long-term Storage for Long Fuel ............................................................... 2-24 Figure 2.3-7 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 3 for Long-term Storage for Long Fuel ............................................................... 2-25 Figure 2.3-8 HI-STORM UMAX MPC-37 Permissible Heat Load for Short and Standard Fuel for Helium Backfill Option3 in Table 3-2 of Appendix A ....................................................................................... 2-26 Figure 2.3-9 HI-STORM UMAX MPC-37 Permissible Heat Load for Long Fuel for Helium Backfill Option 3 in Table 3-2 of Appendix A ......................... 2-27 Figure 2.3-10 HI-STORM UMAX MPC-89 Permissible Heat Load for Long-Term Storage ............................................................................................ 2-28 Figure 2.3-11 HI-STORM UMAX MPC-89 Permissible Heat Load for Helium Backfill Option 2 in Table 3-2 of Appendix A ................................................ 2-29 Figure 2.3-12 HI-STORM UMAX MPC-37 Permissible Threshold Heat Load for VDS High Burnup Fuel in Table 3-1 of Appendix A and Helium Backfill Option 3 in Table 3-2 of Appendix A............................................................. 2-30 Figure 2.3-13 HI-STORM UMAX MPC-89 Permissible Threshold Heat Load for VDS High Burnup Fuel in Table 3-1 of Appendix A and Helium Backfill Option Certificate of Compliance No. 1040 Amendment No. 0 Appendix B i

2 in Table 3-2 of Appendix A............................................................. 2-31 3.0 DESIGN FEATURES ............................................................................................. 3-1 3.1 Site .................................................................................................................... 3-1 3.2 Design Features Important for Criticality Control ............................................... 3-1 3.3 Codes and Standards ........................................................................................ 3-2 3.4 Site Specific Parameters and Analyses ........................................................... 3-10 3.5 Combustible Gas Monitoring During MPC Lid Welding and Cutting ................ 3-16 3.6 Periodic Corrosion Inspections for Underground Systems .............................. 3-16 Figure 3-1 SUBGRADE AND UNDERGRADE SPACE NOMENCLATURE....3-15 Table 3-1 List of ASME Code Alternatives for Multi-Purpose Canisters (MPCs) .... 3-3 Table 3-2 REFERENCE ASME CODE PARAGRAPHS FOR HI-STORM UMAX OVERPACK and HI-TRAC VW TRANSFER CASK, PRIMARY LOAD BEARING PARTS .................................................................................. 3-8 Table 3-3 LOAD COMBINATIONS FOR THE TOP SURFACE PAD, ISFSI PAD, AND SUPPORT FOUNDATION PAD PER ACI-318 (2005) ................. 3-12 Table 3-4 Values of Principal Design Parameters for the Underground ISFSI ...... 3-13 Certificate of Compliance No. 1040 Amendment No. 0 Appendix B ii

Definitions 1.0 1.0 Definitions Refer to Appendix A for Definitions.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 1-1

Approved Contents 2.0 2.0 APPROVED CONTENTS 2.1 Fuel Specifications and Loading Conditions 2.1.1 Fuel to Be Stored in the HI-STORM UMAX Canister Storage System

a. UNDAMAGED FUEL ASSEMBLIES, DAMAGED FUEL ASSEMBLIES, FUEL DEBRIS, and NON-FUEL HARDWARE meeting the limits specified in Table 2.1-1 and other referenced tables may be stored in the HI-STORM UMAX Canister Storage System.

b. All BWR fuel assemblies may be stored with or without ZR channels.

2.1.2 Fuel Loading Figures 2.3-1 through 2.3-7 and 2.3-12 define the unique cell numbers for the MPC-37 and MPC-89 models, respectively, and the maximum allowable heat load per fuel assembly for each cell under multiple loading conditions. Fuel assembly decay heat limits are specified in Section 2.3.1.

Fuel assemblies shall meet all other applicable limits specified in Tables 2.1-1 through 2.1-3.

2.2 Violations If any Fuel Specifications or Loading Conditions of 2.1 are violated, the following actions shall be completed:

2.2.1 The affected fuel assemblies shall be placed in a safe condition.

2.2.2 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify the NRC Operations Center.

2.2.3 Within 30 days, submit a special report which describes the cause of the violation, and actions taken to restore compliance and prevent recurrence.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-1

Approved Contents 2.0 Table 2.1-1 (page 1 of 4)

Fuel Assembly Limits I. MPC MODEL: MPC-37 A. Allowable Contents

1. Uranium oxide PWR UNDAMAGED FUEL ASSEMBLIES, DAMAGED FUEL ASSEMBLIES, and/or FUEL DEBRIS meeting the criteria in Table 2.1-2, with or without NON-FUEL HARDWARE and meeting the following specifications (Note 1):

a. Cladding Type: ZR

b. Maximum Initial Enrichment: 5.0 wt. % U-235 with soluble boron credit per LCO 3.3.1

c. Post-irradiation Cooling Time Cooling Time 3 years and Average Burnup Per Assembly Average Burnup 68.2 GWD/MTU Assembly:

d. Decay Heat Per Fuel Storage As specified in Section 2.3 Location:

e. Fuel Assembly Length: 199.2 inches (nominal design including NON-FUEL HARDWARE and DFC)

f. Fuel Assembly Width: 8.54 inches (nominal design)

g. Fuel Assembly Weight: 2050 lbs (including NON-FUEL HARDWARE and DFC)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-2

Approved Contents 2.0 Table 2.1-1 (page 2 of 4)

Fuel Assembly Limits I. MPC MODEL: MPC-37 (continued)

B. Quantity per MPC: 37 FUEL ASSEMBLIES with up to twelve (12) DAMAGED FUEL ASSEMBLIES or FUEL DEBRIS in DAMAGED FUEL CONTAINERS (DFCs). DFCs may be stored in fuel storage locations 1, 3, 4, 8, 9, 15, 23, 29, 30, 34, 35, and 37 (see Figures 2.3-1 through 2.3-7). The remaining fuel storage locations may be filled with PWR UNDAMAGED FUEL ASSEMBLIES meeting the applicable specifications.

C. One (1) Neutron Source Assembly (NSA) is authorized for loading in the MPC-37.

D. Up to thirty (30) BRPAs are authorized for loading in the MPC-37.

Note 1: Fuel assemblies containing BPRAs, TPDs, WABAs, water displacement guide tube plugs, orifice rod assemblies, or vibration suppressor inserts, with or without ITTRs, may be stored in any fuel storage location. Fuel assemblies containing APSRs, RCCAs, CEAs, CRAs, or NSAs may only be loaded in fuel storage locations 5 through 7, 10 through 14, 17 through 21, 24 through 28, and 31 through 33 (see Figures 2.3-1 through 2.3-7).

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-3

Approved Contents 2.0 Table 2.1-1 (page 3 of 4)

Fuel Assembly Limits II. MPC MODEL: MPC-89 A. Allowable Contents

1. Uranium oxide BWR UNDAMAGED FUEL ASSEMBLIES, DAMAGED FUEL ASSEMBLIES, and/or FUEL DEBRIS meeting the criteria in Table 2.1-3, with or without channels and meeting the following specifications:

a. Cladding Type: ZR

b. Maximum PLANAR-AVERAGE As specified in Table 2.1-3 for the INITIAL ENRICHMENT(Note 1): applicable fuel assembly array/class.

c. Initial Maximum Rod Enrichment 5.0 wt. % U-235

d. Post-irradiation Cooling Time and Average Burnup Per Assembly

i. Array/Class 8x8F Cooling time 10 years and an assembly average burnup 27.5 GWD/MTU.

ii. All Other Array Classes Cooling Time 3 years and an assembly average burnup 65 GWD/MTU

e. Decay Heat Per Assembly

i. Array/Class 8x8F 183.5 Watts ii. All Other Array Classes As specified in Section 2.3

f. Fuel Assembly Length 176.5 inches (nominal design)

g. Fuel Assembly Width 5.95 inches (nominal design)

h. Fuel Assembly Weight 850 lbs, including a DFC as well as a channel Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-4

Approved Contents 2.0 Table 2.1-1 (page 4 of 4)

Fuel Assembly Limits II. MPC MODEL: MPC-89 (continued)

B. Quantity per MPC: 89 FUEL ASSEMBLIES with up to sixteen (16) DAMAGED FUEL ASSEMBLIES or FUEL DEBRIS in DAMAGED FUEL CONTAINERS (DFCs). DFCs may be stored in fuel storage locations 1, 3, 4, 10, 11, 19, 29, 39, 51, 61, 71, 79, 80, 86, 87, and 89 (see Figure 2.3-12). The remaining fuel storage locations may be filled with BWR UNDAMAGED FUEL ASSEMBLIES meeting the applicable specifications.

Note 1: The lowest maximum allowable enrichment of any fuel assembly loaded in an MPC-89, based on fuel array class and fuel classification, is the maximum allowable enrichment for the remainder of the assemblies loaded in that MPC.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-5

Approved Contents 2.0 Table 2.1-2 (page 1 of 3)

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly 14x14 A 14x14 B 14x14 C 15x15 B 15x15 C Array/ Class No. of Fuel Rod 179 179 176 204 204 Locations Fuel Clad O.D. (in.) 0.400 0.417 0.440 0.420 0.417 Fuel Clad I.D. (in.) 0.3514 0.3734 0.3880 0.3736 0.3640 Fuel Pellet Dia. (in.)

0.3444 0.3659 0.3805 0.3671 0.3570 (Note 3)

Fuel Rod Pitch (in.) 0.556 0.556 0.580 0.563 0.563 Active Fuel Length 150 150 150 150 150 (in.)

No. of Guide and/or 5 17 17 21 21 Instrument Tubes (Note 2)

Guide/Instrument 0.017 0.017 0.038 0.015 0.0165 Tube Thickness (in.)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-6

Approved Contents 2.0 Table 2.1-2 (page 2 of 3)

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly 15x15 D 15x15 E 15x15 F 15x15 H 15x15 I Array/Class No. of Fuel Rod 208 208 208 208 216 Locations Fuel Clad O.D. (in.) 0.430 0.428 0.428 0.414 0.413 Fuel Clad I.D. (in.) 0.3800 0.3790 0.3820 0.3700 0.3670 Fuel Pellet Dia. (in.)

0.3735 0.3707 0.3742 0.3622 0.3600 (Note 3)

Fuel Rod Pitch (in.) 0.568 0.568 0.568 0.568 0.550 Active Fuel Length 150 150 150 150 150 (in.)

No. of Guide and/or 17 17 17 17 9 (Note 4)

Instrument Tubes Guide/Instrument 0.0150 0.0140 0.0140 0.0140 0.0140 Tube Thickness (in.)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-7

Approved Contents 2.0 Table 2.1-2 (page 3 of 3)

PWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly 16x16 A 17x17A 17x17 B 17x17 C 17x17 D 17x17 E Array and Class No. of Fuel Rod 236 264 264 264 264 265 Locations Fuel Clad O.D. (in.) 0.382 0.360 0.372 0.377 0.372 0.372 Fuel Clad I.D. (in.) 0.3350 0.3150 0.3310 0.3330 0.3310 0.3310 Fuel Pellet Dia. (in.)

0.3255 0.3088 0.3232 0.3252 0.3232 0.3232 (Note 3)

Fuel Rod Pitch (in.) 0.506 0.496 0.496 0.502 0.496 0.496 Active Fuel length 150 150 150 150 170 170 (in.)

No. of Guide and/or 5 (Note 2) 25 25 25 25 24 Instrument Tubes Guide/Instrument 0.0350 0.016 0.014 0.020 0.014 0.014 Tube Thickness (in.)

Notes:

1. All dimensions are design nominal values. Maximum and minimum dimensions are specified to bound variations in design nominal values among fuel assemblies within a given array/class.

2. Each guide tube replaces four fuel rods.

3. Annular fuel pellets are allowed in the top and bottom 12 of the active fuel length.

4. One Instrument Tube and eight Guide Bars (Solid ZR)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-8

Approved Contents 2.0 Table 2.1-3 (page 1 of 4)

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly Array 7x7 B 8x8 B 8x8 C 8x8 D 8x8 E and Class Maximum Planar-Average Initial

< 4.8 < 4.8 < 4.8 < 4.8 < 4.8 Enrichment (wt.%

235 U) (Note 14)

No. of Fuel Rod Locations (Full Length 49 63 or 64 62 60 or 61 59 or Total/Full Length)

Fuel Clad O.D. (in.) > 0.5630 > 0.4840 > 0.4830 > 0.4830 > 0.4930 Fuel Clad I.D. (in.) < 0.4990 < 0.4295 < 0.4250 < 0.4230 < 0.4250 Fuel Pellet Dia. (in.) < 0.4910 < 0.4195 < 0.4160 < 0.4140 < 0.4160 Fuel Rod Pitch (in.) < 0.738 < 0.642 < 0.641 < 0.640 < 0.640 Design Active Fuel

< 150 < 150 < 150 < 150 < 150 Length (in.)

No. of Water Rods 1-4 0 1 or 0 2 5 (Note 10) (Note 6)

Water Rod Thickness N/A > 0.034 > 0.00 > 0.00 > 0.034 (in.)

Channel Thickness

< 0.120 < 0.120 < 0.120 < 0.120 < 0.100 (in.)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-9

Approved Contents 2.0 Table 2.1-3 (2 of 4)

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly 8x8F 9x9 A 9x9 B 9x9 C 9x9 D Array and Class Maximum Planar-Average Initial < 4.5

< 4.8 < 4.8 < 4.8 < 4.8 Enrichment (wt.% (Note 12) 235 U) (Note 14)

No. of Fuel Rod 74/66 64 72 80 79 Locations (Note 4)

Fuel Clad O.D. (in.) > 0.4576 > 0.4400 > 0.4330 > 0.4230 > 0.4240 Fuel Clad I.D. (in.) < 0.3996 < 0.3840 < 0.3810 < 0.3640 < 0.3640 Fuel Pellet Dia. (in.) < 0.3913 < 0.3760 < 0.3740 < 0.3565 < 0.3565 Fuel Rod Pitch (in.) < 0.609 < 0.566 < 0.572 < 0.572 < 0.572 Design Active Fuel

< 150 < 150 < 150 < 150 < 150 Length (in.)

No. of Water Rods N/A 1 2 1 2 (Note 10) (Note 2) (Note 5)

Water Rod

> 0.0315 > 0.00 > 0.00 > 0.020 > 0.0300 Thickness (in.)

Channel Thickness

< 0.055 < 0.120 < 0.120 < 0.100 < 0.100 (in.)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-10

Approved Contents 2.0 Table 2.1-3 (page 3 of 4)

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly 9x9 E 9x9 F 9x9 G 10x10 A 10x10 B Array and Class (Note 2) (Note 2)

Maximum Planar-

< 4.5 < 4.5 Average Initial (Note (Note < 4.8 < 4.8 < 4.8 Enrichment (wt.%

235 12) 12)

U) (Note 14)

No. of Fuel Rod 92/78 91/83 76 76 72 Locations (Note 7) (Note 8)

Fuel Clad O.D. (in.) >0.4170 >0.4430 >0.4240 >0.4040 >0.3957 Fuel Clad I.D. (in.) <0.3640 <0.3860 <0.3640 < 0.3520 < 0.3480 Fuel Pellet Dia. (in.) <0.3530 <0.3745 <0.3565 < 0.3455 < 0.3420 Fuel Rod Pitch (in.) < 0.572 < 0.572 < 0.572 < 0.510 < 0.510 Design Active Fuel

< 150 < 150 < 150 < 150 < 150 Length (in.)

No. of Water Rods 1 1 5 5 2 (Note 10) (Note 5) (Note 5)

Water Rod Thickness

>0.0120 >0.0120 >0.0320 >0.0300 > 0.00 (in.)

Channel Thickness

< 0.120 < 0.120 < 0.120 < 0.120 < 0.120 (in.)

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-11

Approved Contents 2.0 Table 2.1-3 (page 4 of 4)

BWR FUEL ASSEMBLY CHARACTERISTICS (Note 1)

Fuel Assembly Array and 10x10 C 10x10 F 10x10 G Class Maximum Planar-Average

< 4.7 < 4.6 Initial Enrichment (wt.% 235U) < 4.8 (Note 13) (Note 12)

(Note 14)

No. of Fuel Rod Locations 92/78 96 96/84 (Note 7)

Fuel Clad O.D. (in.) > 0.3780 > 0.4035 > 0.387 Fuel Clad I.D. (in.) < 0.3294 < 0.3570 < 0.340 Fuel Pellet Dia. (in.) < 0.3224 < 0.3500 < 0.334 Fuel Rod Pitch (in.) < 0.488 < 0.510 < 0.512 Design Active Fuel Length (in.) < 150 < 150 < 150 No. of Water Rods (Note 10) 5 5 2

(Note 9) (Note 9)

Water Rod Thickness (in.) > 0.031 > 0.030 > 0.031 Channel Thickness (in.) < 0.055 < 0.120 < 0.060 Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-12

Approved Contents

2.0 NOTES

1. All dimensions are design nominal values. Maximum and minimum dimensions are specified to bound variations in design nominal values among fuel assemblies within a given array/class.

2. This assembly is known as QUAD+. It has four rectangular water cross segments dividing the assembly into four quadrants.

3. For the SPC 9x9-5 fuel assembly, each fuel rod must meet either the 9x9E or the 9x9F set of limits or clad O.D., clad I.D., and pellet diameter.

4. This assembly class contains 74 total rods; 66 full length rods and 8 partial length rods.

5. Square, replacing nine fuel rods.

6. Variable.

7. This assembly contains 92 total fuel rods; 78 full length rods and 14 partial length rods.

8. This assembly class contains 91 total fuel rods; 83 full length rods and 8 partial length rods.

9. One diamond-shaped water rod replacing the four center fuel rods and four rectangular water rods dividing the assembly into four quadrants.

10. These rods may also be sealed at both ends and contain ZR material in lieu of water.

11. Not used.

12. When loading fuel assemblies classified as DAMAGED FUEL, all assemblies in the MPC are limited to 4.0 wt.% U-235.

13. When loading fuel assemblies classified as DAMAGED FUEL, all assemblies in the MPC are limited to 4.6 wt.% U-235.

14. In accordance with the definition of UNDAMAGED FUEL, certain assemblies may be limited to 3.3 wt.% U-235. When loading these fuel assemblies, all assemblies in the MPC are limited to 3.3 wt.% U-235.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-13

Approved Contents 2.0 Table 2.1-4 CLASSIFICATION OF FUEL ASSEMBLY FOR MPC-37 IN THE HI-STORM UMAX ISFSI MPC Type Classification Nominal Active Fuel Length Short Fuel 128 inches < L < 144 inches MPC-37 Standard Fuel 144 inches < L < 168 inches Long Fuel L > 168 inches Note 1: L means "nominal active fuel length".

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-14

Approved Contents 2.0 2.3 Decay Heat Limits This section provides the limits on fuel assembly decay heat for storage in the HI-STORM UMAX Canister Storage System. The method to verify compliance, including examples, is provided in Chapter 13 of the HI-STORM UMAX FSAR.

2.3.1 Fuel Loading Decay Heat Limits Table 2.3-1 provides the maximum permissible decay heat under long-term storage for MPC-37 and MPC-89. Table 2.3-1 also lists the applicable figures providing the permissible decay heat per fuel storage location, including MPCs using the optional helium backfill pressure ranges permitted in Table 3-2 of Appendix A.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-15

Approved Contents 2.0 TABLE 2.3-1 PERMISSIBLE HEAT LOAD FOR LONG-TERM STORAGE Permissible Helium Backfill Permissible Heat Load Aggregate Heat MPC Type Pressure Option Heat Load Per Chart Load, kW (Notes 1,2) Storage Cell (Note 4) 1 1 Figure 2.3-1 33.88 Short Fuel 2 2 Figure 2.3-2 33.70 (Note 3) 3 1 Figure 2.3-3 33.53 1 1 Figure 2.3-1 33.88 Standard Fuel 2 2 Figure 2.3-2 33.70 (Note 3) 3 1 Figure 2.3-4 35.30 1 1 Figure 2.3-5 35.76 Long Fuel MPC-37 2 2 Figure 2.3-6 35.57 (Note 3) 3 1 Figure 2.3-7 37.06 Short Fuel 3 Figure 2.3-8 34.28 (Note 3) 3 Figure 2.3-12 33.46 Standard Fuel 3 Figure 2.3-8 34.28 (Note 3) 3 Figure 2.3-12 33.46 Long Fuel 3 Figure 2.3-9 36.19 (Note 3) 3 Figure 2.3-12 33.46 1 Figure 2.3-10 36.32 MPC-89 2 Figure 2.3-11 36.72 2 Figure 2.3-13 34.75 Notes:

1. For helium backfill pressure option pressure ranges see Appendix A, Table 3-2

2. For the details on the use of VDS to dry High Burnup Fuel see Appendix A, Table 3-1

3. See Table 2.1-4 for fuel length data

4. Aggregate heat load is defined as the sum of heat loads of all stored fuel Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-16

Approved Contents 2.0 assemblies. The permissible aggregate heat load is set to 80% of the design basis heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-17

Approved Contents 2.0 2.3.2 When complying with the maximum fuel storage location decay heat limits, users must account for the decay heat from both the fuel assembly and any NON-FUEL HARDWARE, as applicable for the particular fuel storage location, to ensure the decay heat emitted by all contents in a storage location does not exceed the limit.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-18

Approved Contents 2.0 1 2 3 0.873 0.873 0.873 4 5 6 7 8 0.873 1.602 1.602 1.602 0.873 9 10 11 12 13 14 15 0.873 1.602 1.017 1.017 1.017 1.602 0.873 16 17 18 19 20 21 22 0.873 1.602 1.017 1.017 1.017 1.602 0.873 23 24 25 26 27 28 29 0.873 1.602 1.017 1.017 1.017 1.602 0.873 30 31 32 33 34 0.873 1.602 1.602 1.602 0.873 35 36 37 0.873 0.873 0.873 Legend Cell ID Heat Load, kW Figure 2.3-1 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 1 for Long-term Storage for Short and Standard Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-19

Approved Contents 2.0 1 2 3 1.215 1.215 1.215 4 5 6 7 8 1.215 1.080 1.080 1.080 1.215 9 10 11 12 13 14 15 1.215 1.080 1.080 1.080 1.080 1.080 1.215 16 17 18 19 20 21 22 1.215 1.080 1.080 1.080 1.080 1.080 1.215 23 24 25 26 27 28 29 1.215 1.080 1.080 1.080 1.080 1.080 1.215 30 31 32 33 34 1.215 1.080 1.080 1.080 1.215 35 36 37 1.215 1.215 1.215 Legend Cell ID Heat Load, kW Figure 2.3-2 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 2 for Long-term Storage for Short and Standard Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-20

Approved Contents 2.0 1 2 3 0.922 0.922 0.922 4 5 6 7 8 0.922 1.520 1.520 1.520 0.922 9 10 11 12 13 14 15 0.922 1.710 0.950 0.950 0.950 1.710 0.922 16 17 18 19 20 21 22 0.922 1.520 0.950 0.570 0.950 1.520 0.922 23 24 25 26 27 28 29 0.922 1.710 0.950 0.950 0.950 1.710 0.922 30 31 32 33 34 0.922 1.520 1.520 1.520 0.922 35 36 37 0.922 0.922 0.922 Legend Cell ID Heat Load, kW Figure 2.3-3 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 3 for Long-term Storage for Short Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-21

Approved Contents 2.0 1 2 3 0.970 0.970 0.970 4 5 6 7 8 0.970 1.600 1.600 1.600 0.970 9 10 11 12 13 14 15 0.970 1.800 1.000 1.000 1.000 1.800 0.970 16 17 18 19 20 21 22 0.970 1.600 1.000 0.600 1.000 1.600 0.970 23 24 25 26 27 28 29 0.970 1.800 1.000 1.000 1.000 1.800 0.970 30 31 32 33 34 0.970 1.600 1.600 1.600 0.970 35 36 37 0.970 0.970 0.970 Legend Cell ID Heat Load, kW Figure 2.3-4 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 3 for Long-term Storage for Standard Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-22

Approved Contents 2.0 1 2 3 0.922 0.922 0.922 4 5 6 7 8 0.922 1.691 1.691 1.691 0.922 9 10 11 12 13 14 15 0.922 1.691 1.074 1.074 1.074 1.691 0.922 16 17 18 19 20 21 22 0.922 1.691 1.074 1.074 1.074 1.691 0.922 23 24 25 26 27 28 29 0.922 1.691 1.074 1.074 1.074 1.691 0.922 30 31 32 33 34 0.922 1.691 1.691 1.691 0.922 35 36 37 0.922 0.922 0.922 Legend Cell ID Heat Load, kW Figure 2.3-5 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 1 for Long-term Storage for Long Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-23

Approved Contents 2.0 1 2 3 1.283 1.283 1.283 4 5 6 7 8 1.283 1.140 1.140 1.140 1.283 9 10 11 12 13 14 15 1.283 1.140 1.140 1.140 1.140 1.140 1.283 16 17 18 19 20 21 22 1.283 1.140 1.140 1.140 1.140 1.140 1.283 23 24 25 26 27 28 29 1.283 1.140 1.140 1.140 1.140 1.140 1.283 30 31 32 33 34 1.283 1.140 1.140 1.140 1.283 35 36 37 1.283 1.283 1.283 Legend Cell ID Heat Load, kW Figure 2.3-6 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 2 for Long-term Storage for Long Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-24

Approved Contents 2.0 1 2 3 1.019 1.019 1.019 4 5 6 7 8 1.019 1.680 1.680 1.680 1.019 9 10 11 12 13 14 15 1.019 1.890 1.050 1.050 1.050 1.890 1.019 16 17 18 19 20 21 22 1.019 1.680 1.050 0.630 1.050 1.680 1.019 23 24 25 26 27 28 29 1.019 1.890 1.050 1.050 1.050 1.890 1.019 30 31 32 33 34 1.019 1.680 1.680 1.680 1.019 35 36 37 1.019 1.019 1.019 Legend Cell ID Heat Load, kW Figure 2.3-7 HI-STORM UMAX MPC-37 Permissible Heat Load Chart 3 for Long-term Storage for Long Fuel Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-25

Approved Contents 2.0 1 2 3 0.785 0.785 0.785 4 5 6 7 8 0.785 1.441 1.441 1.441 0.785 9 10 11 12 13 14 15 0.785 1.441 0.915 0.915 0.915 1.441 0.785 16 17 18 19 20 21 22 0.785 1.441 0.915 0.915 0.915 1.441 0.785 23 24 25 26 27 28 29 0.785 1.441 0.915 0.915 0.915 1.441 0.785 30 31 32 33 34 0.785 1.441 1.441 1.441 0.785 35 36 37 0.785 0.785 0.785 Legend Cell ID Heat Load, kW Figure 2.3-8 HI-STORM UMAX MPC-37 Permissible Heat Load for Short and Standard Fuel for Helium Backfill Option 3 in Table 3-2 of Appendix A Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-26

Approved Contents 2.0 1 2 3 0.829 0.829 0.829 4 5 6 7 8 0.829 1.521 1.521 1.521 0.829 9 10 11 12 13 14 15 0.829 1.521 0.966 0.966 0.966 1.521 0.829 16 17 18 19 20 21 22 0.829 1.521 0.966 0.966 0.966 1.521 0.829 23 24 25 26 27 28 29 0.829 1.521 0.966 0.966 0.966 1.521 0.829 30 31 32 33 34 0.829 1.521 1.521 1.521 0.829 35 36 37 0.829 0.829 0.829 Legend Cell ID Heat Load, kW Figure 2.3-9 HI-STORM UMAX MPC-37 Permissible Heat Load for Long Fuel for Helium Backfill Option 3 in Table 3-2 of Appendix A Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-27

Approved Contents 2.0 1 2 3 0.431 0.431 0.431 4 5 6 7 8 9 10 0.431 0.431 0.431 0.607 0.431 0.431 0.431 11 12 13 14 15 16 17 18 19 0.431 0.431 0.607 0.607 0.607 0.607 0.607 0.431 0.431 20 21 22 23 24 25 26 27 28 0.431 0.607 0.607 0.607 0.607 0.607 0.607 0.607 0.431 29 30 31 32 33 34 35 36 37 38 39 0.431 0.431 0.607 0.607 0.431 0.431 0.431 0.607 0.607 0.431 0.431 40 41 42 43 44 45 46 47 48 49 50 0.431 0.607 0.607 0.607 0.431 0.431 0.431 0.607 0.607 0.607 0.431 51 52 53 54 55 56 57 58 59 60 61 0.431 0.431 0.607 0.607 0.431 0.431 0.431 0.607 0.607 0.431 0.431 62 63 64 65 66 67 68 69 70 0.431 0.607 0.607 0.607 0.607 0.607 0.607 0.607 0.431 71 72 73 74 75 76 77 78 79 0.431 0.431 0.607 0.607 0.607 0.607 0.607 0.431 0.431 80 81 82 83 84 85 86 0.431 0.431 0.431 0.607 0.431 0.431 0.431 87 88 89 0.431 0.431 0.431 Legend Figure 2.3-10 Cell ID HI-STORM UMAX MPC-89 Permissible Heat Load for Long-Term Storage Heat Load, kW Note that this figure shows the per cell heat load limit for storage.

The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-28

Approved Contents 2.0 1 2 3 0.387 0.387 0.387 4 5 6 7 8 9 10 0.387 0.387 0.387 0.546 0.387 0.387 0.387 11 12 13 14 15 16 17 18 19 0.387 0.387 0.546 0.546 0.546 0.546 0.546 0.387 0.387 20 21 22 23 24 25 26 27 28 0.387 0.546 0.546 0.546 0.546 0.546 0.546 0.546 0.387 29 30 31 32 33 34 35 36 37 38 39 0.387 0.387 0.546 0.546 0.387 0.387 0.387 0.546 0.546 0.387 0.387 40 41 42 43 44 45 46 47 48 49 50 0.387 0.546 0.546 0.546 0.387 0.387 0.387 0.546 0.546 0.546 0.387 51 52 53 54 55 56 57 58 59 60 61 0.387 0.387 0.546 0.546 0.387 0.387 0.387 0.546 0.546 0.387 0.387 62 63 64 65 66 67 68 69 70 0.387 0.546 0.546 0.546 0.546 0.546 0.546 0.546 0.387 71 72 73 74 75 76 77 78 79 0.387 0.387 0.546 0.546 0.546 0.546 0.546 0.387 0.387 80 81 82 83 84 85 86 0.387 0.387 0.387 0.546 0.387 0.387 0.387 87 88 89 0.387 0.387 0.387 Legend Figure 2.3-11 HI-STORM UMAX MPC-89 Permissible Heat Load for Helium Backfill Cell ID Option 2 in Table 3-2 of Appendix A Heat Load, kW Note that this figure shows the per cell heat load limit for storage.

The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-29

Approved Contents 2.0 1 2 3 0.97 0.97 0.97 4 5 6 7 8 0.97 0.97 0.97 0.97 0.97 9 10 11 12 13 14 15 0.97 0.97 0.7 0.7 0.7 0.97 0.97 16 17 18 19 20 21 22 0.97 0.97 0.7 0.7 0.7 0.97 0.97 23 24 25 26 27 28 29 0.97 0.97 0.7 0.7 0.7 0.97 0.97 30 31 32 33 34 0.97 0.97 0.97 0.97 0.97 35 36 37 0.97 0.97 0.97 Legend Cell ID Heat Load, kW Figure 2.3-12 HI-STORM UMAX MPC-37 Permissible Threshold Heat Load for VDS High Burnup Fuel in Table 3-1 of Appendix A and Helium Backfill Option 3 in Table 3-2 of Appendix A Note that this figure shows the per cell heat load limit for storage. The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-30

Approved Contents 2.0 1 2 3 0.44 0.44 0.44 4 5 6 7 8 9 10 0.44 0.44 0.44 0.35 0.44 0.44 0.44 11 12 13 14 15 16 17 18 19 0.44 0.44 0.35 0.35 0.35 0.35 0.35 0.44 0.44 20 21 22 23 24 25 26 27 28 0.44 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.44 29 30 31 32 33 34 35 36 37 38 39 0.44 0.44 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.44 0.44 40 41 42 43 44 45 46 47 48 49 50 0.44 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.44 51 52 53 54 55 56 57 58 59 60 61 0.44 0.44 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.44 0.44 62 63 64 65 66 67 68 69 70 0.44 0.35 0.35 0.35 0.35 0.35 0.35 0.35 0.44 71 72 73 74 75 76 77 78 79 0.44 0.44 0.35 0.35 0.35 0.35 0.35 0.44 0.44 80 81 82 83 84 85 86 0.44 0.44 0.44 0.35 0.44 0.44 0.44 87 88 89 0.44 0.44 0.44 Legend Figure 2.3-13 Cell ID HI-STORM UMAX MPC-89 Permissible Threshold Heat Load for VDS High Burnup Fuel in Table 3-1 of Appendix A and Helium Backfill Option Heat Load, kW 2 in Table 3-2 of Appendix A Note that this figure shows the per cell heat load limit for storage.

The total permissible aggregate heat load may be less than the sum of each individual cell heat load. See Table 2.3-1 for corresponding permissible aggregate heat load.

Certificate of Compliance No.1040 Amendment No. 0 Appendix B 2-31

Design Features 3.0 3.0 DESIGN FEATURES 3.1 Site 3.1.1 Site Location The HI-STORM UMAX Canister Storage System is authorized for general use by 10 CFR Part 50 license holders at various site locations under the provisions of 10 CFR 72, Subpart K.

3.2 Design Features Important for Criticality Control 3.2.1 MPC-37

1. Basket cell ID: 8.92 in. (min. nominal)

2. Basket cell wall thickness: 0.57 in. (min.nominal )

3. B4C in the Metamic-HT: 10.0 wt % (min. nominal) 3.2.2 MPC-89

1. Basket cell ID: 5.99 in. (min.nominal)

2. Basket cell wall thickness: 0.38 in. (min.nominal)

3. B4C in the Metamic-HT: 10.0 wt % (min. nominal) 3.2.3 Metamic-HT Test Requirements

1. The weight percentage of the boron carbide must be confirmed to be greater than or equal to 10% in each lot of Al/ B4C powder.

2. The areal density of the B-10 isotope corresponding to the 10%

min. weight density in the manufactured Metamic HT panels shall be independently confirmed by the neutron attenuation test method by testing at least one coupon from a randomly selected panel in each lot.

3. If the B- 10 areal density criterion in the tested panel fails to meet the specified minimum, then the manufacturer has the option to reject the entire lot or to test a statistically significant number of panels and perform statistical analysis to show that the minimum areal density in the panels (that comprise the lot) is satisfied with 95% confidence.

4. All test procedures used in demonstrating compliance with the above requirements shall conform to the cask designer's QA program which has been approved by the USNRC under docket number 71-0784.

3.3 Codes and Standards The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), 2007, is the governing Code for the HI-STORM UMAX system MPC as Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-1

Design Features 3.0 clarified in Specification 3.3.1 below, except for Code Sections V and IX. However, the HI-STORM UMAX VVM is structurally qualified per the newer 2010 ASME code. The ASME Code paragraphs applicable to the manufacturing of HI-STORM UMAX VVM and transfer cask are listed in Table 3-2. The latest effective editions of ASME Code Sections V and IX, including addenda, may be used for activities governed by those sections, provided a written reconciliation of the later edition against the applicable edition (including addenda) specified above, is performed by the certificate holder.

American Concrete Institute ACI-318 (2005) is the governing Code for both plain concrete and reinforced concrete as clarified in Chapter 3 of the Final Safety Analysis Report for the HI-STORM 100 UMAX System.

3.3.1 Alternatives to Codes, Standards, and Criteria Table 3-1 lists approved alternatives to the ASME Code for the design of the MPCs of the HI-STORM UMAX Canister Storage System.

3.3.2 Construction/Fabrication Alternatives to Codes, Standards, and Criteria Proposed alternatives to the ASME Code,Section III, 2007 Edition, including modifications to the alternatives allowed by Specification 3.3.1 may be used on a case-specific basis when authorized by the Director of the Office of Nuclear Material Safety and Safeguards or designee. The request for such alternative should demonstrate that:

1. The proposed alternatives would provide an acceptable level of quality and safety, or

2. Compliance with the specified requirements of the ASME Code,Section III, 2007 Edition, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Requests for alternatives shall be submitted in accordance with 10 CFR 72.4.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-2

Design Features 3.0 3.0 DESIGN FEATURES (continued)

TABLE 3-1 List of ASME Code Alternatives for Multi-Purpose Canisters (MPCs)

MPC Subsection General Requirements. Because the MPC is not an ASME Enclosure NCA Requires preparation of a Code stamped vessel, none of the Vessel Design Specification, specifications, reports, certificates, or Design Report, other general requirements specified by Overpressure Protection NCA are required. In lieu of a Design Report, Certification of Specification and Design Report, the Construction Report, Data HI-STORM FSAR includes the design Report, and other criteria, service conditions, and load administrative controls for combinations for the design and an ASME Code stamped operation of the MPCs as well as the vessel. results of the stress analyses to demonstrate that applicable Code stress limits are met. Additionally, the fabricator is not required to have an ASME-certified QA program. All important-to-safety activities are governed by the NRC-approved Holtec QA program.

Because the cask components are not certified to the Code, the terms Certificate Holder and Inspector are not germane to the manufacturing of NRC-certified cask components. To eliminate ambiguity, the responsibilities assigned to the Certificate Holder in the Code, as applicable, shall be interpreted to apply to the NRC Certificate of Compliance (CoC) holder (and by extension, to the component fabricator) if the requirement must be fulfilled. The Code term Inspector means the QA/QC personnel of the CoC holder and its vendors assigned to oversee and inspect the manufacturing process.

MPC NB-1100 Statement of requirements MPC Enclosure Vessel is designed and Enclosure for Code stamping of will be fabricated in accordance with Vessel components. ASME Code,Section III, Subsection NB to the maximum practical extent, but Code stamping is not required.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-3

Design Features 3.0 TABLE 3-1 List of ASME Code Alternatives for Multi-Purpose Canisters (MPCs)

MPC basket NB-1130 NB-1132.2(d) requires that The lugs that are used exclusively for supports the first connecting weld of lifting an empty MPC are welded to the and lift lugs a non-pressure retaining inside of the pressure-retaining MPC structural attachment to a shell, but are not designed in component shall be accordance with Subsection NB. The considered part of the lug-to-Enclosure Vessel Weld is component unless the weld required to meet the stress limits of is more than 2t from the Reg. Guide 3.61 in lieu of Subsection pressure retaining portion NB of the Code.

of the component, where t is the nominal thickness of the pressure retaining material.

NB-1132.2(e) requires that the first connecting weld of a welded nonstructural attachment to a component shall conform to NB-4430 if the connecting weld is within 2t from the pressure retaining portion of the component.

MPC NB-2000 Requires materials to be Materials will be supplied by Holtec Enclosure supplied by ASME- approved suppliers with Certified Vessel approved material supplier. Material Test Reports (CMTRs) in accordance with NB-2000 requirements.

MPC NB-3100 Provides requirements for These requirements are subsumed by Enclosure NF-3100 determining design loading the HI-STORM FW FSAR, serving as Vessel conditions, such as the Design Specification, which pressure, temperature, and establishes the service conditions and mechanical loads. load combinations for the storage system.

MPC NB-4120 NB-4121.2 and NF-4121.2 In-shop operations of short duration that Enclosure provide requirements for apply heat to a component, such as Vessel repetition of tensile or plasma cutting of plate stock, welding, impact tests for material machining, and coating are not, unless subjected to heat treatment explicitly stated by the Code, defined as during fabrication or heat treatment operations.

installation.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-4

Design Features 3.0 TABLE 3-1 List of ASME Code Alternatives for Multi-Purpose Canisters (MPCs)

MPC NB-4220 Requires certain forming The cylindricity measurements on the Enclosure tolerances to be met for rolled shells are not specifically Vessel cylindrical, conical, or recorded in the shop travelers, as would spherical shells of a vessel. be the case for a Code-stamped pressure vessel. Rather, the requirements on inter-component clearances (such as the MPC-to-transfer cask) are guaranteed through fixture-controlled manufacturing. The fabrication specification and shop procedures ensure that all dimensional design objectives, including inter-component annular clearances are satisfied. The dimensions required to be met in fabrication are chosen to meet the functional requirements of the dry storage components. Thus, although the post-forming Code cylindricity requirements are not evaluated for compliance directly, they are indirectly satisfied (actually exceeded) in the final manufactured components.

MPC NB-4122 Implies that with the MPCs are built in lots. Material Enclosure exception of studs, bolts, traceability on raw materials to a heat Vessel nuts and heat exchanger number and corresponding CMTR is tubes, CMTRs must be maintained by Holtec through markings traceable to a specific on the raw material. Where material is piece of material in a cut or processed, markings are component. transferred accordingly to assure traceability. As materials are assembled into the lot of MPCs being manufactured, documentation is maintained to identify the heat numbers of materials being used for that item in the multiple MPCs being manufactured under that lot. A specific item within a specific MPC will have a number of heat numbers identified as possibly being used for the item in that particular MPC of which one or more of those heat numbers (and corresponding CMTRS) will have actually been used. All of the heat numbers identified will comply with the requirements for the particular item.

MPC Lid and NB-4243 Full penetration welds MPC lid and closure ring are not full Closure Ring required for Category C penetration welds. They are welded Welds Joints (flat head to main independently to provide a redundant shell per NB-3352.3) seal.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-5

Design Features 3.0 TABLE 3-1 List of ASME Code Alternatives for Multi-Purpose Canisters (MPCs)

MPC Closure NB-5230 Radiographic (RT) or Root (if more than one weld pass is Ring, Vent and ultrasonic (UT) required) and final liquid penetrant Drain Cover examination required. examination to be performed in Plate Welds accordance with NB-5245. The closure ring provides independent redundant closure for vent and drain cover plates.

Vent and drain port cover plate welds are helium leakage tested.

MPC Lid to NB-5230 Radiographic (RT) or Only progressive liquid penetrant (PT)

Shell Weld ultrasonic (UT) examination is permitted. PT examination required. examination will include the root and final weld layers and each approx. 3/8" of weld depth.

MPC NB-6111 All completed pressure The MPC vessel is welded in the field Enclosure retaining systems shall be following fuel assembly loading. After Vessel and Lid pressure tested. the lid to shell weld is completed, the MPC shall then be pressure tested as defined in Chapter 10. Accessibility for leakage inspections precludes a Code compliant pressure test. Since the shell welds of the MPC cannot be checked for leakage during this pressure test, the shop leakage test to 10-7 ref cc/sec provides reasonable assurance as to its leak tightness. All MPC enclosure vessel welds (except closure ring and vent/drain cover plate) are inspected by volumetric examination. The MPC lid-to-shell weld shall be verified by progressive PT examination. PT must include the root and final layers and each approximately 3/8 inch of weld depth.

The inspection results, including relevant findings (indications) shall be made a permanent part of the users records by video, photographic, of other means which provide an equivalent record of weld integrity. The video or photographic records should be taken during the final interpretation period described in ASME Section V, Article 6, T-676. The vent/drain cover plate and the closure ring welds are confirmed by liquid penetrant examination. The inspection of the weld must be performed by qualified personnel and shall meet the acceptance requirements of ASME Code Section III, NB-5350.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-6

Design Features 3.0 TABLE 3-1 List of ASME Code Alternatives for Multi-Purpose Canisters (MPCs)

MPC NB-7000 Vessels are required to No overpressure protection is provided.

Enclosure have overpressure Function of MPC enclosure vessel is to Vessel protection. contain radioactive contents under normal, off-normal, and accident conditions of storage. MPC vessel is designed to withstand maximum internal pressure considering 100% fuel rod failure and maximum accident temperatures.

MPC NB-8000 States requirements for The HI-STORM UMAX system is to be Enclosure nameplates, stamping and marked and identified in accordance Vessel reports per NCA-8000. with 10CFR71 and 10CFR72 requirements. Code stamping is not required. QA data package to be in accordance with Holtec approved QA program.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-7

Design Features 3.0 Table 3-2 REFERENCE ASME CODE PARAGRAPHS FOR VVM PRIMARY LOAD BEARING PARTS Item Code Explanation and Applicability Paragraph

[2.6.1]

1. Definition of primary and NF-1215 -

secondary members

2. Jurisdictional boundary NF-1133 The VVMs jurisdictional boundary is defined by the bottom surface of the SFP, the top surface of the ISFSI pad and the SES side surfaces.

3. Certification of NF-2130(b) and Materials shall be certified to material(structural) (c) the applicable Section II of the ASME Code or equivalent ASTM Specification.

4. Heat treatment of material NF-2170 and -

NF-2180

5. Storage of welding material NF-2400 -

6. Welding procedure Section IX -

7. Welding material Section II -

8. Loading conditions NF-3111 -

9. Allowable stress values NF-3112.3 -

10. Rolling and sliding NF-3424 -

supports

11. Differential thermal NF-3127 -

expansion

12. Stress analysis NF-3143 Provisions for stress analysis NF-3380 for Class 3 plate and shell NF-3522 supports and for linear supports NF-3523 are applicable for Closure Lid and Container Shell, respectively.

13. Cutting of plate stock NF-4211 -

NF-4211.1

14. Forming NF-4212 -

15. Forming tolerance NF-4221 Applies to the Container Shell

16. Fitting and Aligning Tack NF-4231 -

Welds NF-4231.1

17. Alignment NF-4232 -

18. Storage of Welding NF-4411 -

Materials

19. Cleanliness of Weld NF-4412 Applies to structural and non-Surfaces structural welds

20. Backing Strips, Peening NF-4421 Applies to structural and non-Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-8

Design Features 3.0 Table 3-2 REFERENCE ASME CODE PARAGRAPHS FOR VVM PRIMARY LOAD BEARING PARTS Item Code Explanation and Applicability Paragraph

[2.6.1]

NF-4422 structural welds

21. Pre-heating and Interpass NF-4611 Applies to structural and non-Temperature NF-4612 structural welds NF-4613

22. Non-Destructive NF-5360 InvokesSection V Examination

23. NDE Personnel NF-5522 -

Certification NF-5523 NF-5530 Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-9

Design Features 3.0 3.0 DESIGN FEATURES (continued) 3.4 Site-Specific Parameters and Analyses Site-specific parameters and analyses that will require verification by the system user are, as a minimum, as follows:

1. The temperature of 80o F is the maximum average yearly temperature.

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

3. The resultant zero period acceleration at the top of the grade and at the elevation of the Support Foundation Pad (SFP) at the host site (computed by the Newmarks rule as the sum of A+0.4*B+0.4*C, where A, B, C denote the free field ZPAs in the three orthogonal directions in decreasing magnitude, i.e., A B C) shall be less than or equal to 1.3 and 1.214, respectively.

4. The analyzed flood condition of 15 fps water velocity and a height of 125 feet of water (full submergence of the loaded cask) are not exceeded.

5. The potential for fire and explosion shall be based on site-specific considerations. The user shall demonstrate that the site-specific potential for fire is bounded by the fire conditions analyzed by the Certificate Holder, or an analysis of the site-specific fire considerations shall be performed.

6. The moment and shear capacities of the ISFSI Structures shall meet the structural requirements under the load combinations in Table 3.4-1.

7. Radiation Protection Space (RPS) as defined in Subsection 5.3.9 of Appendix A, is intended to ensure that the subgrade material in and around the lateral space occupied by the VVMs remains essentially intact under all service conditions including during an excavation activity adjacent to the RPS.

8. The SFP for a VVM array established in any one construction campaign shall be of monolithic construction, to the extent practicable, to maximize the physical stability of the underground installation.

9. Excavation activities contiguous to a loaded UMAX ISFSI on the side facing the excavation can occur down to the depth of the bottom surface of the SFP of the loaded ISFSI (i.e. within the area labeled Space B in Figure 3-1) considering that there may be minor variations in the depth due to normal construction practices. For excavation activities which are contiguous to the loaded ISFSI (within a distance W, see Figure 3-1) and below the depth of the bottom surface of the SFP (i.e. within the area labeled Space D in Figure 3-1), a site-specific seismic analysis will be performed to demonstrate the stability of the RPS boundary and structural integrity of the ISFSI structure. This analysis shall be submitted to Holtec International to be incorporated in an amendment request for NRC review and approval prior to any excavation taking place.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-10

Design Features 3.0

10. In cases where engineered features (i.e., berms and shield walls) are used to ensure that the requirements of 10CFR72.104(a) are met, such features are to be considered important-to-safety and must be evaluated to determine the applicable quality assurance category.

11. LOADING OPERATIONS, TRANSPORT OPERATIONS, and UNLOADING OPERATIONS shall only be conducted with working area Ambient Temperature 0o F.

12. For those users whose site-specific design basis includes an event or events (e.g., flood) that result in the blockage of any VVM inlet or outlet air ducts for an extended period of time (i.e., longer than the total Completion Time of LCO 3.1.2), an analysis or evaluation may be performed to demonstrate adequate heat removal is available for the duration of the event. Adequate heat removal is defined as fuel cladding temperatures remaining below the short term temperature limit. If the analysis or evaluation is not performed, or if fuel cladding temperature limits are unable to be demonstrated by analysis or evaluation to remain below the short term temperature limit for the duration of the event, provisions shall be established to provide alternate means of cooling to accomplish this objective.

13. Users shall establish procedural and/or mechanical barriers to ensure that during LOADING OPERATIONS and UNLOADING OPERATIONS, either the fuel cladding is covered by water, or the MPC is filled with an inert gas.

14. The entire haul route shall be evaluated to ensure that the route can support the weight of the loaded transfer cask and its conveyance.

15. The loaded transfer cask and its conveyance shall be evaluated to ensure, under the site specific Design Basis Earthquake, that the cask and its conveyance does not tipover or slide off the haul route.

(continued)

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-11

Design Features 3.0 DESIGN FEATURES (continued)

Table 3-3 LOAD COMBINATIONS FOR THE TOP SURFACE PAD, ISFSI PAD, AND SUPPORT FOUNDATION PAD PER ACI-318 (2005)

Load Combination Case Load Combination LC-1 1.4D LC-2 1.2D + 1.6L LC-3 1.2D + E + L where:

D: Dead Load including long-term differential settlement effects.

L: Live Load E: DBE for the Site DESIGN FEATURES (continued)

Table 3-4 Values of Principal Design Parameters for the Underground ISFSI Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-12

Design Features 3.0 Thickness of the Support Foundation Pad, inch 33 (nominal)

Thickness of the ISFSI Pad, inch (nominal) 34 Thickness of the Top Surface Pad, inch 30 (nominal)

Rebar Size* (min.) and Layout* (max) #11 @ 9" each face, each direction Rebar Concrete Cover (top and bottom)*, inch per 7.7.1 of ACI-318 (2005)

Compressive Strength of Concrete at 28 4500 days*, psi Compressive Strength of Self-hardening 1,000 Engineered Subgrade (SES), psi Lower Bound Shear Wave Velocity in the 1,300 Subgrade lateral to the VVM (Figure 3-1 Space A), fps**

Depth Averaged Density of subgrade in Space 120 A. (Figure 3-1)1 Depth Averaged Density of subgrade in Space 110 B. (Figure 3-1)1 Depth Averaged Density of subgrade in Space 120 C. (Figure 3-1)2 Depth Averaged Density of subgrade in Space 120 D. (Figure 3-1)3 Lower Bound Shear Wave Velocity in the 485 Subgrade below the Support Foundation Pad (Figure 3-1 Space C & D), fps**

Lower Bound Shear Wave Velocity in the 450 Subgrade laterally surrounding the ISFSI (Figure 3-1 Space B), fps**

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-13

Design Features 3.0

Applies to Support Foundation Pad and ISFSI Pad.

- Strain compatible effective shear wave velocities shall be computed using the guidance provided in Section 16 of the International Building Code, 2009 Edition. Users must account for potential variability in the subgrade shear wave velocity in accordance with Section 3.7.2 of NUREG-0800.

Notes:

1. A lower average density value may be used in shielding analysis per FSAR Chapter 5 for conservatism.

2. Not required for shielding.

3. This space will typically contain native soil. Not required for shielding.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-14

Design Features 3.0 Figure 3 SUBGRADE AND UNDERGRADE SPACE NOMENCLATURE Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-15

Design Features 3.0 3.0 DESIGN FEATURES (continued) 3.5 Combustible Gas Monitoring During MPC Lid Welding and Cutting During MPC lid-to-shell welding and cutting operations, combustible gas monitoring of the space under the MPC lid is required, to ensure that there is no combustible mixture present.

3.6 Periodic Corrosion Inspections for Underground Systems HI-STORM UMAX VVM ISFSIs not employing an impressed current cathodic protection system shall be subject to visual and UT inspection of at least one representative VVM to check for significant corrosion of the CEC Container Shell and Bottom Plate at an interval not to exceed 20 years. The VVM chosen for inspection is not required to be in use or to have previously contained a loaded MPC. The VVM considered to be most vulnerable to corrosion degradation shall be selected for inspection. If significant corrosion is identified, either an evaluation to demonstrate sufficient continued structural integrity (sufficient for at least the remainder of the licensing period) shall be performed or the affected VVM shall be promptly scheduled for repair or decommissioning. Through wall corrosion shall not be permitted without promptly scheduling for repair or decommissioning. Promptness of repair or decommissioning shall be commensurate with the extent of degradation of the VVM but shall not exceed 3 years from the date of inspection.

If the representative VVM is determined to require repair or decommissioning, the next most vulnerable VVM shall be selected for inspection. This inspection process shall conclude when a VVM is found that does not require repair or decommissioning. Since the last VVM inspected is considered more prone to corrosion than the remaining un-inspected VVMs, the last VVM inspected becomes the representative VVM for the remaining VVMs.

Inspections Visual Inspection: Visual inspection of the inner surfaces of the CEC Container Shell and Bottom Plate for indications of significant or through wall corrosion (i.e.,

holes).

UT Inspection: The UT inspection or an equivalent method shall be used to measure CEC shell wall thickness to determine the extent of metal loss from corrosion. A minimum of 16 data points shall be obtained, 4 near the top, 4 near the mid-height and 4 near the bottom of the CEC Container Shell all approximately 0, 90, 180, and 270 degrees apart; and 4 on the CEC Bottom Plate near the CEC Container Shell approximately 0, 90, 180, and 270 degrees apart. Locations where visual inspection has identified potentially significant corrosion shall also receive UT inspection. Locations suspected of significant corrosion may receive further UT inspection to determine the extent of corrosion.

Inspection Criteria Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-16

Design Features 3.0 General wall thinning exceeding 1/8 in depth and local pitting exceeding 1/4" in depth are conditions of significant corrosion.

Certificate of Compliance No. 1040 Amendment No. 0 Appendix B 3-17

ML19311C705

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2.0 NOTES

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