ML21011A240

Shine Medical Technologies, LLC - Operating License Application Supplement No. 6 and Response to Request for Additional Information

ML21011A240
Person / Time
Site:	SHINE Medical Technologies
Issue date:	12/15/2020
From:	SHINE Medical Technologies
To:	Office of Nuclear Reactor Regulation
References
2020-SMT-0131
Download: ML21011A240 (176)

Text

ENCLOSURE 3 SHINE MEDICAL TECHNOLOGIES, LLC SHINE MEDICAL TECHNOLOGIES, LLC OPERATING LICENSE APPLICATION SUPPLEMENT NO. 6 AND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PUBLIC VERSION

Page 1 of 130 SHINE MEDICAL TECHNOLOGIES, LLC SHINE MEDICAL TECHNOLOGIES, LLC OPERATING LICENSE APPLICATION SUPPLEMENT NO. 6 AND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NON-PUBLIC VERSION The U.S. Nuclear Regulatory Commission (NRC) staff determined that additional information was required (Reference 1) to enable the continued review of the SHINE Medical Technologies, LLC (SHINE) operating license application (Reference 2). The following information is provided by SHINE in response to the NRC staffs request.

Chapter 2 - Site Characteristics RAI 2.2-1 SHINE analyzed explosive chemicals in FSAR Section 2.2.3.1.1, using NUREG-1537, which states in Section 2.2, Nearby Industrial, Transportation, and Military Facilities, that the information provided in the application should sufficient to support analyses to evaluate potential manmade hazards to the proposed facility due to nearby facilities.

The analyzed explosive chemicals results are presented in FSAR Tables 2.2-15 and 2.2-16. For an ethylene oxide tanker truck carrying 50,000 pounds (lbs) travelling on Highway 51 at a distance of 0.22 miles (mi.) from the facility, the evaluation is concluded to be acceptable by SHINE because it is bounded by a potential explosion of a storage tank of 44,0000 lbs at a distance of 2 mi. from the facility. The NRC staff, however, finds that the minimum safe (standoff) distance determined for this truck transport (0.54 mi.) exceeds the actual distance of 0.22 mi. from the closest point of roadway to the shortest distance to a safety-related structure at the SHINE Facility.

For propane and hydrogen, SHINE only used an unconfined explosion scenario with yield factor of 0.03. However, there is vapor in the tank that could explode as a confined vapor with a 100%

yield factor. The NRC staffs analysis finds that this scenario results in minimum safe distance that exceeds the actual roadway distance of 0.22 mi. for both propane and hydrogen.

(1) Justify and demonstrate how the impact from the oxide tanker truck is bounded by potential explosion of a storage tank impact.

(2) Justify and demonstrate how the confined explosion for propane and hydrogen with a yield factor of 100% is not evaluated.

This information is necessary for the NRC staff to conclude that potential explosions would not cause damage to safety-related equipment at the SHINE facility sufficient to pose undue radiological risks to the SHINE staff, the public, or the environment consistent with the evaluation findings in NUREG-1537, Part 2, Section 2.2. This information is also necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety

Page 2 of 130 functions will be accomplished by equipment that would be potentially impacted by an explosion consistent with 10 CFR 50.34(b)(2).

SHINE Response The SHINE Response to RAI 2.2-1 will be provided by January 31, 2021.

RAI 2.2-2 SHINE evaluated toxic chemicals in FSAR Section 2.2.3.1.3, using NUREG 1537, which states in Section 2.2, that the information provided in the application should be sufficient to support analyses to evaluate potential manmade hazards to the proposed facility due to nearby facilities.

Additionally, SHINE Design Criterion 6, Control Room, states that [a] control room is provided from which actions can be taken to operate the irradiation units safely under normal conditions and to perform required operator actions under postulated accident conditions.

In FSAR Section 2.2.3.1.3, SHINE identified four toxic chemicals that were found to be a potential hazard to the control room of the facility, including Ammonia from US 51, Chlorine fromI-90/39, Propylene oxide from I90/39, and Sodium bisulfite from US 51.

The NRC staff identified five additional toxic chemicals listed in FSAR Table 2.2 19 that have potential to be hazards to control room habitability, including Ethylene Oxide from US 51, Gasoline from US 51, Vinylidene chloride from rail (1.6 mi), Sodium hypochlorite from I-90/39, and Carbon Monoxide from a stationary source. The concentration of each of these chemicals was found by the NRC staff to exceed the respective IDLH (Immediately Dangerous Life and Health) concentrations of chemicals in the control room (The National Institute for Occupational Safety and Health (NIOSH) Table of IDLH Values may be found online at https://www.cdc.gov/niosh/idlh/intridl4.html). As stated in FSAR Section 2.2.3.1.3, a two-minute exposure to NIOSH IDLH chemical concentration limits could result in uninhabitability of the control room, which could prevent operators from having the necessary time (i.e., two minutes) to take required actions.

Provide additional information to demonstrate that the respective chemical potential concentrations from the five additional toxic chemicals do not exceed the respective chemical limiting IDLH concentrations.

This information is necessary for the NRC staff to conclude that potential toxic chemical exposures would not result in the uninhabitability of the control room and prevent the performance of required operator actions, as specified in SHINE Design Criterion 6. The continued habitability of the control room in the event of a toxic chemical release would further demonstrate that operators would be available to take required actions to ensure that safety-related equipment at the SHINE facility would not be damaged sufficient to pose undue radiological risks to the SHINE staff, the public, or the environment consistent with the evaluation findings in NUREG-1537, Part 2, Section 2.2. This information is also necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by toxic chemicals which could be hazards to control room habitability consistent with 10 CFR 50.34(b)(2).

Page 3 of 130 SHINE Response The SHINE Response to RAI 2.2-2 will be provided by January 31, 2021.

RAI 2.2-3 In FSAR Section 2.2.3.1.3.4, SHINE addressed the on-site chemical hazards by referencing FSAR Section 13b.3, using NUREG-1537, which states in Section 2.2, Nearby Industrial, Transportation, and Military Facilities, that the information provided should be sufficient to support analyses to evaluate potential manmade hazards to the proposed facility due to nearby facilities.

Additionally, SHINE Design Criterion 6, Control Room, states that [a] control room is provided from which actions can be taken to operate the irradiation units safely under normal conditions and to perform required operator actions under postulated accident conditions.

In FSAR Section 2.2.3.1.3.4, SHINE addressed on-site toxic chemicals by stating that they are evaluated in FSAR Section 13b.3 (see: Table 13b.3-2). SHINE also stated that worker exposures are representative of exposure to control room personnel. Based on the NRC staff review of the SHINE analyses and results, the evaluation methodology used in FSAR Section 13b.3 is different from that used in FSAR Section 2.2.3.1.3. Using a methodology consistent with that used in FSAR Section 2.2.3.1.2 (i.e., wind speed of 1m/s and Pasquill stability class F; use of IDLH concentration as limiting value), the NRC staff finds the chemicals Ammonia, Nitric acid, Sodium hydroxide could be a potential hazard to control room habitability as each of chemical concentration exceed respective chemical IDLH concentration. As stated in FSAR Section 2.2.3.1.3, a two-minute exposure to NIOSH IDLH chemical concentration limits could result in uninhabitability of the control room, which could prevent operators from having the necessary time (i.e., two minutes) to take required actions.

Provide information to justify in using average meteorological conditions as opposed to 1 m/s wind speed and F stability (representative of 5% percentile met conditions used conservatively),

for the analysis and considering worker exposures representative to the control room operators.

This information is necessary for the NRC staff to conclude that potential toxic chemical exposures would not result in the uninhabitability of the control room and prevent the performance of required operator actions, as specified in SHINE Design Criterion 6. The continued habitability of the control room in the event of a toxic chemical release would further demonstrate that operators would be available to take required actions to ensure that safety-related equipment at the SHINE facility would not be damaged sufficient to pose undue radiological risks to the SHINE staff, the public, or the environment consistent with the evaluation findings in NUREG-1537, Part 2, Section 2.2. This information is also necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by toxic chemicals which could be hazards to control room habitability consistent with 10 CFR 50.34(b)(2).

SHINE Response The SHINE Response to RAI 2.2-3 will be provided by January 31, 2021.

Page 4 of 130 RAI 2.4-1 The evaluation findings in NUREG-1537, Part 2, Section 2.4 state that the information provided by an applicant should be sufficient to support a finding that hydrologic events of credible frequency and consequence have been considered for the site. Additionally, credible hydrologic events have been considered in the development of the design bases for the facility to mitigate or avoid significant damage so that safe operation and shutdown of the facility would not be precluded by a hydrologic event.

Additionally, SHINE Design Criterion 2, Natural Phenomena Hazards, states that [t]he facility structure supports and protects safety-related SSCs and is designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches as necessary to prevent the loss of capability of safety-related SSCs to perform their safety functions.

Section 2.4.2.3, Effect of Local Intense Precipitation (LIP), of the SHINE FSAR states that the site is designed to withstand the effects of a local probable maximum precipitation (PMP) 100-year event, and that the maximum water levels due to local PMP were determined near the safety-related structures of the facilities.

The NRC staff notes that SHINE uses the 1-in-100-year rainfall event in its LIP flood analysis to evaluate the effects of onsite flooding. The World Meteorological Organization (WMO) defines probable maximum precipitation as the greatest depth of precipitation for a given duration meteorologically possible for a design watershed or a given storm area at a particular time of year. PMP depth in general is larger than that of 1-in-100-year rainfall. However, it is unclear to the NRC whether SHINE is applying this WMO definition of PMP to its LIP flood analysis and how this relates to a 1-in-100-year rainfall event.

Confirm the definition of PMP SHINE uses in its LIP flood analysis and describe how this relates to a 1-in-100-year rainfall event. Revise FSAR Section 2.4.2.3 and Table 2.4-7, as necessary to reflect SHINEs definition of PMP.

This information is necessary for the NRC staff to confirm that the SHINE facility is designed to withstand the effects of floods to prevent the loss of capability of safety-related SSCs to perform their safety-related functions, consistent with SHINE Design Criterion 2. This information is also necessary for the NRC staff to conclude that no credible predicted hydrologic event or condition would render the SHINE site unsuitable for operation or safe shutdown of the facility, consistent with the evaluation findings in NUREG-1537, Part 2, Section 2.4. Additionally, this information is necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by floods consistent with 10 CFR 50.34(b)(2).

SHINE Response SHINE defines the probable maximum precipitation (PMP) used in the local intense precipitation (LIP) flood analysis for the SHINE site as those rainfall values and intensities associated with a 1-in-100-year rainfall event. The PMP values and intensities of this 1-in-100-year rainfall event are provided in Table 2.4-7 of the FSAR.

The application guidance (References 3 and 4) does not specify a return interval to be used in defining the credible frequency and consequence of precipitation events, nor does the

Page 5 of 130 application guidance endorse the World Meteorological Organization (WMO) definition of probable maximum precipitation in defining the credible frequency and consequence of precipitation events. Therefore, SHINE chose to apply the PMP values and intensities associated with a 1-in-100-year rainfall event in the determination of the design basis precipitation level and did not apply the WMO definition of probable maximum precipitation.

Design of the stormwater drainage system to carry runoff from the site up to a 1-in-100-year rainfall event is consistent with the performance standards described in Sec.32-103 of the Code of General Ordinances of the City of Janesville, Wisconsin, for the design of stormwater management features. Additionally, use of the 100-year return interval in defining the PMP is consistent with the return intervals specified in Part 1 of NUREG-1537 (Reference 3) for estimating local climatic considerations (i.e., wind speeds and snowpack) in establishing the design basis of site structures.

The LIP flood analysis for the SHINE site determined that the PMP depth does not exceed the finished foundation elevation of the main production facility. As described in Section 3.3 of the FSAR, the PMP rainfall event results in a design basis precipitation level at grade. The main production facility finished foundation elevation is at least 4 inches above grade; therefore, water will not infiltrate the main production facility in the case of a PMP rainfall event.

SHINE has revised Subsection 2.4.2.3 of the FSAR to explicitly state that the SHINE site is designed to withstand those PMP values and intensities associated with the 1-in-100-year rainfall event. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 2.4-2 The evaluation findings in NUREG-1537, Part 2, Section 2.4 state that the information provided by an applicant should be sufficient to support a finding that hydrologic events of credible frequency and consequence have been considered for the site. Additionally, credible hydrologic events have been considered in the development of the design bases for the facility to mitigate or avoid significant damage so that safe operation and shutdown of the facility would not be precluded by a hydrologic event.

Additionally, SHINE Design Criterion 2, Natural Phenomena Hazards, states that [t]he facility structure supports and protects safety-related SSCs and is designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches as necessary to prevent the loss of capability of safety-related SSCs to perform their safety functions.

Section 2.4.2.3, Effect of Local Intense Precipitation (LIP) of the SHINE FSAR describes a drainage system designed to carry onsite and offsite runoffs generated from a 100-year frequency rainfall event. FSAR Figure 2.4-11 displays the elevation contour lines for the post construction ground surface condition and the boundary of offsite drainage basin. FSAR Figure 2.4-12 shows the drainage boundaries for onsite sub-basins with the direction of local runoffs.

However, the resolution of these contour lines is not sufficient to allow the staff to determine the adequacy of basin/sub-basin boundaries and the direction of runoffs, especially the runoffs from the offsite area east to the Onsite Sub-basin Numbers 6 and 9.

To support the NRC staffs understanding of whether SHINE has adequately considered the onsite and offsite drainage pattern in their LIP flood analysis, provide a higher-resolution map or maps showing detailed elevation contour lines for the post-construction ground surface

Page 6 of 130 condition with best available data, particularly at critical off-site areas that may govern the runoff process.

This information is necessary for the NRC staff to confirm that the SHINE facility is designed to withstand the effects of floods to prevent the loss of capability of safety-related SSCs to perform their safety-related functions, consistent with SHINE Design Criterion 2. This information is also necessary for the NRC staff to conclude that no credible predicted hydrologic event or condition would render the SHINE site unsuitable for operation or safe shutdown of the facility, consistent with the evaluation findings in NUREG-1537, Part 2, Section 2.4. Additionally, this information is necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by floods consistent with 10 CFR 50.34(b)(2).

SHINE Response SHINE has revised Figures 2.4-11 and 2.4-12 of the FSAR to provide higher resolution images which show detailed elevation contour lines for the post-construction ground surface condition on site. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

Topographic contours of off-site areas are available in the U.S. Geological Survey (USGS) topographical maps for the Janesville West, Beloit, Janesville East, and Shopiere quadrangles.

Page 7 of 130 Chapter 3 - Design of Structures, Systems, and Components RAI 3.2-1 The NRC staff reviewed the design criteria for the N2PS system as documented in DCD-N2PS-0001, Revision 1. This document describes the N2PS system as a safety-related system that is required for safe shutdown of the facility after a loss of offsite power or station blackout, and it establishes, in part, that SHINEs design criteria for natural phenomena hazards, Criterion 2, is applicable to the N2PS system. The document also describes the N2PS structure as a structure that supports and protects safety-related SSCs, and states that it is designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches as necessary to prevent the loss of capability of safety-related SSCs protected by the structure.

Based on the review of Chapter 3 of the FSAR, it is not clear where the design criteria and parameters for other structures not physically part of the main production facility structure are discussed in the FSAR. The staff notes that design criteria and parameters discussed under Chapter 3 are focused on the main production facility structure and do not clearly identify or discuss the design criteria and parameters applicable to other SSCs that perform an operational or safety function (e.g., N2PS structure). In addition, the staff also noted that Section 3.4.2.6.1 describes the SHINE facility as a boxtype shear wall system of reinforced concrete, which refers to the main production facility structure. However, this description contradicts the description provided in Section 1.4 for the SHINE facility. Therefore, it is not clear what structure(s) are being considered (or need to be considered) in Chapter 3 of the FSAR.

To clarify the issues described above provide the following information, updating the FSAR as necessary:

For each structure identified in FSAR Section 1.4 (i.e., resource building material staging building; storage building; and N2PS structure) that is not part of the main production facility structure and performs, supports, and/or protects a safety function address the following:

(1) Specify the applicable SHINE design criteria(s);

(2) Describe the criteria, parameters and methodology used for its design to ensure that protected safety-related SSCs can withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches (i.e., SHINE Design Criterion 2).

This information is necessary for the NRC staff to conclude that the design bases to protect against meteorological damage provides reasonable assurance that the facility structures, systems, and components will perform the safety functions discussed in the FSAR, consistent with the evaluation findings of Section 3.2, Meteorological Damage, of NUREG-1537, Part 2.

Additionally, this information is necessary for the NRC staff to conclude that SHINE is satisfying its Design Criterion 2, which states that [t]he facility structure supports and protects safety-related SSCs and is designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches as necessary to prevent the loss of capability of safety-related SSCs to perform their safety functions. Further, this information is necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by the effects of natural phenomena consistent with 10 CFR 50.34(b)(2).

Page 8 of 130 SHINE Response The SHINE Response to RAI 3.2-1 will be provided by January 31, 2021.

RAI 3.2-2 Based on the review of Section 3.2 of the FSAR, it is noted that some of the design criteria/parameters used for the SHINE facility are not sufficiently described in the FSAR to make a safety determination. Specifically, the following design criteria/parameters require further clarification:

Section 3.2.1, Wind Loading, defines equation no. 3.2-1 as the equation used to transform the wind speed into an equivalent pressure, however the values applicable to the site for exposure coefficient and other factors used in the equation were not defined in the FSAR. In addition, the values applicable to the site for the referenced gust factor and pressure coefficient were not defined in the FSAR.

Section 3.2.1, Wind Loading, defines V as the basic wind speed (3 second gust) obtained from Figure 61 of ASCE 705 for Wisconsin. However, additional clarification is needed because the ASCE figure provides wind speed for a 50-year return period which is not consistent with the mean recurrence interval of 100 years intended for the design of the SHINE Facility, as identified in FSAR Section 3.2.1.1.

Section 3.2.2, Tornado Loading, states that the design parameters are listed in Table 1 of NRC Regulatory Guide 1.76, however the values applicable to the site for the tornado rotational speed, translation speed, radius of maximum rotation, pressure drop, and rate of pressure drop were not defined in the FSAR. Similarly, Table 2 of NRC Regulatory Guide 1.76 was referenced for the applicable design basis tornado missile spectrum and maximum horizontal speed for the site, however these values were also not defined in the FSAR.

Section 3.2.3, Snow, Ice, and Rain Loading, references Chapter 7 of the ASCE7-05 standards as the applicable design parameter to the SHINE Facility, however it does not specify the snow load, recurrence interval and safety factor applicable to the site. Also, the section defines equation no. 3.2-3 as the equation used to determine the applied forces, however the values applicable to the site for the factors used in the equation were not defined in the FSAR.

For those design criteria and parameters described above, provide the applicable values for the site as considered for in the design of the SHINE facility to cope with meteorological damage.

Update the FSAR as necessary.

This information is necessary for the NRC staff to conclude that the design bases to protect against meteorological damage provides reasonable assurance that the facility structures, systems, and components will perform the safety functions discussed in the FSAR, consistent with the evaluation findings of Section 3.2 of NUREG-1537, Part 2. Additionally, this information is necessary for the NRC staff to conclude that SHINE is satisfying its Design Criterion 2.

Further, this information is necessary to demonstrate that SHINE has performed the appropriate

Page 9 of 130 evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by the effects of natural phenomena consistent with 10 CFR 50.34(b)(2).

SHINE Response Sections 3.2 and 3.4 of the FSAR have been revised to provide the applicable site design parameters, as described below.

Subsection 3.2.1 of the FSAR has been revised to specify the exposure coefficient and other factors used in Equation 3.2-1 of the FSAR, as well as the value applicable to the site for the referenced gust factor and a description of the applied pressure coefficients.

Subsection 3.2.1 of the FSAR has also been revised to specify the basic wind speed for Wisconsin, including a description of the applied factor to account for a 100-year recurrence interval.

Subsection 3.4.2.6.3.8 of the FSAR has been revised to specify the values applicable to the site for the tornado rotational speed, translation speed, differential pressure (i.e., pressure drop), and rate of differential pressure. Subsection 3.2.2.2 of the FSAR was revised to specify the design basis tornado missile spectrum and maximum horizontal speed for the site, and to provide reference to Subsection 3.4.2.6.3.8 of the FSAR for additional discussion of site design parameters related to tornado loading.

Subsection 3.2.3.2 of the FSAR has been revised to specify the values applicable to the site for the factors used in Equation 3.2-3 of the FSAR, including the snow load and recurrence interval.

A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.2-3 The staff evaluates the adequacy of the design criteria of all SSCs that have been identified to perform an operational or safety function by using the guidance and acceptance criteria described in Chapter 3, Design of Structures, System, and Components, of NUREG-1537, Parts 1 and 2. Specifically, Section 3.2 of Part 2 instructs the staff, in part, to ensure that the information provided on meteorological damage include the design criteria and design to provides reasonable assurance that SSCs would continue to perform the safety function under potential meteorological damage conditions.

Section 3.2.2.2 of the FSAR states that the procedure used for transforming the tornado generated missile impact into an effective or equivalent static load on the structures is consistent with Section 3.5.2, Subsection II, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants.

However, based on the NRC staffs review of Section 3.2.2.2 of the FSAR, it is not clear what procedure or criteria were followed to transform the tornado generated missile impact into an effective or equivalent static load on the structures because Section 3.5.2, Subsection II, of NUREG-0800 does not define a procedure or criteria for transforming tornado generated missile impact into an effective or equivalent static load.

Describe the methodology or procedure used for transforming tornado generated missile impact into an effective or equivalent static load on the structures, and state how it is acceptable to

Page 10 of 130 ensure that safety-related SSCs are protected from tornado generated impacts in accordance with SHINE Design Criteria 2 and 4. Update the FSAR as necessary.

This information is necessary for the NRC staff to conclude that the design bases to protect against meteorological damage provides reasonable assurance that the facility structures, systems, and components will perform the safety functions discussed in the FSAR, consistent with the evaluation findings of Section 3.2 of NUREG-1537, Part 2. Additionally, this information is necessary for the NRC staff to conclude that SHINE is satisfying its Design Criteria 2 and 4.

Further, this information is necessary to demonstrate that SHINE has performed the appropriate evaluations required to show that safety functions will be accomplished by equipment that would be potentially impacted by the effects of natural phenomena consistent with 10 CFR 50.34(b)(2).

SHINE Response The methodology used to transform the tornado generated missile impacts into effective or equivalent static loads on the structure is based on the acceptance criteria delineated in Section 3.5.3, Subsection II of NUREG-0800 (Reference 5). The design of the FSTR ensures that safety-related systems, structures, and components (SSCs) can withstand the effects of tornado missiles, as required by SHINE Design Criteria 2 and 4, in conformance with accepted guidance. Subsection 3.2.2.2 of the FSAR has been revised to correct the reference to NUREG-0800, Section 3.5.3, Subsection II, and to add additional information related to the tornado missiles considered in the analysis. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.3-1 Section 3.3 of NUREG-1537, Part 2, notes that facility designs should provide reasonable assurance that structures, systems, and components will continue to perform required safety functions under water damage conditions.

FSAR Section 3.3 notes that the bounding internal flood is due to fire protection and that it will result in a maximum depth of water in the radiologically controlled area (RCA) of 2 inches. The FSAR further notes that water sensitive safety-related equipment is raised 8 inches from the floor; however, Table 9-1 in calculation CALC-2020-0001 summarizes water depths in the RCA during manual fire suppression and notes values greater than 8 inches. The NRC staff notes that internal flood levels in excess of 8 inches could damage and impact the performance of water-sensitive safety-related equipment.

CALC-2020-0001 identifies internal flood water depths in the radioisotope process facility cooling system (RPCS) room based on breaks in either the RPCS or process chilled water system (PCHS) line in the RPCS room. The calculation further notes that these water depths are not a concern because the RPCS room has an appropriately sized manual flood barrier.

(1) Explain how water-sensitive safety-related equipment in the RCA will be protected from internal flood waters that may rise above 8 inches. Update the FSAR, as necessary.

(2) Explain how water-sensitive safety-related equipment within the RPCS room will be protected from water that rises to the depths identified in CALC-2020-0001. If the manual flood barrier is being relied on to keep RPCS or PCHS leakage from leaving the RPCS room, explain how it is ensured that the barrier will be in place if there is an accident. Update the FSAR, as necessary.

Page 11 of 130 Consistent with the evaluation findings in NUREG-1537, Part 2, Section 3.3 and 10 CFR 50.34(b)(2), the this information is requested for the NRC staff to conclude that the design bases of the SHINE facility protects against potential hydrological damage and provides reasonable assurance that the facility structures, systems, and components (SSCs) will perform the functions necessary to allow any required operation to continue safely, to allow safe shutdown, and to protect the health and safety of the public from radioactive materials and radiation exposure.

SHINE Response (1) Evaluation of the bounding internal flood volume of 15,000 gallons distributed over the minimum open floor area of the grade level floor space within the irradiation facility (IF) produces the bounding flood scenario. The resulting maximum flood height for this scenario is approximately 11.7 inches. Subsection 3.3.1.1.2 of the FSAR has been revised to raise the minimum height of water sensitive safety-related equipment to 12 inches off the floor in the radiologically controlled area (RCA). Subsection 3.3.1.1.2 of the FSAR has also been revised to clarify the resulting maximum flood height when the discharge volume is distributed over the entire RCA. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

(2) The bounding flood scenario for the radioisotope process facility cooling system (RPCS) room results in a flood height of approximately 22.9 inches. The minimum height of water sensitive safety-related equipment in the RPCS room is 24 inches. Subsection 3.3.1.1.2 of the FSAR has been revised to include a minimum height for water sensitive safety-related equipment in the RPCS room of 24 inches. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

The manual flood barrier within the RPCS room is not relied upon within the SHINE safety analysis to keep RPCS or process chilled water system (PCHS) leakage from leaving the RPCS room. The manual flood barrier within the RPCS room is a defense-in-depth (DID) measure, designed to slow the release of potential flood water to the transfer aisle to avoid any surge that could challenge the height of the elevated floor in the uranium receipt and storage system (URSS) and target solution preparation system (TSPS) rooms. In an RPCS flood scenario, the water will be slowed by both the normally closed door to the RPCS as well as the manual flood barrier. As the flood water enters the transfer aisle, it will continue to spread into the radioisotope production facility (RPF). SHINE has revised Table 5.5.4 of the technical specifications to remove the RPCS room manual flood barrier from the list of main production facility structure controls. A mark-up of the technical specifications incorporating these changes is provided as Attachment 2. SHINE will provide a revision to the technical specifications incorporating the mark-up by February 28, 2021.

RAI 3.3-2 Section 3.3 of NUREG-1537, Part 2, notes that facility designs should provide reasonable assurance that structures, systems, and components will continue to perform required safety-related functions under water damage conditions.

FSAR Section 3.3.1.1.2 notes that the uninterruptible electrical power supply system (UPSS) has two redundant and isolated trains to prevent both trains from being damaged by discharge of the fire protection system (FPS). However, the FSAR does not discuss how other

Page 12 of 130 water-sensitive safety-related equipment in the RCA is protected from damage due to discharge of the FPS.

Explain how safety-related, water sensitive equipment in the RCA is protected from damage due to discharge of the FPS. Update the FSAR, as necessary.

Consistent with the evaluation findings in NUREG-1537, Part 2, Section 3.3 and 10 CFR 50.34(b)(2), the this information is requested for the NRC staff to conclude that the design bases of the SHINE facility protects against potential hydrological damage and provides reasonable assurance that the facility SSCs will perform the functions necessary to allow any required operation to continue safely, to allow safe shutdown, and to protect the health and safety of the public from radioactive materials and radiation exposure.

SHINE Response Discharge of the fire suppression system within the RCA consists of manual discharge via fire hoses from dry standpipes, except in those areas of the RCA in which gaseous fire suppression is provided, as described in Section 9a2.3 of the FSAR. As described in the SHINE Response to RAI 3.3-1, water sensitive safety-related equipment within the RCA is raised at least 12 inches above floor height to avoid potential damage due to submergence or partial submergence from flooding. The safety-related function(s) of equipment that are subject to the effects of a discharge of the fire suppression system are appropriately protected by redundancy and separation, where practicable. Where redundant equipment is unable to be effectively separated, fire response plans are established to ensure redundant trains of water sensitive safety-related equipment are not both subject to damage due to discharge of the fire suppression system.

SHINE has revised Subsection 3.3.1.1 of the FSAR to provide a description of how water sensitive safety-related equipment within the RCA is protected from damage due to discharge of the fire suppression system. A mark-up of the FSAR incorporating these changes is provided as.

RAI 3.4-1 In FSAR Section 3.4.1, Seismic Input, SHINE discusses how the design time histories for the seismic analysis of the SHINE facility structures (FSTR) are generated and states that the structural damping values for various structural elements used in the seismic analysis are provided in Section 1.1 of Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Revision 1. The applicant further states that, in the modal analysis of structures composed of different materials (having different damping values), the composite modal damping is calculated using either the stiffness-weighted method or mass-weighted method based on NUREG-0800, SRP Section 3.7.2. In FSAR Section 3.4.2.2, Soil-Structure Interaction Analysis, SHINE discusses three bounding soil properties (best estimate (BE),

upper bound (UB), and lower bound (LB)) used in the seismic analysis of the FSTR accounting for potential variations in in-situ and backfill soil conditions around the building. However, the staff notes that SHINE did not provide actual numerical data for these input parameters used in the seismic analysis of the FSTR. This information is important for the staff to assess the adequacy of the input used in the seismic analysis of the FSTR. Therefore, SHINE is requested to provide the following information, updating the FSAR with a summary of results, as necessary:

Page 13 of 130 (1) Numerical data (in figures or tabular form) for the input ground motion time histories used in the seismic analysis of the FSTR. Also, a comparison of the response spectra obtained from the input ground motion time histories with the target design response spectra (i.e., the Safe Shutdown Earthquake or SSE), demonstrating that the enveloping criteria of NUREG-0800, SRP 3.7.1 are satisfied, as applicable.

(2) Critical damping values used for various structural elements (or element groups) and the composite modal damping method used in the seismic analysis of the FSTR.

(3) Numerical data (in figures or tabular form) for the three bounding soil columns (BE, UB, and LB) used in the seismic analysis of the FSTR.

SHINE Response (1) Numerical data for the input ground motion time histories (i.e., artificial acceleration time histories) used in the seismic analysis of the main production facility structure (FSTR) are provided in Figures 3.4-1-1 through 3.4-1-3. Three safe shutdown earthquake (SSE) consistent mutually orthogonal artificial acceleration time histories (two horizontal and one vertical) are generated, with seed recorded time histories from the 1940 El Centro earthquake. These artificial acceleration time histories are used as the free-field (outcrop) input motion in the soil column site response analysis to determine strain-compatible soil properties and in-profile ground motions used in the seismic analysis of the FSTR.

The target design response spectra used in the seismic analysis of the FSTR are provided in Figures 3.4-1-4 and 3.4-1-5. Comparison of the response spectra obtained from the input ground motion time histories with the target design response spectra, provided in Figures 3.4-1-6 through 3.4-1-8, illustrate that the enveloping criteria of Section 3.7.1 of NUREG-0800 (SRP 3.7.1) (Reference 6) are satisfied. Subsection 3.4.1.2 of the FSAR has been revised to include a summary of these results. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

The seismic design parameters used in the seismic analysis of the FSTR are consistent with SRP 3.7.1 acceptance criteria, including:

Each artificial acceleration time history is generated starting with a seed recorded time history.

The 5 percent target spectral acceleration response spectra are computed at 100 points per frequency decade, uniformly spaced over the log frequency scaled from 0.1 Hz to 50 Hz.

The total duration of the time histories are 40 seconds, well above the 20 second requirement.

The 5 percent damped acceleration response spectra of the artificial acceleration time histories do not fall below 90 percent of the target acceleration response spectra at any one frequency and no more than nine adjacent frequency points fall below the target acceleration response spectra.

The computed 5 percent damped acceleration response spectra of artificial acceleration time histories do not exceed the target acceleration response spectra at any frequency by more than 30 percent in the frequency range of 0.1 Hz to 50 Hz.

The durations of strong motion are 13.27 seconds, 18.785 seconds, and 14.02 seconds for the artificial acceleration time histories H1, H2, and V, respectively. These durations of strong motions exceed the minimum acceptable strong motion duration of 6 seconds.

Page 14 of 130 The calculated absolute values of correlation coefficients are 0.058, 0.082, and 0.069 for time history pairs H1, H2; H1, V; and H2, V. These values are less than the maximum allowable correlation coefficient of 0.16.

(2) Damping values for various structural elements used in the seismic analyses are provided in Section 1.1 of Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plans (Reference 7). In the seismic analysis of the FSTR, the uncracked analysis cases utilize the operating basis earthquake (OBE) damping values of 4 percent for concrete and 3 percent for steel. The cracked analysis case utilizes the SSE damping values of 7 percent for concrete and 4 percent for steel. The SSE damping values are associated with high seismic response of the structure where concrete cracking and relatively large deformations are expected. Therefore, the SSE damping values are appropriate for the cracked seismic analysis case in which cracking is assumed to have occurred.

The SHINE seismic soil-structure interaction (SSI) analysis of the FSTR is performed in the program SASSI2010, System for Analysis of Soil-Structure Interaction, which performs the analysis in the frequency domain. The variations in damping are accounted for in the seismic SSI analysis through the complex frequency response analysis method which incorporates damping as an imaginary component in the stiffness matrix. The SASSI2010 SSI analysis captures the interaction between the embedded structure and the surrounding soil through modeling the equivalent linear strain-dependent soil properties, the excavated soil volume, and the FSTR. In SASSI2010, the damping of the system is not represented by the viscous damping matrix (i.e., Equation 3.4-1 of the FSAR, [C]), but as an imaginary component in the complex stiffness matrix. Thus, Equation 3.4-1 of the FSAR becomes complex and must be solved in the frequency domain. Subsection 3.4.1.3 of the FSAR has been revised to clarify damping as applied via SASSI2010 in the SHINE SSI analysis of the FSTR. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

(3) Numerical data for the three bounding soil columns (i.e., lower bound [LB], best estimate

[BE], and upper bound [UB]) used in the seismic analysis of the FSTR are provided in Tables 3.4-1-1 through 3.4-1-3.

Page 15 of 130 Table 3.4-1 Lower Bound Strain Dependent Soil Properties Layer No.

Thickness (ft)

Unit Weight (kcf)

ShearWave VelocityVs(ft/s)

Compression WaveVelocityVp (ft/s)

Dampingfor Vs(%)

Dampingfor Vp(%)

1 2.50 0.119 330.37 674.72 8.2 8.2 2

2.50 0.119 347.48 754.13 11.4 11.4 3

4.00 0.119 578.19 1126.14 10.0 10.0 4

3.00 0.119 726.07 1382.83 9.8 9.8 5

4.00 0.119 683.04 1413.27 11.5 11.5 6

3.00 0.119 585.32 1317.78 13.9 13.9 7

4.00 0.119 499.35 1077.86 15.0 15.0 8

2.25 0.119 301.77 545.93 15.0 15.0 9

3.25 0.119 527.95 838.38 15.0 15.0 10 3.25 0.119 543.92 850.06 15.0 15.0 11 3.25 0.119 528.83 861.92 15.0 15.0 12 3.25 0.119 534.02 913.02 15.0 15.0 13 3.25 0.119 512.51 876.23 15.0 15.0 14 3.25 0.119 491.74 840.72 15.0 15.0 15 3.25 0.119 471.80 806.64 15.0 15.0 16 3.25 0.119 455.07 802.07 15.0 15.0 17 3.25 0.119 442.12 846.50 15.0 15.0 18 3.25 0.119 434.26 907.16 15.0 15.0 19 3.25 0.119 429.26 1661.89 15.0 15.0 20 3.25 0.119 425.00 2167.09 15.0 15.0 21 3.25 0.119 421.50 2149.24 15.0 15.0 22 3.25 0.119 418.54 2134.16 15.0 15.0 23 3.25 0.119 416.80 2125.25 15.0 15.0 24 3.25 0.119 415.95 2120.91 15.0 15.0 25 3.25 0.119 414.43 2113.21 15.0 15.0 26 3.25 0.119 410.49 2093.10 15.0 15.0 27 3.25 0.119 406.65 2073.53 15.0 15.0 28 3.25 0.119 470.78 2400.54 15.0 15.0 29 3.25 0.119 633.23 3228.84 15.0 15.0 30 7.00 0.119 928.47 4577.43 14.8 14.8 31 7.00 0.119 951.52 4800.00 15.0 15.0 32 7.00 0.119 914.56 4660.13 15.0 15.0 33 7.00 0.119 1158.96 4788.94 14.6 14.6 34 7.00 0.119 1218.37 4800.00 14.3 14.3 35 7.00 0.119 1243.55 4800.00 14.4 14.4 36 7.00 0.119 1109.49 4800.00 15.0 15.0 37 7.00 0.119 1162.38 4800.00 15.0 15.0 38 7.00 0.119 1129.10 4800.00 15.0 15.0 39 7.00 0.119 941.47 4800.00 15.0 15.0 40 10.00 0.119 1469.01 4800.00 13.8 13.8 41 10.00 0.119 1556.94 4800.00 13.7 13.7 42 10.00 0.122 1279.93 4800.00 12.4 12.4 43 10.00 0.119 1440.19 4800.00 14.7 14.7 44 10.00 0.119 1481.83 4800.00 15.0 15.0 45 10.00 0.119 1481.83 4800.00 15.0 15.0 46 10.00 0.119 1481.83 4800.00 15.0 15.0 47 10.00 0.119 1396.24 4800.00 15.0 15.0 48 10.00 0.119 1382.15 4800.00 15.0 15.0 49 10.00 0.119 1382.15 4800.00 15.0 15.0 50 10.00 0.119 1382.15 4800.00 15.0 15.0 51 10.00 0.119 1382.15 4800.00 15.0 15.0 52 10.00 0.119 1382.15 4800.00 15.0 15.0 53 6.50 0.119 1382.15 4800.00 15.0 15.0

halfspace 0.15 6370.04 13260.29 1.1 1.1 Note that damping values higher than 15% are limited to 15% in the seismic SSI analysis.

Page 16 of 130 Table 3.4-1 Best Estimate Strain Dependent Soil Properties Layer No.

Thickness (ft)

Unit Weight (kcf)

ShearWave VelocityVs(ft/s)

Compression WaveVelocityVp (ft/s)

Dampingfor Vs(%)

Dampingfor Vp(%)

1 2.50 0.119 446.63 909.31 4.0 4.0 2

2.50 0.119 498.66 1081.81 5.5 5.5 3

4.00 0.119 806.23 1570.15 4.8 4.8 4

3.00 0.119 1009.20 1921.99 4.7 4.7 5

4.00 0.119 986.91 2039.54 5.5 5.5 6

3.00 0.119 896.01 2016.51 6.6 6.6 7

4.00 0.119 814.72 1771.63 7.9 7.9 8

2.25 0.119 576.52 1042.97 11.3 11.3 9

3.25 0.119 858.20 1358.97 8.4 8.4 10 3.25 0.119 874.42 1366.53 8.7 8.7 11 3.25 0.119 861.33 1403.84 9.2 9.2 12 3.25 0.119 875.47 1496.79 9.4 9.4 13 3.25 0.119 858.94 1468.53 9.9 9.9 14 3.25 0.119 843.55 1442.21 10.3 10.3 15 3.25 0.119 829.15 1417.60 10.7 10.7 16 3.25 0.119 815.76 1437.50 11.1 11.1 17 3.25 0.119 803.19 1537.44 11.4 11.4 18 3.25 0.119 791.37 1653.22 11.8 11.8 19 3.25 0.119 779.97 3021.46 12.1 12.1 20 3.25 0.119 767.90 3915.54 12.5 12.5 21 3.25 0.119 756.75 3858.68 12.8 12.8 22 3.25 0.119 745.45 3801.04 13.2 13.2 23 3.25 0.119 736.16 3753.69 13.4 13.4 24 3.25 0.119 727.59 3709.99 13.7 13.7 25 3.25 0.119 719.69 3669.71 13.9 13.9 26 3.25 0.119 712.61 3633.60 14.1 14.1 27 3.25 0.119 706.14 3600.60 14.3 14.3 28 3.25 0.119 794.41 4050.71 13.1 13.1 29 3.25 0.119 998.46 4704.01 11.0 11.0 30 7.00 0.119 1376.00 4800.00 8.6 8.6 31 7.00 0.119 1450.17 4800.00 8.4 8.4 32 7.00 0.119 1453.34 4800.00 8.5 8.5 33 7.00 0.119 1795.87 4800.00 7.0 7.0 34 7.00 0.119 1890.38 4800.00 6.8 6.8 35 7.00 0.119 1944.56 4800.00 6.7 6.7 36 7.00 0.119 1808.39 4800.00 7.3 7.3 37 7.00 0.119 1891.31 4800.00 7.0 7.0 38 7.00 0.119 1871.29 4800.00 7.1 7.1 39 7.00 0.119 1654.24 4800.00 8.2 8.2 40 10.00 0.119 2315.30 5123.65 5.9 5.9 41 10.00 0.119 2439.41 5078.05 5.7 5.7 42 10.00 0.122 1870.97 4837.22 5.1 5.1 43 10.00 0.119 2329.49 5066.10 5.8 5.8 44 10.00 0.119 2457.72 5116.15 6.0 6.0 45 10.00 0.119 2457.72 5116.15 6.0 6.0 46 10.00 0.119 2457.72 5116.15 6.0 6.0 47 10.00 0.119 2410.91 5018.72 6.5 6.5 48 10.00 0.119 2402.84 5001.91 6.6 6.6 49 10.00 0.119 2402.84 5001.91 6.6 6.6 50 10.00 0.119 2402.84 5001.91 6.6 6.6 51 10.00 0.119 2402.84 5001.91 6.6 6.6 52 10.00 0.119 2402.84 5001.91 6.6 6.6 53 6.50 0.119 2402.84 5001.91 6.6 6.6

halfspace 0.15 7963.27 16576.87 0.8 0.8

Page 17 of 130 Table 3.4-1 Upper Bound Strain Dependent Soil Properties Layer No.

Thickness (ft)

Unit Weight (kcf)

ShearWave VelocityVs(ft/s)

Compression WaveVelocityVp (ft/s)

Dampingfor Vs(%)

Dampingfor Vp(%)

1 2.50 0.119 580.92 1180.80 1.2 1.2 2

2.50 0.119 670.60 1454.61 1.6 1.6 3

4.00 0.119 1070.00 2083.61 1.4 1.4 4

3.00 0.119 1332.95 2538.50 1.4 1.4 5

4.00 0.119 1323.23 2733.32 1.7 1.7 6

3.00 0.119 1227.55 2762.24 2.1 2.1 7

4.00 0.119 1157.97 2526.74 2.4 2.4 8

2.25 0.119 890.33 1610.67 3.4 3.4 9

3.25 0.119 1233.94 1951.79 2.5 2.5 10 3.25 0.119 1267.14 1980.19 2.5 2.5 11 3.25 0.119 1266.25 2063.75 2.6 2.6 12 3.25 0.119 1293.47 2211.45 2.7 2.7 13 3.25 0.119 1281.22 2190.51 2.9 2.9 14 3.25 0.119 1268.18 2168.21 3.1 3.1 15 3.25 0.119 1256.10 2147.56 3.3 3.3 16 3.25 0.119 1244.89 2193.51 3.4 3.4 17 3.25 0.119 1234.47 2362.64 3.6 3.6 18 3.25 0.119 1224.75 2558.31 3.8 3.8 19 3.25 0.119 1215.62 4022.89 3.9 3.9 20 3.25 0.119 1207.08 4800.00 4.0 4.0 21 3.25 0.119 1199.11 4800.00 4.1 4.1 22 3.25 0.119 1191.15 4800.00 4.3 4.3 23 3.25 0.119 1184.13 4800.00 4.3 4.3 24 3.25 0.119 1177.58 4829.59 4.4 4.4 25 3.25 0.119 1171.44 4923.79 4.5 4.5 26 3.25 0.119 1165.66 4899.49 4.6 4.6 27 3.25 0.119 1160.26 5308.16 4.7 4.7 28 3.25 0.119 1267.13 6461.12 4.3 4.3 29 3.25 0.119 1508.66 6847.38 3.7 3.7 30 7.00 0.119 1967.93 6777.61 2.7 2.7 31 7.00 0.119 2062.63 5484.20 2.7 2.7 32 7.00 0.119 2074.37 4800.00 2.7 2.7 33 7.00 0.119 2463.82 5172.98 2.3 2.3 34 7.00 0.119 2572.23 5354.53 2.3 2.3 35 7.00 0.119 2636.82 5488.98 2.3 2.3 36 7.00 0.119 2490.29 5183.95 2.4 2.4 37 7.00 0.119 2584.23 5379.50 2.4 2.4 38 7.00 0.119 2564.76 5338.98 2.4 2.4 39 7.00 0.119 2323.41 4836.56 2.7 2.7 40 10.00 0.119 3102.68 6458.75 2.1 2.1 41 10.00 0.119 3237.76 6739.94 2.0 2.0 42 10.00 0.122 2355.52 5014.79 1.8 1.8 43 10.00 0.119 3060.57 6403.70 2.1 2.1 44 10.00 0.119 3268.98 6804.92 2.2 2.2 45 10.00 0.119 3268.98 6804.92 2.2 2.2 46 10.00 0.119 3268.98 6804.92 2.2 2.2 47 10.00 0.119 3248.48 6762.25 2.3 2.3 48 10.00 0.119 3244.89 6754.78 2.3 2.3 49 10.00 0.119 3244.89 6754.78 2.3 2.3 50 10.00 0.119 3244.89 6754.78 2.3 2.3 51 10.00 0.119 3244.89 6754.78 2.3 2.3 52 10.00 0.119 3244.89 6754.78 2.3 2.3 53 6.50 0.119 3244.89 6754.78 2.3 2.3

halfspace 0.15 9763.21 20323.75 0.8 0.8

Page 18 of 130 Figure 3.4-1 Artificial (Synthetic) Acceleration Time History H1 (Modification of Imperial Valley 1940 El Centro Array #9, 180)

Page 19 of 130 Figure 3.4-1 Artificial (Synthetic) Acceleration Time History H2 (Modification of Imperial Valley 1940 El Centro Array #9, 270)

Page 20 of 130 Figure 3.4-1 Artificial (Synthetic) Acceleration Time History V (Modification of Imperial Valley 1940 El Centro Array #9, UP)

Page 21 of 130 Figure 3.4-1 5% Damped Horizontal SSE Target Acceleration Response Spectrum Frequency (Hz)

Acceleration (g)

Point A 33 0.2 Point B 9

0.522 Point C 2.5 0.626 Point D 0.25 0.094 Point E 0.1 0.015

Page 22 of 130 Figure 3.4-1 5% Damped Vertical SSE Target Acceleration Response Spectrum Frequency (Hz)

Acceleration (g)

Point A 33 0.2 Point B 9

0.522 Point C 3.5 0.596 Point D 0.25 0.063 Point E 0.1 0.01

Page 23 of 130 Figure 3.4-1 Acceleration Response Spectra Comparison between the Generated, Target, 130% Target, and 90% Target Spectra for Horizontal Artificial Time-History H1

Page 24 of 130 Figure 3.4-1 Acceleration Response Spectra Comparison between the Generated, Target, 130% Target, and 90% Target Spectra for Horizontal Artificial Time-History H2

Page 25 of 130 Figure 3.4-1 Acceleration Response Spectra Comparison between the Generated, Target, 130% Target, and 90% Target Spectra for Vertical Artificial Time-History V

Page 26 of 130 RAI 3.4-2 In FSAR Section 3.4.2.1, Seismic Analysis Methods, SHINE discusses the general equations of motion and the finite element model used for the seismic analysis of the FSTR. However, the staff notes that SHINE did not specifically identify the seismic analysis methods used in the seismic analysis of the FSTR. In FSAR Section 3.4.2.1, the applicant also indicates that the finite element model consists of shell, solid, beam, or a combination of these elements. In FSAR Section 3.4.2.2, the applicant explains that shell elements are used to represent concrete slabs and walls and beam elements to represent steel members; however, the applicant did not explain how the solid elements are used. Therefore, SHINE is requested to address the following and update the FSAR as appropriate:

(1) Identify and describe in the FSAR the seismic analysis methods (e.g., response spectrum method, time history method, equivalent static load method, etc.) used in the seismic analysis of the FSTR.

(2) Update the FSAR by describing the usage of solid elements in finite element discretization of the soil-structure interactive system for the safety-related SSCs of the SHINE facility.

SHINE Response (1) The SHINE seismic analysis of the FSTR is performed in the program SASSI2010, System for Analysis of Soil-Structure Interaction, and SAP2000, integrated software for structural analysis and design. SASSI2010 performs a complex frequency response analysis using input acceleration time-histories consistent with the defined SSE response spectra to determine the seismic response of the structure and generate required in-structure response spectra (ISRS), element forces and moments, and nodal accelerations. The element forces and moments are permuted with the static analysis results in the element-based design evaluations for the structure. The ISRS provide acceleration spectra at key locations which are used in the design/evaluation of systems and components within the FSTR. The bounding nodal accelerations are exported, and an equivalent static analysis is performed using the program SAP2000 to evaluate wall in-plane shear force, wall overturning, and building stability.

Subsection 3.4.2.1 of the FSAR has been updated to identify the use of SASSI2010 and SAP2000 for the seismic analysis of the FSTR. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

(2) Solid elements are only used in the modeling of the excavated soil volume. No solid elements are used in the modeling of the building structure. The main components of the SASSI2010 model are the horizontal soil layers, the excavated soil model, and the structural model. Horizontal soil layers are used to model the soil column down to an elastic half space which represents bedrock. Soil layer thickness is determined to ensure transmittal of the highest cut-off frequency of interest in the analysis. The excavated soil is modeled using solid elements. The solid elements are modeled such that the vertical thickness of the elements match the height of the corresponding soil layers. The properties of the soil solid elements match the equivalent linear strain-compatible soil properties determined for the site. Soil layer thickness and maximum soil solid element dimensions are defined such that they are fine enough to ensure transmittal of the highest frequency of interest in the analysis.

Page 27 of 130 Subsection 3.4.2.2 of the FSAR has been updated to describe the use of solid elements in the SSI analysis of the FSTR. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.4-3 In FSAR Section 3.4.2.2, Soil-Structure Interaction Analysis, the applicant states that, in addition to self-weight of the structure, floor loads and equipment loads are converted to mass and included in the model and that a portion of the loads are considered mass sources in the following manner according to NUREG-0800, SRP Section 3.7.2: Dead Load - 100 percent; Live Load - 25 percent; Snow Load - 75 percent. The staff notes Acceptance Criteria 3.D of SRP Section 3.7.2 states, in part, that In addition to the structural mass, mass equivalent to a floor load of 50 pounds per square foot should be included, to represent miscellaneous dead weights such as minor equipment, piping, and raceways. Also, mass equivalent to 25 percent of the floor design live load and 75 percent of the roof design snow load, as applicable, should be included.

In view of these SRP acceptance criteria, it appears that the applicant did not consider the mass equivalent to a floor load of 50 pounds per square foot to represent miscellaneous dead weights on the floor. The staff also notes that FSAR Section 3.4.2.6.4.5 addresses this topic but includes two additional items in the bulleted list there: Parked Crane Load - 100 percent; Hydrodynamic Load - 100 percent. Therefore, the applicant is requested to address the following questions and update the FSAR as appropriate:

(1) Explain whether the mass equivalent to a floor load of 50 pounds per square foot to represent miscellaneous dead weights on the floor is considered in the seismic analysis of the FSTR; if not, provide justification for not considering it.

(2) Explain an apparent discrepancy between FSAR Sections 3.4.2.2 and 3.4.2.6.4.5 in the information about percentages of the loads considered as mass sources in the seismic analysis of the FSTR.

SHINE Response (1) A mass equivalent floor load of 50 pounds per square foot to represent miscellaneous dead weights on the floor is considered in the seismic analysis of the FSTR. Subsections 3.4.2.2 and 3.4.2.6.4.5 of the FSAR have been revised to clarify that this miscellaneous equipment floor load was considered in the seismic analysis. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

(2) The seismic analysis of the FSTR included 100 percent of the hydrodynamic mass of the water in the irradiation unit (IU) cells and 100 percent of the parked crane mass.

Subsection 3.4.2.2 of the FSAR has been revised to identify that both the hydrodynamic and crane masses were considered as mass sources in the seismic analysis of the FSTR, consistent with the discussion in Subsection 3.4.2.6.4.5 of the FSAR. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.4-4 In FSAR Section 3.4.2.4, Seismic Analysis Results, the applicant discusses seismic loads for structural design and in-structure response spectra for sizing equipment and components.

However, the staff notes that the applicant did not include analysis results (in tabular form or figures) which provide design-basis demands for the seismic design of the FSTR and seismic

Page 28 of 130 qualification of safety-related equipment. Further, the staff notes SHINEs creation of an item in the Issues Management Report (IMR; documented in NUREG-2189, Safety Evaluation Report related to SHINE Construction Permit Application) System, which is associated with RAI 3.4-1 from the staffs construction permit application review. This IMR item, which tracks the inclusion of the final seismic analysis results into the FSAR, is an applicants regulatory commitment and the staff verifies its implementation during the review of the SHINE Operating License (OL) application. However, the staff notes that final seismic analysis results for the FSTR are not included in FSAR Section 3.4.2.4 or any other location in the FSAR.

Therefore, update the SHINE FSAR by including final results from the seismic analysis of the FSTR, such as element forces and moments, nodal accelerations, seismic soil pressures, in-structure response spectra, and any other response quantities as appropriate, for representative structural elements and at key equipment locations.

SHINE Response The results from the seismic analysis of the FSTR are summarized in Subsections 3.4.2.4, 3.4.2.5, and 3.4.2.6 of the FSAR. The seismic loads are applied to the structural analysis model as described in Subsection 3.4.2.6 of the FSAR and utilized to develop in-structure response spectra of the facility for use in sizing equipment and components. Response spectra accelerations are output from SASSI2010 at the 75 standard frequencies between 0.2 Hz and 34 Hz, consistent with Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components (Reference 8). In addition, response spectra accelerations are specified to be output at frequencies of 37 Hz, 40 Hz, 43 Hz, 46 Hz and 50Hz. These spectra are generated at six critical damping ratios (2 percent, 3 percent, 4 percent, 5 percent, 7 percent, and 10 percent).

To account for the responses of the structure subjected to the three directional excitations, the co-directional responses (i.e., X response due to X input motion, X response due to Y input motion, X response due to Z input motion) at each node are combined consistent with the square root of the sum of the squares (SRSS) method described in Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis (Reference 9). The SRSSd responses at various nodes for the LB, BE, UB, and cracked seismic analysis cases are enveloped to determine the bounding responses for selected locations and structural elements. Figures 3.4-4-1 through 3.4-4-9 show the selected nodal locations within the analysis model at which response spectra are generated. The output response spectra are grouped and enveloped to determine bounding spectra for the 39 structural elements/locations specified in Table 3.4-4-1. Peaks of the bounding acceleration response spectra are subsequently widened by 15 percent to account for uncertainties in parameters used in the structural analysis and approximations in the modeling techniques.

ISRS plots for these groups are provided in Figures 3.4-4-10 through 3.4-4-48. Peak accelerations (zero period acceleration [ZPA]) for these 39 groups are provided in Table 3.4-4-2. The peak acceleration for each group is taken from the maximum frequency point in the corresponding ISRS.

The element forces and moments due to each direction of input motion (X, Y, and Z) are combined using the 100-40-40 rule in the design of the structure. This method results in 24 permutations of the seismic responses. Select force and moment contour plots are provided to illustrate the distribution of seismic forces and moments. Seismic in-plane shear (F12) and vertical axial (F22) shell element force contour plots for the south exterior wall subjected to the 40X-100Y-40Z combination are provided in Figures 3.4-4-49 and Figure 3.4-4-50. Seismic

Page 29 of 130 in-plane shear (F12) and vertical axial (F22) shell element force contour plots for the west wall and G-line wall subjected to the 100X-40Y-40Z combination are provided in Figures 3.4-4-51 through 3.4-4-54. Seismic out-of-plane moment (M11 and M22) shell element contour plots for each at-grade slab and roof slab subjected to the 40X-40Y-100Z combination are provided in Figures 3.4-4-55 through 3.4-4-58. SASSI2010 has no graphical user interface in which force and moment distributions/contours can be shown. Therefore, the plots are generated from the SAP2000 equivalent static analysis performed using the nodal accelerations output from SASSI2010. The contour plots are only intended to be representative of the general distribution of the seismic forces and moments within the structure.

The seismic loads are applied to the structural analysis model as described in Subsection 3.4.2.6 of the FSAR. Earthquake load is applied in a SAP2000 model on an equivalent static basis to evaluate wall in-plane shear, wall overturning, and building stability.

The equivalent static model represents the soil as dynamic springs, developed in accordance with American Society of Civil Engineers (ASCE) Standard ASCE 4-98, Seismic Analysis of Safety-Related Nuclear Structures and Commentary (Reference 10). Maximum seismic accelerations at each node of the structure necessary for determination of the equivalent static nodal forces are output by the SASSI2010 SSI analysis. Axial load, out-of-plane shear, and out-of-plane moment design for the concrete walls and slabs are performed on an element basis using the seismic forces and moments output by the SASSI2010 SSI seismic analysis. The seismic forces and moments due to each direction of input motion (X, Y, and Z) are combined using the 100-40-40 method. The SAP2000 and SASSI2010 models are both three-dimensional models that represent the structural elements with equivalent mass and stiffness properties. The lumped masses at each node of the SAP2000 analysis are multiplied by the peak accelerations determined from the SSI analysis to determine equivalent static earthquake loads at each node.

The SASSI2010 SSI analysis produces three directions of acceleration response due to each direction of input motion (i.e., x-direction, y-direction, and z-direction accelerations due to x-direction input motion) resulting in nine equivalent static seismic force terms at each node.

The design forces and moments output from the seismic analysis model include the effect of the seismic soil pressure due to the interaction between the modeled excavated soil volume and the structure. To ensure the maximum potential soil pressure is designed for, bounding soil pressures are calculated and applied to the subgrade walls in the SAP2000 design model where applicable. Sub-grade walls of the FSTR are designed to resist static lateral earth pressure loads, compaction loads, static earth pressure, dynamic surcharge loads, and elastic dynamic soil pressure loads. Static earth pressure consists of at-rest, active, and passive soil pressure loads, which are applied as required to ensure the stability of the building. Conservatively, this applied pressure is in addition to any forces/moments due to soil pressures resulting from the seismic analysis model. The applied pressures are the envelope of the static at-rest pressures, dynamic at-rest pressures, static active pressures, dynamic active pressures, and the stabilizing passive pressures.

Subsection 3.4.2.4 of the FSAR has been revised to state that the results of the seismic analysis demonstrate that the design of the FSTR meets the seismic requirements of SHINE Design Criterion 2. A mark-up of the FSAR incorporating these changes is provided as.

Page 30 of 130 Table 3.4-4 Response Spectra Group Identification and Corresponding Figures (Sheet 1 of 2)

Group No.

Description Figure No.

1 Below Grade Slabs - Elevations -23 ft to -12 ft - Horizontal 3.4-4-10 2

Below Grade Slabs - Elevations -23 ft to -12 ft - Vertical 3.4-4-11 3

Grade Slab - Elevation 0 ft - Horizontal 3.4-4-12 4

Grade Slab - Elevation 0 ft - Vertical 3.4-4-13 5

Supercell Location - Elevation 0 ft - Horizontal 3.4-4-14 6

Supercell Location - Elevation 0 ft - Vertical 3.4-4-15 7

TPS Room Ceiling Slab - Elevation 19 ft - Horizontal 3.4-4-16 8

TPS Room Ceiling Slab - Elevation 19 ft - Vertical 3.4-4-17 9

Mezzanine Slab - Elevation 22 ft - Horizontal 3.4-4-18 10 Mezzanine Slab - Elevation 22 ft - Vertical 3.4-4-19 11 West Lower Roof Slab - Elevation 22 ft - Horizontal 3.4-4-20 12 West Lower Roof Slab - Elevation 22 ft - Vertical 3.4-4-21 13 Main Roof Slab - Elevation 44.75 ft to 56 ft - Horizontal 3.4-4-22 14 Main Roof Slab - Elevation 44.75 ft to 56 ft - Vertical 3.4-4-23 15 Below Grade Production Area Walls - Elevation -16 ft to 0 ft

- In-Plane (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 3.4-4-24 16 Below Grade Production Area Walls - Elevation -16 ft to 0 ft

- Out-of-Plane (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 3.4-4-25 17 Below Grade Production Area Walls - Elevation -16 ft to 0 ft

- Vertical (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 3.4-4-26 18 Below Grade Production Area Walls - Elevation -23 ft to 0 ft

- In-Plane (Carbon Delay Beds and RDS Sumps) 3.4-4-27 19 Below Grade Production Area Walls - Elevation -23 ft to 0 ft

- Out-of-Plane (Carbon Delay Beds and RDS Sumps) 3.4-4-28 20 Below Grade Production Area Walls - Elevation -23 ft to 0 ft

- Vertical (Carbon Delay Beds and RDS Sumps) 3.4-4-29 21 IU Cell Walls - Elevation -16 ft to 19 ft - In-Plane 3.4-4-30 22 IU Cell Walls - Elevation -16 ft to 19 ft - Out-of-Plane 3.4-4-31 23 IU Cell Walls - Elevation -16 ft to 19 ft - Vertical 3.4-4-32 24 Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - In-Plane 3.4-4-33 25 Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - Out-of-Plane 3.4-4-34 26 Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - Vertical 3.4-4-35 27 Interior Walls - Elevation 0 ft to 22 ft - In-Plane (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 3.4-4-36

Page 31 of 130 Table 3.4-4 Response Spectra Group Identification and Corresponding Figures (Sheet 2 of 2)

Group No.

Description Figure No.

28 Interior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 3.4-4-37 29 Interior Walls - Elevation 0 ft to 22 ft - Vertical (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 3.4-4-38 30 Exterior Walls - Elevation 0 ft to 22 ft - In-Plane 3.4-4-39 31 Exterior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (Excluding North and South Walls Between Column Lines G and K) 3.4-4-40 32 Exterior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (North and South Walls Between Column Lines G and K Only) 3.4-4-41 33 Exterior Walls - Elevation 0 ft to 22 ft - Vertical 3.4-4-42 34 All Walls - Elevation 22 ft to 56 ft - In-Plane 3.4-4-43 35 All Walls - Elevation 22 ft to 56 ft - Out-of-Plane 3.4-4-44 36 All Walls - Elevation 22 ft to 56 ft - Vertical 3.4-4-45 37 Crane Supports - Elevations 26 ft and 36 ft - NS Direction 3.4-4-46 38 Crane Supports - Elevations 26 ft and 36 ft - EW Direction 3.4-4-47 39 Crane Supports - Elevations 26 ft and 36 ft - Vertical 3.4-4-48

Page 32 of 130 Table 3.4-4 Peak Acceleration (ZPA) for Each Group (Sheet 1 of 2)

Group No.

Description Peak Acceleration (ZPA) (g) 1 Below Grade Slabs - Elevations -23 ft to -12 ft - Horizontal 0.2831 2

Below Grade Slabs - Elevations -23 ft to -12 ft - Vertical 0.3321 3

Grade Slab - Elevation 0 ft - Horizontal 0.3530 4

Grade Slab - Elevation 0 ft - Vertical 0.3892 5

Supercell Location - Elevation 0 ft - Horizontal 0.2850 6

Supercell Location - Elevation 0 ft - Vertical 0.3429 7

TPS Room Ceiling Slab - Elevation 19 ft - Horizontal 0.3857 8

TPS Room Ceiling Slab - Elevation 19 ft - Vertical 0.3177 9

Mezzanine Slab - Elevation 22 ft - Horizontal 0.5750 10 Mezzanine Slab - Elevation 22 ft - Vertical 0.8578 11 West Lower Roof Slab - Elevation 22 ft - Horizontal 0.4913 12 West Lower Roof Slab - Elevation 22 ft - Vertical 1.0160 13 Main Roof Slab - Elevation 44.75 ft to 56 ft - Horizontal 0.4673 14 Main Roof Slab - Elevation 44.75 ft to 56 ft - Vertical 1.0698 15 Below Grade Production Area Walls - Elevation -16 ft to 0 ft

- In-Plane (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 0.3291 16 Below Grade Production Area Walls - Elevation -16 ft to 0 ft

- Out-of-Plane (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 0.7476 17 Below Grade Production Area Walls - Elevation -16 ft to 0 ft

- Vertical (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 0.3567 18 Below Grade Walls - Elevation -23 ft to 0 ft - In-Plane (Carbon Delay Beds and RDS Sumps) 0.3062 19 Below Grade Walls - Elevation -23 ft to 0 ft - Out-of-Plane (Carbon Delay Beds and RDS Sumps) 0.3958 20 Below Grade Walls - Elevation -23 ft to 0 ft - Vertical (Carbon Delay Beds and RDS Sumps) 0.2918 21 IU Cell Walls - Elevation -16 ft to 19 ft - In-Plane 0.3533 22 IU Cell Walls - Elevation -16 ft to 19 ft - Out-of-Plane 0.3711 23 IU Cell Walls - Elevation -16 ft to 19 ft - Vertical 0.3028 24 Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - In-Plane 0.5483 25 Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - Out-of-Plane 0.5247 26 Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - Vertical 0.3402 27 Interior Walls - Elevation 0 ft to 22 ft - In-Plane (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 0.3486

Page 33 of 130 Table 3.4-4 Peak Acceleration (ZPA) for Each Group (Sheet 2 of 2)

Group No.

Description Peak Acceleration (ZPA) (g) 28 Interior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 0.4848 29 Interior Walls - Elevation 0 ft to 22 ft - Vertical (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 0.3605 30 Exterior Walls - Elevation 0 ft to 22 ft - In-Plane 0.3927 31 Exterior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (Excluding North and South Walls Between Column Lines G and K) 0.8440 32 Exterior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (North and South Walls Between Column Lines G and K Only) 1.2564 33 Exterior Walls - Elevation 0 ft to 22 ft - Vertical 0.3892 34 All Walls - Elevation 22 ft to 56 ft - In-Plane 0.3922 35 All Walls - Elevation 22 ft to 56 ft - Out-of-Plane 1.3846 36 All Walls - Elevation 22 ft to 56 ft - Vertical 0.3978 37 Crane Supports - Elevations 26 ft and 36 ft - NS Direction 0.3362 38 Crane Supports - Elevations 26 ft and 36 ft - EW Direction 1.0841 39 Crane Supports - Elevations 26 ft and 36 ft - Vertical 0.3720 Note: The reported peak accelerations are the lowest acceleration value at 57.5 Hz from the six in-structure response spectra (various dampings) for each group. These values may be slightly higher than the actual nodal peak accelerations since the in-structure response spectra output by SASSI asymptotically approach the nodal peak acceleration as frequency increases.

Page 34 of 130 Figure 3.4-4 Selected Response Spectra Locations, Exterior (Looking South-East)

Page 35 of 130 Figure 3.4-4 Selected Response Spectra Locations, Exterior (Looking North-West)

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 36 of 130 Figure 3.4-4 Selected Response Spectra Locations, At Grade Slab

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 37 of 130 Figure 3.4-4 Selected Response Spectra Locations, Below Grade Slab

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 38 of 130 Figure 3.4-4 Selected Response Spectra Locations, Interior Walls - West of Column Line G

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 39 of 130 Figure 3.4-4 Selected Response Spectra Locations, Interior Walls - Carbon Delay Bed and RDS Sumps

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 40 of 130 Figure 3.4-4 Selected Response Spectra Locations, Interior Walls and Miscellaneous Slabs - East of Column Line G

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 41 of 130 Figure 3.4-4 Selected Response Spectra Locations, Interior Walls - East of Column Line G

Security-Related Information - Withheld from public disclosure under 10 CFR 2.390(d)

Page 42 of 130 Figure 3.4-4 Selected Response Spectra Locations, Crane Support Locations

Page 43 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Slabs - Elevations -23 ft to -12 ft - Horizontal 0.0 2.0 4.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group1,2%Damping Group1,3%Damping Group1,4%Damping Group1,5%Damping Group1,7%Damping Group1,10%Damping

Page 44 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Slabs - Elevations -23 ft to -12 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group2,2%Damping Group2,3%Damping Group2,4%Damping Group2,5%Damping Group2,7%Damping Group2,10%Damping

Page 45 of 130 Figure 3.4-4 Enveloped Response Spectra - Grade Slab - Elevation 0 ft - Horizontal 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group3,2%Damping Group3,3%Damping Group3,4%Damping Group3,5%Damping Group3,7%Damping Group3,10%Damping

Page 46 of 130 Figure 3.4-4 Enveloped Response Spectra - Grade Slab - Elevation 0 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group4,2%Damping Group4,3%Damping Group4,4%Damping Group4,5%Damping Group4,7%Damping Group4,10%Damping

Page 47 of 130 Figure 3.4-4 Enveloped Response Spectra - Supercell Location - Elevation 0 ft - Horizontal 0.0 2.0 4.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group5,2%Damping Group5,3%Damping Group5,4%Damping Group5,5%Damping Group5,7%Damping Group5,10%Damping

Page 48 of 130 Figure 3.4-4 Enveloped Response Spectra - Supercell Location - Elevation 0 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group6,2%Damping Group6,3%Damping Group6,4%Damping Group6,5%Damping Group6,7%Damping Group6,10%Damping

Page 49 of 130 Figure 3.4-4 Enveloped Response Spectra - TPS Room Ceiling Slab - Elevation 19 ft - Horizontal 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group7,2%Damping Group7,3%Damping Group7,4%Damping Group7,5%Damping Group7,7%Damping Group7,10%Damping

Page 50 of 130 Figure 3.4-4 Enveloped Response Spectra - TPS Room Ceiling Slab - Elevation 19 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group8,2%Damping Group8,3%Damping Group8,4%Damping Group8,5%Damping Group8,7%Damping Group8,10%Damping

Page 51 of 130 Figure 3.4-4 Enveloped Response Spectra - Mezzanine Slab - Elevation 22 ft - Horizontal 0.0 2.0 4.0 6.0 8.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group9,2%Damping Group9,3%Damping Group9,4%Damping Group9,5%Damping Group9,7%Damping Group9,10%Damping

Page 52 of 130 Figure 3.4-4 Enveloped Response Spectra - Mezzanine Slab - Elevation 22 ft - Vertical 0.0 2.0 4.0 6.0 8.0 10.0 12.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group10,2%Damping Group10,3%Damping Group10,4%Damping Group10,5%Damping Group10,7%Damping Group10,10%Damping

Page 53 of 130 Figure 3.4-4 Enveloped Response Spectra - West Lower Roof Slab - Elevation 22 ft - Horizontal 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group11,2%Damping Group11,3%Damping Group11,4%Damping Group11,5%Damping Group11,7%Damping Group11,10%Damping

Page 54 of 130 Figure 3.4-4 Enveloped Response Spectra - West Lower Roof Slab - Elevation 22 ft - Vertical 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group12,2%Damping Group12,3%Damping Group12,4%Damping Group12,5%Damping Group12,7%Damping Group12,10%Damping

Page 55 of 130 Figure 3.4-4 Enveloped Response Spectra - Main Roof Slab - Elevation 44.75 ft to 56 ft - Horizontal 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group13,2%Damping Group13,3%Damping Group13,4%Damping Group13,5%Damping Group13,7%Damping Group13,10%Damping

Page 56 of 130 Figure 3.4-4 Enveloped Response Spectra - Main Roof Slab - Elevation 44.75 ft to 56 ft - Vertical 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group14,2%Damping Group14,3%Damping Group14,4%Damping Group14,5%Damping Group14,7%Damping Group14,10%Damping

Page 57 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Production Area Walls - Elevation -16 ft to 0 ft - In-Plane (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 0.0 2.0 4.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group15,2%Damping Group15,3%Damping Group15,4%Damping Group15,5%Damping Group15,7%Damping Group15,10%Damping

Page 58 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Production Area Walls - Elevation -16 ft to 0 ft - Out-of-Plane (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 0.0 2.0 4.0 6.0 8.0 10.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group16,2%Damping Group16,3%Damping Group16,4%Damping Group16,5%Damping Group16,7%Damping Group16,10%Damping

Page 59 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Production Area Walls - Elevation -16 ft to 0 ft - Vertical (Hold Tanks, Valve Pits, Storage Tanks, Waste Tanks, Gamma Gates) 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group17,2%Damping Group17,3%Damping Group17,4%Damping Group17,5%Damping Group17,7%Damping Group17,10%Damping

Page 60 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Production Area Walls - Elevation -23 ft to 0 ft - In-Plane (Carbon Delay Beds and RDS Sumps) 0.0 2.0 4.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group18,2%Damping Group18,3%Damping Group18,4%Damping Group18,5%Damping Group18,7%Damping Group18,10%Damping

Page 61 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Production Area Walls - Elevation -23 ft to 0 ft - Out-of-Plane (Carbon Delay Beds and RDS Sumps) 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group19,2%Damping Group19,3%Damping Group19,4%Damping Group19,5%Damping Group19,7%Damping Group19,10%Damping

Page 62 of 130 Figure 3.4-4 Enveloped Response Spectra - Below Grade Production Area Walls - Elevation -23 ft to 0 ft - Vertical (Carbon Delay Beds and RDS Sumps) 0.0 2.0 4.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group20,2%Damping Group20,3%Damping Group20,4%Damping Group20,5%Damping Group20,7%Damping Group20,10%Damping

Page 63 of 130 Figure 3.4-4 Enveloped Response Spectra - IU Cell Walls - Elevation -16 ft to 19 ft - In-Plane 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group21,2%Damping Group21,3%Damping Group21,4%Damping Group21,5%Damping Group21,7%Damping Group21,10%Damping

Page 64 of 130 Figure 3.4-4 Enveloped Response Spectra - IU Cell Walls - Elevation -16 ft to 19 ft - Out-of-Plane 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group22,2%Damping Group22,3%Damping Group22,4%Damping Group22,5%Damping Group22,7%Damping Group22,10%Damping

Page 65 of 130 Figure 3.4-4 Enveloped Response Spectra - IU Cell Walls - Elevation -16 ft to 19 ft - Vertical 0.0 2.0 4.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group23,2%Damping Group23,3%Damping Group23,4%Damping Group23,5%Damping Group23,7%Damping Group23,10%Damping

Page 66 of 130 Figure 3.4-4 Enveloped Response Spectra - Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - In-Plane 0.0 2.0 4.0 6.0 8.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group24,2%Damping Group24,3%Damping Group24,4%Damping Group24,5%Damping Group24,7%Damping Group24,10%Damping

Page 67 of 130 Figure 3.4-4 Enveloped Response Spectra - Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - Out-of-Plane 0.0 2.0 4.0 6.0 8.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group25,2%Damping Group25,3%Damping Group25,4%Damping Group25,5%Damping Group25,7%Damping Group25,10%Damping

Page 68 of 130 Figure 3.4-4 Enveloped Response Spectra - Off Gas Cells and Cooling Room Walls - Elevation 0 ft to 19 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group26,2%Damping Group26,3%Damping Group26,4%Damping Group26,5%Damping Group26,7%Damping Group26,10%Damping

Page 69 of 130 Figure 3.4-4 Enveloped Response Spectra - Interior Walls - Elevation 0 ft to 22 ft - In-Plane (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group27,2%Damping Group27,3%Damping Group27,4%Damping Group27,5%Damping Group27,7%Damping Group27,10%Damping

Page 70 of 130 Figure 3.4-4 Enveloped Response Spectra - Interior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group28,2%Damping Group28,3%Damping Group28,4%Damping Group28,5%Damping Group28,7%Damping Group28,10%Damping

Page 71 of 130 Figure 3.4-4 Enveloped Response Spectra - Interior Walls - Elevation 0 ft to 22 ft - Vertical (Excluding Walls of IU Cells, Off Gas Cells, and Cooling Rooms) 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group29,2%Damping Group29,3%Damping Group29,4%Damping Group29,5%Damping Group29,7%Damping Group29,10%Damping

Page 72 of 130 Figure 3.4-4 Enveloped Response Spectra - Exterior Walls - Elevation 0 ft to 22 ft - In-Plane 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group30,2%Damping Group30,3%Damping Group30,4%Damping Group30,5%Damping Group30,7%Damping Group30,10%Damping

Page 73 of 130 Figure 3.4-4 Enveloped Response Spectra - Exterior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (Excluding North and South Walls Between Column Lines G and K) 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group31,2%Damping Group31,3%Damping Group31,4%Damping Group31,5%Damping Group31,7%Damping Group31,10%Damping

Page 74 of 130 Figure 3.4-4 Enveloped Response Spectra - Exterior Walls - Elevation 0 ft to 22 ft - Out-of-Plane (North and South Walls Between Column Lines G and K Only) 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group32,2%Damping Group32,3%Damping Group32,4%Damping Group32,5%Damping Group32,7%Damping Group32,10%Damping

Page 75 of 130 Figure 3.4-4 Enveloped Response Spectra - Exterior Walls - Elevation 0 ft to 22 ft - Vertical 0.0 2.0 4.0 6.0 8.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group33,2%Damping Group33,3%Damping Group33,4%Damping Group33,5%Damping Group33,7%Damping Group33,10%Damping

Page 76 of 130 Figure 3.4-4 Enveloped Response Spectra - All Walls - Elevation 22 ft to 56 ft - In-Plane 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group34,2%Damping Group34,3%Damping Group34,4%Damping Group34,5%Damping Group34,7%Damping Group34,10%Damping

Page 77 of 130 Figure 3.4-4 Enveloped Response Spectra - All Walls - Elevation 22 ft to 56 ft - Out-of-Plane 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 20.0 22.0 24.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group35,2%Damping Group35,3%Damping Group35,4%Damping Group35,5%Damping Group35,7%Damping Group35,10%Damping

Page 78 of 130 Figure 3.4-4 Enveloped Response Spectra - All Walls - Elevation 22 ft to 56 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group36,2%Damping Group36,3%Damping Group36,4%Damping Group36,5%Damping Group36,7%Damping Group36,10%Damping

Page 79 of 130 Figure 3.4-4 Enveloped Response Spectra - Crane Supports - Elevations 26 ft and 36 ft - NS Direction 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group37,2%Damping Group37,3%Damping Group37,4%Damping Group37,5%Damping Group37,7%Damping Group37,10%Damping

Page 80 of 130 Figure 3.4-4 Enveloped Response Spectra - Crane Supports - Elevations 26 ft and 36 ft - EW Direction 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group38,2%Damping Group38,3%Damping Group38,4%Damping Group38,5%Damping Group38,7%Damping Group38,10%Damping

Page 81 of 130 Figure 3.4-4 Enveloped Response Spectra - Crane Supports - Elevations 26 ft and 36 ft - Vertical 0.0 2.0 4.0 6.0 0.1 1.0 10.0 100.0 ACCELERATION - g FREQUENCY - Hz Group39,2%Damping Group39,3%Damping Group39,4%Damping Group39,5%Damping Group39,7%Damping Group39,10%Damping

Page 82 of 130 Figure 3.4-4 South Wall Shell Elements Resultant F12 Diagram

Page 83 of 130 Figure 3.4-4 South Wall Shell Elements Resultant F22 Diagram

Page 84 of 130 Figure 3.4-4 West Wall Shell Elements Resultant F12 Diagram

Page 85 of 130 Figure 3.4-4 West Wall Shell Elements Resultant F22 Diagram

Page 86 of 130 Figure 3.4-4 G-Line Wall Shell Elements Resultant F12 Diagram

Page 87 of 130 Figure 3.4-4 G-Line Wall Shell Elements Resultant F22 Diagram

Page 88 of 130 Figure 3.4-4 At-Grade Slab Shell Elements Resultant M11 Diagram

Page 89 of 130 Figure 3.4-4 At-Grade Slab Shell Elements Resultant M22 Diagram

Page 90 of 130 Figure 3.4-4 Roof Slab Shell Elements Resultant M11 Diagram

Page 91 of 130 Figure 3.4-4 Roof Slab Shell Elements Resultant M22 Diagram

Page 92 of 130 RAI 3.4-5 In FSAR Section 3.4.2.5, Assessment of Structural Seismic Stability, the applicant states that the seismic stability of the SHINE facility is evaluated for sliding and overturning considering the load combinations and factors of safety in accordance with American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) Standard 43-05 and NUREG-0800, SRP Section 3.8.5. However, the staff notes that the applicant did not provide a summary of the results or conclusions of such evaluation, which are needed for the staff to make its safety findings with respect to the stability of the SHINE facility structures during the design-basis earthquake.

Therefore, provide in the FSAR a summary of the results or conclusions from the applicants seismic stability assessment of the SHINE facility structures.

SHINE Response The main production facility is evaluated for stability as described in Subsection 3.4.2.5 of the FSAR and is considered stable. The computed factors of safety against sliding and overturning are greater than the minimum values required. Subsection 3.4.2.5 of the FSAR has been revised to include summarized results and conclusions from the stability evaluation. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.4-6 Section 3.4.2.6.3.1, Soil Parameters, of the SHINE operating license application does not include all necessary parameters and does not provide sufficient information regarding the stability of the foundations and subsurface materials for the SHINE facility for NRC staff to confirm the acceptability of the site. The section provided some soil parameters that were used in a soil-structure interaction (SSI) analysis, such as the minimum average shear wave velocity, minimum unit weight, and Poissons ratio. It also provided other parameters, such as net allowable static bearing pressure at 3 feet below grade and net allowable static bearing pressure at 17 feet below grade. However, the soil parameters necessary for the use of an SSI analysis and foundation stability assessment do not include information on how these net allowable static bearing pressures were determined. In addition, there are no details on the safety-related foundation settlements (total and differential settlements) evaluation.

In order for the NRC staff to determine whether SHINE has adequately evaluated the stability of the foundations and subsurface materials for SHINE facility, update the application to provide information on allowable soil bearing capacities at designated elevations and allowable settlements (i.e., total and differential settlements) for the specific designed structures, and a comparison of maximum structural foundation responses with soil/foundation capacities (e.g. maximum foundation pressure vs. allowable soil bearing capacity, maximum foundation settlements vs allowable settlements).

SHINE Response The evaluation of the FSTR foundation and subgrade materials beneath the foundation utilizes a single allowable soil bearing pressure based on the average depth of foundation for the FSTR considering two cases - shallow foundations at 3 ft. below grade and deep foundations at 17 ft.

below grade. The allowable soil bearing pressure for the building foundation is at least 6000 pounds per square foot (psf) with a factor of safety of 3. The method used to

Page 93 of 130 determine the allowable soil bearing capacity is consistent with Chapter 5 of Federal Highway Administration Report No. FHWA-SA-02-054, Geotechnical Engineering Circular No. 6 - Shallow Foundations (Reference 11). This allowable soil bearing pressure is higher than the foundation contact pressures (net allowable static bearing pressures) provided in Subsection 3.4.2.6.3.1 of the FSAR.

Differential settlement of the FSTR is implicitly evaluated within the structural finite analysis model. The structural model is developed with soil springs supporting the foundation. The stiffness of the soil springs, which represents the stiffness of the soil, is developed based on soil subgrade moduli of the soil beneath the two foundation types (i.e., 3 ft. feet below grade and 17 ft. below grade). Deflections of the FSTR foundation (settlements), which are determined in the structural analysis model, produce moments and shears in the foundation based on and accounting for the soil stiffness. Therefore, an acceptable design of the FSTR based on the structural analysis confirms settlements (differential and total) are not beyond maximum allowable values. The maximum settlements for normal loading determined by the structural analysis (approximately 0.9 in. and 0.65 in. at 3 ft. below grade and 17 ft. below grade, respectively) are similar to those estimated to determine the subgrade modulus (approximately 0.9 in. and 0.35 in. at 3 ft. below grade and 17 ft. below grade, respectively) and therefore confirm the subgrade moduli that were used in the FSTR analysis are acceptable.

RAI 3.4-7 Section 3.4, Seismic Damage, of NUREG-1537 Parts 1 and 2, as well as the ISG Augmenting NUREG-1537, state that seismic design for non-power reactors should, at a minimum, be consistent with local building codes and other applicable standards to provide assurance that significant damage to the facility and associated safety functions is unlikely.

Section 2.5.5.3, 2015 International Building Code Seismic Design Ground Motion Parameters, of the SHINE operating license application (OLA) references the International Building Code (IBC) of the International Code Council (ICC) and the American Society of Civil Engineers/Structural Engineering Institute (ASCE/SEI) 7, Minimum Design Loads for Buildings and Other Structures. The local building code Wisconsin Administrative Code for Safety and Professional Services (SPS), Chapter 362, Building and Structures, also references ASCE/SEI 7 and IBC for the design of building and structures.

Section 3.4 of the SHINE OLA, however, supplements the requirements listed in the above local and national codes and standards for the design of the FSTR and its safety-related SSCs with IAEA-TECDOC-1347, Consideration of External Events [EE] in the Design of Nuclear Facilities other than Nuclear Power Plants, with Emphasis on Earthquakes, for seismic analysis criteria

[and] generic requirements and guidance for the seismic design of nuclear facilities other than nuclear power plants. Section 3.4 also references additional national codes and standards, as well as regulatory guides (RGs) and nuclear regulatory reports (NUREGs) used in the structural design of the FSTR and its SSCs.

The staff reviewed IAEA-TECDOC-1347, but was not clear whether the applicant has used its generic requirements to supplement or replace requirements imposed by IBC and/or ASCE/SEI 7-05 referenced in SPS 362 or other pertinent local and national building codes for the overall design, including seismic design, of FSTR and its SSCs. It also was not clear to the NRC staff to what extent the guidance of IAEA-TECDOC-1347 was used in lieu of that contained in the referenced NRC RGs and NUREGs.

Page 94 of 130 Clarify to what extent IAEA-TECDOC-1347 has been used in the analysis and design, including seismic design, of the FSTR and its SSCs. If it was used in lieu of the SHINE OLA referenced RGs and NUREGs or to supplement requirements delineated in local building codes, such as SPS 362 or other local and national codes standards, state where. Update the FSAR, as appropriate.

SHINE Response International Atomic Energy Agency (IAEA) IAEA-TECDOC-1347, Consideration of External Events in the Design of Nuclear Facilities other than Nuclear Power Plants, with Emphasis on Earthquakes (Reference 12), is not used to replace, supplement, or take precedence over requirements or guidance provided by the referenced Regulatory Guides, NUREGs, local building codes, or national codes and standards. IAEA-TECDOC-1347 is not referenced in either the seismic or structural analyses of the facility. Sections 3.4 and 3.6 of the FSAR have been revised to remove the reference to IAEA-TECDOC-1347. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.4-8 Section 3.4, Seismic Damage, of the SHINE OLA states that the FSTR includes the irradiation facility (IF), the radioisotope production facility (RPF), the non-radiologically controlled seismic area, and a non-safety-related area. The section succinctly describes the FSTR to be built as a reinforced concrete box shear wall system on soil, with a mezzanine floor, and a roof slab supported by steel trusses. The FSTR design includes SSCs such as a tall exhaust stack, below grade reinforced concrete vaults, tanks, and supercell(s). Additional details for FSTR SSCs can be found, for example, in:

Section 1.2.1, Consequences from the Operation and Use of the Facility, for internal structures including supercells.

Section 2.1.1.2, Boundary and Zone Area Maps, for the free-standing exhaust stack to discharge filtered air (e.g., see Chapters 4, 6, 7, 9, 11, 13) to the atmosphere.

Section 3.4.2.6.4.1, Dead Load, for precast tanks in radioisotope production facility (RPF) vaults.

Section 4b.2.2.2, Geometry and Configuration, for pits, trenches, and cover plugs.

Tables 7.7-2 and 7.7-3 of the SHINE OLA, Radiation Area Monitor Locations and Continuous Airborne Monitor Locations, respectively, for the facility mezzanine safety-related area.

Section 3.4, Seismic Damage, of the SHINE OLA also references IAEA-TECDOC-1347, Consideration of External Events in the Design of Nuclear Facilities other than Nuclear Power Plants, with Emphasis on Earthquakes. Chapter 6, Building Design, of IAEA-TECDOC-1347 provides seismic analysis criteria [and] generic requirements and guidance for the seismic design. In part, it states:

Inverted pendulum structures are not allowed for EEC1 [External Event Class 1 safety structures which during and after an external event interact with other safety related

Page 95 of 130 SSCs] in the structures [] Precast panels (or other prefabricated elements) need to be connected in such a way that they behave as an integral unit during an earthquake.

Some non-structural elements may affect the dynamic behavior of the structure and its capacity.

It is typically the case of masonry filling in framed reinforced concrete structures which can lead to shear damage (and rupture) of the so called short column configuration. In the case of design of new buildings, this solution needs to be avoided.

The TECDOC also states, in part, that the probability of an EE to generate a radiological consequence depends on characteristics of the facility and the EE, particularly for facilities subjected to frequent configuration and layout changes (such as activities associated to new product developments).

Section 3.4 of the SHINE OLA provides limited information to assess the adequacy of structural design of the FSTR and its structural safety significant SSCs (or non-safety SSCs that could affect those that are safety related) so that a reasonable assurance for safety determination can be made. It does not state how these structural SSCs are integrated in the FSTR seismic design to provide a defense-in-depth against radiological release and to provide reasonable assurance that significant damage to the facility and associated safety functions is unlikely.

As noted in IAEA-TECDOC-1347, the probability of the FSTR to generate a radiological consequence depends on characteristics of the facility and the EE, particularly for facilities subjected to frequent configuration and layout changes (such as activities associated to new product developments). It is not clear how structural design changes at the FSTR and its SSCs during plant operation would be assessed for regulatory compliance.

The SHINE OLA does not state what specific construction materials (e.g., ASTM designations, their yield/compressive/tensile strengths, etc.) have been used in the current design configuration, including seismic design, for the construction of the FSTR and its safety-related SSCs and how future configuration changes will be controlled. It is not clear whether adequate safety margins were introduced to accommodate structural alterations/configuration changes during facility operation so that defense-in-depth will be maintained for potential new configurations.

Additionally, it is not clear whether, in the current design configuration, the exhaust stack is designed as an isolated self-supporting cantilever structure or one framed/anchored in part or in whole into the FSTR. It is also not clear what specific materials have been selected for its construction. It is not clear whether the stack is designed as a cast in place/precast concrete structure, a steel framed cantilever structure, or a composite structure to resist seismic forces.

In addition, it is not clear whether and how the identified precast tanks, supercells, cover plugs are integrated in the current design configuration of the FSTR to sustain abnormal loads (seismic, aircraft impact, blast effects). Furthermore, it is not clear to what extent masonry structures have been used in the design of the SHINE facility and if so, whether they are safety related and further integrated within the FSTR seismic design. Similar arguments are made for the FSTR safety related area of mezzanine floor, referenced in Table 7.7-3, of the SHINE OLA, and its ability as a diaphragm to carry loads and distribute seismic forces.

(1) Clarify how future configuration/layout changes, if any, to the FSTR and (safety/non-safety affecting safety) SSCs noted above for activities associated with process/product developments will be controlled.

Page 96 of 130 (2) Provide a complete description of the current FSTR and aforementioned SSCs configuration that includes descriptions and locations of the exhaust stack, precast tanks, supercells, cover plugs, and masonry structures. Include in the description information such as the dimensions of major safety related structural components and structural materials used in their design/construction and how they were integrated in the overall FSTR design to resist seismic, aircraft impact, and blast loads.

(3) For non-safety structural components that could affect a safety function that are not integrated in the overall structural design of the FSTR, describe their capacity to resist seismic, aircraft impact, and blast loads without undue risk to health and safety of the public and damage to the environment.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response (1) The SHINE configuration management program, as described in Section 5.5.4 of the technical specifications, is used to evaluate each change to the SHINE facility for the potential to affect safety-related SSCs. Future configuration and layout changes to the FSTR, other safety-related SSCs, and nonsafety-related SSCs that may affect safety are evaluated under the SHINE configuration management program, which includes an evaluation of changes to the SHINE facility in accordance with the requirements of 10 CFR 50.59.

(2) Additional information on the requested SSCs, as it relates to seismic damage, is provided below.

Exhaust Stack: The exhaust stack is a nonsafety-related SSC designed as an isolated self-supporting steel cantilever structure that is mounted to a reinforced concrete foundation. The exhaust stack is located east of the FSTR and connected to the nonsafety-related area by ductwork which is designed to allow adequate movement to not impart any loads on the FSTR.

Tanks on the RPF below-grade slab: The RPF includes a series of below grade tanks and pipe trenches, most of which are prefabricated metal tanks separated by conventional reinforced concrete walls on all sides. Conventional reinforced concrete walls that separate the RPF tanks are explicitly modeled. The reinforced concrete walls that separate the RPF tanks support removable precast concrete shield plugs (see shield plugs discussion below). A precast concrete base with an embedded stainless steel drip pan is located at the bottom of each tank vault. Previously designed precast tank vaults were changed to conventional concrete walls to improve constructability.

Supercell: The Supercell is explicitly modeled in the design model. The Supercell has approximate external dimensions of 136 feet long by 21 feet wide by 26 feet high. A series of interconnected frames made of massless rigid links are used to model the stiffness of the Supercell shielding walls. The total weight of the Supercell is calculated and converted to a line load (over the length of the Supercell) that is applied to the rigid link structure at the horizontal and vertical centers of gravity.

Page 97 of 130 Shield Plugs: Some interior rooms within the IF (irradiation unit cells, cooling rooms, and off-gas rooms) and RPF (tanks, pits, carbon delay bed, and trenches) have concrete cover blocks (or shield plugs). Shield plugs made from one or more prefabricated concrete pieces and, depending on the installed location, are approximately 4 to 6 feet thick when fully assembled. Since the reinforced concrete walls that separate the rooms or tanks are explicitly considered, the weight of the cover blocks is distributed amongst the perimeter joints at the top of the walls. Dead loads associated with cover blocks for tank basins and trenches are calculated using applicable dimensions and the density of concrete. Live loads are considered to be the maximum of any applied external operating loads (i.e., forklift) or the weight associated with the placement of other cover blocks on top of adjacent cover blocks.

Interior Partition Walls: There are a number of walls which are not part of the seismic force resisting system in either the vertical or horizontal direction and exist to create separation or partitions between various rooms within the facility. These partitions include walls that are designed as partially grouted 8 inch-thick reinforced concrete masonry units (CMUs). The weights of these CMU walls were included in the design model. The masonry walls are not positioned to restrain adjacent reinforced concrete walls in a manner that would cause seismic damage.

(3) The safety-related portions of the FSTR are seismically isolated from the nonsafety-related portions except for a small section of the nonsafety-related portion of the FSTR which is built on the southwest corner of the safety-related basemat. The mass and stiffness of this nonsafety-related portion of the FSTR is considered in the design of the safety-related basemat. A seismic gap has been sized between the nonsafety-related and safety-related portions of the FSTR structure, such that no seismic interaction will occur during their respective design basis seismic events.

Nonsafety-related structures that are co-located with a Seismic Category I SSC are designated as Seismic Category II, as described in Subsection 3.4.3.1 of the FSAR, and must maintain structural integrity in the event of a SSE to prevent unacceptable interactions with a Seismic Category I SSC.

The walls and slabs that envelope the safety-related portion of the FSTR are designed for aircraft impact as described in Subsection 3.4.5.1 of the FSAR. The nonsafety-related portions of the FSTR are not designed for aircraft impact because they are not credited with protection of any safety-related SSCs.

Similarly, the effects of blast on the walls and slabs that envelope the safety-related portion of the FSTR are addressed in Subsection 3.4.5.2 of the FSAR. The nonsafety-related portions of the FSTR are not designed for any blast effects because they are not credited with protection of any safety-related SSCs.

Section 3.4 of the FSAR has been updated to more clearly describe the structural systems of the FSTR and how they relate to one another. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 3.4-9 Section 3.4.2.1, Seismic Analysis Methods, of the SHINE OLA states that the finite element analysis (FEA) methodology was used as part of the facility seismic analysis. It also states that

Page 98 of 130 the finite element model consists of plate/shell, solid, beam, or a combination of finite elements. Section 3.4.2.2, Soil-Structure Interaction (SSI) Analysis, of the SHINE OLA, however, states that the FEA model uses thick shell elements to represent concrete slabs and walls, and beam elements to represent steel members, mostly comprising the truss components in the facility. Section 3.3.1.1.2, Flood Protection from Internal Sources, of the SHINE OLA states that the light water pool is approximately 4 feet thick.

The SRP acceptance criteria to NUREG-0800, Section 3.7.2, Seismic System Analysis, referenced by the applicant in section 3.4.2.1 as guidance to FEA modeling, identify finite element modeling as a method acceptable to the NRC for seismic analysis, and states:

The type of finite element used for modeling a structural system should depend on the structural details, the purpose of the analysis, and the theoretical formulation upon which the element is based. The mathematical discretization of the structure should consider the effect of element size, shape, and aspect ratio on solution accuracy. The element mesh size should be selected on the basis that further refinement has only a negligible effect on the solution results [] In general, three-dimensional models should be used for seismic analyses.

However, simpler models can be used if justification can be provided that the coupling effects of those degrees of freedom that are omitted from the three-dimensional models are not significant

[] The effects of concrete cracking on membrane, bending, and shear stiffness should be considered as appropriate in the mathematical model. Because the effect of cracking on the stiffness of concrete members is complex and depends on a number of factors, the approach used should be shown to be conservative.

As noted in NUREG-0800, there are distinct differences in the mathematical formulation of finite elements. Element selection and discretization of the structural domain (modeling of structure) should be made to fit the characteristics of the structure and the loading conditions.

It is not clear where solid elements are used in the finite element analysis model of the soil and of the structure. It is also not clear whether the simpler modeling of concrete structural components (e.g., walls/slabs) with plate/shell elements is adequate for analyses of field effects (e.g., distribution of internal forces, damage, cracking, reduction in overall FSTR structural stiffness) due to seismic, aircraft impact, and blast loads. In addition, it is not clear whether the same model was sufficiently discretized to capture the salient features of applied loads due to seismic, aircraft impact, and blast effects.

(1) Discuss the discretized FEA model and elements used (including solid elements) in the analysis of the FSTR, its connected structures, foundations, and elastic half-space to predict field quantities (forces/stresses, moments) including deformation, cracking, damage, consistent with Section 3.7.2 of NUREG-0800.

(2) Justify where the approach departs from Section 3.7.2 of NUREG-0800, and if so, whether it aligns with broadly accepted engineering practices.

(3) State whether the same FEA model was used for seismic, aircraft impact, and blast effects analyses. If so, justify the FEA model sufficiency to capture the salient effects for each of the applied loads.

Update the FSAR as appropriate to reflect the above clarifications and justifications.

Page 99 of 130 SHINE Response (1) The seismic analysis of the FSTR utilizes the program SASSI2010, System for Analysis of Soil-Structure Interaction. The main components of the SASSI2010 model are the horizontal soil layers, the excavated soil model, and the structural model. Horizontal soil layers are used to model the soil column down to an elastic half space which represents bedrock. Soil layer thickness is determined to ensure transmittal of the highest cut-off frequency of interest in the analysis. The excavated soil is modeled using solid elements. The solid elements are modeled such that the vertical thickness of the elements match the height of the corresponding soil layers. The properties of the soil solid elements match the equivalent linear strain-compatible soil properties determined for the site. The structural model consists of the building structure which is modeled using thick shell elements and beam elements following standard 3D finite element modeling practices.

Figures 3.4-1 and 3.4-2 of the FSAR show the finite element model used for the FSTR.

Discretization in the finite element analysis (FEA) is based on generally accepted engineering practices and practical limitations in the modeling and analysis of a structure of this size. The finite element mesh size and aspect ratio are carefully selected to ensure accuracy of the analysis results. The massive interior components referred to as the Hot Cell and the Supercell are included to capture the effect of their eccentricities on the at-grade slab. Based on the mass ratio of the overhead cranes to the supporting system, a coupled analysis is not required, and the cranes are not explicitly modeled other than their seismic mass.

Major structural elements of the FSTR, including walls, slabs, beams, and columns, are modeled with appropriate mass and stiffness properties. Major openings within walls and slabs are included in the SSI model. The model uses thick shell elements to represent concrete slabs and walls, and beam elements to represent steel members, mostly comprising the truss components in the facility.

The SSI analyses are performed for mean (BE), UB, and LB soil properties to represent potential variations in in-situ and backfill soil conditions around the building consistent with Section 3.7.2 of NUREG-0800 (SRP 3.7.2) (Reference 13). SSI analysis requires detailed input of the soil layers supporting the structure. Discrete soil layers are defined with an elastic half-space below for the LB, BE, and UB soil conditions. The properties of the solid elements that compose the excavated soil volume are defined to match the equivalent linear strain-dependent properties of the soil layers in the analysis of each soil condition. Soil layer thickness and maximum soil solid element dimensions are defined such that they are fine enough to ensure transmittal of the highest frequency of interest in the analysis. Equation 3.4-9-1 is used to compare the soil layer thicknesses and element dimensions against 20 percent of the shear wavelength at the highest frequency of interest (24 or 33 Hz).

Page 100 of 130

max 5

f V

h s

(Equation 3.4-9-1)

where, s

V = the shear wave velocity in the soil layer media max f

= the maximum frequency of interest With the layer thicknesses used in the analysis, the BE and UB soil profile properties satisfy this requirement at a frequency of 33 Hz. However, the properties of some of the LB soil profile layers only satisfy the passing frequency requirement up to 24 Hz. Section 3.7.1 of NUREG-0800 (Reference 6) shows the target power spectral density of the Regulatory Guide 1.60 (Reference 14) spectrum above 24 Hz is so low that the effect in this range is inconsequential. The power spectral density of the generated acceleration time-histories confirms there is minimal power in the region beyond 24 Hz. Therefore, the effect of the soil layer/element size satisfying the passing frequency requirement only up to 24 Hz for the LB case on the overall seismic analysis is determined to be insignificant.

As described in Subsection 3.4.2.2 of the FSAR, strain dependent soil properties were determined from geotechnical investigations and free field site response analysis. The free-field site response analysis is performed for the LB, BE, and UB soil properties.

Consistent with SRP 3.7.2 (Reference 13), the UB and LB values of the soil shear modulus, G, are obtained in terms of the BE properties through the equations shown below.

Equations 3.4-2 and 3.4-3 of the FSAR are used to calculate the LB and UB low strain properties. The final in-profile soil properties are calculated using the program SHAKE2000, version 3.5.

A COV of 0.5 is used because the site is well investigated. The soil column is defined to a depth of 300 ft. below grade with an elastic half-space below to represent bedrock. Damping values for each soil layer are limited to 15 percent in the SASSI2010 model to meet SRP 3.7.2 (Reference 13) acceptance criteria.

To account for the effects of concrete cracking, a cracked analysis is performed with the BE soil profile where a reduced modulus of elasticity is used for the modeled concrete shell elements. The reduced concrete modulus of elasticity is defined as 50 percent of its nominal value, based on guidance in ASCE/Structural Engineering Institute (SEI) Standard ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities (Reference 15). In addition, in the cracked analysis the concrete damping and steel damping are increased to the Regulatory Guide 1.61 (Reference 7) SSE damping values of 7 percent and 4 percent, respectively. All other seismic analysis cases (LB, BE, and UB) considering uncracked concrete utilize the lower Regulatory Guide 1.61 (Reference 7) OBE damping values of 4 percent for concrete and 3 percent for steel. The LB, BE, UB, and cracked analysis cases are enveloped to determine the bounding seismic responses.

The SSI analysis is summarized in Section 3.4.2.2 of the FSAR, including discussion of the finite element model and the elements used in the analysis. Therefore, no changes to the FSAR are necessary.

Page 101 of 130 (2) The SASSI2010 model and the methodology for the seismic analysis were developed such that they would be in accordance with the applicable portions of SRP 3.7.2 (Reference 13).

The approaches used align with SRP 3.7.2 and standard modeling and FEA practices.

Therefore, no additional justification is required and no changes to the FSAR are necessary.

(3) The SASSI2010 FEA model is only used for the seismic analysis. The building structure portion of the SASSI2010 model has identical finite element geometry and properties to the SAP2000 model (with minor adjustments to account for differences between the two programs) that is used to perform static loading analyses for use in the design of the structure. The element mesh size and aspect ratio are carefully selected to ensure accuracy of the seismic analysis results. The SASSI2010 and SAP2000 seismic FEA models are not used for evaluating aircraft impact and blast loads (although ancillary models of specific structural elements of the facility are used to determine deflections due to unit loads for aircraft impact evaluation). Therefore, no changes to the FSAR are necessary.

RAI 3.4-10 Section 3.4.2.6.2, Applicable Codes and Standards, of the SHINE OLA states that SHINE designed the FSTR consistent with the national code/standard ACI 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary.

Consistent with NRC RG 1.69, Revision 1, Concrete Radiation Shields and Generic Shield Testing for Nuclear Power Plants, Section 4a2.5.4.1.6, Exceptions for Use of ACI 349-13, of the SHINE OLA itemizes several exceptions to ACI 349-13. For the exception taken to Section 5.6.2.3 of ACI 349-13, the application states that Regulatory Position 5 of Revision 2 to NRC Regulatory Guide (RG) 1.142, Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments) for concrete strength testing is used.

The NRC endorses national codes and standards, such ACI 349, through regulatory guides and enhances the performance of standards with certain provisions/regulatory positions so that defense-in-depth is maintained in the design of applicable nuclear facilities. Regulatory Guide 1.142, in addition to the enhancement for concrete strength testing that the applicant chose to follow, also provides guidance as regulatory positions, to further strengthen the code philosophy that design of nuclear facility structures other than reactors have an increased capacity to function as a direct barrier or support a direct barrier against the release of radioactivity to the atmosphere for code addressed loads and loading conditions.

RG 1.142, Revision 2, endorses ACI-349-97, Code Requirements for Nuclear Safety Related Concrete Structures, and provides additional guidance to licensees and applicants through regulatory positions on methods acceptable to the NRC staff for complying with the NRCs regulations in the design, evaluation, and quality assurance of safety-related nuclear concrete structures.

It is not clear whether sections of ACI 349-13, other than those itemized in Section 4a2.5.4.1.6 of the FSAR, as modified or considered inapplicable to the FSTR concrete structural design (e.g., loading combinations, load factors, seismic detailing) are consistent with the NRC philosophy for defense-in-depth promulgated in RG 1.142 and review and acceptance procedures of NUREG-1537. It is not clear whether regulatory positions in Revision 2 to RG 1.142 are included in the design of the FSTR so that facility defense-in-depth can be maintained for seismic, aircraft impact, blast loading designs.

Page 102 of 130 (1) State whether any modifications and/or exceptions taken to ACI 349-13 (other than those itemized in Section 4a2.5.4.1.6 of the FSAR) for the design of the FSTR and its associated concrete SSCs.

(2) If none, state how the current concrete design of the FSTR and its SSCs as implemented based on ACI 349-13 provides an appropriate level of conservatism with successive levels of protection (defense-in-depth), such that health and safety of the public is not wholly dependent upon a structural failure of a single element of the design for seismic, aircraft impact, and blast (explosion effects) loads.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response (1) The design of the safety-related reinforced concrete portions of the FSTR is consistent with American Concrete Institute (ACI) 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures & Commentary (Reference 16), with the following modifications and exceptions:

Where interconnected structural components, such as walls, slabs, and foundations, exhibit a structural response consistent with the response of structural frames, such components should conform to the requirements of Chapters 10, 11, and 21 of ACI 349-13, in addition to Chapters 13, 14, and 15 as appropriate. Treat the response of structural components in a manner consistent with the response of structural frames if the flexural moment from seismic loads is equal to or exceeds two-thirds of the design flexural capacity of the section in the absence of axial forces.

From Section 7.12.4 of ACI 349-13, the application of one-third greater than that required by analysis provision to the minimum reinforcement is not considered.

Section 7.12.4 of ACI 349-13 does not address minimum reinforcement for the foundation base mat. The ratio of non-prestressed reinforcement area to gross concrete area, on a tension face of a foundation base mat, where a calculated reinforcement requirement exists, should not be less than 0.0018 in the direction of the span under consideration.

To determine the shear strength of slabs and walls for concentrated loads or reactions perpendicular to the plane of the slabs and walls, the effective width of the critical section for the beam action condition is the zone of influence induced by the concentrated loads instead of the entire width of the slab as specified in Section 11.11.1.1 of ACI 318-08, Building Code Requirements for Structural Concrete and Commentary (Reference 17), which is incorporated by reference in Section 11.9 of ACI 349-13. The effective zone of influence may be determined, for example, by analysis.

Hydrodynamic loads associated with seismic loads (i.e., the impulsive and sloshing loads for fluids in tanks) are to be considered as part of Ess in load combinations (9-6) and (9-9). Hydrodynamic loads associated with seismic loads are to be considered as part of EO in load combination (9-4). All other hydrodynamic loads should be taken as Yj in load combination (9-9).

Design for loads due to accidental explosions, or accidental vehicle impacts, or small aircraft impacts should use load combination (9-7) with those loads in lieu of Wt, and further guidance provided in ACI 349-13, Appendix F, Special Provisions for Impulsive and Impactive Effects. The Wt load in the load combination (9-7) should be evaluated for both tornado and hurricane loads applicable to the site.

Page 103 of 130 For load combinations (9-6) and (9-9), the load effects of the seismic load (Ess) are not reduced by 10 percent for the conditions in Section 9.2.10 of ACI 349-13.

In the case of the reaction shear (beam action) at the supports in section F.5 of ACI 349-13, effective width of the critical section for the shear capacity at the supports is to be determined according to the zone of influence induced by the local loads instead of the entire width of the support. The zone of influence induced by the concentrated loads may be determined, for example, by an analysis.

(2) As stated in Part 1 of the SHINE Response to RAI 3.4-10, the design of the safety-related reinforced concrete portions of the FSTR is consistent with ACI 349-13, with the identified modifications and exceptions. This, along with the conservative nature of the design methodology described in Section 3.4 of the FSAR, ensures that the structure provides successive levels of protection such that structural failure of a single element of the design for seismic, aircraft impact, and blast (explosion effects) loads does not impair the health and safety of the public.

RAI 3.4-11 Section 1.2.2, Safety Considerations, of the SHINE OLA states that [t]he building structure is robust enough to remain intact following an aircraft impact. Section 3.4, Seismic Damage, states that [t]he roof of the facility is supported by a steel truss system. Section 3.4.2.6.2, Applicable Codes and Standards, states that ANSI/AISC N690-12, Specification for Safety-Related Steel Structures for Nuclear Facilities is the applicable code and standard for the design of the SHINE main production facility structural steel SSCs.

The SHINE OLA referenced U.S. Department of Energy (DOE) Standard DOE-STD-3014-2006, Accident Analysis for Aircraft Crash into Hazardous Facilities, which discusses aircraft impacts and potentially ensuing fuel fires and explosions and the capacity of existing (or proposed) barriers to dissipate their energy. It emphasizes that fire can spread through ducts and along wiring conduits. It limits fire barriers and breaks credits to those that remain undamaged by the crash. The standard further, states:

The basis for taking credit (e.g., short duration of the fire) should be documented.

Therefore, a characterization of fire duration will almost certainly be required, although the level of detail will depend on how much sophistication is required to determine the duration of the fire relative to the capability of the fire barriers. Due to the difficulty of demonstrating that active systems can function following a crash, credit should not be allowed for fire suppression systems unless an explicit analysis shows that they will remain effective [] In calculating an effective [aircraft impact/skidding target] area, the analyst needs to be cognizant of the critical areas of the facility. Critical areas are locations in a facility that contain hazardous material and/or locations that, once impacted by a crash, can lead to cascading failures, e.g., a fire, collapse, and/or explosion that would impact the hazardous material.

It is not clear to the NRC staff whether applicable specifications of ANSI/AISC N690-12, are met in their entirety, including those related to fire in Appendix N4 to ANSI/AISC N690-12. The Appendix discusses carbon steel material properties at elevated temperatures. It states, material properties at elevated temperatures are short-term properties intended for fire design by analysis only. It also states that the specification does not address either Important to Safety structural steel members or loading conditions associated with a facility fire. It is not clear to the NRC staff whether a fuel fire due to an aircraft impact was considered for an aircraft

Page 104 of 130 global response analysis but ruled out because of its potential short duration and lack of damage to the external concrete building envelope (external walls, roof, exhaust stack). If so, it is not clear how that was determined so that fuel fire and aircraft combustible material would remain localized and external to the FSTR and its SSCs. If not, it is not clear whether a structural steel analysis was performed taking into consideration aircraft fuel fire. If so, it is not clear whether the analysis appropriately considered for the roof steel truss system (including steel decking, if any) thermal effects and material properties at elevated temperatures. Given the proximity/connection of the exhaust stack to the FSTR, it is also not clear whether the stack could function as an intake duct for the spreading of fuel fire in the facility that could affect critical areas that contain hazardous/radioactive material and/or locations that, once impacted by a crash, can lead to cascading failures. This information is requested to verify that SHINE has performed the necessary evaluations required to show that safety functions will be accomplished, as required by 10 CFR 50.34(b)(2).

(1) Clarify, whether any deviations were made to applicable specifications of ANSI/AISC N690-12 in the design of the SHINE main production facility structural steel SSCs. If so, justify their exclusion.

(2) Clarify whether an aircraft fuel fire was considered in aircraft impact global response analysis but ruled out. Justify ruling out such fires that could affect the integrity of FSTR structural steel members and steel decking, if any.

(3) If the aircraft impact global response analysis included an aircraft fuel fire, discuss whether requirements of ANSI/AISC N690-12 for structural steel performance at elevated temperatures were considered in the steel design of FSTR and its Important to Safety structural steel members. If excluded, justify the exclusion.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response (1) The only safety-related structural steel SSCs in the FSTR are the trusses that support vertical roof slab loading and the RCA mezzanine structural steel that supports vertical mezzanine slab loading. Provisions of American National Standards Institute/American Institute of Steel Construction (ANSI/AISC) N690-12, Specification for Safety-Related Steel Structures for Nuclear Facilities (Reference 18), that are applicable based on the applied loading are considered in the design of these SSCs without exception.

(2) U.S. Department of Energy (DOE) Standard DOE-STD-3014-2006, Accident Analysis for Aircraft Crash into Hazardous Facilities (Reference 19), states that it contains several interrelated analytical modules, intended to provide the user with a framework of step-wise increases in analytical sophistication aimed at eliciting only that amount of analysis needed to demonstrate that aircraft crash either does or does not exceed a risk level of concern This step-wise approach allows the user to determine that the aircraft crash does not exceed a risk level of concern at the end of each analytical module. The third analysis module defined by the standard is the Structural Screening and Evaluation, which is described in Subsection 3.4.5.1 of the FSAR.

The structural screening and evaluation of the FSTR demonstrates that the building components meet the structural screening and evaluation guidelines at all impact locations

Page 105 of 130 and there is no safety-related equipment supported from the building in the vicinity of the postulated impact; therefore, the risk is deemed small and the results are documented.

The discussion of evaluation of the facility for fuel fires and explosions due to aircraft impact are included in the fourth and fifth modules, Release Frequency Screening and Release Frequency Evaluation, respectively. The user of DOE-STD-3014-2006 does not need to perform modules four and five if the third module, Structural Screening and Evaluation, indicates that the risk is deemed small.

Therefore, the SHINE FSTR is not evaluated for the risks of aircraft fuel fires and explosions in the aircraft impact global response because it has been demonstrated, in accordance with DOE-STD-3014-2006, that no impact results in structure damage (i.e., damage does not exceed the structural response guidelines).

(3) An aircraft fuel fire was excluded from the aircraft impact global response as described in Part 2 of the SHINE Response to RAI 3.4-11. Therefore, the provisions of Appendix N4 to ANSI/AISC N690-12 for fire conditions are not applicable and are not considered for the steel design of the FSTR and its structural steel members.

RAI 3.4-12 Section 1.3.3.3, Facility Systems, of the SHINE OLA states that the neutron driver assembly system (NDAS) is an accelerator-based assembly that accelerates a deuterium ion beam into a tritium gas target chamber. The resulting fusion reaction produces 14 million electron volt (MeV) neutrons, which move outward from the tritium target chamber in all directions.

Section 4a2.3, Neutron Driver Assembly System, states that [s]tructural support beams support the neutron driver in the IU cell, with components installed above and adjacent to safety-related equipment. Neutron driver components within the IU cell are classified as a Seismic Category II component. It also states that [t]he target chamber generates up to 1.5E+14 neutrons per second (n/s) during operation.

Guidance documents such as NUREG-7171, A Review of the Effects of Radiation on Microstructure and Properties of Concretes Used in Nuclear Power Plants, discussed in 4a2.5.3.2 Radiation Damage, and referenced by the applicant in Chapter 4 of the SHINE OLA and the industry standard ACI 349.3R-18, Report on Evaluation and Repair of Existing Nuclear Safety-Related Concrete Structures, provide radiation thresholds for concrete and insights beyond which the compressive strength of concrete appears to rapidly decline while its crack density increases.

To effectively accomplish their intended function, nuclear safety-related SSCs are designed to resist operating loads, severe environments such as seismic events, and postulated accidents.

To prevent lifetime-radiation related degradation of concrete SSCs and maintain an acceptable level of serviceability, NUREG-7171 limits the lifetime reinforced concrete neutron fluence exposure of 0.1 (and above) MeVs to 1 x 1019 neutrons/cm2 and for gamma dose to 1010 rads.

ACI 349.3R-18, which is more relevant to long term operation of nuclear facilities, also states that neutron fluence can change the mechanical properties of carbon steel resulting in an increase in yield strength and a rise in the ductile to brittle transition temperature.

Given the projected hours of operation for the SHINE facility, the high neutron flux and gamma dose exposures at NDAS or at other locations exposed to intense radiation within the facility, it

Page 106 of 130 is not clear whether concrete or steel structural support members or components, such as the IU driver supporting beams, have been evaluated for neutron fluence and gamma dose damage for the life of the facility. It is also not clear whether conservatively a reduction in strength due to radiation for materials used in the construction of the facility was considered and factored where applicable in the concrete or structural steel designs for seismic, aircraft impact, blast loadings.

(1) Discuss whether the radiation limits provided in NUREG-7171 were used to determine that safety related concrete or steel support structures or SSCs exposed to radiation (e.g., the IU driver support beams) will maintain their safety function during seismic, aircraft impact, or blast loading scenarios during the intended licensing period. Provide a discussion of any relevant evaluations used to support a conclusion that irradiation will not affect the safety functions described above.

(2) If applicable, state what actions are taken to ensure that potential damage to safety related concrete or steel support structures or SSCs that have radiation exposure above the previously discussed radiation limits will not adversely affect safe facility operability and its defense-in-depth.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response (1) The neutron fluence and gamma dose rate to the concrete support structures in the facility were calculated using Monte Carlo N-Particle Transport Code 5 (MCNP5) and were shown to be under the acceptance criteria provided in NUREG/CR-7171 (i.e., 1019 n/cm2 and 1010 rad, respectively) (Reference 20). The neutron fluence to the concrete support structures were also under the limit provided by ACI 349.3R-18 (i.e., 1020 n/cm2)

(Reference 21), where steel reinforcement could experience reduced ductility. The incident energy flux was also calculated for the concrete and was compared to the American National Standards Institute/American Nuclear Society (ANSI/ANS)-6.4-2006 (Reference 22) acceptance criterion that nuclear heating shall be considered for concrete shields that are exposed to incident energy flux great than 1010 MeV/cm2 and that will operate at a temperature of 65°C or greater. The energy flux was calculated to be less than 1010 MeV/cm2 in most cases, and where it was greater than 1010 MeV/cm2, the temperature of the concrete was determined to be less than 65°C. Because the radiation exposure to the concrete structures is below allowable limits during the intended licensing period, no damage affecting the safety function of the safety-related concrete structures during seismic, aircraft impact, blast loading, or other accident scenarios is anticipated.

Steel structures are outside the scope of NUREG/CR-7171. To support the conclusion that irradiation will not affect the safety functions of safety-related steel structures located in the IU cells, lifetime neutron fluence to steel components close to the target solution vessel (TSV) and neutron driver assembly system (NDAS) target chamber were calculated using MCNP5. The structural component with the highest lifetime neutron fluence is the subcritical assembly support structure (SASS), which includes both pressure vessel and structural elements (e.g., support legs) and is fabricated from austenitic stainless steel. The highest fluence to the SASS occurs at the pressure vessel portion of the SASS adjacent to the TSV and is up to approximately [

]PROP/ECI. The fluence to the SASS was compared to literature data (Reference 23), which indicates an increase in yield and ultimate tensile strength and a reduction in ductility of austenitic stainless steel. An increase of yield and Proprietary Information - Withheld from public disclosure under 10 CFR 2.390(a)(4)

Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Page 107 of 130 ultimate tensile strength by itself does not affect the ability of a steel structure from performing its safety function. However, a decrease in ductility, if excessive, could lead to brittle fracture during normal or accident loadings. To determine if the loss of ductility could be excessive, literature data was consulted to determine if austenitic stainless steel retains adequate ductility when irradiated with comparable fluence. Literature data (Reference 23) indicates that irradiated austenitic stainless steel loses some ductility but still retains approximately twice the energy (60 lb-ft) to fracture than the commonly accepted criterion of 30 lb-ft and therefore remains ductile. Literature (References 23 and 24) also indicates that austenitic stainless steel SASS components will remain ductile for temperatures greater than -50°C throughout their service life. The SASS is applicable as the limiting case for safety-related steel structures within the IU cell because such structures are made of austenitic stainless steel. Because the limiting case of irradiated steel structural components indicates an increase in strength and retention of adequate ductility, no damage affecting the safety function of the safety-related steel structures is anticipated.

Radiation impacts on structural components of the IU cell concrete are summarized in Subsection 4a2.5.3.2.1 of the FSAR. Section 4a2.2.5 of the FSAR has been revised to provide additional details regarding the radiation impacts on the SASS. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

(2) No actions are needed to ensure that radiation exposure will not adversely affect safety-related SSCs because radiation exposure is below applicable radiation limits.

Radiation exposure of safety-related concrete or steel support structures or SSCs will not adversely affect safe facility operability and its defense-in-depth.

RAI 3.4-13 Section 3.4.2.6.4.6, Crane Load, of the SHINE OLA states that the building is evaluated for loads associated with two overhead bridge cranes, one servicing the Irradiation Facility (Irradiation Unit cell) area (IF/IU) and one servicing the Radioisotope Production Facility area (RPF). It also states that crane loading is evaluated in accordance with American Society for Mechanical Engineers (ASME) NOG-1, Rules for Construction of Overhead and Gantry Cranes (ASME, 2004).

Section 9b.7.2, Material Handling, of the SHINE OLA states that crane hooks are rated at 40 and 15-ton lifts and for the IF and RPF designated as ASME NOG-1 Type I and II, respectively.

Sections 1000, Introduction, to ASME NOG-1 define/discuss crane loads as superimposed weight and credible critical loads. The Non-Mandatory Appendix B Commentary to ASME NOG-1, further clarifies that crane loads the structure sustains can be assessed either deterministically or probabilistically (credible critical loads). ASME NOG-1 also states that probabilistic calculations establish the weight of lifted load that should be considered in combination with OBE, and that should be considered in combination with SSE, or of specifying the range of loads that should be considered for varying magnitudes of earthquakes, from magnitude less than OBE up to SSE.

Section 4000, Requirements for Structural Components, of ASME NOG-1 further defines loads, loading combinations, restraint conditions at nodes to be used in static, dynamic, seismic, and abnormal events analyses and design of crane hardware systems. It also provides added guidance to calculate the maximum structural response values for the three-directional components of an earthquake motion. Guidance on loading combinations and structural

Page 108 of 130 responses include impactive vertical loads, and horizontal (i.e., longitudinal and transverse) loads. When performing seismic analyses, the ASME NOG-1 provides specific criteria to decouple the crane from its runway.

As noted in Chapters 2 and 3 of the SHINE OLA, structural design of FSTR, including its design of the structures for seismic and abnormal loads, is in accordance with ASCE 7/IBC, ACI 349, and AISC N690-12 national codes and standards.

The descriptions provided in the SHINE OLA do not provide adequate information of how the crane loads were derived (deterministically or probabilistically) and subsequently used in seismic and other abnormal load (dynamic/impact) analyses consistent with ASME NOG-1. In addition, structural codes and national standards used in the design of the FSTR address crane loads (e.g., Chapters 4 of ASCE 7/1607 of IBC, with loading combinations further elaborated in Chapters 9 and Appendix C of ACI 349-13 and Chapter NB of ANSI/AISC N690-12). These codes/national standards differ in some respects (e.g., impactive loads) with the ASME NOG-1 in the assessment of crane loads and loading combinations. It is not clear whether the FSTR was designed based on IF and RPF crane loads derived consistent to ASME NOG-1 or the structural design codes/national standards.

(1) Clarify how SHINE is applying ASME NOG-1 to crane loads at the facility. As applicable, consistent with ASME NOG-1, provide the following:

(a) Describe whether the IF and RPF crane loads were derived deterministically or probabilistically and included as such in the loading combinations used. Justify the approach taken, discuss their use, adequacy and conservatism for seismic and abnormal load analyses/design.

(b) Describe whether the IF and RPF crane response was decoupled from their respective runways for seismic analyses. If so, state type of loads considered (deterministic or probabilistic) in the decoupling and whether such selection provided conservatism in crane/FSTR structural analyses and design.

(c) Describe whether the IF and RPF crane decoupling from their runways was limited only to seismic analyses and if so, why. Otherwise, describe how the (deterministic, probabilistic) crane loads were integrated in the facility analysis and design for seismic as well as for abnormal (aircraft impact, blast effects) load structural analyses.

(2) If ASME NOG-1 derived crane loads were applied to the FSTR design, clarify whether the use of such loads provide an additional design conservatism than those derived based on the aforementioned structural codes and standards.

(3) If credible critical crane (probabilistic) loads were used in the structural and seismic analyses and design of the FSTR, describe how they are integrated with the load resistant factor design philosophy of ACI and AISC structural design codes and standards.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response The SHINE Response to RAI 3.4-13 will be provided by January 31, 2021.

Page 109 of 130 RAI 3.4-14 Section 1.2.1, Consequences from the Operation and Use of the Facility, of the SHINE OLA states that [t]he IF and RPF within the main production facility constitute the radiologically controlled area (RCA) where radioactive materials are present. Section 4b.4, Special Nuclear Material [SNM] Processing and Storage, states that SNM is used throughout the radioisotope production facility (RPF) radiologically controlled area (RCA) in both unirradiated and irradiated forms.

Section 3.4.2.6.4.6, Crane Load, of the SHINE OLA states that the building is evaluated for loads associated with two overhead bridge cranes, one servicing the IU cell area and one servicing the RPF area. It also states that [c]rane loading is evaluated in accordance with []

ASME NOG-1.

Section 9b.7.2, Material Handling, of the SHINE OLA states that the IF and RPF cranes are ASME NOG-1 Type I and Type II cranes rated at 40 and 15-tons, respectively. Both cranes are to perform at a Service Level B (Light Service) consistent with CMAA 70. Consistent also with ASME NOG-1, the IF crane includes in its design single failure-proof features while the RPF crane does not and hence it may not support a critical load during a seismic event. This Section also states that safeguards consistent to NUREG-0612, Section 5.1.1, would be developed to limit consequences of radiological release as promulgated in 10 CFR Part 20, if a heavy load was dropped on safety-related SSCs.

Sections 1000, Introduction, and 5000, Mechanical, and others to ASME NOG-1 address exposure of cranes to radiation and other environmental conditions that could reduce normal life of their components including the loss of single failure-proof features potentially could result in load drops. In the RCA, such unexpected crane load drops could result in radiological consequences. Such code sections also state that select crane components need to be designed to withstand maximum facility lifetime radiation exposure. In its non-mandatory Appendix B, the Code emphasizes that nuclear facilities of new or unforeseen designs should consider special fracture toughness acceptance criteria for ASME NOG-1 acceptable materials (reference ASME NOG-1, Table 4212-1) used as structural components, including bolts and welds, exposed to unusual radiation. It is not clear whether the IF and RPF cranes were designed consistent with ASME NOG-1 Section 1000 guidance to withstand lifetime gamma and/or neutron radiation.

Because of the unpredictability of such failures, it is not clear whether planned safeguards consistent with Section 5.1.1 of NUREG-0612 would be adequate to limit effects of unexpected drops of heavy loads on safety related SSCs due to radiation induced critical crane component malfunctions/brittle fractures during normal operation of the FSTR as well as during a safe shutdown seismic event so that that defense-in-depth is maintained.

(1) Discuss, consistent with guidance (for example, Sections 1000, Introduction to ASME NOG-1), whether detrimental effects of lifetime radiation on critical and structural components (including fracture toughness assessments, for members, fasteners, and welds) for RCA cranes were considered.

(2) State whether RCA cranes would be radiation hardened and/or periodically inspected for effects of radiation to ensure that their service life (lift cycles allowed) consistent with CMAA-70 remain as noted in the FSAR and their critical/structural components, particularly those associated with single failure-proof features of the IF crane, remain unaffected from

Page 110 of 130 radiation throughout their operating life to minimize potential drops on FSTR critical structural components and its safety related SSCs and defense-in-depth of the facility remains during seismic and other abnormal loading conditions.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response (1) Effects of lifetime radiation were considered in the design of the IF and the RPF cranes.

Consistent with Section 1141 of American Society of Mechanical Engineers (ASME)

NOG-1-2004, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) (Reference 25), the purchase specifications for both the IF and the RPF cranes includes environmental conditions within the design criteria which specifies the accumulated lifetime radiation dose (i.e., 100 rad over 30 years for the IF crane and 75 rad over 30 years for the RPF crane) and the accident dose rate (i.e., 1E+08 mrem/hr for both cranes).

(2) The IF crane is designed and constructed in accordance with ASME NOG-1 for Type I cranes, including fracture toughness requirements (e.g., Table 4212-1, Table 4222-1, and Table 4232-1) and requirements for radiation resistance of mechanical and electrical components (e.g., Sections 5413, 5461, 5540, 6150, 6422, and 6481). The applicable testing requirements for the critical structural and mechanical components of the IF crane are listed in Section 7000 of ASME NOG-1.

The RPF crane is designed and constructed in accordance with Crane Manufactures Association of America, Inc. (CMAA) 70-2004, Specifications for Top Running Bridge & Gantry Type Multiple Girder Electric Overhead Traveling Cranes (Reference 26).

The RPF crane is seismically designed and constructed in accordance with ASME NOG-1, Type II criteria to ensure that it will remain in place with or without a load during an SSE; however, the crane need not support the load nor be operational during and after such an event.

The IF and RPF cranes are inspected, tested, and maintained in accordance with ASME B30.2, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist) (Reference 27) as described in Subsection 9b.7.2.3 of the FSAR. Inspection and testing of special lifting devices are performed in accordance with ANSI N14.6-1993, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More (Reference 28), as described in Subsection 9b.7.2.3 of the FSAR. Implementation of these standards provides assurance that the service life of the IF crane and RPF crane critical components remain as described in the FSAR. Implementation of these standards also provides assurance that the IF crane and RPF crane critical/structural components, particularly those associated with single failure-proof features of the IF crane, remain capable of performing their intended functions throughout their operating life.

Page 111 of 130 RAI 3.4-15 Section 3.4.2.6.4.8, Fluid Load, of the SHINE OLA states that [t]he hydrostatic loading is calculated based on the actual dimensions of the IU cells and applied in the model as lateral hydrostatic pressure on the walls and vertical hydrostatic pressure on the bottom slabs.

Section 3.4.2.6.4.5, Earthquake Load, states that 100 percent of hydrodynamic loads are accounted for in earthquake analysis.

Section 1.2.1, Consequences from the Operation and Use of the Facility states that [w]ithin the irradiation facility (IF), the [low enriched uranium] LEU in the target solution is in the form of a uranyl sulfate.

Section 4.a2, Irradiation Facility Description, states that the stainless-steel target solution vessel (TSV) contains the uranyl sulfate target solution undergoing irradiation to produce Mo-99 and other fission products and is attached to the floor of the light water stainless steel lined pool via seismic anchorages. It also states that the subcritical assembly support structure (SASS) and primary system boundary (PSB) components are designed to withstand the design basis loads, including thermal, seismic, and hydrodynamic loads imposed by the light water pool during a seismic event. In addition, it states that The SASS does not normally contact the target solution. In the event of a breach in the TSV, the SASS provides a defense-in-depth fission product boundary between the target solution and the light water pool. The section also states that the pool has minimum acceptable water levels that are assumed for safety analysis accident scenarios for normal operation and for loss of cooling conditions and that the target stage of the neutron driver is partially submerged.

The staff noted that Sections 3.4.2.6.4.5, 3.4.2.6.4.8, and 4.a2 of the FSAR discuss general application of hydrodynamic loads, hydrostatic loads for the pool, and pool submerged/semi-submerged equipment, under seismic conditions. It is not clear whether the light water pool is the only area of concern in the FSTR where hydrostatic and hydrodynamic loads are applied. It is also not clear where else in the FSTR fluid-equipment/structure interaction may have been addressed. Chapter 3 of FSAR references ASCE 4-98, Seismic Analysis of Safety-Related Nuclear Structures and Commentary. The standard addresses horizontal and vertical fluid motion (hydrodynamic loads - impulsive and convective) and their effects on submerged structures in a basin. It is not clear, however, whether ASCE 4-98 was used for hydrodynamic load estimations and for structural stability analyses in fluid-structure interactions during seismic events. It is not clear what was the method followed to derive the hydrodynamic loads and their effects on the stability of totally or partially submerged structures during seismic loads or other abnormal events.

It is also not clear whether the hydrostatic/hydrodynamic loads/analyses were limited to water as a fluid or extended to include dilution of uranyl sulfate solution into the pool. Material DATA Sheets indicate that uranyl sulfate in its solid form has a specific gravity of 3.28. Consistent with this data, the uranyl sulfate solution specific gravity is anticipated to be greater than that of water. Although it is noted that the SASS provides a defense-in-depth against uranyl sulfate solution release, it is not clear what constitute the echelons of defense to preclude leakage of uranyl sulfate solution into the pool during a seismic or other abnormal loading events that could alter hydrodynamic loads on the pool and their effects on submerged structures during a design basis earthquake.

(1) Discuss the method followed/standard(s) used to derive the hydrodynamic loads, their effects on submerged equipment/structures within the pool, and analyses performed for

Page 112 of 130 seismic or other abnormal loading events. Identify other locations in the FSTR where these loads/analyses were applied.

(2) Briefly describe the echelons of defense claimed in SASS/TSV equipment/structure against the release of uranyl sulfate solution into the pool. Justify their adequacy consistent with the defense-in-depth philosophy of NUREG-1537 during seismic or other abnormal loading events.

(3) If a breach of the SASS/TSV equipment/structure would occur, discuss whether effects of an increased density fluid on hydrodynamic loads and semi/submerged structures were considered for seismic or other abnormal loadings and steps taken so that defense in depth of the facility remains.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response (1) Hydrodynamic loads are analyzed for the IU cell structure and for safety-related equipment submerged in the light water pool. Hydrodynamic loads are not applied to other areas of the FSTR.

The IU cell structure is designed to resist hydrodynamic fluid loads. The hydrodynamic loading is applied by considering hydrodynamic masses rigidly attached to the cells in accordance with Section 3.1.6.3 of ASCE 4-98, Seismic Analysis of Safety-Related Nuclear Structures (Reference 10), and Chapter 6 of TID-7024, Nuclear Reactors and Earthquakes (Reference 29). As required by the referenced documents, the impulsive and convective (sloshing) masses are applied to capture the dynamic effects on the IU cell due to seismic motion.

Safety-related equipment submerged in the light water pool is designed to resist hydrodynamic fluid loads. Two categories of hydrodynamic loads on such equipment are considered: hydrodynamic added mass and hydrodynamic forces from sloshing.

The hydrodynamic added masses applied to submerged components within the light water pool account for the additional water mass that moves with the components during seismic excitation. These masses are calculated based on the geometry of the component in question. For example, the added mass for a circular pipe submerged in water is taken to be equal to the mass of the water displaced by the pipe. The added masses are used to calculate forces based on the acceleration of the components.

The hydrodynamic drag forces applied to submerged components within the light water pool account for light water pool water sloshing against components. The maximum vertical displacement of sloshing pool water is calculated using the methods in Section 9 of ACI 350.3-06, Seismic Design of Liquid-Containing Concrete Structures and Commentary (Reference 30). The vertical and horizontal water velocities are calculated using the maximum vertical displacement of the pool water. Using these velocities, the vertical and horizontal drag forces are calculated.

Hydrodynamic loads on submerged components are combined with seismic self-weight excitation loads and other loads in accordance with ANSI/AISC N690-12, Specification for Safety-Related Steel Structures for Nuclear Facilities (Reference 18). Earthquake loads,

Page 113 of 130 including hydrodynamic effects, are the only abnormal loads generally analyzed for the equipment submerged within the light water pool.

Subsections 3.4.2.6.4.8 and 4a2.2.5 of the FSAR have been revised to clarify the analyses of hydrodynamic loads for the IU cell structure and for safety-related equipment submerged in the light water pool, respectively.

(2) The primary system boundary (PSB), which includes the TSV, forms the single barrier against the release of uranyl sulfate target solution into the light water pool, as described in Subsection 4a2.2.1.4 of the FSAR. The PSB is designed to retain pressure boundary integrity during seismic and other abnormal loading events, as described in Subsection 4a2.2.5 of the FSAR. A mark-up of the FSAR incorporating these changes is provided as.

The primary confinement boundary, which encompasses the light water pool, provides a secondary barrier against the release of uranyl sulfate target solution and associated fission products, as described in Subsection 6a2.2.1.1 of the FSAR. In the event of a failure of the PSB, the primary confinement boundary mitigates the release from the PSB, maintaining accident doses below allowable limits, as described in Chapter 13 of the FSAR. The primary confinement boundary is qualified to withstand seismic and other abnormal loading.

The SASS surrounds the TSV, with the exception of the TSV nozzles which pass through the SASS pressure boundary to connect to subcritical assembly system (SCAS) piping and instrumentation. The SASS has a safety function to provide structural support to the TSV and other safety-related SCAS equipment but does not have a safety function to retain pressure boundary integrity. The SASS is not relied upon to contain uranyl sulfate target solution during a seismic event; however, the SASS provides a defense in depth barrier around the portions of the TSV.

(3) A seismic event combined with a failure of the PSB and release of uranyl sulfate target solution is not a postulated event for the SHINE facility because the PSB is a passive barrier qualified to withstand seismic loading.

If a failure of the PSB were to occur (such as a breach of the TSV) and uranyl sulfate target solution were to leak into the light water pool, the increase in pool water density would be negligible (i.e., approximately [ ]PROP/ECI), which would result in insignificant changes to seismic and other abnormal loadings.

RAI 3.4-16 Section 1.2.2, Safety Considerations, of the SHINE OLA states that [t]he building structure is robust enough to remain intact following an aircraft impact as described in Section 3.4.

Section 3.4.5.1, Aircraft Impact Analysis, of the SHINE OLA states that safety-related structures at the SHINE facility are evaluated for global and local aircraft impact loadings resulting from small aircraft. It also states that the Challenger 605 was selected as a design basis aircraft impact based on airport operation data. It further states that the performed global impact response analysis the energy balance method was used with ductility limits in accordance with ACI 349-13 and ANSI/AISC N690-12, for reinforced concrete elements and steel truss members, respectively. For the local impact it states that [b]ecause engine diameter and engine weight are both critical for the local evaluation, the local impact evaluation was performed for the Hawker 400 as well as the Challenger 605 aircraft [which were...] evaluated Proprietary Information - Withheld from public disclosure under 10 CFR 2.390(a)(4)

Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Page 114 of 130 as design basis aircraft impacts. The section subsequently references U.S. Department of Energy (DOE) Standard DOE-STD-3014-2006, which provides guidance for screening and evaluating global, local, and vibration damage to FSTR and its SSCs.

The DOE Standard provides functional assessments of safety related SSCs for fuel fire, missile impact shock, and structural damage/collapse that when followed could minimize the risk posed to the health and safety of the public and onsite workers from a release of hazardous material following an aircraft crash. The standard considers deformable/soft (e.g., entire aircraft, wings, fuselage) and nondeformable/rigid (e.g., landing gear, engine shaft) aircraft components (missiles) that could impact a target directly or after skidding. It states that the selection of missiles should be bounding based on all applicable categories/subcategories and types of aircraft having the highest kinetic energy and provides methodologies to evaluate global and local impact damage. It further recommends impact assessments to include for global response aircraft-target interactions (including soil-structure interaction - SSI) and for local spalling, scabbing, perforation.

The applicants statement in Section 1.2.2 regarding the design robustness of the building structure, stems from its summary of an aircraft impact analysis based on the DOE Standard described in Section 3.4.5.1. In evaluating this section, however, the staff was not clear on how the two aircrafts were selected or what methodology was used to assess missile mass and velocities at impact for momentum transfer and kinetic energy estimations for global and local impact damage. This information is requested to verify that SHINE has performed the necessary evaluations required to show that safety functions will be accomplished, as required by 10 CFR 50.34(b)(2).

(1) Clarify what airport operation data were used and how such data were applicable for the selection of Challenger 605 as a global impactor.

(2) Clarify which mode of Challenger 605 global impact at contact with the walls and roof of the FSTR (e.g., direct impact, oblique impact, skidding) was considered in the global response/damage analysis. For the excluded modes, justify their exclusion.

(3) For global response mode of impact having a horizontal velocity component, state whether its effects/traction were considered in the design of the concrete roof and supporting steel truss, including stability analyses of truss compression flanges.

(4) State whether the global response/damage analysis included the total aircraft mass (including fuel), aircraft-target interaction, and SSI. If not, state reasons for exclusions.

(5) Clarify when considering rigid impactors for local damage analysis, whether their mass was reduced. If so, justify the basis for the mass reduction.

Update the FSAR as appropriate to reflect the above requested information.

Page 115 of 130 SHINE Response (1) The following airport operation data is used to evaluate aircraft hazards associated with the FSTR:

Historical flight data for Southern Wisconsin Regional Airport (KJVL) retrieved from the Air Traffic Activity Data System (ATADS) for airport operations for 1990-2018 (i.e., actual values and not historical forecasts).

Forecasts for future flight traffic for KJVL for 2019-2045 generated in 2019 from the Federal Aviation Administration (FAA) Terminal Area Forecast (TAF).

The division of operation at KJVL airport into various runways obtained from communication with the KJVL airport manager.

The division of military operation at KJVL (i.e., 1 percent small aircraft and 99 percent helicopters) obtained from communication with the KJVL air traffic manager.

Civil air traffic operations at KJVL airport (i.e., all small aircraft operations, less than 30 seats), obtained from communication with the KJVL airport manager.

This data is used in the aircraft hazard probability analysis, described in Section 2.2.2 of the FSAR. DOE Standard DOE-STD-3014-96, Accident Analysis for Aircraft Crash into Hazardous Facilities (Reference 19), provides a screening value of 1E-6 per year, where the risk of an aircraft accident is considered acceptable if the frequency of occurrence is less than 1E-6 per year. The calculated crash probability for small non-military aircraft does not meet this criterion; therefore, the safety-related structures of the facility credited to prevent release in excess of regulatory limits are designed to withstand the impact of a small non-military aircraft.

The overall impact response for small non-military aircraft depends on the engine weight and the impact velocity. As all aircraft under consideration are general aviation aircraft, they have the same impact velocity (UCRL-ID-123577, Structures, Systems, and Components Evaluation Technical Support Documents [Reference 31]). Therefore, the critical aircraft for impact analysis is the aircraft with the heaviest engines. Per communication with the KJVL airport manager, the Challenger 605 was determined to have the heaviest engines of aircraft that frequent KJVL and meet the small aircraft criteria (i.e., less than 30 seats or payloads less than 7500 lbs.). Therefore, the global impact analysis is based on the Challenger 605 aircraft.

(2) A direct impact of the Challenger 605 aircraft at contact with the walls and roof is considered for global response and evaluation of the FSTR. The aircraft is postulated to strike any wall or roof that bounds the exterior of the safety-related areas of the structure as well as several interior walls that act as barriers due to openings in exterior walls.

Oblique or skidding impacts of the aircraft were not explicitly evaluated because a direct impact was determined to be controlling. This is because non-direct impacts would impart a portion of their energy perpendicular to the plane of wall or roof element (similar but less than the energy due to direct impact) and a portion into the plane of the wall or roof diaphragm, which is designed to be robust enough to distribute in-plane forces to adjacent supporting structural members. Non-direct impacts acting nearly in-plane with the impacted structural elements would likely glance off thereby imparting less overall energy to the structure.

Page 116 of 130 (3) As described in Part (2) of this response, only a direct impact of the aircraft is postulated at the roof. The global response mode of impact at the roof having a horizontal velocity component is determined to be controlled by the direct impact mode. As the roof diaphragm is 12 in. thick and the truss is not designed to be composite with the roof, any horizontal component of an aircraft strike would be dissipated through the roof diaphragm directly which is much stiffer against in-plane (horizontal) loading than the truss elements. The concrete roof and truss elements were not explicitly designed for this horizontal force component as it is not considered to be a critical design load.

(4) The total mass of the aircraft (including fuel) was not specifically considered in the impact analysis. Guidance from Section 6.3.3.1 of DOE-STD-3014-2006 (Reference 19) states that for small aircraft with airframes that are flexible relative to the target structure, the effective missile mass may be conservatively estimated as twice the combined mass of the engines.

This estimated mass is utilized for the effective aircraft mass in the impact analysis.

The global response evaluation of the FSTR was performed using the energy-balance method. Regarding aircraft-target interaction and SSI; as described in Section 6.3.3 of DOE-STD-3014-2006, the objective of the global response evaluation is to determine if the impact of the aircraft results in the excessive deformation or collapse of the target structure.

Global response evaluation of a building or structure typically involves characterization of the nonlinear behavior of both the aircraft and the target, including SSI. Global response evaluation can be performed by either the energy-balance method or the time-history analysis method. In the case of the FSTR, the energy-balance method was utilized as the conditions for use set forth in DOE-STD-3014-2006 were met. Namely that: (1) there are no safety-related SSCs supported by the target in the vicinity of the impact; (2) the configuration of the target is simple, such that the overall dynamic characteristics of the structure can be adequately represented by a single-degree-of-freedom (SDOF) nonlinear energy-absorbing system; and (3) the resulting response of the SDOF system would be compatible with the strength and ductility limits of the various components and supports of the impacted structure.

(5) When considering rigid impactors for local damage analysis, their mass was not reduced.

Local damage for concrete elements is considered due to perforation, scabbing, or punching shear following impact by the aircraft. As the aircraft is relatively deformable, Table I of DOE-STD-3014-2006 (Reference 19) permits a 40 percent reduction on the scabbing thickness and a 30 percent reduction on the perforation thickness of the impacted concrete elements. Notes 4 and 5 of Table I recommend a 10 percent and 20 percent increase on the minimum scabbing and perforation thickness, respectively. Section F.7.2.2 and Section F.7.2.1 of ACI 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary (Reference 16), require a 20 percent increase in the thickness of the impacted concrete element to prevent scabbing and perforation, respectively.

Therefore, the required concrete element thickness accounts for both the permissible reduction as well as the specified increase (using the maximum of either ACI 349-13

[Reference 16] or Notes 4 and 5 in Table I of DOE-STD-3014-2006 [Reference 19]).

Page 117 of 130 RAI 3.4-17 Section 3.4.5.2, Explosion Hazards, of the SHINE OLA states that the maximum overpressure at any safety-related area of the facility from any credible external source is discussed in its Section 2.2.3, which states:

Regulatory Guide 1.91 cites 1 pound per square inch differential pressure (psid)

(6.9 kilopascal [kPa]) as a conservative value of peak positive incident overpressure, below which no significant damage would be expected. Regulatory Guide 1.91 defines this standoff distance by the relationship R kW1/3 where R is the distance in feet from an exploding charge of W pounds of trinitrotoluene (TNT); and the value k is a constant.

The TNT mass equivalent, W, was determined by comparing the heat of combustion of the chemical to the heat of combustion of TNT.

ALOHA was used to model the worst-case accidental vapor cloud explosion, including the standoff distances and overpressure effects at the nearest SHINE safety-related area.

Section 2.2.3 of the SHINE OLA states that in addition to multiple external explosion sources, their yield and overpressures on the SHINE facility were evaluated. It also states that a liquid nitrogen storage tank [is] located outside the facility buildings. The tank and its associated process piping are designed in accordance with applicable codes, including overpressure protection. The Section further states that safety-related areas are designed to withstand a peak positive overpressure of at least 1 psid (6.9 kPa) without loss of function/significant damage [] Conservative assumptions were used to determine a standoff distance, or minimum separation distance, required for an explosion to have less than 1 psid (6.9 kPa) peak incident pressure.

Section 3.4.5.2 of the SHINE OLA then concludes by stating [t]he seismic area is protected by outer walls and roofs consisting of reinforced concrete robust enough to withstand credible external explosions, as defined in RG 1.91, Revision 2.

It is not clear what guidance, or applied safety factors, were considered to increase the degree of conservatism for the blast loads applied to the FSTR in order to account for the uncertainties involved in calculating the TNT equivalent mass for each evaluated chemical explosion and the standoff distance for 1 psid incident overpressure. It is also not clear whether the applicant used reflected peak pressure for the nearby chemical explosions and associated impulse for the review of FSTR seismic design effectiveness to resist blast loads. In addition, it is not clear what codes have been used for the design of the external nitrogen tank in proximity to the FSTR for overpressure protection and whether a consideration was given for additional blast loads to the FSTR, in case of its accidental explosion. This information is requested to verify that SHINE has performed the necessary evaluations required to show that safety functions will be accomplished, as required by 10 CFR 50.34(b)(2).

(1) Clarify how it was concluded that the FSTR reinforced concrete [seismic design is] robust enough to withstand credible external explosions, given the uncertainties involved in calculating external blast loads, their time scale in comparison to those associated with a seismic disturbance, and the philosophical differences between the approaches for seismic and blast load designs. State what specific design guidance was followed, safety factors applied, or specific analyses performed to reach that conclusion.

Page 118 of 130 (2) State what codes have been used for the design of the external nitrogen tank in proximity to the FSTR for overpressure and fragment protection of safety-related areas when evaluating adequacy of the FSTR seismic design, in case of accidental tank explosion.

Update the FSAR as appropriate to reflect the above requested information.

SHINE Response The SHINE Response to RAI 3.4-17 will be provided by January 31, 2021.

Page 119 of 130 Chapter 8 - Electrical Power Systems RAI 8-1 Section 8a2.1, Normal Electrical Power Supply System, of the SHINE FSAR provides a general description of the SHINE normal electrical power supply system (NPSS).

Section 8a2.1.1, Design Basis, states that:

The design of the NPSS provides sufficient, reliable power to facility and site electrical equipment as required for operation of the SHINE facility and to comply with applicable codes and standards.

SHINE states that National Fire Protection Association (NFPA) 70-2017, National Electrical Code (NEC) is used as the code for the design of the NPSS. However, it is not clear to the NRC staff to what extent SHINE is applying or taking exception to NFPA 70-2017 and other referenced standards in the design of its NPSS and emergency electrical power systems.

Additionally, during the May 11 to May 15, 2020 regulatory audit of SHINEs electrical power systems, SHINE indicated that it intends to partially conform to Regulatory Guide (RG) 1.180, Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation and Control Systems, which provides guidance to licensees and applicants on additional methods acceptable to the NRC staff for addressing the effects of electromagnetic interference and radiofrequency interference (EMI/RFI) and power surges on safety-related instrumentation and control (I&C) systems. However, it is not clear to the NRC staff to what extent SHINE is applying or taking exception to this regulatory guide. It is also not clear to the NRC staff how use of the NEC and other referenced standards satisfy SHINEs design criteria 27 and 28.

Provide additional detail on how SHINE is applying codes and standards to the design of its NPSS and emergency electrical power system. Specifically, provide references in the FSAR to documents that calculate and/or evaluate electrical design such that correlation is evident that demonstrates how the design of its NPSS and emergency electrical power system satisfy its principal design criteria 27 and 28. Such information could include descriptions of how standards, calculations, methodologies, and analyses are used in order to determine whether the design of the electrical systems meet the applicable regulations and is commensurate with the design bases of the facility. Clarify what calculations and studies were performed. If SHINE is not performing one or more the following calculations, provide justification why the calculation or study is not applicable for the electrical design of the SHINE facility:

Load Flow/Voltage Regulation Studies and Under/Overvoltage Protection; Short-Circuit Studies (alternating current (AC) and direct current (DC) systems),

including faults on cables in the penetrations to ensure that confinement integrity is maintained; Equipment Sizing Studies; Equipment Protection and Coordination Studies; Insulation Coordination (Surge and Lightning Protection);

Power Quality Limits (Harmonic Analysis);

Grounding Grid studies; Grid Stability studies; and Electromagnetic interference and radiofrequency interference, including conformance to RG 1.180, as applicable.

Page 120 of 130 This information is important for the NRC staff to determine how SHINE is satisfying its design criteria 27 and 28. The above is a list of specific calculations of interest to the NRC staff that would assist in the evaluation of SHINEs electrical design to ensure that on-site uninterruptible electric power supply and distribution system has sufficient independence, redundancy, testability, capacity, and capability to perform its safety functions consistent with SHINEs design criterion 27.

SHINE Response The SHINE Response to RAI 8-1 will be provided by January 31, 2021.

RAI 8-2 Section 8a2.1.3, Normal Electrical Power Supply System Description, provides a description of the protection of safety-related systems, which includes undervoltage trip enclosed breakers for the Neutron Driver Assembly System (NDAS), the vacuum transfer system (VTS), extraction feed pumps in the molybdenum extraction and purification system (MEPS), and the radiological ventilation exhaust fans (RVZ1, RVZ2, and RVZ3). Figure 8a2.1-1, Electrical Distribution System (Simplified), provides a simplified diagram of the overall electrical power supply system. The diagram shows two safety-related breakers connected to the non-safety-related NDAS. Section 8a2.1 of the FSAR states the following:

The NPSS is sized for safe operation of the facility. The largest loads on the NPSS are the process chilled water system (PCHS), neutron driver assembly system (NDAS), and the facility chilled water system (FCHS); however, those loads are not required for safe shutdown of the facility. Refer to Section 8a2.2 for a tabulation of emergency electrical load requirements.

Section 8a2.1.3, Normal Electrical Power Supply System Description, provides a list of safety related equipment in the NPSS. However, it is not clear to the NRC staff why two safety related breakers are connected to a non-safety-related NDAS, the VTS, the MEPS, and the RVZs.

Provide a detailed description of why the two circuit breakers connected to the systems mentioned above are categorized as safety-related, why the safety related breakers are specified only for undervoltage protection, and how these circuit breakers are important to providing and maintaining a safe shutdown condition of the facility. This information is necessary for the NRC staff to determine how SHINE is satisfying its design criteria 27 and 28.

Update the FSAR, as necessary.

SHINE Response The safety functions performed by the safety-related breakers specified in Subsection 8a2.1.3 of the FSAR are related to preventing actions that could initiate or increase the consequences of an accident. The equipment tied to these breakers does not perform an active safety function.

Redundant breakers are provided to ensure that the safety function can still be performed in the event of a single active failure. The following is a description of each of the safety functions being performed by the respective breakers:

Page 121 of 130 Redundant breakers for the NDAS high voltage power supply (HVPS) are provided to ensure that the neutron driver is de-energized under certain conditions to prevent TSV overpower events, which may cause overheating of the target solution or exceed the TSV off-gas system (TOGS) design basis.

The vacuum transfer system (VTS) vacuum transfer pump breakers prevent the vacuum transfer pump from moving target solution during an accident scenario, limiting the amount of radioactive material that is potentially released.

The molybdenum extraction and purification system (MEPS) extraction feed pump breakers prevent the extraction feed pump from moving target solution during an accident scenario, limiting the amount of radioactive material that is potentially released.

The radiological ventilation zone 1 (RVZ1) and radiological ventilation zone 2 (RVZ2) exhaust fan breakers and RVZ2 supply fan breakers prevent damage or excessive leakage of the isolation dampers in the event of an RCA Isolation.

The NDAS breakers are opened as part of the sequence to reach a safe shutdown condition in an IU, as the unit is transitioned to Mode 3. The VTS, MEPS, and RVZ breakers are not involved in reaching a safe shutdown condition.

The safety-related breakers are not designed for undervoltage protection. The use of the word undervoltage refers to the undervoltage release on the control circuit, which opens the breaker when the TSV reactivity protection system (TRPS) or engineered safety features actuation system (ESFAS) determine that trip conditions exist. Subsections 7.4.4.1 and 7.5.4.1 of the FSAR provide a discussion of the monitored variables used to determine trip conditions. An undervoltage release was selected over a shunt trip because it results in the breakers failing to a safe position if there is a failure in the control circuit.

Subsection 8a2.1.3 has been revised to remove the misleading characterization of the safety-related breakers and to clarify the functions performed by the safety-related breakers.

A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 8-3 Section 8a2.2, Emergency Electrical Power System, states the following:

The emergency electrical power systems for the SHINE facility consist of the safety-related uninterruptible electrical power supply system (UPSS),

the nonsafety-related standby generator system (SGS), and nonsafety-related local power supplies and unit batteries. The UPSS provides reliable power for the safety-related equipment required to prevent or mitigate the consequences of design basis events.

Section 8a2.2.2, Uninterruptible Electrical Power Supply System Codes and Standards, provides the list of standards used for the design of the UPSS. However, SHINE does not provide standards used for the maintenance, testing, installation and qualification for the safety-related batteries used in the DC system. In addition, for the battery chargers, maintenance, testing, and qualification of the battery chargers is not addressed in the FSAR.

Page 122 of 130 Describe the standards and/or methodologies used to perform maintenance, testing, installation, and qualification for the safety-related batteries in the DC system used in the UPSS. In addition, Describe the maintenance, testing, and qualification of the battery chargers. This information is necessary for the NRC staff to determine how SHINE is satisfying its design criteria 27 and 28.

SHINE Response The SHINE Response to RAI 8-3 will be provided by January 31, 2021.

RAI 8-4 It is not clear to the NRC staff how SHINE is applying its Principal Design Criterion 4, Environmental and dynamic effects, to the safety-related SSCs associated with its electrical power systems. This information is necessary for the NRC staff to ensure that the SHINE facility will be maintained in a safe condition during and following design-basis events.

Provide information describing how SHINE will apply its Principal Design Criterion 4 for the environmental qualification of electrical equipment. In addition, provide a list of equipment or the types of equipment that will be qualified, including the environmental conditions to which the equipment will be subjected. Indicate any methodologies and standards used for the environmental qualification of electrical equipment. Update the FSAR, as necessary.

SHINE Response Safety-related SSCs associated with the electrical power systems are located in a mild environment, are not subject to harsh environmental conditions during normal operation or transient conditions, and have no significant aging mechanisms. SHINE applies the guidance in Sections 4.1, 5.1, 6.1, and 7 of the Institute of Electrical and Electronics Engineers (IEEE)

Standard 323-2003, IEEE Standards for Qualifying Class 1E Equipment for Nuclear Power Generating Stations (Reference 32), in the design and qualification of safety-related electrical equipment. By applying these sections of this standard, SHINE ensures that the safety-related SSCs associated with the electrical power systems are designed and qualified to meet the requirements of SHINE Design Criterion 4.

The safety-related SSCs associated with the Normal Electrical Power Supply System (NPSS) consist of the safety-related breakers identified in Subsection 8a2.1.3 of the FSAR. The equipment specifications for these safety-related breakers explicitly state the environmental conditions (i.e., normal and transient temperature, pressure, and relative humidity) for which the equipment is required to operate within. These safety-related breakers are qualified to the environmental parameters provided in Tables 7.2-2 and 7.2-3 of the FSAR.

The safety-related portions of the uninterruptible electrical power supply system (UPSS) consist of distribution equipment, inverters, bypass transformers, battery chargers, and batteries. The equipment specifications for the safety-related UPSS equipment explicitly state the environmental conditions (i.e., normal and transient temperature, pressure, and relative humidity) for which the equipment is required to operate within. The safety-related UPSS equipment is qualified to the environmental parameters provided in Tables 7.2-2 and 7.2-3 of the FSAR.

Page 123 of 130 Specifications for safety-related NPSS and UPSS equipment describe the qualification and equipment interface requirements, safety functions, and the normal, abnormal, and design basis event conditions the equipment is required to function within.

SHINE has revised Sections 8a2.1 and 8a2.2 of the FSAR to describe the environmental qualification of safety-related electrical equipment. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 8-5 SHINE states in Section 8a2.1.3, Normal Electrical Power Supply System Description, that the NPSS operates as two separate branches, and that the branches automatically physically disconnect from the utility by opening the utility power supply breakers on a loss of phase, phase reversal, or sustained overvoltage or undervoltage as detected by protection relays for each utility transformer. However, SHINE does not address the electric power system design vulnerability to open phase conditions in the FSAR. This information is necessary to ensure that SHINE has designed its electrical power systems consistent with its Principal Design Criterion 27 to permit functioning of safety-related SSCs and minimize the probability of losing electric power from the uninterruptible power supply as a result of or coincident with, the loss of power from the off-site electric power system.

Provide additional information on how SHINE has considered the impact of open phase conditions on the safe operation of its facility, including clarification of the location of the loss of phase protection relays and whether there is an alarm in the control room to indicate an open phase condition (for reference, the NRC staff has considered electric power system design vulnerability to open phase conditions in offsite electric power systems at nuclear power plants in Bulletin 2012-01, Design Vulnerability in Electric Power System [ADAMS Accession No. ML12074A115] and subsequently issued Branch Technical Position 8-9, Open Phase Conditions in Electric Power System [ADAMS Accession No. ML15057A085], dated July 2015).

Update the FSAR, as necessary.

SHINE Response Electrical power is supplied to the main production facility from the utility via two transformers in the NPSS. This NPSS power supply to the main production facility operates as two branches that automatically physically disconnect from the utility by opening the utility power supply breakers. The NPSS monitors for loss of phase, for all three phases, at the line side of the incoming breakers using a negative sequence relay. Upon a loss of phase to any one of the three incoming phases, the respective utility supply breaker will open, and the control room will receive an alarm via the process integrated control system (PICS), as described in FSAR subsection 7.3.1.4.1.

Opening of a utility power supply breaker results in a loss of offsite power (LOOP) for the respective NPSS loads, as described in Subsection 8a2.1.6 of the FSAR. Although not required by accident analysis, the standby generator system (SGS) will start and provide a nonsafety-related source of backup electrical power to emergency buses. In the event of a LOOP in conjunction with SGS unavailability, the UPSS provides reliable power for safety-related equipment required to prevent or mitigate the consequences of design basis events.

Page 124 of 130 SHINE has revised Subsection 8a2.1.3 of the FSAR to describe the loss of phase protection provided for the NPSS. A mark-up of the FSAR incorporating these changes is provided as.

RAI 8-6 Section 8a2.2, Emergency Electrical Power System, states, The UPSS consists of a 125-volt direct current (VDC) battery subsystem, inverters, bypass transformers, distribution panels, and other distribution equipment necessary to feed safety-related alternating current (AC) and direct current (DC) loads and select non-safety-related AC and DC loads. However, SHINE does not provide a description of the technical specifications for the electrical equipment comprising the UPSS. This information is necessary for the NRC staff to determine how SHINE is satisfying its design criteria 27 and 28.

Provide a description of the specifications for the electrical equipment comprising the UPSS.

The information should include voltage, current, and frequency specifications including acceptable tolerances for these parameters. In addition, provide a description of how SHINE will ensure the failure of nonsafety-related loads do not impact safety-related loads. Update the FSAR and technical specifications, as necessary.

SHINE Response The electrical equipment comprising the UPSS consists of distribution equipment, inverters, bypass transformers, battery chargers, and batteries. Specifications for this equipment, including voltage, current, and frequency where applicable, are provided below:

Batteries Each division battery has sufficient capacity to support a 160 kilo-volt-amperes (kVA) alternating current (AC) load and a 45 kVA direct current (DC) load. As described in Subsection 8a2.2.3 of the FSAR, UPSS batteries are sized using the sizing guidance provided in Sections 6.1.1, 6.2.1, 6.2.2, 6.2.3, 6.2.4, 6.3.2 and 6.3.3 of IEEE 485-2010, Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications (Reference 33). Table 8a2.2-2 of the FSAR provides a summary of the UPSS battery sizing. Table 8-6-1 provides detailed UPSS battery specifications, including battery voltage (VDC) and volts per cell (VPC).

Table 8-6-1: UPSS Battery Specifications Type Lead Calcium Number of Cells 60 per bank Nominal Voltage 125 VDC Minimum Voltage 1.75 VPC / 105 VDC Float Voltage 2.26 VPC / 135.6 VDC Equalizing Voltage 2.38 VPC / 142.8 VDC Specific Gravity 1.210 to 1.300 at 77°F Battery Chargers Battery chargers are capable of supplying AC and DC system loads and recharging a fully discharged battery within 10 to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. AC inverter loads are 160 kVA while DC loads are

Page 125 of 130 estimated to be 45 kVA. Battery charger sizing is based upon supplying the DC loads and AC loads via the inverter during normal operating conditions and charging current. The load, less charging current, is approximately 1730 DC amps. Charging current required to be supplied by the charger is calculated to be approximately 175 DC amps at the 10-hour rate. The battery charger is sized at 240 kilowatts (kW) to meet the system requirements. Table 8-6-2 provides detailed UPSS battery charger specifications.

Table 8-6-2: UPSS Battery Charger Specifications AC Input 480 VAC, 3-phase, 60 Hz Nominal DC Output 125 VDC DC Output Rating 240 kW Normal Charge Output Voltage 132-137 VDC Equalizing Charge Output Voltage 138-143 VDC Equalizing Charge Timer 0-100 hours Inverters The minimum DC input voltage is lower than the minimum battery voltage to allow for voltage drop from the battery to the input of the inverter. The AC inverter loads are 160 kVA.

Table 8-6-3 provides detailed UPSS inverter specifications.

Table 8-6-3: UPSS Inverter Specifications DC Input Range 100-143 VDC AC Output - Voltage 208Y/120 VAC +/- 2%

AC Output - Frequency 60 Hz +/- 1 Hz AC Output - Rating 160 kVA Distribution Panels The AC and DC distribution panels are provided with overcurrent breaker protection, remote trip capability, and manual reclosure.

DC distribution panels include:

125 VDC switchgear, with circuit breakers; 125 VDC motor control centers, with circuit breakers; and 125 VDC panelboards, with circuit breakers.

AC distribution panels include:

208Y/120 VAC motor control centers, with circuit breakers; and 208Y/120 VAC panelboards, with circuit breakers.

Bypass Transformers The bypass transformers provide an alternate power source for the AC distribution panels when the inverter is not available. Table 8-6-4 provides detailed UPSS bypass transformer specifications.

Page 126 of 130 Table 8-6-4: UPSS Bypass Transformer Specifications AC Input 480 VAC, 3-phase, 60 Hz Nominal AC Output - Voltage 208Y/120 VAC (grounded WYE)

AC Output - Rating 160 kVA (to match the inverter output capability)

A Division of UPSS is considered Operable per Limiting Condition of Operation (LCO) 3.6.1 of the technical specifications if the associated battery, battery charger, inverter, and distribution panels are Operable; the inverter is being supplied by the DC distribution panel and is supplying power to the AC distribution panel; and the battery and battery charger are connected to the DC distribution panel. In this configuration, the UPSS Division is capable of supplying required loads during a loss of offsite power. The UPSS bypass transformer is not required for Operability.

The parameters for UPSS equipment that are required to ensure technical specification Operability are voltage, frequency, and battery specific gravity. These parameters establish the lowest functional capability or performance levels for safe operation. SHINE has revised LCO 3.6.1 of the technical specifications to explicitly define the equipment required for a division of the UPSS to be considered Operable. A mark-up of the technical specifications, including bases, is provided as Attachment 2. SHINE will provide a revision to the technical specifications incorporating the mark-up by February 28, 2021.

Nonsafety-related loads which are fed from the safety-related UPSS equipment are required to be fed by an isolation device which meets the requirements of Section 6.1.2 of IEEE 384-2008, IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits (Reference 34).

This ensures that a failure of nonsafety-related loads does not impact safety-related loads.

SHINE has revised Subsection 8a2.2.3 of the FSAR to address the isolation of nonsafety-related and safety-related loads. A mark-up of the FSAR incorporating these changes is provided as Attachment 1.

RAI 8-7 Section 8a2.2.2, Uninterruptible Electrical Power System Codes and Standards, states the UPSS is designed in accordance with IEEE Standard 384-2008, Standard Criteria for Independence of Class 1E Equipment. IEEE Std. 384-2008, Section 3.6 defines Class 1E as, the safety classification of the electric equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or are otherwise essential in preventing a significant release of radioactive material to the environment. For electrical systems, the staff considers safety-related and Class 1E as synonymous terms and classifications. However, it is unclear to the NRC staff whether SHINE is classifying its UPSS as Class 1E. This information is necessary for the NRC staff to determine how SHINE is satisfying its design criteria 27 and 28.

Clarify whether SHINE classifies the UPSS as Class 1E. If the UPSS is not considered Class 1E, describe why not and how the criteria or standards, including IEEE Std. 384-2008, are applied to the design and classification of the UPSS. Update the FSAR, as necessary.

SHINE Response SHINE does not classify the UPSS as a Class 1E system. The NRC has not endorsed conformance with IEEE standards related to Class 1E power systems in satisfying the NRCs

Page 127 of 130 regulations with respect to the design, operation, and testing of safety-related power systems at non-power production and utilization facilities, like the NRC has for nuclear power plants.

Additionally, the NRC has previously taken the position that NRC regulations do not require the use of Class 1E equipment at non-power production and utilization facilities (References 35 and 36).

IEEE Standard 384-2008 (Reference 34) provides a power reactor-specific classification for Class 1E equipment, while SHINE uses a facility-specific classification for safety-related SSCs, including electrical power systems, as described in Section 3.1 of the FSAR. Application of this SHINE-specific classification for safety-related SSCs results in the identification of those physical SSCs whose intended functions are to prevent accidents that could cause undue risk to health and safety of workers and the public, and to control or mitigate the consequences of such accidents.

While SHINE does not classify the UPSS as a Class 1E system and apply the full-scope of Class 1E-related standards to the UPSS, portions of Class 1E-related standards are applied to the design of the UPSS in order to satisfy applicable SHINE design criteria. The SHINE Response to RAI 8-1 provides additional detail on how codes and standards, including those portions of Class 1E-related standards, are applied to the SHINE design of the UPSS to satisfy the applicable SHINE design criteria.

SHINE has revised Subsection 8a2.2.2 of the FSAR to clarify that the UPSS is not classified as a Class 1E system. A mark-up of the FSAR incorporating these changes is provided as.

Page 128 of 130 References

1. NRC letter to SHINE Medical Technologies, LLC, Issuance of Request for Additional Information Related to the SHINE Medical Technologies, LLC Operating License Application (EPID No. L-2019-NEW-0004), dated October 16, 2020

2. SHINE Medical Technologies, LLC letter to the NRC, SHINE Medical Technologies, LLC Application for an Operating License, dated July 17, 2019 (ML19211C143)

3. U.S. Nuclear Regulatory Commission, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Format and Content," NUREG-1537, Part 1, February 1996 (ML042430055)

4. U.S. Nuclear Regulatory Commission, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review Plan and Acceptance Criteria,"

NUREG 1537, Part 2, February 1996 (ML042430048)

5. U.S. Nuclear Regulatory Commission, Barrier Design Procedures, NUREG-0800, Section 3.5.3, March 2007 (ML070570004)

6. U.S. Nuclear Regulatory Commission, Seismic Design Parameters, NUREG-0800, Section 3.7.1, December 2014 (ML14198A460)

7. U.S. Nuclear Regulatory Commission, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61, Revision 1, March 2007 (ML070260029)

8. U.S. Nuclear Regulatory Commission, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Regulatory Guide 1.122, Revision 1, February 1978 (ML003739367)

9. U.S. Nuclear Regulatory Commission, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Regulatory Guide 1.92, Revision 3, October 2012 (ML12220A043)

10. American Society of Civil Engineers, Seismic Analysis of Safety-Related Nuclear Structures and Commentary, ASCE 4-98, January 2000

11. Federal Highway Administration, Geotechnical Engineering Circular No. 6 - Shallow Foundations, Report No. FHWA-SA-02-054, September 2002

12. International Atomic Energy Agency, Consideration of External Events in the Design of Nuclear Facilities other than Nuclear Power Plants, with Emphasis on Earthquakes, IAEA-TECDOC-1347, 2003

13. U.S. Nuclear Regulatory Commission, Seismic System Analysis, NUREG-0800, Section 3.7.2, September 2013 (ML13198A223)

14. U.S. Nuclear Regulatory Commission, Design Response Spectra for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.60, Revision 2, July 2014 (ML13210A432)

Page 129 of 130

15. American Society of Civil Engineers/Structural Engineering Institute, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE/SEI 43-05, Reston, VA

16. American Concrete Institute, Code Requirements for Nuclear Safety-Related Concrete Structures & Commentary, ACI 349-13, 2014

17. American Concrete Institute, Building Code Requirements for Structural Concrete and Commentary, ACI 318-08, 2008

18. American National Standards Institute/American Institute of Steel Construction, Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC N690-12, 2012

19. U.S. Department of Energy, Accident Analysis for Aircraft Crash into Hazardous Facilities, DOE-STD-3014-2006, 2006

20. U.S. Nuclear Regulatory Commission, A Review of the Effects of Radiation on Microstructure and Properties of Concretes Used in Nuclear Power Plants, NUREG/CR-7171, November 2013

21. American Concrete Institute, Report on Evaluation and Repair of Existing Nuclear Safety-Related Concrete Structures, ACI 349.3R-18, Farmington Hills, MI

22. American National Standards Institute/American Nuclear Society, Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants, ANSI/ANS-6.4-2006 (R2016), La Grange Park, IL

23. Radiation Effects Information Center, The Effects of Neutron Radiation on Structural Materials, REIC Report No. 45, Kangilaski, M., June 30, 1967

24. National Aeronautics and Space Administration, "Radiation Effects Design Handbook:

Section 7. Structural Alloys", NASA CR-1873, Kangilaski, M., October 1971

25. American Society of Mechanical Engineers, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), ASME NOG-1-2004, 2004

26. Crane Manufactures Association of America, Inc., Specifications for Top Running Bridge &

Gantry Type Multiple Girder Electric Overhead Traveling Cranes, CMAA 70-2004, 2004

27. American Society of Mechanical Engineers, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist), ASME B30.2-2011, 2011

28. American National Standards Institute, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More, ANSI N14.6-1993, 1993

29. United States Atomic Energy Commission, Nuclear Reactors and Earthquakes, TID-7024, August 1963

30. American Concrete Institute, Seismic Design of Liquid-Containing Concrete Structures and Commentary, ACI 350.3-06, 2006

Page 130 of 130

31. Lawrence Livermore National Laboratory, Structures, Systems, and Components Evaluation Technical Support Documents, UCRL-ID-123577, 1997

32. Institute of Electrical and Electronics Engineers, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, IEEE Std. 323-2003, New York, NY

33. Institute of Electrical and Electronics Engineers, Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications, IEEE Std. 485-2010, New York, NY

34. Institute of Electrical and Electronics Engineers, IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits, IEEE Std. 384-2008, New York, NY

35. NRC letter to Purdue University, dated April 1, 2019, Purdue University - Issuance of Amendment No. 14 to Renewed Facility Operating License No. R-87 for the Purdue University Research Reactor Facility Digital Instrumentation and Controls Upgrade (EPID No. L-2017-LLA-0251) (ML18275A090)

36. NRC letter to Massachusetts Institute of Technology, dated December 4, 2019, Massachusetts Institute of Technology - Issuance of Amendment No. 42 to Renewed Facility Operating License No. R-37 for the Massachusetts Institute of Technology Research Reactor Regarding the Nuclear Safety System Digital Upgrade (EPID No. L-2016-LLA-0003) (ML19123A212)

39 pages follow ENCLOSURE 3 ATTACHMENT 1 SHINE MEDICAL TECHNOLOGIES, LLC SHINE MEDICAL TECHNOLOGIES, LLC OPERATING LICENSE APPLICATION SUPPLEMENT NO. 6 AND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FINAL SAFETY ANALYSIS REPORT CHANGES PUBLIC VERSION (MARK-UP)

Chapter 2 - Site Characteristics Hydrology SHINE Medical Technologies 2.4-7 Rev. 2 2.4.2.3 Effect of Local Intense Precipitation The effect of the local probable maximum precipitation (PMP) on the areas adjacent to safety-related structures of the facility, including the drainage from the roofs of the structures, was evaluated. The maximum water levels due to local PMP were estimated near the safety-related structures of the facility based on the site topographic survey map.

All elevations in this subsection are referenced to the NAVD 88.

A drainage system designed to carry runoff from the site up to a 100-year precipitation event consists of conveying water from roofs, as well as runoff from the site and adjacent areas, to peripheral ditches. The facility is surrounded by berms with interior ditches along the berms. The plant site is graded such that the high point of grade is set at Elevation 827 ft. (252.1 m). The grade around the structures slopes towards the peripheral ditches. The storm water drains into the peripheral ditches. A plan showing the delineated off-site drainage area is presented in Figure 2.4-11. Peripheral diversion swales and berms north and east of the site are provided to divert the off-site runoff around the facility area. During a local PMP event, the storm water drainage system is conservatively assumed to be not functional. No active surface water drainage waterway exists which flows towards the site. PMP runoff from the off-site area northeast of the site flows towards the site. The off-site area is relatively flat.

The finished site grade elevation is approximately 825 ft. (251.46 m), and the top of the finished foundation elevation is at least 4 inches (in.) above grade; therefore, water will not infiltrate the door openings in the case of a local PMP event.

The site is designed to withstand the effects of a local probable maximum precipitation (PMP),

defined as the rainfall values and intensities associated with a 1-in-100-year rainfall event. The PMP values and intensities associated with a 1-in-100-year rainfall event are provided in Table 2.4-7. The PMP values and intensities were determined from the 100-year rainfall intensity-duration-frequency curve for Madison, Wisconsin (WDOT, 1979) (Figure 2.4-10).

The effect of the PMP event on the areas adjacent to safety-related structures of the facility, including the drainage from the roofs of the structures, was evaluated. The maximum water levels due to local PMP were determined near the safety-related structures of the facility based on site topographic survey maps.

The site is protected from PMP flooding by a developed drainage channel on the north and east sides of the site and an existing drainage channel east and southeast of the site (Figure 2.4-11)

(Figure 2.4-12). Off-site runoff approaches the site from the north or northeast (Figure 2.4-11).

The developed drainage channel on the north and east sides of the facility directs off-site runoff away from the facility. Off-site runoff that flows from the north towards the site is captured by the channel which directs flow to an uncontrolled sub-basin on the west side of the site (Figure 2.4-11). The runoff flow rate was calculated as 42 cfs. The upstream bank elevation of the channel is 827 ft. The channel is approximately 1100 ft. long with a 0.8% slope. In the event of a 100-year storm, the water surface elevation at the upstream end of the channel reaches a maximum height of 826.3 ft., which is below the bank height.

Off-site runoff that flows from the northeast towards the site is captured by an existing channel southeast of the site that flows to an unnamed tributary approximately one mile south of the site

Chapter 2 - Site Characteristics Hydrology SHINE Medical Technologies 2.4-45 Rev. 2 Figure 2.4 PMP Site Drainage Area Tsc Tch Ts Tsc Ts Tch Tch Tch Tsc WaA WaA WaA WaA WaA WaA WaA WaA WaB WaB WaB WaB WaB WaB WaB WaB LoC2 LoC2 LoC2 LoC2 WaB N. RIVERSIDE DRIVE -

SOUTH US HWY 51 Tch POI 1 Tch Ts Tsc EXISTING CHANNEL

Chapter 2 - Site Characteristics Hydrology SHINE Medical Technologies 2.4-46 Rev. 2 Figure 2.4 PMP 100-Year Event Facility Drainagea

a. Figure displays location of (sSix) localized low points (inlets 3A, 6, 12, and 13 and trench drains 8 and 9) are subject to impoundment if drainage is assumed blocked. Surface elevation of impounded areas during a 100-year PMP event remain below the ground floor elevation of the main production facility and material staging building.

12 6

9 3A 8

823.00 822.19 824.30 823.60 824.10 820.66 821.10 818.61 819.10 822.70 823.30

Chapter 3 - Design of Structures, Systems, and Components Meteorological Damage SHINE Medical Technologies 3.2-1 Rev. 2 3.2 METEOROLOGICAL DAMAGE 3.2.1 WIND LOADING This subsection discusses the criteria used to design the main production facility for protection from wind loading conditions.

3.2.1.1 Applicable Design Parameters The main production facility structure is designed to withstand wind pressures based on a basic wind velocity of 90 miles per hour (mph) (145 kilometers per hour [kph]) adjusted for a mean recurrence interval of 100 years, per Figure 6-1 and Table C6-7 of American Society of Civil Engineers/Structural Engineering Institute (ASCE), Standard 7-05, Minimum Design Loads for Buildings and Other Structures (ASCE, 2006).

3.2.1.2 Determination of Applied Forces The design wind velocity is converted to velocity pressure in accordance with Equation 6-15 of ASCE 7-05 (ASCE, 2006):

qz = 0.00256KzKztKdV2I (pounds per square foot [lb/ft2])

(Equation 3.2-1)

Where:

Kz

= velocity pressure exposure coefficient evaluated at height (z) in Table 6-3 of ASCE 7-05 equal to 1.13 Kzt

=

topographic factor as defined in Section 6.5.7 of ASCE 7-05 equal to 1.0 Kd

=

wind directionality factor in Table 6-4 of ASCE 7-05 equal to 0.85 V

=

basic wind speed (3-second gust) obtained from Figure 6-1 of ASCE 7-05 for Wisconsin equal to 90 mph and increased by a factor of 1.07 to account for a 100-year recurrence interval I

= importance factor =equal to 1.15 Additional discussion of site design parameters related to wind loading is provided in Subsection 3.4.2.6.3.7.

The design wind pressures and forces for the building at various heights above ground are obtained in accordance with Section 6.5.12.2.1 of ASCE 7-05 (ASCE, 2006) by multiplying the velocity pressure by the appropriate pressure coefficients, gust factors, accounting for sloped surfaces (i.e., the roof of the building). The building is categorized as an enclosed building according to Section 6.2 of ASCE 7-05 (ASCE, 2006) and, as a result, both external and internal pressures are applied to the structure. A positive and negative internal pressure is applied to the internal surfaces of the exterior walls as well as the roof.

For external wind pressures, a Gust Effect Factor (G) of 0.85 for rigid structures is used per Section 6.5.8.1 of ASCE 7-05. External pressure coefficients are determined for windward,

Chapter 3 - Design of Structures, Systems, and Components Meteorological Damage SHINE Medical Technologies 3.2-2 Rev. 2 leeward, and roof wind pressures according to Figure 6-6 of ASCE 7-05 (ASCE, 2006). For internal wind pressures, an internal pressure coefficient (GCpi) of +/-18 for enclosed buildings is used per Figure 6-5 of ASCE 7-05. Wind pressures are combined and iterated in multiple load cases to ensure the worst-case wind loading is considered in the building design.

3.2.2 TORNADO LOADING This subsection discusses the criteria used to design the main production facility to withstand the effects of a design-basis tornado phenomenon.

3.2.2.1 Applicable Design Parameters The design-basis tornado characteristics are described in Regulatory Guide 1.76, Design Basis Tornado for Nuclear Power Plants (USNRC, 2007a):

a.

Design-basis tornado characteristics are listed in Table 1 of Regulatory Guide 1.76 for Region I.

b.

The design-basis tornado missile spectrum and maximum horizontal missile speeds are given in Table 2 of Regulatory Guide 1.76.

3.2.2.2 Determination of Applied Forces The maximum tornado wind speed is converted to velocity pressure in accordance with Equation 6-15 of ASCE 7-05 (ASCE, 2006):

qz = 0.00256KzKztKdV2I (lb/ft2)

(Equation 3.2-2)

Where:

Kz

=

velocity pressure exposure coefficient equal to 0.87 Kzt

=

topographic factor equal to 1.0 Kd

= wind directionality factor equal to 1.0 V

=

maximum tornado wind speed equal to 230 mph (370 kph) for Region I I

=

importance factor equal to 1.15 Additional discussion of site design parameters related to tornado loading is provided in Subsection 3.4.2.6.3.8.

The tornado differential pressure is defined in Regulatory Guide 1.76, Table 1 as 1.2 pounds per square inch (psi) (8.3 kilopascals [kPa]) for Region I (USNRC, 2007a). The tornado differential pressure is applied as an outward pressure to the exterior walls of the building, as well as the roof, because the structure is categorized as an enclosed building in accordance with Section 6.2 of ASCE 7-05 (ASCE, 2006).

The procedure used for transforming the tornado-generated missile impact into an effective or equivalent static load on the structure is consistent with NUREG-0800, Standard Review Plan for

Chapter 3 - Design of Structures, Systems, and Components Meteorological Damage SHINE Medical Technologies 3.2-3 Rev. 2 the Review of Safety Analysis for Nuclear Power Plants (SRP) Section 3.5.23, Subsection II (USNRC, 2007b). Tornado missile loading applied to the structure is derived considering the effects of the following missile tyeps and maximum horizontal speeds:

Schedule 40 pipe at 135 feet per second (ft./sec)

Automobile (4000 lb.) at 135 ft./sec Solid steel sphere at 26 ft./sec The loading combinations of the individual tornado loading components and the load factors are in accordance with SRP Section 3.3.2 (USNRC, 2007c).

3.2.2.3 Effect of Failure of Structures, Systems, or Components Not Designed for Tornado Loads SSCs whose failure during a tornado event could affect the safety-related portions of the facility are either designed to resist the tornado loading or the effect on the safety-related structures from the failure of these SSCs or portions thereof are shown to be bounded by the tornado missile or aircraft impact evaluations.

The Seismic Category I boundary provides missile walls to protect safety-related systems from damage due to tornado missiles. SSCs that are credited to prevent or mitigate potential accidents caused by a tornado event are protected by the design of the enclosed structure. The structural analysis does not credit venting of the Seismic Category I boundary during a tornado event. The differential pressure on all surfaces as an enclosed structure results in higher pressures, and the differential pressure would be reduced by the effects of venting. Therefore, there are no consequences to venting the building during a tornado event.

3.2.3 SNOW, ICE, AND RAIN LOADING This subsection discusses the criteria used to design the main production facility to withstand conditions due to snow, ice, and rain loading. Rain loading is not considered in the structural design of the building as the sloped roofs do not result in rain accumulation. As a result of the lack of rain accumulation, load due to ice is anticipated to be minimal and is enveloped by the design snow load.

3.2.3.1 Applicable Design Parameters Snow load design parameters pertinent to the main production facility are provided in Chapter 7 of ASCE 7-05 (ASCE, 2006) and adjusted for a mean recurrence interval of 100 years, per Table C7.3 of ASCE 7.05 (ASCE, 2006).

3.2.3.2 Determination of Applied Forces The sloped roof snow load is calculated in accordance with Sections 7.3 and 7.4 of ASCE 7-05 (ASCE, 2006). The combined equation utilized to calculate the sloped roof load is:

ps = 0.7CsCeCtIpg (Equation 3.2-3)

Chapter 3 - Design of Structures, Systems, and Components Meteorological Damage SHINE Medical Technologies 3.2-4 Rev. 2 Where:

Cs

=

roof slope factor as determined by Sections 7.4.1 through 7.4.4 of ASCE 7-05 equal to 1.0 Ce

=

exposure factor as determined by Table 7-2 of ASCE 7-05 equal to 1.0 Ct

=

thermal factor as determined by Table 7-3 of ASCE 7-05 equal to 1.0 I

= importance factor as determined by Table 7-4 of ASCE 7-05 equal to 1.2 pg

= ground snow load as set forth in Figure 7-1 of ASCE 7-05 equal to 30 pounds per square foot (psf) and increased by a factor of 1.22 to account for a 100-year recurrence interval Additional discussion of site design parameters related to snow loading is provided in Subsection 3.4.2.6.3.4.

Unbalanced roof snow loads are computed in accordance with Section 7.6 of ASCE 7-05 (ASCE, 2006). The design snow drift surcharge loads are computed in accordance with Section 7.7.1 of ASCE 7-05 (ASCE, 2006).

Chapter 3 - Design of Structures, Systems, and Components Water Damage SHINE Medical Technologies 3.3-2 Rev. 2 3.3.1.1 Flood Protection Measures for Structures, Systems, and Components Postulated flooding from component failures in the building compartments is prevented from adversely affecting plant safety or posing any hazard to the public. Exterior or access openings and penetrations into the main production facility are above the maximum postulated flooding level and thus do not require protection against flooding.

3.3.1.1.1 Flood Protection from External Sources Safety-related components located below the design (PMP) flood level are protected using the hardened protection approach described below. The safety-related systems and components are flood-protected because they are enclosed in a reinforced concrete safety-related structure, which has the following features:

a.

Exterior walls below flood level are not less than 2 ft. (0.61 m) thick.

b.

Water stops are provided in construction joints below flood level.

c.

Waterproofing is applied to external surfaces exposed to flood level.

d.

Roofs are designed to prevent pooling of large amounts of water.

Waterproofing of foundations and walls of Seismic Category I structures below grade is accomplished principally by the use of water stops at construction joints.

In addition to water stops, waterproofing of the main production facility is provided up to 4 in.

(10.2 cm) above the plant ground level to protect the external surfaces from exposure to water.

There is no fire protection piping in the RCA general area.

3.3.1.1.2 Flood Protection from Internal Sources Fire suppression systems within the RCA consist of manual discharge via fire hoses from dry standpipes, except in those areas of the RCA in which gaseous fire suppression is provided, as described in Section 9a2.3. The total discharge from the fire protection discharge consists of the combined volume from any firefighting hoses. In accordance with National Fire Protection Association (NFPA) 801, Section 5.10 (NFPA, 2008), the credible volume of discharge is sized for a manual fire-fighting flow rate of 500 gallons per minute (1893 liters per minute) for a duration of 30 minutes (min.). Therefore, the total discharge volume is 15,000 gallons (56,782 liters). The resulting flooded water depth in the RCA from fire protection discharge is less than 2 in. This bounds the total water available in the PCHS and RPCS cooling systems that could cause internal flooding. When the total discharge volume of fire water is distributed over the entire RCA, the depth is less than 2 in. (5.1 cm). When the total discharge volume of fire water is distributed only over the minimum open floor area in the irradiation facility (IF), the depth is less than 12 in. (30.5 cm).

The safety-related function(s) of systems within the RCA that are subject to the effects of a discharge of the fire suppression system are appropriately protected by redundancy and separation. Where redundant equipment is unable to be effectively separated, fire response plans are established to ensure redundant trains of water sensitive safety-related equipment are not both subject to damage due to discharge of the fire suppression system. The floors of the URSS/TSPS rooms are elevated to prevent water intrusion in the event of an internal flood.

Water sensitive safety-related equipment is raised from the floor 8a minimum of 12 in. (230.35

Chapter 3 - Design of Structures, Systems, and Components Water Damage SHINE Medical Technologies 3.3-3 Rev. 2 cm) in the RCA, with the exception of the RPCS room, where water sensitive safety-related equipment is raised a minimum of 24 in. (61.0 cm) from the floor to provide defense in depth.

Therefore, the depth of water due to fire protection discharge is less than the elevation that water sensitive safety-related equipment is raised from the floor.

Outside of the RCA there is limited water discharge from fire protection systems. The safety-related function(s) of systems outside the RCA that are subject to the effects of a discharge of the fire suppression system are appropriately protected by redundancy and separation. The uninterruptible electrical power supply system (UPSS) has two trains to provide redundancy.

These trains are isolated from each other to prevent one train from being damaged by discharge of the fire protection system in the vicinity of the other train. Any water sensitive safety-related equipment outside the RCA is installed a minimum of 8 in. (20.3 cm) above the floor slab at grade.

Flood scenarios have been considered for the pipe trenches and vaults. Process piping, vessels, and tanks containing special nuclear material (SNM) or radioactive liquids are seismically qualified. There is no high-energy piping within these areas. Any pipe or tank rupture in the radioisotope production facility (RPF) vaults is routed to the radioactive drain system (RDS). The RDS is sized for the maximum postulated pipe or tank failure as described in Subsection 9b.7.6.

The design of the shield plugs over the pipe trenches and vaults prevents bulk leakage of liquid into the vaults from postulated flooding events within the remainder of the RCA.

The light water pool in the irradiation unit cell (IU) is filled to an elevation approximately equal to the top of the surrounding area floor slab. Given the robust design of the light water pool (approximately 4 ft. thick reinforced concrete) and the stainless steel liner, loss of a significant amount of pool water is not credible.

3.3.1.2 Permanent Dewatering System There is no permanent dewatering system provided for the flood design.

3.3.2 STRUCTURAL DESIGN FOR FLOODING Since the design PMP elevation is at the finished plant grade and the PMF elevation is approximately 50 ft. (15.2 m) below grade, there is no dynamic force due to precipitation or flooding.

The load from build-up of water due to discharge of fire water in the RCA is supported by slabs on grade, with the exception of the mezzanine floor. Openings that are provided in the mezzanine ensure that the mezzanine slab is not significantly loaded. The mezzanine floor slab is designed to a live load of 250 pounds per square foot (1221 kilograms per square meter).

Therefore, the mezzanine floor slab is capable of withstanding temporary water collection that may occur while water is draining from the mezzanine floor.

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-1 Rev. 4 3.4 SEISMIC DAMAGE Seismic analysis criteria for the main production facility conform to IAEA-TECDOC-1347, Consideration of External Events in the Design of Nuclear Facilities other than Nuclear Power Plants, with Emphasis on Earthquakes (IAEA, 2003), which provides generic requirements and structure (FSTR) are supported by the detailed guidance for the seismic design of nuclear facilities other than nuclear power plants. Additional criteria provided inby the referenced Regulatory Guides and sections of NUREG-0800, Standard Review Plan for the Review of Safety Analysis for Nuclear Power Plants (SRP), provide more detailed guidance in the seismic analysis of the main production facility structure (FSTR).

The dimensions of the FSTR at grade level are approximately 212feet(ft.) (64.6 meters[m]) in the north-south (N/S) direction and 158ft. (48.2m) in the east-west (E/W) direction. The main production facility is a single-story building with a mezzanine, with a roof height of approximately 58ft. (17.7m). The FSTR also includes an exhaust stack with a height of approximately 67ft.

(20.4m). The main production facility main floor has below grade reinforced concrete vaults for housing equipment. The roof of the facility is supported by a steel truss system.

The FSTR building is a box-type shear wall system of reinforced concrete. The major structural elements include the foundation mat, mezzanine floor, roof slab supported by roof trusses, and shear walls. The exterior building walls of the majority of the FSTR are thick cast-in-place concrete, and are designed to protect the people, materials, and equipment inside the facility from natural and manmade accidents.

The FSTR includes the irradiation facility (IF), the radioisotope production facility (RPF), the non-radiologically controlled seismic area, and a nonsafety-related area. The IF contains the irradiation units (IUs) and tritium purification system(TPS), and the RPF contains the supercell and below-grade tanks. The non-radiologically controlled seismic area contains the control room, battery rooms, uninterruptible electrical power supply rooms, and other miscellaneous support rooms. The RPF, IF, and non-radiologically controlled seismic area are within the seismic boundary and are classified as Seismic Category I. These areas contain the safety-related structures, systems, and components (SSCs). To the south of the seismic boundary are the shipping and receiving areas, as well as other areas that contain nonsafety-related support systems and equipment. This part of the structure is not Seismic Category I. The areas outside the seismic boundary do not contain safety-related SSCs.

The IF, RPF, and non-radiologically controlled seismic area comprise the safety-related portion of the FSTR. The dimensions of the safety-related portion of the FSTR at grade level are approximately 212 feet (ft.) (64.6 meters [m]) in the north-south direction and 158 ft. (48.2 m) in the east-west direction. Each of the three main areas of the safety-related portion of the FSTR is a parallel, single-story box-type structure designed with cast-in-place reinforced concrete shear walls. The major structural elements include the foundation mat, mezzanine floor, roof slab, and shear walls. The roof slabs of the IF and RPF are supported by steel roof truss systems. The reinforced concrete mezzanine slab on metal deck is vertically supported by structural steel beams and columns, and laterally restrained by interior reinforced concrete walls. A large section of the basemat in the RPF is recessed below grade, where a series of below grade tanks, valve pits, and other mechanical systems are located. Each of these tanks is separated by cast-in-place reinforced concrete walls and is covered by a series of precast concrete shield plugs that create a removable slab at the same elevation as the rest of the basemat. Depending on their

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-2 Rev. 4 function, interior walls are cast-in-place reinforced concrete, reinforced masonry, or gypsum mounted to metal studs.

The dimensions of the nonsafety-related portion of the FSTR at grade are approximately 77 ft.

(23.5 m) in the north-south direction and 158 ft. (48.2 m) in the east-west direction. Additionally, the southwest corner of the safety-related basemat contains a part of the nonsafety-related portion of the FSTR. The dimensions of this nonsafety-related part are approximately 63 ft.

(19.2m) in the north-south direction and 32 ft. (9.8 m) in the east-west direction. The safety-related basemat and structures on the safety-related basemat are seismically isolated from the nonsafety-related portion of the FSTR.

The nonsafety-related portion of the FSTR is a two-story steel framed structure with a roof height of approximately 40 ft. (12.2 m). The concrete on metal deck mezzanine slab and metal deck roof slab are diaphragms that transfer the lateral loads to a series of vertical brace systems. The FSTR also includes a nonsafety-related, isolated, self-supporting steel on reinforced-concrete foundation cantilevered exhaust stack with a height of approximately 67 ft. (20.4 m) located east of the nonsafety-related portion of the FSTR.

The FSTR is modeled to the analyses described in this chapter. The concrete walls, slabs, and basemat are modeled using thick shell elements. The steel structural members are modeled using three-dimensional beam elements. Interior partition walls made of concrete are modeled using thick shell elements. Interior partition walls made of masonry or gypsum are isolated from the lateral load resisting system of the building and are not explicitly modeled, but their mass is accounted for. The excavated soil volume of the soil-structure interaction (SSI) analysis is modeled using solid elements. Seismic mass is considered in the model in accordance with SRP Section 3.7.2 (USNRC, 2013a). Figure3.4-1 and Figure3.4-2 provide three-dimensional views of the structural model.

Certain material in this section provides information that is used in the technical specifications, including conditions for operation and design features. In addition, significant material is also applicable to, and may be referenced by, the bases that are described in the technical specifications.

3.4.1 SEISMIC INPUT 3.4.1.1 Design Response Spectra The safe shutdown earthquake (SSE) ground motion is defined with a maximum ground acceleration of 0.2 g and design response spectra in accordance with Regulatory Guide 1.60, Revision 2, Design Response Spectra for Seismic Design of Nuclear Power Plants (USNRC, 2014a).

Consistent with SRP Section 3.7.2 (USNRC, 2013a), the location of the ground motion should be at the ground surface. The competent material (material with a minimum shear wave velocity of 1,000 feet per second [ft./sec] [305 meters per second {m/s}]) is 7.5 ft. (2.3 m) below the ground surface for the site. Hence, the SSE response spectra are defined as an outcrop at a depth of 7.5ft. (2.3 m) below grade.

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-3 Rev. 4 3.4.1.2 Design Time Histories For soil-structure interaction (SSI) analysis and for generating in-structure response spectra, design acceleration time histories are required. Synthetic acceleration time histories are generated to envelop the design response spectra. Mutually orthogonal synthetic acceleration time histories are generated for each horizontal direction and one for the vertical direction. Each of these time histories meets the design response spectra enveloping requirements consistent with Approach 2, Option 1 of SRP Section 3.7.1 (USNRC, 2014b). The specifics of each of these time histories are:

Each synthetic time history has been generated starting with seed recorded earthquake time histories.

The strong motion durations (Arias intensity to rise from 5 percent to 75 percent) of synthetic time histories are greater than a minimum of 6 seconds.

The time history has a sufficiently small increment and sufficiently long duration. Records shall have a Nyquist frequency of at least 50 hertz (Hz) and a total duration of at least 20seconds. The time step increment will be 0.005 seconds, which meets the Nyquist requirement for frequencies up to 100 Hz.

Spectral acceleration at 5 percent damping is computed at a minimum of 100 points per frequency decade, uniformly spaced over the log frequency scale from 0.1 Hz to 50 Hz or the Nyquist frequency.

Comparison of the response spectrum obtained from the synthetic time history with the target response spectrum shall be made at each frequency computed in the frequency range of interest.

The computed 5 percent damped response spectrum of the acceleration time history shall not fall more than 10 percent below the target response spectrum at any one frequency and shall have no more than 9 adjacent frequency points falling below the target response spectrum.

The computed 5 percent damped response spectrum of the artificial time history shall not exceed the target spectrum at any frequency by more than 30 percent in the frequency range of interest.

Comparison of the response spectra obtained from the artificial acceleration time histories with the target design response spectra illustrates that the enveloping criteria of SRP Section 3.7.1 (USNRC, 2014b) are satisfied. The seismic design parameters used in the seismic analysis of the FSTR, including the artificial acceleration time histories, target design response spectra, and response spectra obtained from artificial acceleration time histories, are consistent with the SRP Section 3.7.1 acceptance criteria.

3.4.1.3 Critical Damping Values Structural damping values for various structural elements used in the seismic analyses are provided in Section 1.1 of Regulatory Guide 1.61, Revision 1, Damping Values for Seismic Design of Nuclear Power Plants (USNRC, 2007d). In the modal analysis, for structures composed of different materials (having different damping values) the composite modal damping is calculated using either the stiffness-weighted method or mass-weighted method, based on SRP Section 3.7.2 (USNRC, 2013a). This applies to either the response spectrum method or the time history method.Seismic SSI analysis of the FSTR is performed in the program SASSI2010, Version 1.0, which performs the analysis in the frequency domain. The variations in damping are

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-4 Rev. 4 accounted for in the seismic SSI analysis through the complex frequency response analysis method which incorporates damping as an imaginary component in the stiffness matrix.

3.4.2 SEISMIC ANALYSIS OF FACILITY STRUCTURES 3.4.2.1 Seismic Analysis Methods The general equation of motion (as seen below) is used regardless of the method selected for the seismic analysis.

(Equation 3.4-1)

Where:

[M]

=

mass matrix

[C]

=

damping matrix

[K]

=

stiffness matrix

=

column vector of relative accelerations

=

column vector of relative velocities

=

column vector of relative displacements

=

ground acceleration Analytical models are represented by finite element models. Consistent with SRP Section 3.7.2 (USNRC, 2013a), SRP Acceptance Criterion 3.C, finite element models are acceptable if the following guidelines are met:

The type of finite element used for modeling a structural system should depend on structural details, the purpose of analysis, and the theoretical formulation upon which the element is based. The mathematical discretization of the structure should consider the effect of element size, shape, and aspect ratio on solution accuracy.

In developing a finite element model for dynamic response, it is necessary to consider that local regions of the structure, such as individual floor slabs or walls, may have fundamental vibration modes that can be excited by the dynamic seismic loading. These local vibration modes are represented in the dynamic response model, in order to ensure that the in-structure response spectra include the additional amplification.

The seismic analysis of the FSTR is performed in the program SASSI2010, Version 1.0, System for Analysis of Soil-Structure Interaction, and SAP 2000, integrated software for structural analysis and design. The finite element model consists of plate/shell, solid, beam, or a combination of finite elements.

3.4.2.2 Soil-Structure Interaction (SSI) Analysis The SSI model provides structural responses for design basis level seismic loading of the main production facility, including transfer functions, maximum seismic acceleration (zero period

g u

M x

K x

C x

M

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-5 Rev. 4 acceleration [ZPA]), and in-structure response spectra (ISRS) (horizontal and vertical directions) for various damping values. The SSI model is developed using the computer program Structural Analysis Software System Interface (SASSI2010), version 1.0.

Solid elements are only used in the modelng of the excavated soil volume. No solid elements are used in the modeling of the building structure. Major structural elements of the main production facility, including walls, slabs, beams and columns, are modeled with appropriate mass and stiffness properties. Major openings within walls and slabs are included in the SSI model. The model uses thick shell elements to represent concrete slabs and walls, and beam elements to represent steel members, mostly comprising the truss components in the facility. Elements are modeled at the geometric centerline of the structural member they represent with the following exceptions:

The below grade and mezzanine slabs are modeled at their actual top-of-slab elevation.

Minor adjustments are made to the dimensions and locations of wall openings to maximize mesh regularity in the model.

Roof truss locations are adjusted to align with the roof shell element mesh.

In addition to self-weight of the structure, floor loads and equipment loads are converted to mass and included in the model. A mass equivalent to a 50 pounds per square foot (psf) floor load is added to floor slabs to represent miscellaneous loads from minor equipment, piping, and raceways. A portion of the loads are considered mass sources in the following manner according to SRP Section 3.7.2 (USNRC, 2013a):

Dead Load 100 percent Live Load25 percent Snow Load.75 percent In addition to the loads that are converted to mass, 100 percent of the hydrodynamic mass of the water in the IU cells and 100 percent of the parked crane mass is included.

The SSI analyses are performed separately on an equivalent linear-elastic basis for mean (best estimate [BE]), upper bound (UB), and lower bound (LB) soil properties to represent potential variations in in-situ and backfill soil conditions around the building in accordance with SRP Section 3.7.2 (USNRC, 2013a). SSI analysis requires detailed input of the soil layers supporting the structure. Strain dependent soil properties were determined from geotechnical investigations and free field site response analysis. The free-field site response analysis is performed for the LB, BE, and UB soil properties. In accordance with SRP Section 3.7.2, the UB and LB values of the soil shear modulus, G, are obtained in terms of their BE through the equations shown below.

Equations 3.4-2 and 3.4-3 are used to calculate the low strain properties for the LB and UB. The final soil properties are calculated from the SHAKE2000 program, version 3.5.

(Equation 3.4-2)

(Equation 3.4-3)

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-6 Rev. 4 Where, COV is the coefficient of variation. A COV of 0.5 is used because the site is well-investigated.

3.4.2.3 Combination of Earthquake Components In order to account for the responses of the structures subjected to the three directional (two horizontal and the vertical) excitations, the maximum co-directional responses are combined using either the square root of the sum of the squares (SRSS) method or the 100-40-40 rule as described in Section 2.1 of Regulatory Guide 1.92, Revision 3, Combining Modal Responses and Spatial Components in Seismic Response Analysis (USNRC, 2012).

3.4.2.4 Seismic Analysis Results The seismic loads are applied to the structural analysis model as described in Subsection3.4.2.6 and utilized to develop in-structure response spectra of the facility for use in sizing equipment and components. Response spectra accelerations are output from SASSI at the 75 standard frequencies between 0.2 Hz and 34 Hz as suggested by Regulatory Guide1.122, Revision1, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components (USNRC, 1978). In addition, response spectra accelerations are specified to be output at frequencies of 37 Hz, 40 Hz, 43 Hz, 46 Hz and 50Hz.

The results of the seismic analysis demonstrate that the design of the FSTR meets the seismic requirements of SHINE Design Criterion 2.

3.4.2.5 Assessment of Structural Seismic Stability The stability of the main production facility is evaluated for sliding and overturning considering the following load combinations and factors of safety in accordance with Section 7.2 of American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) Standard 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities (ASCE/SEI, 2005) and SRP Section 3.8.5 (USNRC, 2013b):

1.1 1.1 (Equation 3.4-4) 1.1 1.1 (Equation 3.4-5) 1.5 1.5 (Equation 3.4-6)

Where:

D

=

Dead Load H

=

Lateral Earth Pressures E

=

Earthquake Load Wt

=

Tornado Load W

=

Wind Load Minimum Factor of Safety Load Combination Sliding Overturning

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-7 Rev. 4 The base reactions due to seismic forces envelop the reactions due to wind and tornado loading; therefore, a stability analysis for wind and tornado is not required. Seismic excitation in each direction is considered using the 100-40-40 percent combination rule as specified in Subsection3.4.2.3 above.

The lateral driving forces applicable to the seismic stability evaluation of the main production facility include active lateral soil force, static surcharge lateral soil force, dynamic surcharge lateral soil, dynamic lateral soil force, and seismic lateral inertial force. The resistance for sliding is due to the static friction at the soil-basemat interface for sliding evaluation and passive lateral soil resistance. The self-weight of the structure is considered in the resistance to overturning effects.

The seismic stability evaluation of the main production facility determined the minimum factor of safety against sliding to be 1.11 and the minimum factor of safety against overturning to be 1.99.

As such, the main production facility is considered stable.

3.4.2.6 Structural Analysis of Facility 3.4.2.6.1 Description of the Structures The main production facility is a box-type shear wall system of reinforced concrete with reinforced concrete floor slabs. The major structural elements in the main production facility include the shear walls, the floor and roof slabs, and the foundation mat.

3.4.2.6.2 Applicable Codes and Standards ACI 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary (ACI,2014)

ANSI/AISC N690-12, Specification for Safety-Related Steel Structures for Nuclear Facilities (ANSI/AISC,2012) 3.4.2.6.3 Site Design Parameters The following subsections provide the site-specific parameters for the design of the facility.

3.4.2.6.3.1 Soil Parameters The soil parameters for the facility are provided below.

Net allowable static bearing pressure at 3 ft. below grade: 2380 pounds per square foot (psf) (114 kilopascal [kPa]).

Net allowable static bearing pressure at 17 ft. below grade: 1230 psf (58.9 kPa).

Minimum average shear wave velocity: 459 ft./sec (140 m/s).

Minimum unit weight: 117 pounds per cubic foot (lb/ft3) (1874 kilograms per cubic meters

[kg/m3]).

3.4.2.6.3.2 Maximum Ground Water Level 50 ft. (15.2 m) below grade level.

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-8 Rev. 4 3.4.2.6.3.3 Maximum Flood Level Section2.4 describes the probable maximum precipitation (PMP).

Section2.4 describes the probable maximum flood (PMF).

3.4.2.6.3.4 Snow Load Snow load: 30 psf (1.44 kPa) (50-year recurrence interval).

A factor of 1.22 is used to account for the 100-year recurrence interval required.

3.4.2.6.3.5 Design Temperatures The winter dry-bulb temperature (-7°F [-22°C]).

The summer dry bulb temperature (88°F [31°C]).

3.4.2.6.3.6 Seismology SSE peak ground acceleration (PGA): 0.20 g (for both horizontal and vertical directions).

SSE response spectra: per Regulatory Guide 1.60 (USNRC, 2014a).

SSE time history: envelope SSE response spectra in accordance with SRP Section3.7.1 (USNRC, 2014b).

3.4.2.6.3.7 Extreme Wind Basic wind speed for Wisconsin: 90 miles per hour (mph) (145 kilometers per hour [kph])

(50-year recurrence interval).

A factor of 1.07 is used to account for the 100-year recurrence interval required.

Exposure Category C.

3.4.2.6.3.8 Tornado Maximum tornado wind speed (Region 1): 230 mph (370 kph).

Maximum tornado rotational speed (Region 1): 184 mph (82 m/s).

Maximum tornado translational speed (Region 1): 46 mph (21 m/s).

Radius of maximum rotational speed: 150 ft. (45.7 m).

Tornado differential pressure: 1.2 pounds per square inch (psi) (8.3 kPa).

Rate of tornado differential pressure: 0.5 psi/s (3.7 kPa/s).

Missile Spectrum: see Table 2 of Regulatory Guide 1.76 (USNRC, 2007a).

3.4.2.6.3.9 Rainfall The main production facility's sloped roof and building configuration preclude accumulation of rainwater; therefore, rain loads are not considered in this evaluation.

3.4.2.6.4 Design Loads and Loading Combinations 3.4.2.6.4.1 Dead Load Dead loads consist of the weight of all materials of construction incorporated into the building, as well as the following:

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-9 Rev. 4 Concrete cover blocks for below grade tanks and trenches.

Fixed equipment (includes tanks and hot cells).

Partition walls.

Precast tank vault bases in the RPF.

Weight of commodities attached to structural elements.

Crane dead loads as described in Subsection3.4.2.6.4.6.

3.4.2.6.4.2 Live Load The building is evaluated for live loads consistent with the use of and occupancy of the facility.

This includes minimum live loads driven by occupancy and non-permanent loads caused by equipment or required during plant operations.

The following categories encompass the live loads for the main production facility:

A distributed live load of 125 psf (5.99 kPa) is used for areas designated as light manufacturing.

A distributed live load of 250 psf (12.0 kPa) is used for areas designated as heavy manufacturing.

Additionally, the following categories are considered as live loads in the areas where they occur:

Concrete cover block laydown load.

Supercell drum export system and shield gate live load.

Forklift live load associated with the movement of a shipping container throughout the radiologically controlled area (RCA).

Roof live load.

Equipment live loading.

3.4.2.6.4.3 Snow Load The snow load is based on a ground snow load of 30 psf (1.44 kPa) with an importance factor of 1.2 and a mean recurrence interval of 100 years.

3.4.2.6.4.4 Wind Load The wind load is based on a basic wind speed of 90 mph (145 kph) with an importance factor of 1.15 and a mean recurrence interval of 100 years.

3.4.2.6.4.5 Earthquake Load Dynamic analysis is conducted with a portion of the loads considered as mass sources in the following manner according to SRP Section 3.7.2 (USNRC, 2013a):

Dead Load 100 percent Miscellaneous Load.100 percent Live Load25 percent Snow Load.75 percent Parked Crane Load.100 percent Hydrodynamic Load100 percent

Chapter 3 - Design of Structures, Systems, and Components Seismic Damage SHINE Medical Technologies 3.4-10 Rev. 4 Earthquake load is applied in a SAP2000 model (version 17.2) on an equivalent static basis. The equivalent static model represents the soil as dynamic springs, developed in accordance with ASCE 4-98 (ASCE, 2000). Maximum seismic acceleration at each node of the structure is determined by SSI analysis using SASSI2010, as discussed in Subsection3.4.2.2. Figures3.4-3 through 3.4-6 show selected response spectra locations throughout the FSTR.

The SAP2000 and SASSI2010 models are both three-dimensional models that represent the structural elements with equivalent mass and stiffness properties. The lumped masses at each node of the SAP2000 analysis are multiplied by the peak accelerations determined from the SSI analysis to determine an equivalent static earthquake load at each node. The direction of load application is iterated to obtain nine seismic force terms.

3.4.2.6.4.6 Crane Load The building is evaluated for loads associated with two overhead bridge cranes, one servicing the IU cell area and one servicing the RPF area. Crane loading is evaluated in accordance with American Society for Mechanical Engineers (ASME) NOG-1, Rules for Construction of Overhead and Gantry Cranes (ASME, 2004).

3.4.2.6.4.7 Soil Pressure Sub-grade walls of the main production facility are designed to resist static lateral earth pressure loads, compaction loads, static earth pressure, dynamic surcharge loads, and elastic dynamic soil pressure loads. Static earth pressure consists of at-rest, active, and passive soil pressure loads, which are applied as required to ensure the stability of the building.

3.4.2.6.4.8 Fluid Load The hydrostatic loading is calculated based on the actual dimensions of the IU cells and applied in the model as lateral hydrostatic pressure on the walls and vertical hydrostatic pressure on the bottom slabs.

The hydrodynamic loading is applied to the model by considering hydrodynamic masses rigidly attached to the IU cells in accordance with Section 3.1.6.3 of ASCE 4-98 (ASCE, 2000) and Chapter 6 of TID-7024, Nuclear Reactors and Earthquakes (AEC, 1963). The provisions, as outlined in the referenced documents, require that the impulsive and convective masses be applied to the model to capture the dynamic effects due to seismic motion.

3.4.2.6.4.9 Tornado Load The tornado load is based on a tornado wind speed of 230 mph (370 kph) and a tornado missile spectrum as described in Table 2 of Regulatory Guide 1.76 (USNRC, 2007a). The tornado load, Wt, is further defined by the following combinations:

Wt = Wp (Equation 3.4-7)

Wt = Ww + 0.5Wp (Equation 3.4-8)

Wt = Ww + 0.5Wp + Wm (Equation 3.4-9)

Chapter 3 - Design of Structures, Systems, and Components References SHINE Medical Technologies 3.6-1 Rev. 1

3.6 REFERENCES

ACI, 2014. Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, ACI 349-13, American Concrete Institute, 2014.

AEC, 1963. Nuclear Reactors and Earthquakes, TID-7024, U.S. Atomic Energy Commission, August 1963.

ANSI/AISC, 2012. Specification for Safety-Related Steel Structures for Nuclear Facilities, ANSI/AISC-N690, American National Standards Institute/American Institute of Steel Construction, 2012.

ASCE, 2000. Seismic Analysis of Safety-Related Nuclear Structures and Commentary, ASCE 4-98, American Society of Civil Engineers, 2000.

ASCE, 2006. Minimum Design Loads for Buildings and Other Structures, ASCE 7-05, American Society of Civil Engineers, 2006.

ASCE/SEI, 2005. Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE 43-05, American Society of Civil Engineers, 2005.

ASME, 2004. Rules for Construction of Overhead and Gantry Cranes, ASME NOG-1, American Society of Mechanical Engineers, 2004.

DOE, 2006. Accident Analysis for Aircraft Crash into Hazardous Facilities, DOE-STD-3014-2006, U.S. Department of Energy, 2006.

IAEA, 2003. Consideration of External Events in the Design of Nuclear Facilities other than Nuclear Power Plants, with Emphasis on Earthquakes, IAEA-TECDOC-1347, International Atomic Energy Agency, 2003.

NFPA, 2008. Standard for Fire Protection for Facilities Handling Radioactive Materials, NFPA 801-2008, National Fire Protection Association, 2008.

UCRL, 1997. Hossain, Q.A., R.P. Kennedy, R.C. Murray, K. Mutreja, and B.P. Tripathi, Structures, Systems, and Components Evaluation Technical Support Documents, DOE Standard, Accident Analysis for Aircraft Crash into Hazardous Facilities, UCRL-ID-123577, Lawrence Livermore National Laboratory, 1997.

USNRC, 1978. Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Regulatory Guide 1.122, Revision 1, U.S. Nuclear Regulatory Commission, 1978.

USNRC, 2007a. Design Basis Tornado and Tornado Missiles for Nuclear Power Plants, Regulatory Guide 1.76, Revision 1, U.S. Nuclear Regulatory Commission, 2007.

USNRC, 2007b. Structures, Systems, and Components to be Protected from Externally-Generated MissilesBarrier Design Procedures, NUREG-0800, Subsection 3.5.23, Revision 3, U.S. Nuclear Regulatory Commission, 2007.

Proprietary Information - Withheld from public disclosure under 10 CFR 2.390(a)(4)

Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Subcritical Assembly SHINE Medical Technologies 4a2.2-10 Rev. 3 4a2.2.4.3 Source Strength The source is required to provide greater than 5.0E+05 neutrons per second (n/s).

4a2.2.4.4 Interaction with the System The placement of the source is inside of a stainless steel capsule that is located below the tritium target chamber. This capsule is accessible when the target chamber is removed and the source is able to be inserted and removed using long-handled tools.

4a2.2.4.5 Physical Environment The nominal temperature of the cooling water surrounding the source is approximately 68°F (20°C). The neutron source will be exposed to external neutron radiation up to [

]PROP/ECI and external gamma radiation up to [

]PROP/ECI.

4a2.2.4.6 Verification of Integrity and Performance Leak and contamination tests of the subcritical multiplication source are performed prior to use in the SHINE facility. Neutron strength measurements are made to ensure the stated activity prior to operation using the source.

4a2.2.4.7 Technical Specifications There are no technical specifications applicable to the subcritical multiplication source.

4a2.2.5 SUBCRITICAL ASSEMBLY SUPPORT STRUCTURE The TSV maintains the location and shape of the target solution during irradiation. The SASS positions the TSV relative to the neutron driver, neutron multiplier, subcritical multiplication source, and neutron flux detectors as shown in Figure 4a2.1-2. The SASS contains the TSV and supports TSV dump lines, TSV overflow lines, TOGS components, and associated instrumentation.

The SASS channels cooling water around the TSV and neutron multiplier. The PCLS is attached to the SASS upper and lower plenums. The PCLS forces cooling water to pass [

]PROP/ECI along the TSV inner and outer shells, and around the neutron multiplier to remove heat from the TSV and neutron multiplier during operation.

The SASS and PSB components are designed to withstand the design basis loads, including thermal, seismic, and hydrodynamic loads imposed by the light water pool during a seismic event. Hydrodynamic loads to safety-related equipment submerged within the light water pools were applied considering hydrodynamic added mass and drag forces from sloshing pool water.

These hydrodynamic loads were calculated using the maximum vertical displacement of sloshing pool water determined using the methods in Section 9 of ACI 350.3-06, Seismic Design of Liquid-Containing Concrete Structures and Commentary (ACI, 2006). In addition, the SASS and supported PSB components are designed to withstand normal operating loads imposed by the

Proprietary Information - Withheld from public disclosure under 10 CFR 2.390(a)(4)

Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Subcritical Assembly SHINE Medical Technologies 4a2.2-11 Rev. 3 primary cooling water and target solution, including hydraulic and thermal stresses. The entire subcritical assembly is submerged in the light water pool.

The SASS is supported on top of the TSV dump tank which is attached to the floor of the light water pool via seismic anchorages that establish the alignment of the SASS relative to the neutron driver. The SASS provides vertical and lateral support for the subcritical assembly components. Nozzle loads imposed on the SASS are accommodated by the design of the SASS, ensuring that stresses and displacements resulting from these loads do not result in stresses that exceed design allowables, or displacements that affect the function, or neutronics of the overall system.

The SASS is operated at a higher pressure than the TSV to minimize leakage of target solution into the PCLS in the event of a loss of TSV integrity. The SASS is designed for an internal pressure of 100 psi (689 kPa) to provide a defense-in-depth fission product boundary in the event of a TSV breach.

Surrounding the TSV, the SASS is exposed to neutron fluxes of up to approximately

[ ]PROP/ECI. The SASS does not normally contact the target solution. In the event of a breach in the TSV, the SASS provides a defense-in-depth fission product boundary between the target solution and the light water pool.

The material of construction for the SASS is 304L stainless steel. The properties and behavior of 304L stainless steel in the expected neutron fluences are well-documented and are accounted for in the design of the SASS, ensuring its safe reliable operation over its 30 year design life. The highest fluence to the SASS occurs adjacent to the TSV and is up to approximately

[

]PROP/ECI. The fluence to the SASS was compared to literature data (Kangilaski, 1967), which indicates an increase in yield and ultimate tensile strength and a reduction in ductility of austenitic stainless steel. Literature (Kangilaski, 1967; Kangilaski, 1971) indicates that while irradiated austenitic stainless steel loses some ductility, it will remain ductile for temperatures greater than -58°F (-50°C) when exposed to applicable fluence.

Neutron flux detectors are supported using brackets attached to the SASS outer shell. These brackets serve to locate the flux detectors in a fixed location relative to the TSV, ensuring flux profiles are measured consistently. Flux detectors are positioned around the SASS at nominally 120 degree intervals.

4a2.2.5.1 Technical Specifications There are no technical specifications applicable to the SASS.

4a2.2.6 NEUTRON MULTIPLIER The neutron multiplier is an annulus (approximately [

]PROP/ECI tall) of aluminum-clad uranium metal that serves to moderate and multiply the fast neutrons coming from the fusion reactions initiated by the neutron driver. The multiplier consists of natural uranium metal (uranium that has not been enriched or depleted in U-235) with a thickness of approximately

[ ]PROP/ECI, clad in approximately [ ]PROP/ECI thick aluminum. The

Proprietary Information - Withheld from public disclosure under 10 CFR 2.390(a)(4)

Export Controlled Information - Withheld from public disclosure under 10 CFR 2.390(a)(3)

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description Subcritical Assembly SHINE Medical Technologies 4a2.2-12 Rev. 3 design life of the neutron multiplier is 30 years, but it is designed to allow remote replacement should physical damage occur to it, or a distortion that is outside of acceptable limits.

The multiplier is manufactured by casting natural uranium metal sections, machining the sections, and then placing the sections in a machined cladding. During the casting, the natural uranium is alloyed with a small weight fraction of silicon to assist in obtaining small, randomly-oriented grains to help reduced irradiation-induced growth. The uranium is cast in two axial pieces, with one piece supported on top of the other piece when installed in the cladding. The cast uranium is machined to final dimensions, and then is inserted into the aluminum cladding.

The aluminum cladding is type 6061 aluminum. The aluminum cladding is welded closed after the uranium core is inserted. [

]PROP/ECI The cladding is leak-tested following fabrication.

The fast fusion neutrons that collide with the uranium metal can cause several high energy reactions to occur. The most common reactions in the multiplier include fission, and to a lesser extent (n,2n) and (n,3n) reactions with U-235 and U-238. The resulting spectrum of fast fission, fusion, epithermal, and thermal neutrons then enter the TSV.

The aluminum cladding contains fission products created within the uranium (the cladding thickness is much greater than the distance traveled by a fission fragment). In the event of a cladding failure, there are no consequences that would affect the safe operation and shutdown of the irradiation system. There is potential that the uranium metal could form surface oxidization, releasing hydrogen gas. [

]PROP/ECI A cladding failure could also result in fission products being released into the primary cooling water, leading to contamination in the PCLS. Sampling the PCLS detects such a breach via the increased radioactive contamination present in the water. Additionally, radiation monitors on the radiological ventilation zone 1 exhaust subsystem (RVZ1e) line ventilating the PCLS expansion tank can detect fission products leaving the PCLS cooling water. The TRPS initiates an IU Cell Safety Actuation if radiation levels exceed predetermined limits, resulting in the isolation of the PCLS supply and return lines and the RVZ1e IU cell line as described in Section 7.4.

Radiation damage and burnup are not expected to impact operation of the multiplier for the lifetime of the plant. The maximum fast neutron fluence (greater than approximately 100 thousand electron volts [keV]) of the multiplier over a 30 year period of continuous operation is calculated to be less than [ ]PROP/ECI. Nuclear parameters of the subcritical assembly at the end-of-life for the multiplier have been calculated and do not affect the safety of the subcritical assembly. Nuclear parameters are described in Section 4a2.6.

Bounding fission product gas generation for the lifetime of the multiplier has been incorporated into the design. [

]PROP/ECI Overall heat generation rate in the multiplier is approximately 15 kW (50,000 Btu/hr) during operation of the TSV at the licensed power limit. Heat generation in the multiplier is from fissions occurring within the multiplier and radiation absorbed from the fission process in the TSV. Most fission energy is short range (fission products, betas) and is deposited locally. Some energy from long range products (gammas and fast neutrons) is deposited in the multiplier, with the balance

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description References SHINE Medical Technologies 4a2.9-1 Rev. 2 4a

2.9 REFERENCES

ACI, 2006. Seismic Design of Liquid-Containing Concrete Structures and Commentary, ACI 350.3-06, American Concrete Institute, 2006.

ACI, 2007. Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, ACI 349-06, American Concrete Institute, 2007.

ACI, 2008. Building Code Requirements for Structural Concrete and Commentary, ACI 318-08, American Concrete Institute, 2008.

ACI, 2014. Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, ACI 349-13, American Concrete Institute, 2014.

ANSI/ANS, 2006. Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants, ANSI/ANS-6.4-2006, American National Standards Institute, 2006.

ANSI/ANS, 2015. Program for Testing Radiation Shields in Light Water Reactors (LWR),

ANSI/ANS-6.3.1-1987 (R2015), American National Standards Institute, 2015.

ASME, 2010. Boiler & Pressure Vessel Code - Rules for Construction of Pressure Vessels,Section VIII, American Society of Mechanical Engineers, July 1, 2010.

ASME, 2013. Code for Pressure Piping, B31.3-2012, American Society of Mechanical Engineers, January 10, 2013.

Baker, 1944. Water Boiler, LA Report 134, Los Alamos National Laboratory, Baker, C., et al.,

September 4, 1944.

Barbry Francis, 2007. French CEA Experience on Homogeneous Aqueous Solution Nuclear Reactors, Commissariat a l Energie Atomique CEA/France, June 2007.

Beall, 1954. The Homogeneous Reactor Experiment - A Chemical Engineering Pilot Plant, Chemical Engineering Progress, Vol. 50, No. 5, Beall, S., et al., May 1954.

BNL, 2010. Aqueous Homogenous Reactor Technical Panel Report, BNL-94462-2010, Brookhaven National Laboratory, December 10, 2010.

Bull, 2014. Effects of Gas Bubble Production on Heat Transfer from a Volumetrically Heated Liquid Pool, University of Wisconsin-Madison, Bull, G., 2014.

Clayton, 1985. Neutron Source Multiplication Method, Pacific Northwest National Laboratory, Proceedings of the Workshop on Subcritical Reactivity Measurements, Clayton, E.D., 1985.

Gamble, 1959. A Proposed Model of Bubble Growth during Fast Transients in the KEWB Reactor, Paper 25-3, American Nuclear Society, Gamble, D., June 15, 1959.

Chapter 4 - Irradiation Unit and Radioisotope Production Facility Description References SHINE Medical Technologies 4a2.9-2 Rev. 2 IAEA, 2008. Homogeneous Aqueous Solution Nuclear Reactors for the Production of Mo-99 and Other Short Lived Radioisotopes, TECDOC-1601, International Atomic Energy Agency, September 2008.

Kangilaski, 1967. The Effects of Neutron Radiation on Structural Materials, Radiation Effects Information Center, Battelle Memorial Institute, REIC Report No. 45, Kangilaski, M., June 30, 1967.

Kangilaski, 1971. Radiation Effects Design Handbook: Section 7. Structural Alloys, National Aeronautics and Space Administration, NASA CR-1873, Kangilaski, M., October 1971.

Lane, 1958. Fluid Fuel Reactors, Part 1, Aqueous Homogeneous Reactors, Oak Ridge National Laboratory, Lane, J., June 1958.

LANL, 2011. Monte-Carlo N-Particle Transport Code MCNP5-1.60 Release & Verification, LA-UR-11-00230, Los Alamos National Laboratory, F.B. Brown, B.C. Kiedrowski, J.S. Bull, M.A. Gonzales, N.A. Gibson, 2011.

LANL, 2013a. A Generic System Model for a Fissile Solution Fueled Assembly, LA-UR-13-22033, Los Alamos National Laboratory, Kimpland, R., Klein, S., 2013.

LANL, 2013b. A Generic System Model for a Fissile Solution Fueled Assembly - Part II, LA-UR-13-28572, Los Alamos National Laboratory, Kimpland, R., Klein, S., November 2013.

LANL, 2014a. Aqueous Homogeneous Reactor (AHR) Benchmarks, LA-UR-14-21529, Los Alamos National Laboratory, Kimpland, R., Klein, S., March 2014.

LANL, 2014b. Discussion Regarding Aqueous Homogeneous Reactor (AHR) Benchmarks, LA-UR-14-23768, Los Alamos National Laboratory, Kimpland, R., Klein, S., May 2014.

Lee, 1952. The Density of Uranyl Sulfate Solutions and the Determination of Uranium Concentration by Density Measurements, ORNL-1332, Oak Ridge National Laboratory, Lee, J.,

et al., June 18, 1952.

NCSU, 1955. Raleigh Research Reactor Critical on September 1953, NCSCR-1, North Carolina State University, June 1955.

NEA, 2014a. International Handbook of Evaluated Reactor Physics Benchmark Experiments, Nuclear Energy Agency, NEA/NSC/DOC(2006)1, March 2014.

NEA, 2014b. International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Energy Agency, NEA/NSC/DOC(95)03, September 2014.

NRC, 1979. Hydrodynamics and Heat Transfer Characteristics of Liquid Pools with Bubble Agitation, NUREG/CR-0944, United States Nuclear Regulatory Commission, Blottner, F.,

November 1979.

NRC, 2001a. Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments), Regulatory Guide 1.142, Revision 2, United States Nuclear Regulatory Commission, November 2001.

Chapter 8 - Electrical Power Systems Normal Electrical Power Supply System SHINE Medical Technologies 8a2.1-1 Rev. 2 8a2 IRRADIATION FACILITY ELECTRICAL POWER SYSTEMS 8a2.1 NORMAL ELECTRICAL POWER SUPPLY SYSTEM A single overall electrical power system serves the main production facility, including both the irradiation facility and the radioisotope production facility, as well as the site and support buildings. The normal electrical power supply system (NPSS) for the SHINE facility consists of the normal power service entrances from the electric utility and a distribution system providing twothree utilization voltages, 480Y/277, 400Y/230, and 208Y/120 volts alternating current (VAC),

3-phase, 60hertz. Grounding and lightning protection is provided.

The NPSS receives off-site power service from the local utility, Alliant Energy, at 480Y/277 VAC in twothrough five separate transformer feeds. Portions of the NPSS that comprise the emergency electrical power system can also receive power from the standby generator system (SGS). The NPSS is used for normal operation and normal shutdown of the facility.

The NPSS is sized for safe operation of the facility. The largest loads on the NPSS are the process chilled water system (PCHS), neutron driver assembly system (NDAS), and the facility chilled water system (FCHS); however, those loads are not required for safe shutdown of the facility. Refer to Section8a2.2 for a tabulation of emergency electrical load requirements.

A simplified diagram of the overall electrical power system is provided in Figure8a2.1-1.

8a2.1.1 DESIGN BASIS The design of the NPSS is based on Criterion 27, Electrical power systems, and Criterion 28, Inspection and testing of electric power systems, of the SHINE design criteria. The SHINE design criteria are described in Section3.1.

The design of the NPSS provides sufficient, reliable power to facility and site electrical equipment as required for operation of the SHINE facility and to comply with applicable codes and standards. The NPSS is designed such that it:

Does not prevent the ability of safety-related SSCs to perform their safety functions; Provides for the separation or isolation of safety-related circuits from nonsafety-related circuits, including the avoidance of electromagnetic interference with safety-related instrumentation and control functions; Fails to a safe configuration upon a loss of off-site power (LOOP);

Provides the normal source of power supply to the safety-related electrical buses; Provides the safety-related function of removing power from select components when demanded by the safety-related engineered safety features actuation system (ESFAS) or target solution vessel (TSV) reactivity protection system (TRPS); and Is able to be inspected, tested, and maintained to meet the above design bases.

The following codes and standards are used in the design of the NPSS:

National Fire Protection Association (NFPA) 70-2017, National Electrical Code (NFPA, 2017), as adopted by the State of Wisconsin (Chapter SPS316 of the Wisconsin Administrative Code, Electrical)

Chapter 8 - Electrical Power Systems Normal Electrical Power Supply System SHINE Medical Technologies 8a2.1-2 Rev. 2 Institute of Electrical and Electronics Engineers (IEEE) 384-2008, Standard Criteria for Independence of Class 1E Equipment and Circuits (IEEE, 2008), invoked for isolation and separation of nonsafety-related circuits from safety-related circuits, as described in Subsections8a2.1.3 and 8a2.1.5.

IEEE Standard 323-2003, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations (IEEE, 2003), invoked for environmental qualification of safety-related equipment as described in Subsection8a2.1.3.

8a2.1.2 OFF-SITE POWER SUPPLY DESCRIPTION The SHINE facility is connected to atwo single power circuits from the off-site transmission electric network. The power circuits isare shared with other utility customers. Thise two power circuits feeds twofive local outdoor 12.47 kilovolt (kV) - 480Y/277 VAC 3-phase transformers.

The 12.47 kV feeders originates from the Alliant Energy Tripp Road substation, about 2.8 circuit miles from the SHINE facility, and the Alliant Energy Venture substation, about 2.3 circuit miles from the SHINE facility.

Each of the tTwo transformers isare each connected to one of the SHINE facility's two main 480 VAC switchgear buses. Figure8a2.1-1 depicts the off-site connections to the SHINE facility.

8a2.1.3 NORMAL ELECTRICAL POWER SUPPLY SYSTEM DESCRIPTION The NPSS operates as twofive separate branches, each receiving utility power at 480Y/277 VAC. The branches automatically physically disconnect from the utility by opening the associated utility power (UP) supply breakers (UP BKR 1, and UP BKR 2, UP BKR 3, or UP BKR 4) on a loss of phase, phase reversal, or sustained overvoltage or undervoltage as detected by protection relays for each utility transformer. This function is not required for safe shutdown, as described in Subsection8a2.1.6. UP BRK 5, which provides isolation for the resource building, provides overcurrent and surge protection. UP BKR 5 disconnecting from the utility is not required for safe shutdown since it does not impact safety-related equipment in the main production facility.

The two branches, serving loads in the main production facility and the nitrogen purge system (N2PS) structure, can be cross-connected by manually opening one of the UP breakers and manually closing both bus tie (BT) breakers (BT BKR 1 and BT BKR 2) in the event of the loss of a single utility 480Y/277 VAC feed. This cross-connection would be performed at reduced loading and administratively controlled to ensure the remaining utility feed is not overloaded.

The distribution system serving the main production facility and the N2PS structure consists of two line-ups of 480 volts (V) switchgear, two emergency 480V buses (that are supported by the standby generator), and isolation and cross-tie breakers. The two switchgear line-ups each feed an individual emergency bus and the single SGS switchgear. The two emergency 480 V buses are nonsafety-related, but each provides power to a safety-related uninterruptible electrical power supply system (UPSS) division via division-specific battery chargers and bypass transformers. The SGS and the UPSS are further described in Section8a2.2.

The distribution system serving the material staging building, storage building, and facility chillers consists of two 480 V switchgear with isolation and bus tie breakers (BT BKR 3 and BT BKR 4).

A single distribution system serves the resource building. There are no safety-related loads powered from these distribution systems.

Chapter 8 - Electrical Power Systems Normal Electrical Power Supply System SHINE Medical Technologies 8a2.1-3 Rev. 2 Surge protection is provided at each electrical service entrance to limit voltage spikes and electrical noise. The electrical services are monitored for voltage, frequency, and loss of phase.

When an electrical service exceeds prescribed limits, the facility is disconnected from the utility to prevent damage.

Loss of phase protection is provided by use of a negative sequence relay. The NPSS monitors each phase and disconnects from utility power on a loss of any one of the three incoming phases. Refer to Section8a2.2 for further discussion of facility response to transient events.

The NPSS complies with NFPA 70 (NFPA, 2017), as adopted by the State of Wisconsin (ChapterSPS 316 of the Wisconsin Administrative Code, Electrical); with Sections 6.1.2.1, 6.1.2.2, and 6.1.2.3 of IEEE 384 (IEEE, 2008) for isolation; and with Section 5.1.1.2, Table 1 of Section5.1.3.3, and Table 2 of Section 5.1.4 of IEEE 384 (IEEE, 2008) for physical separation between nonsafety-related circuits and safety-related circuits.

The NPSS contains the following safety-related equipment:

Two undervoltage trip enclosedsafety-related breakers are provided for each instance of the NDAS to provide the redundant ability to disconnect power.

Two undervoltage trip enclosedsafety-related breakers per vacuum pump to provide the redundant ability to disconnect power from each vacuum pump in the vacuum transfer system (VTS).

Two undervoltage trip enclosedsafety-related breakers per extraction feed pump to provide the redundant ability to disconnect power from each (of three) extraction feed pumps in the molybdenum extraction and purification system (MEPS).

Two undervoltage trip enclosedsafety-related breakers providing the redundant ability to disconnect power from the radiological ventilation zone 1 (RVZ1) exhaust fans, radiological ventilation zone 2 (RVZ2) exhaust fans and RVZ2 supply air handling units.

The safety functions performed by the specified breakers are related to preventing actions that could initiate or increase the consequences of an accident. The equipment tied to these breakers does not perform an active safety function. Redundant breakers are provided to ensure that the safety function can still be performed in the event of a single active failure.

Safety-related NPSS equipment is located in a mild environment, is not subject to harsh environmental conditions during normal operation or transient conditions, and has no significant aging mechanisms. This equipment is designed and qualified by applying the guidance of Sections 4.1, 5.1, 6.1, and 7 of IEEE 323 (IEEE, 2003), and is qualified to the environmental parameters provided in Tables7.2-2 and 7.2-3.

8a2.1.4 GROUNDING AND LIGHTNING PROTECTION Equipment ground conductors, driven electrodes, buried conductors, and ground bars provide a conductive connection between facility SSCs and earth. These components, when taken together, provide intentional low impedance conductive paths for facility SSCs as required to ensure personnel safety, equipment protection, proper component function, electrical noise reduction and signal integrity.

The facility grounding system complies with NFPA 70 (NFPA, 2017). The facility grounding equipment provides no safety-related function.

Chapter 8 - Electrical Power Systems Normal Electrical Power Supply System SHINE Medical Technologies 8a2.1-4 Rev. 2 Lightning protection equipment provides low impedance paths to ground that minimize the effects of potential lightning strikes on personnel, equipment, and the facility structure. It provides no safety-related function.

8a2.1.5 RACEWAY AND CABLE ROUTING There are four separation groups for cables and raceways for the SHINE facility: Group A, GroupB, Group C, and Group N. Spatial separation between groups is in accordance with Section5.1.1.2, Table 1 of Section 5.1.3.3, and Table 2 of Section5.1.4 of IEEE384 (IEEE, 2008).

Separation Group A contains safety-related power circuits from UPSS Division A and safety-related control circuits from TRPS, NFDS, and ESFAS Division A.

Separation Group B contains safety-related power circuits from UPSS Division B and safety-related control circuits from TRPS, NFDS, and ESFAS Division B.

Separation Group C contains safety-related control circuits from TRPS and ESFAS Division C. For additional information on the Division C circuits see Section7.4.

Group N contains the facility nonsafety-related cables, including NPSS and SGS power circuits and process integrated control system (PICS) control circuits.

Nonsafety-related circuits are electrically isolated from safety-related circuits by isolation devices in accordance with Sections 6.1.2.1, 6.1.2.2, and 6.1.2.3 of IEEE 384 (IEEE, 2008). See Chapter7 for additional discussion of safety-related control systems.

8a2.1.6 LOSS OF OFF-SITE POWER A LOOP is defined as zero voltage/power supplied by the utility, loss of a phase, phase reversal, sustained overvoltage or sustained undervoltage. When there is loss of phase, phase reversal, sustained overvoltage or sustained undervoltage, the facility automatically disconnects from the utility. For the plant equipment, all the scenarios result in zero voltage/power supplied by the utility.

IUs in Mode 0 (Solution Removed) are unaffected by the LOOP - the neutron driver is not operating and target solution is not present in the IU.

TSV filling operations for IUs in Mode 1 (Filling) will be stopped via the loss of power to the VTS, which causes the VTS vacuum pumps to shut down and the VTS vacuum breaker valve to open.

Neutron flux monitoring and safety-related protection systems will remain operational, powered via the UPSS (see Section8a2.2). If the SGS is available, the SGS will auto start to provide backup power to the TSV off-gas system (TOGS), allowing the TOGS to continue to operate and mitigate hydrogen generated by radiolysis from decay radiation in the target solution. If the SGS is not available, the TOGS will continue to operate for five minutes, powered by the UPSS. Three minutes after loss of external power to the UPSS, before the TOGS blowers are unloaded from the UPSS, TRPS will initiate an IU Cell Nitrogen Purge, and the nitrogen purge system (N2PS) will inject nitrogen into the TSV dump tank to provide hydrogen control in the IU. The PCLS pumps are not powered by the UPSS or SGS; therefore, PCLS flow to the TSV will be lost. Loss of PCLS flow starts a three minute timer. If PCLS flow is not restored within the three minute duration, TRPS will initiate an IU Cell Safety Actuation, resulting in the TSV dump valves opening and the target solution draining from the TSV to the TSV dump tank. Once in the TSV dump tank, the decay heat is passively removed from the target solution via natural convection to the light

Chapter 8 - Electrical Power Systems Normal Electrical Power Supply System SHINE Medical Technologies 8a2.1-6 Rev. 2 Figure 8a2.1 Electrical Distribution System (Simplified)

UP BKR 1 UP BKR 2 UP XFMR 1 UP XFMR 2 NV BKR 1 NV BKR 2 480V SWGR A 480V SWGR B BT BKR 1 BT BKR 2 EMERG. BUS-A EMERG. BKR 1 EMERG. BKR 2 BATT CHGR BKR 1 BATT CHGR BKR 2 Bypass XFMR BKR 1 Bypass XFMR BKR 2 125VDC UPSS A 125VDC UPSS B Battery A

BATT BKR 1 UPS BKR 1 Battery B

BATT BKR 2 UPS BKR 2 AC UPSS A AC UPSS B AC Loads NDAS AC Loads DC Loads DC Loads AC Loads AC Loads Safety-Related Equipment 12.47kV 480Y/

277VAC 125VDC 208Y/

120VAC BATT CHGR A BATT CHGR B BYPASS XFMR B BYPASS XFMR A UPSS NPSS Nonsafety-Related Equipment NEC 700 NEC 701 NEC 702 SAFETY RELATED SAFETY RELATED NDAS SAFETY RELATED SAFETY RELATED TYP. 4 TYP. 4 BYPASS XFMR SEC. BRKR BYPASS XFMR SEC. BKR UP XFMR 5 UP XFMR 3 UP XFMR 4 UP BKR 5 Resource Bldg Service UP BKR 4 480V SWGR D UP BKR 3 480V SWGR C BT BKR 3 BT BKR 4 CHILLER TYP (3)

CHILLER TYP (3)

NPSS SG SG ISO BKR SGS SWGR EMERG. BUS-B Transfer Bus B SWGR Transfer Bus A SWGR From SGS SWGR To Transfer Bus A To Transfer Bus B Serves Outbuildings N.O.

N.O.

N.O.

N.O.

NORMARLLY DE - ENGERGIZED CONTROLLED BY UTILTITY NORMARLLY DE - ENGERGIZED CONTROLLED BY UTILTITY Alliant 12.47kV Utility Power CKT 1 Alliant 12.47kV Utility Power CKT 2 Storage Outbuilding Material Staging Outbuilding

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-1 Rev. 3 8a2.2 EMERGENCY ELECTRICAL POWER SYSTEMS The emergency electrical power systems for the SHINE facility consist of the safety-related uninterruptible electrical power supply system (UPSS), the nonsafety-related standby generator system (SGS), and nonsafety-related local power supplies and unit batteries. The UPSS provides reliable power for the safety-related equipment required to prevent or mitigate the consequences of design basis events. The UPSS consists of a 125-volt direct current (VDC) battery subsystem, inverters, bypass transformers, distribution panels, and other distribution equipment necessary to feed safety-related alternating current (AC) and direct current (DC) loads and select nonsafety-related AC and DC loads.

The SGS consists of a single natural gas-driven generator and, associated breakers, transfer switches, and distribution equipment. The SGS provides an alternate source of power for UPSS loads. Additionally, emergency power is provided by the SGS for facility physical security control systems and information and communications systems. Unit batteries provide power for egress and exit lights, switchgear control (station control batteries), and nonsafety-related local uninterruptible power supplies which provide back-up power for communications, data systems, and nonsafety-related control systems. The SGS provides an alternate source of power for the unit batteries and their associated loads.

Nonsafety-related local power supplies for the process integrated control system (PICS) and the facility data and communications systems (FDCS) are described in Sections 7.6 and 9a2.4, respectively.

8a2.2.1 UNINTERRUPTIBLE ELECTRICAL POWER SUPPLY SYSTEM DESIGN BASIS The design of the UPSS is based on Criterion 27, Electrical power systems, and Criterion 28, Inspection and testing of electric power systems, of the SHINE design criteria. The SHINE design criteria are described in Section3.1.

The purpose of the UPSS is to provide a safety-related source of power to equipment required to ensure and maintain safe facility shutdown and prevent or mitigate the consequences of design basis events.

The UPSS:

Provides power at a sufficient capacity and capability to allow safety-related SSCs to perform their safety functions; Is designed, fabricated, erected, tested, operated, and maintained to quality standards commensurate with the importance of the safety functions to be performed; Is designed to withstand the effects of design basis natural phenomena without loss of capability to perform its safety functions; Is located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions; Has sufficient independence, redundancy, and testability to perform its safety functions assuming a single failure; Incorporates provisions to minimize the probability of failure as a result of or coincident with the loss of power from the transmission network; and Permits appropriate periodic inspection and testing to assess the continuity of the system and the condition of components.

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-2 Rev. 3 8a2.2.2 UNINTERRUPTIBLE ELECTRICAL POWER SUPPLY SYSTEM CODES AND STANDARDS The UPSS is designed in accordance with the following codes and standards:

National Fire Protection Association (NFPA) 70-2017, National Electrical Code (NFPA, 2017), as adopted by the State of Wisconsin (Chapter SPS 316 of the Wisconsin Administrative Code, Electrical)

IEEE Standard 344 - 200413, Recommended PracticeIEEE Standard for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating SystemsStations (IEEE, 200413); invoked to meet seismic requirements, as described in Subsection8a2.2.3 IEEE Standard 384 - 2008, Standard Criteria for Independence of Class 1E Equipment &

Circuits (IEEE, 2008); invoked for separation and isolation of safety-related and nonsafety-related cables and raceways and for associated equipment, as described in Subsection8a2.2.3 IEEE Standard 485 - 2010, Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications (IEEE, 2010); invoked for battery sizing of UPSS loads, as described in Subsection8a2.2.3 IEEE Standard 323-2003, Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations (IEEE, 2003); invoked for environmental qualification of safety-related equipment as described in Subsection8a2.2.3 While the UPSS is not classified as a Class 1E system, portions of Class 1E-related standards, as described in this section, are applied to the design of the UPSS in order to satisfy applicable SHINE design criteria.

8a2.2.3 UNINTERRUPTIBLE ELECTRICAL POWER SUPPLY SYSTEM DESCRIPTION The safety-related UPSS provides a reliable source of power to the redundant divisions of AC and DC components on the safety-related power buses. Each division of the UPSS consists of a 125 VDC battery subsystem, 125 VDC to 208Y/120 volts alternating current (VAC) inverter, rectifier (battery charger), bypass transformer, static switch and a manual bypass switch, 208Y/120 VAC and 125 VDC distribution panels, and a nonsafety-related 208Y/120 VAC bus system isolated from the safety-related portion of the system by breakers or isolating fuses which meet Section 6.1.2 requirements of IEEE 384 (IEEE, 2008) for isolation devices, ensuring that a failure of nonsafety-related loads does not impact safety-related loads..

Distribution wiring from each division of the UPSS is isolated and separated from the other division per Sections 6.1.2.1, 6.1.2.2, and 6.1.2.3 of IEEE 384 (IEEE, 2008) for isolation and with Section 5.1.1.2, Table 1 of Section 5.1.3.3, and Table 2 of Section 5.1.4 of IEEE 384 (IEEE, 2008) for physical separation.

A simplified diagram of the UPSS is provided in Figure8a2.2-1.

Each division of UPSS is normally powered by an emergency 480 VAC NPSS bus via a division-specific battery charger. The emergency 480 VAC NPSS buses can also be powered by the SGS, providing an alternate source of power to the UPSS. The SGS is described in Subsection8a2.2.4.

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-4 Rev. 3 separate fire area in the safety-related, seismic portion of the main production facility. The UPSS is required to perform its safety function before, during, and after a seismic event, and is qualified by one of the testing methods described in Chapter 8 of IEEE 344 (IEEE, 200413).

The battery sizing for the UPSS loads is shown in Table8a2.2-2, using the sizing guidance provided in Sections 6.1.1, 6.2.1, 6.2.2, 6.2.3, 6.2.4, 6.3.2 and 6.3.3 of IEEE 485 (IEEE, 2010).

Batteries are vented lead-acid. Transfer of loads from the NPSS to the UPSS is automatic and requires no control power.

The required reserve for loads is listed in Table8a2.2-2. 15 percent of the total is reserved to accommodate variations of power during equipment procurement and an additional 10 percent is initially reserved for future needs that may be identified during the lifetime of the facility.

The run time requirements in Table8a2.2-1 are based on:

1) Equipment required to prevent hydrogen deflagration is powered for five minutes,

2) Equipment used to minimize transient effects on the facility due to short duration power loss is powered for five minutes,

3) Equipment used to provide alerts for facility personnel and monitor the status of the facility during immediate recovery efforts is powered for two hours, or

4) Defense-in-depth power for nonsafety-related equipment used to monitor and reduce the tritium source term in the tritium confinement is powered for six hours.

The UPSS is designed and tested to be resistant to the electromagnetic interference (EMI)/radio frequency interference (RFI) environment. When equipment (e.g., portable radios) poses risks to the UPSS equipment or distribution wiring, administrative controls prevent the use of the equipment where it can adversely affect the UPSS.

Safety-related UPSS equipment is located in a mild environment, is not subject to harsh environmental conditions during normal operation or transient conditions, and has no significant aging mechanisms. This equipment is designed and qualified by applying the guidance of Sections 4.1, 5.1, 6.1, and 7 of IEEE 323 (IEEE, 2003), and is qualified to the environmental parameters provided in Tables7.2-2 and 7.2-3.

8a2.2.4 STANDBY GENERATOR SYSTEM DESIGN BASIS The design of the SGS is based on Criterion 27, Electrical power systems, and Criterion 28, Inspection and testing of electric power systems, of the SHINE design criteria. The SHINE design criteria are described in Section3.1.

The purpose of the SGS is to provide a temporary source of nonsafety-related alternate power to the UPSS and selected additional loads for operational convenience and defense-in-depth.

The SGS:

Will provide for the separation or isolation of safety-related circuits from nonsafety-related circuits, including the avoidance of electromagnetic interference with safety-related instrumentation and control functions; Will provide an alternate source of power for the safety-related electrical buses;

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-5 Rev. 3 Will provide an alternate source of power to systems required for life-safety or important for facility monitoring; Will automatically start and supply loads upon a loss of off-site power; and Permits appropriate periodic inspection and testing to assess the continuity of the system and the condition of components.

8a2.2.5 STANDBY GENERATOR SYSTEM CODES AND STANDARDS The SGS is designed in accordance with NFPA 70 - 2017, National Electrical Code (NFPA, 2017) as adopted by the State of Wisconsin (Chapter SPS 316 of the Wisconsin Administrative Code, Electrical).

8a2.2.6 STANDBY GENERATOR SYSTEM DESCRIPTION The SGS consists of a 480Y/277 VAC, 60 Hertz (Hz) natural gas-driven generator, a 480 VAC switchgear, and two SGS cross-tie breakerstransfer switches to allow the SGS switchgear to be connected to either or both emergency 480 VAC NPSS buses. Upon a loss of off-site power (LOOP) (i.e., undervoltage or overvoltage sensed on utility service), the SGS automatically starts, both non-vital breakers (NVBKR 1 and NV BKR 2) automatically open, and the associated SGS cross-tie breakers (SGS BKR 1 and SGS BKR 2) automatically closetransfer switches operate to provide power to the associated emergency 480 VAC NPSS bus. Upon a loss of normal power to any transfer switch, the SGS automatically starts, the associated non-vital breaker (NV BKR 1 or NVBKR 2) automatically opens, and the associated transfer switch operates to provide power to the associated emergency 480 VAC NPSS bus.

The loads supplied by the SGS include the loads supplied by the UPSS (see Table8a2.2-1), as well as the following facility loads:

Emergency lighting Facility data and communications system (FDCS) equipment Radiation area monitoring system (RAMS) detectors Continuous air monitoring system (CAMS) detectors Facility fire detection and suppression system (FFPS)

Hot cell fire detection and suppression system (HCFD)

PICS equipment PVVS equipment TPS SEC heaters Switchgear station batteries (NPSS, SGS)

Facility access control system (FACS)

Facility ventilation zone 4 (FVZ4) UPSS battery room and equipment room exhaust fans FDCS dedicated cooling systems FDCS equipment, PICS equipment, and the FFPS contain nonsafety-related unit batteries or local uninterruptible power supplies to provide power to span the time between the LOOP event and the start of the SGS.

Emergency lighting located inside the main production facility is provided with unit batteries capable of supplying 90 minutes of illumination.

Operation of the SGS is not required for any safety function at the SHINE facility.

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-7 Rev. 3 Table 8a2.2 UPSS Load List (Sheet 1 of 2)

Load Description kVA Loads UPS-A kVA Loads UPS-B Required Runtime Target solution vessel (TSV) off-gas system (TOGS)

Blowers (8) 48.175.2 48.175.2 5 Min Recombiner heaters (16) 32.08 32.08 5 Min Nitrogen purge system (N2PS) valves 2.20.5 2.20.5 5 Min TSV dump valves 3.0.4 3.0.4 5 Min Neutron flux detection system (NFDS) 12.0 12.0 120 Min TSV reactivity protection system (TRPS) 1.5 1.5 120 Min TRPS radiation monitors 7.7 7.7 120 Min Engineered safety features actuation system(ESFAS) radiation monitors 7.7 7.7 120 Min Neutron driver assembly system (NDAS) hold circuits (8)

Vacuum transfer system (VTS) hold circuits (3)

Molybdenum extraction and purification system (MEPS) pump hold circuits (3)

Radiological ventilation exhaust and supply fans hold circuit 0.51 0.51 120 Min 0.2 0.2 120 Min 0.2 0.2 120 Min 0.4 0.4 120 Min ESFAS 0.5 0.45 6 Hrs Tritium purification system (TPS) tritium monitors (9) 1.2.4 1.2.4 6 Hrs Criticality accident alarm system (CAAS),

nonsafety-related 0.98 0.98 120 Min Stack release monitoring system (SRMS),

nonsafety-related 0.0 3.8 120 Min

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-8 Rev. 3 TPS secondary enclosure cleanup (SEC) blowers (3), nonsafety-related 0.21.6 0.38 6 Hrs Note: Required charger kVA does not include battery charging Total:

11843.2 1246.2 Required Reserve:

314.3 3214.6 Minimum Charger kVA:

14957.5 15560.8 Table 8a2.2 UPSS Load List (Sheet 2 of 2)

Load Description kVA Loads UPS-A kVA Loads UPS-B Required Runtime

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-9 Rev. 3 Table 8a2.2 UPSS Battery Sizing (Sheet 1 of 2)

Load Description Amp-Hours Battery A Amp-Hours Battery B Target solution vessel (TSV) off-gas system (TOGS)

Blowers (8) 581 581 Recombiner heaters (16) 312 312 Nitrogen purge system (N2PS) valves 21 21 TSV dump valves 31 31 Neutron flux detection system (NFDS) 3510 3510 TSV reactivity protection system (TRPS) 34 34 TRPS radiation monitors 224198 224198 Engineered safety features actuation system(ESFAS) radiation monitors 224198 224198 Neutron driver assembly system (NDAS) hold circuits (8)

Vacuum transfer system (VTS) hold circuits (3)

Molybdenum extraction and purification system(MEPS) pump hold circuits (3)

Radiological ventilation exhaust and supply fans hold circuit 163 163 6

6 6

6 12 12 ESFAS 334 2937 Tritium purification system (TPS) tritium monitors (9) 10586 1405 Criticality accident alarm system (CAAS), nonsafety-related 261 261 Stack release monitoring system (SRMS), nonsafety-related 0

11299

Chapter 8 - Electrical Power Systems Emergency Electrical Power Systems SHINE Medical Technologies 8a2.2-10 Rev. 3 TPS secondary enclosure cleanup (SEC) subsystem, nonsafety-related Blowers 615 2961 Note: Total amp-hours include inverter efficiency Subtotal:

114463 126718 Subtotal with 1.25% aging factor:

14530 158422 Total with 10% reserve and 80% dischargemargin for future loads:

19671 217751 Table 8a2.2 UPSS Battery Sizing (Sheet 2 of 2)

Load Description Amp-Hours Battery A Amp-Hours Battery B

Chapter 8 - Electrical Power Systems References SHINE Medical Technologies 8a2.3-1 Rev. 1 8a

2.3 REFERENCES

IEEE, 2004. Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Systems, IEEE 344-2004, Institute of Electrical and Electronics Engineers, 2004.

IEEE, 2003. Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, IEEE 323-2003, Institute of Electrical and Electronics Engineers, 2003.

IEEE, 2008. Standard Criteria for Independence of Class 1E Equipment and Circuits, IEEE 384-2008, Institute of Electrical and Electronics Engineers, 2008.

IEEE, 2010. Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications, IEEE 485-2010, Institute of Electrical and Electronics Engineers, 2010.

IEEE, 200413. Recommended PracticeIEEE Standard for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating SystemsStations, IEEE 344-200413, Institute of Electrical and Electronics Engineers, 200413.

NFPA, 2017. National Electrical Code, NFPA 70, National Fire Protection Association, 2017.

4 pages follow ENCLOSURE 3 ATTACHMENT 2 SHINE MEDICAL TECHNOLOGIES, LLC SHINE MEDICAL TECHNOLOGIES, LLC OPERATING LICENSE APPLICATION SUPPLEMENT NO. 6 AND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATIONS CHANGES PUBLIC VERSION (MARK-UP)

Page 3.6-1 Revision 34 3.6 Emergency Power Objective:

To ensure that safety-related emergency power is available to prevent or mitigate the consequences of design basis accidents.

LCO 3.6.1 Two Divisions of the UPSS shall be Operable. A Division of UPSS is considered Operable if:

1. The battery, battery charger, inverter, AC distribution panel, and DC distribution panel are Operable,

2. The inverter is supplied by the DC distribution panel and is supplying power to the AC distribution panel, and 1.3.

The battery and battery charger are connected to the DC distribution panel.

Applicability Facility not Secure Action According to Table 3.6.1 SR 3.6.1

1. UPSS battery voltage and specific gravity shall be checked semi-annually.

2. UPSS battery charger and inverter voltage shall be checked semi-annually.

2.3.

UPSS discharge test shall be performed every five years.

Page B 3.6-1 Revision 34 Basis 3.6.1 LCO The safety-related uninterruptible electrical power supply system (UPSS) for the facility consists of two redundant Divisions of 125-volt direct current (VDC) batteries, inverters, bypass transformers, distribution panels, and other breakers and distribution equipment necessary to feed safety-related alternating current (AC) and direct current (DC) loads, as described in FSAR Section 8a2.2. The 48V power supplies for the TRPS and ESFAS cabinets are also within the scope of this LCO for the UPSS distribution system. The UPSS provides an emergency back-up power supply for safety-related equipment and monitoring which protects against a total or partial loss of normal facility power.

The UPSS minimum Operable Divisions ensures there is adequate backup battery power for postulated accident scenarios, as described in FSAR Subsection 8a2.2.3. A Division of the UPSS is Operable if it is capable of supplying power to its safety-related AC and DC loads from the battery, the breakers integral to the UPSS are capable of protecting the UPSS from an upstream fault, and the battery is sufficiently charged such that it is capable of supplying power for the minimum runtimes specified in FSAR Table 8a2.2-1.

A battery is considered Operable when battery specific gravity is in the range of 1.210 and 1.300 at 77°F and battery voltage is at or greater than the minimum battery voltage provided in Table B-3.6.1.

A battery charger is considered Operable when it is energized to the voltage provided in Table B-3.6.1 and is connected to its associated DC distribution panel.

An inverter is considered Operable when it is energized to the voltage provided in Table B-3.6.1 at a frequency of 60 Hz +/- 1 Hz.

A DC distribution panel is the switchgear to which the battery, battery charger and inverter connect. A DC distribution panel is considered Operable when it is energized from either the battery or battery charger.

An AC distribution panel is the switchgear which the inverter supplies. An AC distribution panel is considered Operable when it is energized by the inverter.

Additionally, for a UPSS Division to be Operable, the inverter must be supplied by the DC distribution panel and must be supplying power to the AC distribution panel and the battery and battery charger must be connected to the DC distribution panel. This configuration ensures availability of required power on a loss of off-site power.

A single overall electrical power system serves the main production facility, including both the irradiation facility and the radioisotope production facility, as well as the site and support buildings, as described in FSAR Section 8a2.1. The normal electrical power supply system receives off-site power from the local utility in two separate feeds for improved reliability.

The standby generator system (SGS) consists of a 480Y/277 VAC, 60 Hertz natural gas-driven generator, as described in FSAR Subsection 8a2.2.6.

Although not required by the accident analysis, the SGS is designed to automatically start and begin step loading within one minute of and complete power transfers within five minutes of the loss of off-site power (LOOP). The SGS is sized to carry the full load of both Divisions of the UPSS. The SGS supplies

Page B 3.6-2 Revision 34 power to the UPSS buses, re-charges the UPSS batteries, supplies additional loads used for life-safety or facility monitoring, and allows operational flexibility while responding to the LOOP.

One Division of UPSS may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to perform corrective or preventative maintenance. The 72-hour completion time is based on the availability of off-site power and the SGS. This provides a reasonable time to restore the UPSS to Operable status with an acceptably low risk. It also provides sufficient time to prepare and implement an orderly and safe facility shutdown if the UPSS is not restored to Operable status.

With both Divisions of UPSS inoperable, or if one Division is inoperable for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, facility processes are required to be shut down to the extent practicable to minimize the risk of an accident coincident with a loss of off-site power. Placing all IUs undergoing irradiation in Mode 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> minimizes the hydrogen production in IUs via radiolysis and may be performed via a manual IU Cell Safety Actuation or by cycling the IU Operating Mode through Mode 2 to Mode 3 to perform a normal shutdown. Opening the VTS vacuum pump breakers and at least one VTS vacuum break valve within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> stops the transfer of radioactive liquids throughout the facility and may be performed via a manual VTS Safety Actuation or by performing the actions individually. Placing tritium in the TPS gloveboxes in its storage location within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> minimizes the risk of a release; otherwise, isolating the TPS gloveboxes within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> via a TPS Train Isolation places the TPS gloveboxes in an isolated condition. The likelihood of an event requiring the UPSS to be used during the allowed completion time is low. The completion times allow for adequate time to place the affected components in a safe condition and minimize the risk of extended operation with both divisions of UPSS inoperable.

SR The battery voltage, specific gravity, and discharge surveillance requirements ensure the operability of the UPSS and are consistent with the frequencies stated in ANSI/ANS 15.1-2007.

The battery charger and inverter voltage surveillance requirements ensure that this equipment is functioning properly to support the operability of the UPSS.

Table B-3.6.1 UPSS Voltage Ranges Component Voltage Requirements

a.

Battery 105 VDC

b.

Battery charger 132 and 143 VDC

c.

Inverter 204 and 212 VAC Basis 3.6.2 LCO The safety-related breakers are required to be Operable to support the safety functions described in FSAR Sections 7.4.3.1 and 7.5.3.1. Safety-related

Page 5.0-9 Revision 34 Table 5.5.4 Controls Category Characteristic Main Production Facility Structure The facility structure is designed to protect safety-related SSCs from tornado winds and missile loads, heavy snow and ice loading, and the impact of design basis (small) aircraft, including the effects of fire, as described in FSAR Section 3.4.

The design of the facility includes provisions for the prevention of adverse effects of flooding from external sources, barriers in the RPCS room prevent flooding from leaving the RPCS room, and the floor of the URSS and TSPS room is elevated to prevent water intrusion in the event of internal flood or use of water for fire suppression in other areas of the facility. The RPF sub-grade vaults and trenches are sealed to resist water intrusion that could compromise the function of the RDS sump.

The interior design includes shield walls for line of sight trajectory paths from the overhead doors to the interior structure to protect against possible intrusion of tornado missiles.

The overhead crane in the irradiation facility is designed as single-failure proof to protect against heavy load drops.

The RPF shielding cover plugs are designed to maintain their structural integrity, to protect equipment located in the vaults, trench, and pits, in the event of a heavy load drop, or a specific evaluation of the potential for and effects of a dropped load is performed.

The TPS-NDAS Interface lines are protected from damage due to external impacts by 1) the majority of tube lengths are run through subgrade sleeves and are protected by rebar re-enforced concrete, and

2) sections that are above grade in the TPS rooms are protected by mechanical guards.

Irradiation Unit The PSB piping and structural supports for SCAS and TOGS safety-related equipment are seismically qualified.

The VTS is designed such that the TSV fill lift tank has a drain path to the target solution hold tank The external design pressure of the TSV is greater than the maximum design cooling water pressure of the PCLS.

The TSV fill line is designed to prevent reverse flow of target solution from the TSV from occurring.

The PSB contains redundant pressure relief valves and redundant vacuum relief valves to limit pressure transients.