ML22069B099
| ML22069B099 | |
| Person / Time | |
|---|---|
| Site: | Hermes |
| Issue date: | 03/10/2022 |
| From: | Kairos Power |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22069B097 | List: |
| References | |
| KP-NRC-2203-004 | |
| Download: ML22069B099 (9) | |
Text
Enclosure 1 Kairos Power Response to NRC Question 3.21, 3.23, 3.24, and 3.51 (NonProprietary)
Question Number: 3.21 Section 3.2.3.2 of the PSAR states that the maximum hurricane wind speed (V) is taken from RG 1.221. In the PSAR this value is identified as 130 mph. Reading Figure 2 in RG 1.221 it appears V should be 140 mph. Please explain how it was determined that 130 mph was the proper value in accordance with RG 1.221.
Kairos Power Response:
Figure 1 (attached) shows the location of the Hermes site superimposed on Figure 2 from RG 1.221, DesignBasis Hurricane Windspeeds for the Eastern Gulf of Mexico and Southeastern Atlantic U.S.
Coastline Representing Exceedance Probabilities of 107 per Year. Values are nominal 3second gust windspeeds in miles per hour (meters per second) at 33 ft (10 m) above ground over open terrain (reproduced from NUREG/CR7005). The 130mph line from RG 1.221, Figure 2, is almost directly above the Hermes site. Paragraph C.1 of RG 1.221 states that [l]inear interpolation for sites located between two wind contour lines is permitted. Linear interpolation would change the hurricane wind speed by about 1 mph, which would not affect the results of the structural analysis or design.
For this reason, Kairos Power affirms that it is reasonable to use 130 mph as the maximum hurricane wind speed.
Impact on Licensing Document:
This response has no impact on the content of the Hermes NonPower Reactor Preliminary Safety Analysis Report.
References:
- 1. Regulatory Guide 1.221, DesignBasis Hurricane and Hurricane Missiles for Nuclear Power Plants, Revision 0, October 2011 Page 1 of 2
Figure 1. Location of the Hermes Site Superimposed on RG 1.221, Figure 2 Page 2 of 2
Question Number: 3.23 PSAR Section 3.2.4.2 lists the ground snow load, pg, in Eq. 3.22 as 10 pounds per square foot (psf) and notes that is taken from Fig. 71 of American Society of Civil Engineers (ASCE) 710. Section 7.2 of ASCE 710 notes that pg is based on a 50year recurrence interval. Preliminary Safety Analysis Report (PSAR) Section 2.3.1.11 notes that the value of 10 psf must be adjusted by a factor of 1.22 to determine the 100year return value of 12.2 psf. PSAR 2.3.1.11 further notes that the normal winter precipitation event at the site is determined to be 21.9 psf. Explain why 10 psf was taken as the ground snow load instead of 12.2 psf or 21.9 psf.
Kairos Power Response:
Kairos Power will change the ground snow load in Section 3.2.4.2 from 10 psf to 21.9 psf for consistency with Section 2.3.1.11.
Impact on Licensing Document:
This response impacts Sections 3.2.4.2 of the Hermes NonPower Reactor Preliminary Safety Analysis Report. A markup of the affected sections is provided with this response.
References:
None Page 1 of 1
Question Number: 3.24 PSAR Section 3.2.4 There is no discussion of rainonsnow surcharge loads as discussed in ASCE 7 10 Section 7.10. Explain how the rainonsnow surcharge is addressed or why it does not apply.
Kairos Power Response:
Kairos Power intends to apply ASCE 710 concerning rainonsnow surcharge loads if such loads apply to the final roof design. Kairos Power will add content in Section 3.2.4.2 stating that the roof design will be consistent with ASCE 710, including Section 7.10 if applicable.
Impact on Licensing Document:
This response impacts Section 3.2.4.2 of the Hermes NonPower Reactor Preliminary Safety Analysis Report. A markup of the affected sections is provided with this response.
References:
None Page 1 of 1
Question Number: 3.5-1 PSAR Section 3.5.3.2.2 explains how safetyrelated SSCs will be protected from internal flooding due to fire protection; however, there is no discussion of how SSCs will be protected from the actuation of the fire protection system. Explain how safetyrelated SSCs will be protected from fire protection water.
Kairos Power Response:
Kairos Powers intent was that safetyrelated structures, systems, and components would be protected both from actuation of the fire protection system (i.e., spray) and from fire protection water on the floor (i.e., pooling). To clarify this in the PSAR, Kairos Power will to add content to Section 3.5.3.2.2 stating that intent.
Impact on Licensing Document:
This response impacts Section 3.5.3.2.2 of the Hermes NonPower Reactor Preliminary Safety Analysis Report. A markup of the affected sections is provided with this response.
References:
None Page 1 of 1
Preliminary Safety Analysis Report Design of Structures, Systems, and Components 3.2.4.1 Applicable Design Parameters Based on Risk Category IV characterization (See Section 3.2.1.1) and site location, Chapters 1 and 7 of ASCE/SEI 710 provide snow load design parameters to be applied to the safetyrelated portions of the Reactor Building.
3.2.4.2 Determination of Applied Forces The sloped roof (balanced) snow load is calculated by Equation 3.23 as derived from ASCE/SEI 710, Section 7.3 and Section 7.4 using the ground snow load specified in Section 2.3.1.11.
ps = 0.7CsCeCtIspg (Equation 3.22)
- Where, Cs = roof slope factor as determined by Sections 7.4.1 through Section 7.4.4 of ASCE/SEI7 10 corresponding to the geometry of the roof Ce = exposure factor as determined by Table 72 of ASCE/SEI 710 equal to 1.0 Ct = thermal factor as determined by Table 73 of ASCE/SEI 710 equal to 1.0 Is = importance factor as determined by Table 1.51 of ASCE/SEI 710 and 1.52 of ASCE/SEI 710 equal to 1.2 pg = ground snow load as set forth in Figure 71 of ASCE/SEI 710 consistent with Section 2.3.1.11 equal to 1021.9 psf Unbalanced snow loads on the ceiling of the safetyrelated portion of the Reactor Building are determined in accordance with Section 7.6 of ASCE/SEI 710. The design snow drift loads are determined in accordance with Section 7.7 of ASCE/SEI 710. If applicable to the roof design, rain on snow surcharge loads are determined in accordance with Section 7.10 of ASCE/SEI 710.
3.5-1 References
- 1. American Society of Civil Engineers, Seismic Engineering Institute, ASCE/SEI 710, Minimum Design Loads for Buildings and Other Structures. 2010.
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Preliminary Safety Analysis Report Design of Structures, Systems, and Components The facility is a passively dry site with respect to external flooding hazards. Section 3.3 describes that in the PMF event, there are no loads on the safetyrelated portion of the Reactor Building that is above grade. The basement containing the seismic isolator units is about 20 feet below grade. The safety related portion of the Reactor Building is designed to withstand buoyant forces and groundwater associated with the PMF.
No SSCs located in the basement are credited to mitigate the effects of a postulated external flood event. The basemat of the safety related portion of the Reactor Building, which is supported by the base isolators, as discussed in Section 3.5.1, is at grade level and no safetyrelated SSCs in the safetyrelated portion of the reactor building are below that basemat elevation. Therefore, PDC 2 is met for PMF events based on the location above grade level of all safetyrelated SSCs that are credited to mitigate the effects of a postulated external flood.
Although they do not perform a safety function to mitigate the adverse effects of a postulated external flood event, the seismic isolator units are on elevated pedestals above the foundation slab. The base isolation basement is a reinforced concrete safetyrelated structure with the following features:
Water stops are provided in construction joints below flood level.
External surfaces exposed to flood level have waterproof coating.
Furthermore, the safetyrelated portion of the Reactor Building is a reinforced concrete structure designed to meet ACI 3492013. ACI 3492013 is specific to the design of safetyrelated nuclear structures and has builtin margin. ACI 349 is used to design a structure that can withstand the postulated external flooding water loads from Section 3.3. With respect to buoyant forces from a postulated external flood event on the basement area of the safetyrelated portion of the Reactor Building, based on a flood level no higher than grade, the weight of the building offsets the potential buoyant forces on the basement. By designing in accordance with ACI 3492013, the safetyrelated portion of the Reactor Building satisfies PDC 2 for design basis loads from external flooding as discussed in Section 3.3.
Finally, consistent with PDC 2, grading and drainage on the site preclude loads from precipitation affecting the safetyrelated portion of the Reactor Building. Specific grading and drainage features will be described in the application for an Operating License.
3.5.3.2.2 Internal Flood and Spray Design Features This section describes the design features that satisfy PDC 2 with respect to protection from internal flooding for safetyrelated SSCs. Safetyrelated SSCs that are vulnerable to water damage from internal spray or floods are elevated above the floor and shielded, or otherwise protected, from potential spray.
Water is directed away from enclosures for safetyrelated equipment and sloped floors and curbs preclude water entry into these areas. Where there is a potential for pebbles to be on a sloped or curbed floor, features prevent pebbles from rolling so that pebbles on the floor of the safetyrelated portion of the Reactor Building maintain a geometrically safe configuration for criticality.
Internal flooding or spraying in the safetyrelated portion of the Reactor Building has three potential sources: water system with SSCs located in the safetyrelated portion of the reactor building, water system SSCs located in the nonsafety related portion of the Reactor Building, and fire protection water.
For water systems with SSCs located in the safetyrelated portion of the Reactor Building, the amount of water is limited by design. The maximum flow rate and the volume of water available for release from a break in the safetyrelated portion of the Reactor Building, is used to determine the effect of internal flooding or spraying on safetyrelated equipment. The quantity and flow rate of water is limited to the Kairos Power Hermes Reactor 324 Revision 0
Preliminary Safety Analysis Report Design of Structures, Systems, and Components gravitydriven pressure head above the break location. A pump trip in a water system is assumed to terminate the flow and a constrained amount of fluid is assumed to spill into the facility.
For water sources external to the safetyrelated portion of the Reactor Building (e.g., fire water),
automatic or a manual termination of flow will be specified in the application for the Operating License.
The fire protection system implements NFPA 801, Standard for Fire Protection for Facilities Handling Radioactive Materials (Reference 3). The water collection due to the potential failure of the fire protection piping is bounded by the total discharge from the operation of the fire protection system.
The water collection system can accommodate the total firefighting water volume. Sloped floors and curbs prevent fire protection water from draining into the radioactive waste handling system drains.
Spray shields, or similar, prevent fire protection water from spraying safetyrelated SSCs that would be sensitive to water spray.
Safetyrelated SSCs are protected from spilled Flibe and Flibebearing components are also protected from water to prevent interaction between water and Flibe. Features include steel liners, catch pans or troughs, or similar design solutions.
Those pipes, vessels, and tanks with the potential to flood or spray safetyrelated portions of the Reactor Building are seismically qualified in accordance with local building code and consistent with the seismic design category based on the SSCs safety classification. There is no high energy piping in the safetyrelated portion of the Reactor Building, therefore a high energy break is not considered.
Further information on the analysis of the impacts of internal flooding and spraying will be provided with the application for an Operating License.
3.5.3.3 Conformance with PDC 2 for Earthquakes Section 3.4 discussed the design basis earthquake characteristics that are the input for the design of the safetyrelated portion of the Reactor Building. The safetyrelated portion of the reactor building is designed consistent with the graded approach in ASCE 4319 (Reference 4). See Section 3.4 for more information about the graded approach. By meeting ASCE 4319, the safetyrelated portion of the Reactor Building provides protection for safetyrelated SSCs from design basis earthquakes, consistent with PDC 2.
The safetyrelated portion of the Reactor Building uses base isolation as described in Section 3.5.1. The seismic isolation system is designed to limit the loads from design basis earthquakes on safetyrelated SSC, consistent with PDC 2.
3.5.3.3.1 Seismic Design of the SafetyRelated Portion of the Reactor Building Seismic qualification of SDC3 structures follows the requirements of Section 5 of ASCE 4319. Structural demands are determined based on the results of the response analysis outlined in Section 3.4.1. In addition to the seismic effects, the effects from gravity, operating loads, and other concurrent loading (e.g. snow) are considered on the structural demands.
Seismic acceptance is checked for both strength and displacementbased criteria summarized in Section 5.2.2 and 5.2.3 of ASCE 4319, respectively, for the applicable limit states. Strengthbased qualification of structural elements utilize, when appropriate, the inelastic energy absorption factors discussed in Section 5.1.3 of ASCE 4319 and summarized in Table 51 of ASCE 4319. Allowable drift and rotation limits are based on the discussion in Section 5.2.3 of ASCE 4319 and summarized in Tables 52 and 53 of ASCE 4319.
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