RAIO-0818-61631, LLC Response to NRC Request for Additional Information No. 489 (Erai No. 9534) on the NuScale Design Certification Application
| ML18242A685 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 08/30/2018 |
| From: | Rad Z NuScale |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| RAIO-0818-61631 | |
| Download: ML18242A685 (11) | |
Text
RAIO-0818-61631 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com August 30, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Response to NRC Request for Additional Information No.
489 (eRAI No. 9534) on the NuScale Design Certification Application
REFERENCE:
U.S. Nuclear Regulatory Commission, "Request for Additional Information No.
489 (eRAI No. 9534)," dated June 15, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).
The Enclosure to this letter contains NuScale's response to the following RAI Questions from NRC eRAI No. 9534:
06.04-4 06.04-5 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions on this response, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.
Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution:
Gregory Cranston, NRC, OWFN-8G9A Omid Tabatabai, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9534 Sincerely, Zackary W. Rad Director, Regulatory Affairs
RAIO-0818-61631 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com :
NuScale Response to NRC Request for Additional Information eRAI No. 9534
Response to Request for Additional Information Docket No.52-048 eRAI No.: 9534 Date of RAI Issue: 06/15/2018 NRC Question No.: 06.04-4 Regulatory Basis:
10 CFR 52.47(a)(2) requires that a standard design certification application include a final safety analysis report (FSAR) that describes the design of the facility including the principal design criteria for the facility, for which NuScale used the 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
General Design Criterion (GDC) 19 requires that a control room be provided with adequate radiation protection to permit access and occupancy of the control room under accident conditions without the personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
Question:
The applicant's response to request for information (RAI) 9079 discusses the basis for the dose analysis modeling assumption that after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (after the control room habitability system (CRHS) is exhausted) the normal control room heating ventilation and air conditioning system (CRVS) is assumed to operate in supplemental filtration mode. The discussion of the CRVS reliability and operability is mainly focused on isolation and filtration component capabilities and backup power. The RAI response did not discuss augmented quality with respect to the capability to recover the CRVS for reasons other than loss of power. NuScale does not consider failure of the CRVS to operate post-72 hours concurrent with a design basis accident (DBA) to be within the design basis for evaluation of the radiological consequences of DBAs. However, if recovery of the CRVS supplemental filtration mode capability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is not sufficiently and reliably shown to be ensured for accident conditions, then the FSAR Chapter 15 dose analysis assumptions on filtration and removal of radioactive material in the control room ventilation intake are not justified and the dose results may exceed the dose criterion of GDC 19.
Considering that the NuScale FSAR does not include technical specifications for the CRVS, specific testing and inspection requirements for the CRVS are left to the combined license NuScale Nonproprietary
(COL) applicant (COL Item 9.4-1), and the CRVS is not classified as Seismic Category I except for the components that isolate the control room, the staff requires additional information regarding the CRVS supplemental filtration capability to limit dose to control room operators under accident conditions. Specifically, the staff requests the following information in order to complete its review by fully evaluating the importance of the post-72 hours operation of the CRVS supplemental filtration mode on the NuScale design ability to meet the requirements of GDC 19:
Provide a sensitivity analysis, including both a qualitative and quantitative assessment, evaluating the effect on the control room operator dose for DBAs for the case where after the CRHS is exhausted, the CRVS supplemental filtration mode is not recovered within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as assumed in the DBA control room dose analyses described in FSAR Chapter 15.0.3. Describe the analysis assumptions and inputs, as well as the dose results. For this sensitivity case, would the GDC 19 dose criterion of 5 rem TEDE be met for all DBAs without credit for CRVS filtration after the CRHS is exhausted?
NuScale Response:
A sensitivity study has been performed to evaluate the effect on the control room operator dose for DBA scenarios where after the CRHS is activated and exhausted, the CRVS supplemental filtration mode is not recovered within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as assumed in the design basis dose analysis models. This sensitivity study, which is for informational purposes only and not considered to be a source of design basis dose results, evaluates whether the GDC 19 control room dose criterion of 5 rem TEDE is met for all DBA scenarios presented in Chapter 15 without credit for CRVS filtration after the CRHS is exhausted. Additionally, this sensitivity study evaluates whether the GDC 19 control room dose criterion of 5 rem TEDE is met for all DBA scenarios in the event of total CRVS failure without CRHS activation and exhaustion.
The inputs and assumptions used in this sensitivity study, including the assumed accident event progression phases, are unchanged from those used in the design basis dose consequence models, with the exception of all active-CRVS-dependent flow rates being set to zero for the duration of the evaluated event. This CRVS flow rate perturbation facilitates the simulation of total CRVS failure from event onset throughout the event duration.
It is further noted that the calculation revisions necessitated by this sensitivity study incorporated updated primary coolant source term input to the applicable steam generator tube failure, main steam line break, and small line break DBA evaluations. Accordingly, updated design basis control room dose consequence values are provided with the results of this sensitivity study, for comparison purposes. Updated design basis values and sensitivity case values are provided for comparison purposes here with a degree of precision consistent with that of final reported design basis values. Given the low magnitude of certain calculated values, NuScale Nonproprietary
a qualitative comparison between sensitivity case values and design basis values may not be practical. In this respect, a comparison of all calculated values against the 5 rem TEDE acceptance criterion is considered the key observation/conclusion of this study.
For sensitivity baselining purposes, updated control room dose calculations are provided in Table 1 and Table 2 of this RAI response. Table 1 provides calculated dose consequences for a scenario of uninterrupted power supply with continuous filtered airflow to the control room for the event duration. Table 2 provides calculated dose consequences for a scenario of loss of power with CRHS activation and restored filtered airflow to control room envelope at time of CRHS depletion. The maximum of Table 1 and Table 2 results for a given DBA will be taken as design basis values, with the exception of the core damage maximum hypothetical accident results, which are not currently planned for inclusion in the FSAR.
Table 1.
Results summary continuous filtered airflow Event Control Room Dose (rem TEDE)
Failure of Small Lines Carrying Primary Coolant Outside Containment 0.07 Steam Generator Tube Failure (pre-incident iodine spike) 0.16 Steam Generator Tube Failure (coincident iodine spike)
<0.01 Main Steam Line Break (pre-incident iodine spike) 0.01 Main Steam Line Break (coincident iodine spike)
<0.01 Fuel Handling Accident 0.89 Maximum Hypothetical Accident (significant core damage) 2.14 Maximum Hypothetical Accident (pre-incident iodine spike)
<0.01 Maximum Hypothetical Accident (coincident iodine spike)
<0.01 Table 2.
Results summary 72-hour power loss with CRHS activation Event Control Room Dose (rem TEDE)
Failure of Small Lines Carrying Primary Coolant Outside Containment 0.08 Steam Generator Tube Failure (pre-incident iodine spike) 0.20 Steam Generator Tube Failure (coincident iodine spike)
<0.01 Main Steam Line Break (pre-incident iodine spike)
<0.01 Main Steam Line Break (coincident iodine spike)
<0.01 Fuel Handling Accident 0.71 Maximum Hypothetical Accident (significant core damage) 1.44 Maximum Hypothetical Accident (pre-incident iodine spike)
<0.01 Maximum Hypothetical Accident (coincident iodine spike)
<0.01 NuScale Nonproprietary
Sensitivity study control room dose calculations are provided in Table 3 and Table 4 of this RAI response. Table 3 provides calculated dose consequences for a scenario in which total CRVS failure occurs without CRHS activation and exhaustion. Table 4 provides calculated dose consequences for a scenario in which total CRVS failure occurs with CRHS activation.
Table 3.
Results summary total CRVS failure Event Control Room Dose (rem TEDE)
Failure of Small Lines Carrying Primary Coolant Outside Containment 0.10 Steam Generator Tube Failure (pre-incident iodine spike) 0.20 Steam Generator Tube Failure (coincident iodine spike)
<0.01 Main Steam Line Break (pre-incident iodine spike) 0.02 Main Steam Line Break (coincident iodine spike)
<0.01 Fuel Handling Accident 1.38 Maximum Hypothetical Accident (significant core damage) 3.94 Maximum Hypothetical Accident (pre-incident iodine spike)
<0.01 Maximum Hypothetical Accident (coincident iodine spike)
<0.01 Table 4.
Results summary total CRVS failure with CRHS activation Event Control Room Dose (rem TEDE)
Failure of Small Lines Carrying Primary Coolant Outside Containment 0.04 Steam Generator Tube Failure (pre-incident iodine spike) 0.09 Steam Generator Tube Failure (coincident iodine spike)
<0.01 Main Steam Line Break (pre-incident iodine spike)
<0.01 Main Steam Line Break (coincident iodine spike)
<0.01 Fuel Handling Accident 0.57 Maximum Hypothetical Accident (significant core damage) 2.36 Maximum Hypothetical Accident (pre-incident iodine spike)
<0.01 Maximum Hypothetical Accident (coincident iodine spike)
<0.01 As observed from these results, the 5 rem TEDE acceptance criterion for control room dose is satisfied for all evaluated total CRVS failure sensitivity scenarios.
NuScale Nonproprietary
Impact on DCA:
Table 15.0-12: Radiological Dose Consequences for Design Basis Analyses, has been revised as described in the response above and as shown in the markup provided in this response.
NuScale Nonproprietary
NuScale Final Safety Analysis Report Transient and Accident Analyses Tier 2 15.0-69 Draft Revision 2 RAI 02.03.04-1, RAI 06.04-4, RAI 15.00.03-1, RAI 15.00.03-5, RAI 15.00.03-8 Table 15.0-12: Radiological Dose Consequences for Design Basis Analyses Event Location Acceptance Criteria (rem TEDE)
Failure of Small Lines Carrying Primary Coolant Outside Containment EAB 6.3 0.110.02 LPZ 6.3 0.200.04 CR 5.0 0.440.08 Steam Generator Tube Failure (pre-incident iodine spike)
EAB 25.0 0.430.08 LPZ 25.0 0.440.08 CR 5.0 1.110.20 Steam Generator Tube Failure (coincident iodine spike)
EAB 2.5
<0.01 LPZ 2.5
<0.01 CR 5.0
<0.01 Main Steam Line Break (pre-incident iodine spike)
EAB 25.0
<0.01 LPZ 25.0 0.03<0.01 CR 5.0 0.060.01 Main Steam Line Break (coincident iodine spike)
EAB 2.5
<0.01 LPZ 2.5
<0.01 CR 5.0
<0.01 Fuel Handling Accident EAB 6.3 0.55 LPZ 6.3 0.55 CR 5.0 0.89 Design Basis Source Term (significant core damage)
EAB 25.0 0.22 LPZ 25.0 0.99 CR 5.0 1.43
Response to Request for Additional Information Docket No.52-048 eRAI No.: 9534 Date of RAI Issue: 06/15/2018 NRC Question No.: 06.04-5 Regulatory Basis:
10 CFR 52.47(a)(2) requires that a standard design certification application include an FSAR that describes the design of the facility including the principal design criteria for the facility, for which NuScale used the 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants."
General Design Criterion (GDC) 19 requires that a control room be provided with adequate radiation protection to permit access and occupancy of the control room under accident conditions without the personnel receiving radiation exposures in excess of 0.05 Sv (5 rem)
TEDE for the duration of the accident.
10 CFR Part 20 Subpart C ³Occupational Dose Limits,' states in part that the licensee shall control the occupational dose to individual adults, except for planned special exposures under 20.1206, to the following dose limits. (1) An annual limit, which is the more limiting of:
The total effective dose equivalent being equal to 5 rems (0.05 Sv) or The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 rems (0.5 Sv).
10 CFR Part 20 Subpart C ³Occupational Dose Limits,' requires consideration dose resulting from external radiation sources and dose due to the inhalation of radionuclides.
10 CFR Part 20 Subpart H 'Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas,' states that if the licensee assigns or permits the use of respiratory protection equipment to limit the intake of radioactive material, that the licensee will implement a respiratory protection program that includes the following elements:
Supervision and training of respirator users
Fit testing
NuScale Nonproprietary
Determination by a physician that the individual user is medically fit to use respiratory protection equipment:
Before the initial fitting of a face sealing respirator
Either every 12 months thereafter, or periodically at a frequency determined by a physician.
Question :
In response to RAI 9079, question (e), the applicant states that as added protection against radiation overexposure, if the area radiation monitor radiation level exceeds a limit (to be set by the licensee), the operators would trip any operating reactors, initiate decay heat removal and containment isolation, and vacate the control room. In their application, NuScale has also identified that the presence of toxic gas may result in a condition where the Main Control Room (MCR) operators may need to vacate the MCR.
In implementing its statutory authority under the Atomic Energy Act, the NRC preempts the application of the Occupational Safety and Health Act for working conditions that involve radioactive materials. That is, the training, medical, fit test, etc. requirements are provided by the NRC in 10 CFR Part 20 Subpart H. However, a Memorandum of Understanding (MOU) between the Occupational Safety and Health Administration and the NRC, states that if an NRC licensee is using respiratory protection to protect workers against non-radiological hazards (i.e.,
toxic gas), the OSHA requirements apply.
Since the action of the operators to leave the confines of the MCR would potentially expose the operators to concentrations of radiological contaminants or toxic gases, that could result in the operators exceeding the regulatory limits in the short time required to reach the alternate safe location for the MCR operators, the MCR operators need to be able to wear the appropriate respiratory protection device. The applicant has identified in COL item 6.4-1 that the COL applicant has the responsibility to ensure that the MCR operators are able to use respiratory protection equipment, which should include consideration of radiological contaminants as well as toxic gases. The NuScale FSAR does not state that respiratory protection equipment (e.g., self-contained breathing apparatus (SCBA)) would be staged in the control room as a backup if the control room becomes uninhabitable. COL item 6.4-1 tells the COL applicant to comply with RG 1.78, ³Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release,' to evaluate control room habitability for hazardous chemical releases. RG 1.78 discusses use of respiratory protection equipment for hazardous chemical releases, but does not explicitly discuss respiratory protection for radiological releases.
Therefore, the staff requires the following information in order to complete its review:
Provide a COL Item requiring the applicant to provide a storage location for SCBAs that allows the MCR operators to access, don, place the facility in a safe condition and move to a safe location without exceeding the radiation exposure limits of 10 CFR Part 20, and the exposure to toxic substance guidance in RG 1.78, in the event that both MCR ventilation systems do not function. Or provide reasoning (1)
NuScale Nonproprietary
why an alternative method would be acceptable, or (2) why SCBA would not be needed.
NuScale Response:
The simultaneous failure of the normal control room HVAC system (CRVS) and the control room habitability system (CRHS) coincident with a design basis event is not considered credible. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of CRHS operation, the normal control room ventilation system would be available for use, except after a seismic event. After a seismic event the normal control room ventilation system may not be available for use but dose consequences are not applicable because mitigating SSCs are Seismic Category I and capable of performing their safety-related and nonsafety-related functions during and following the event. Therefore a COL Item requiring a storage location for SCBA in the event that both MCR ventilation systems do not function is not needed.
In addition, as described in the response to RAI 9079 (NuScale letter RAIO-1017-56676 dated October 18, 2017, ML17291A672 ), the design specifications for the Main Control Room Ventilation systems and accompanying COL Items e.g., 6.4-1, 6.4-5 and 9.4-1, provide a high degree of system reliability as well as chemical and radiation protection for control room personnel under accident conditions. A significant beyond basis event would be required to defeat the current design.
The emergency response organization would be in place providing direction for response actions during design basis and beyond design basis events. If a high radiation or toxic chemical condition challenging the control room occurred plant personnel would respond in accordance with Severe Accident Management Guidelines. Any supplies needed to support control room activities would be provided through emergency plan resources. Based on actual conditions the emergency plan organization would determine whether control room evacuation was the safest response.
The plant design could support a control room evacuation if that was determined to be the safest response. The operators would trip any operating reactors, initiate decay heat removal and containment isolation, and vacate the control room using the necessary protective equipment as determined by the emergency plan organization. Each reactor module would then reach safe shutdown conditions without operator action.
Impact on DCA:
There are no impacts to the DCA as a result of this response.
NuScale Nonproprietary