ML18275A426

LLC Supplemental Response to NRC Request for Additional Information No. 489 (Erai No. 9534) on the NuScale Design Certification Application

ML18275A426
Person / Time
Site:	NuScale
Issue date:	10/02/2018
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-1018-62021
Download: ML18275A426 (14)

Text

RAIO-1018-62021 October 02, 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 Supplemental Response to NRC Request for Additional Information No. 489 (eRAI No. 9534) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 489 (eRAI No. 9534)," dated June 15, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 489 (eRAI No.9534)," dated August 30, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 9534:

• 06.04-4 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 Rani Franovich, NRC, OWFN-8H12 Enclosure 1: NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9534 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com

RAIO-1018-62021 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 9534 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

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 NuScale Nonproprietary

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 (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:

Background:

NuScale agreed to submit a supplemental response during a meeting on September 9th, 2018 to address the following NRC requests for clarification.

• Clarify the assumptions used for the source term sensitivity analysis.

• Clarify the following sentence in the response, 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.

• Clarify the assumptions used for the 3 MHA cases. For example, what was used for the aerosol deposition coefficients, and also, what coolant activity concentration were used, and what were the iodine spiking values?

NuScale Nonproprietary

The original RAI response has been updated to reflect these clarifications as shown below.

Supplemental Response:

A sensitivity study has been performed to evaluate the effect on the control room (CR) 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.

It is 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. This update of primary coolant source term input is pursuant to the failed fuel fraction update performed in response to RAI 9161 and the latest I-131 dose equivalent and Xe-133 dose equivalent values which were updated in response to RAI 8759. Accordingly, FSAR Table 15.0-14 was updated to provide the latest primary coolant source term input assume for dose analyses, and FSAR Table 15.0-12 was updated to provide latest dose consequence results based on the latest primary coolant source term. Additionally, FSAR Section 15.0.3.8 was updated to provide the latest I-131 dose equivalent and Xe-133 dose equivalent values used in the applicable steam generator tube failure, main steam line break, and small line break DBA evaluations.

It is further noted that the calculation revisions necessitated by this sensitivity study incorporated updated aerosol removal coefficient input to the applicable core damage MHA evaluation. This update of aerosol removal coefficient input is pursuant to the containment aerosol transport and removal evaluation update performed in response to RAI 9224.

Accordingly, FSAR Table 15.0-12 has been updated to provide the latest core damage MHA dose consequence results based on the latest aerosol removal coefficients. Aerosol removal coefficients applied in the latest core damage MHA evaluation and the core damage MHA sensitivity model for this response are provided in the following table:

NuScale Nonproprietary

Table 1. Tabulated aerosol removal rate for radiological consequence input Time (hr) Removal Rate (1/hr) 0.00E+00 0.00E+00 3.80E+00 2.20E+01 3.99E+00 9.54E+00 4.18E+00 3.66E+00 4.42E+00 2.06E+00 4.67E+00 1.83E+00 4.87E+00 1.73E+00 5.15E+00 1.98E+00 6.19E+00 1.76E+00 2.59E+01 0.00E+00 Apart from the above mentioned updates, 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, as described in FSAR Section 15.0.3, with the exception of all active-CRVS-dependent flow rates being set to zero for the duration of the evaluated event.

Design basis control room and TSC flow rate values are listed in FSAR Table 15.0-15 and FSAR Table 15.0-18, respectively. Of the flow rate values listed in FSAR Table 15.0-15 and FSAR Table 15.0-18, the following parameters are set equal to zero for the simulated event duration for the purposes of this sensitivity study:

• Control room normal flow rate,

• Control room recirculation flow rate,

• Control room unfiltered inleakage,

• TSC normal flow rate,

• TSC recirculation flow rate, and

• TSC unfiltered inleakage.

Additionally, for the core damage MHA, which has been recognized as the limiting radiological consequence model for purposes of defining accident radiation post-filtration monitor analytical limits in response to RAI 8941, two additional control room model pathways were developed:

• Environment to pre-filter CRVS flow rate (compartment volume 150 ft3; flow rate 4000 cfm), and

• Pre-filter CRVS to post-filter CRVS flow rate (compartment volume 3,709 ft3; flow rate 24,532 cfm).

NuScale Nonproprietary

These two core damage MHA model pathway flows, which are depicted in FSAR Figure 15.0-3 pursuant to updates performed in response to RAI 8941, are set to zero along with the aforementioned CR/TSC pathway flows for the purposes of this sensitivity study.

The described CRVS flow rate perturbations facilitate the simulation of total CRVS failure from event onset throughout the event duration. As described in FSAR Section 6.4.4, the only source of unfiltered leakage into the control room envelope (CRE) is due to ingress and egress; up to 5 cfm is considered for this type of inleakage. The dose analysis conservatively includes an additional 10 cfm of inleakage for the CRHS pathway when the CRHS is modeled as operational. No additional inleakage sources were considered for the purposes of this sensitivity study. As described in the NuScale response to RAI 9079, all CRE isolation dampers are paired (redundant design) and Seismic Category 1, and in the event of a failure of the CRVS coincident with a design basis event, the redundant CRE isolation dampers would close. Because this sensitivity study stipulates total CRVS failure from event onset, all active-CRVS-driven flows would be slack (0 cfm) from event onset. This condition precludes the CRVS-forced flow of unfiltered radioactive material into the CRE prior to isolation.

Updated design basis control room dose consequence values are provided, where applicable, 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, 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 2 and Table 3 of this RAI response. Table 2 provides calculated dose consequences for a scenario of uninterrupted power supply with continuous filtered airflow to the control room for the event duration. Table 3 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 2 and Table 3 results for a given DBA will be taken as design basis values.

NuScale Nonproprietary

Table 2. Results summary - continuous filtered airflow Control Room Dose Event (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 Table 3. Results summary hour power loss with CRHS activation Control Room Dose Event (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 Sensitivity study control room dose calculations are provided in Table 4 and Table 5 of this RAI response. Table 4 provides calculated dose consequences for a scenario in which total CRVS failure occurs without CRHS activation and exhaustion. Table 5 provides calculated dose consequences for a scenario in which total CRVS failure occurs with CRHS activation.

Table 4. Results summary - total CRVS failure Control Room Dose Event (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 NuScale Nonproprietary

Table 5. Results summary - total CRVS failure with CRHS activation Control Room Dose Event (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 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.

Impact on DCA:

Section 15.0.3.8, Transient and Accident Analyses, Table 15.0-12: Radiological Dose Consequences for Design Basis Analyses and Table 15.0-14: Primary Coolant Source Term have been revised as described in the response above and as shown in the markup provided in this response.

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NuScale Final Safety Analysis Report Transient and Accident Analyses

• maximum mass release - double-ended break of the CVCS letdown line

• maximum time of iodine spiking - equivalent 100 percent cross-sectional area break of the CVCS makeup line RAI 15.00.03-3 Table 15.6-5 provides the assumed mass released from the reactor and the break isolation times for the two scenarios. The total mass released from the event is the sum of the mass released from the reactor provided in Table 15.6-5 and the primary coolant from CVCS equipment and piping discussed above.

Before containment isolation occurs, primary coolant flows out of the reactor vessel through the break at a rate and duration as described in Section 15.6.2. The coolant flow results in a time-dependent release of activity in the RXB that is conservatively modeled as a direct release to the environment. After containment isolation, primary coolant leaks through one containment isolation valve (the redundant in-series valve is assumed to fail open) at the maximum leak rate allowed by design basis limits. The activity from this leak path is also assumed to flow directly to the environment with no mitigation or reduction by intervening structures. After 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, the reactor is assumed to be shut down and depressurized, and releases through the containment isolation valve stop.

The following is a summary of the assumptions used from Appendix E (main steam line break) of RG 1.183:

• coincident iodine spiking factor- 500

• duration of coincident iodine spike- 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

• iodine chemical form- 97 percent elemental iodine and 3 percent organic iodide

• activity released from the fuel due to the iodine spike is assumed to mix instantaneously and homogeneously within the primary coolant in the reactor vessel

• no reduction or mitigation of noble gas radionuclides released from the primary system RAI 06.04-4S1 The primary coolant in the reactor vessel and CVCS equipment and piping in the RXB initially contains the allowable concentration of dose equivalent (DE) I-131 of 0.23.7E-02 Ci/gm and DE Xe-133 of 6010 Ci/gm.

There are no single failures for this event that affect the thermal-hydraulic response of the NPM. However, the failure of one of the two containment isolation valves on the faulted line is assumed in the dose consequence analysis.

RADTRAD is used to determine the dose, as outlined in Section 15.0.3.3.5. The control room model is described in Section 15.0.3.7.1. The potential radiological consequences of the small lines carrying primary coolant break outside containment event are presented in Table 15.0-12.

Tier 2 15.0-31 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses 15.0.3.8.2 Steam Generator Tube Failure Radiological consequences of the SGTF are calculated based on the guidance provided in Appendix F of RG 1.183.

Section 15.6.3 describes the sequence of events and thermal-hydraulic response to an SGTF. The SGTF analysis shows that the reactor core remains covered and no fuel failures occur.

This radiological consequence analysis considers the SGTF event with two different initial iodine concentrations, one based on a pre-incident iodine spike and the other based on a coincident iodine spike. A description of the scenario evaluated is summarized as follows:

1) An SGTF occurs in one of the two SGs.

RAI 06.04-4S1

2) For each of the iodine spiking scenarios, the iodine and noble gas coolant activity is calculated based on the maximum concentrations allowed by design basis limits. The primary coolant contains a concentration of 0.23.7E-02 Ci/gm DE I-131 for the coincident Iodine spike scenario and 122.2 Ci/gm DE I-131 for the pre-incident Iodine spike scenario. For both iodine spiking scenarios, the primary coolant contains 6010 Ci/gm DE Xe-133.

3) Primary coolant flows into the secondary coolant through the failed SG tube at a rate and duration defined by the transient analysis described in Section 15.6.3.

4) Primary coolant leaks into the secondary side of the intact SGs at the maximum leak rate of 150 gallons per day allowed by design basis limits. The leakage continues until the primary system pressure is less than the secondary system pressure.

5) A time-dependent release is modeled that results in releasing the activity directly to the environment through the break.

6) Once secondary system isolation occurs, both steam lines continue to release small quantities of radioactivity through valve leakage into the RXB which is assumed to go directly into the environment without any source term reduction.

7) At 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, the primary and secondary systems equalize and valve leakage stops.

Assumptions used from Appendix F of RG 1.183 are:

• coincident iodine spiking factor- 335

• duration of coincident iodine spike- 8 hr

• density for leak rate conversion- 62.4 lbm/ft3 Tier 2 15.0-32 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses

• iodine chemical form- 97 percent elemental iodine and 3 percent organic iodide

• no reduction or mitigation of noble gas radionuclides released from the primary system As described in Section 15.6.3, a single failure of the main steam isolation valve (MSIV) for the faulted SG delays isolating the steamline, resulting in a larger release.

A loss of normal AC power causes the steamline to isolate earlier, limiting the release. Therefore, a loss of normal AC power is not assumed to occur for the portion of the SGTF analysis used to determine the radiological releases. However, a loss of normal AC power is assumed to occur for the thermal-hydraulic portion of the SGTF analysis used to determine peak pressures presented in Section 15.6.3.

Doses are determined at the EAB, LPZ, and for personnel in the control room and TSC. The control room model is described in Section 15.0.3.7.1. The dose results for the SGTF event are presented in Table 15.0-12.

15.0.3.8.3 Main Steam Line Break Outside Containment Accident Radiological consequences of the MSLB outside containment accident are calculated based on the guidance provided in Appendix E of RG 1.183. Section 15.1.5 describes the sequence of events and thermal-hydraulic response to a MSLB outside containment.

The radiological dose consequence analysis considers the MSLB event with two different initial iodine concentrations, one based on a pre-incident iodine spike and the other based on a coincident iodine spike. A description of the scenario evaluated is summarized as follows

1) An MSLB occurs in one of the two main steam lines.

RAI 06.04-4s1

2) The iodine and noble gas coolant activity is calculated based on the maximum concentrations allowed by design basis limits for each of the iodine spiking scenarios. The primary coolant contains a concentration of 0.23.7E-02 Ci/gm DE I-131 for the coincident Iodine spike scenario and 122.2 Ci/gm DE I-131 for the pre-incident Iodine spike scenario. For both iodine spiking scenarios, the primary coolant contains 6010 Ci/gm DE Xe-133.

RAI 15.00.03-4

3) Primary coolant leaks into the secondary side of the intact SGs at the maximum leak rate of 150 gallons per day allowed by design basis limits. The leakage continues until the primary system pressure is less than the secondary system pressure.

4) A time-dependent release is modeled that effectively releases the activity directly to the environment through the break.

5) The non-faulted steam line continues to release a small quantity of radiation through valve leakage.

Tier 2 15.0-33 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 02.03.04-1, RAI 06.04-4, RAI 06.04-4S1, 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 Dose (rem TEDE)

(rem TEDE)

Failure of Small Lines Carrying Primary Coolant EAB 6.3 0.110.02 Outside Containment LPZ 6.3 0.200.04 CR 5.0 0.440.08 Steam Generator Tube Failure EAB 25.0 0.430.08 (pre-incident iodine spike) LPZ 25.0 0.440.08 CR 5.0 1.110.20 Steam Generator Tube Failure EAB 2.5 <0.01 (coincident iodine spike) LPZ 2.5 <0.01 CR 5.0 <0.01 Main Steam Line Break EAB 25.0 <0.01 (pre-incident iodine spike) LPZ 25.0 0.03<0.01 CR 5.0 0.060.01 Main Steam Line Break EAB 2.5 <0.01 (coincident iodine spike) 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 EAB 25.0 0.220.63 (significant core damage) LPZ 25.0 0.991.37 CR 5.0 1.432.14 Tier 2 15.0-68 Draft Revision 2

NuScale Final Safety Analysis Report Transient and Accident Analyses RAI 06.04-4S1 Table 15.0-14: Primary Coolant Source Term Nuclide Primary Activity (Ci)

I-131 7.181E-011.322E+00 I-132 2.854E-016.031E-01 I-133 1.021E+001.992E+00 I-134 1.558E-013.545E-01 I-135 5.997E-011.254E+00 Kr-85m 4.098E-019.660E-01 Kr-85 1.278E+022.865E+02 Kr-87 2.238E-015.275E-01 Kr-88 6.518E-011.536E+00 Xe-133 1.078E+022.536E+02 Xe-135 3.703E+008.569E+00 Tier 2 15.0-70 Draft Revision 2