ML19112A220

Submittal of Changes to Final Safety Analysis Report, Chapter 1, Introduction and General Description of the Plant, Chapter 3, Design of Structures, Systems, Components and Equipment, Chapter 12, Radiation Protection, & Chapter 15, ...

ML19112A220
Person / Time
Site:	NuScale
Issue date:	04/19/2019
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
LO-0419-65280
Download: ML19112A220 (58)

Text

LO-0419-65280 April 19, 2019 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 Submittal of Changes to Final Safety Analysis Report, Chapters 1, Introduction and General Description of the Plant, Chapter 3, Design of Structures, Systems, Components and Equipment, Chapter 12, Radiation Protection, and Chapter 15, Transient and Accident Analyses

REFERENCES:

1. NuScale Topical Report, "Accident Source Term Methodology," TR-0915-17565, Revision 3, dated April 2019

2. Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, NuScale Power, LLC Submittal of Presentation Materials Entitled NuScale Source Term Revision, Revision 0, PM-0118-58201, dated January 23, 2018 (ML18019A163)

3. U.S. Nuclear Regulatory Commission, Category 1 Public Meeting announcement, NuScale Power Design Certification Application Teleconference, dated April 9, 2019 (ML19098B624)

4. Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application, Revision 2, dated October 30, 2018 (ML18311A006)

NuScale Power, LLC (NuScale) has modified the approach to the development of radiological accident source terms as presented in the Accident Source Term Methodology Topical Report (Reference 1).

NuScale has discussed its intention to modify the approach with the NRC staff in a series of public meetings ranging from January 23, 2018 (Reference 2) to April 9, 2019 (Reference 3). As a result of these discussions, NuScale has updated the subject chapters of the Final Safety Analysis Report (FSAR). The Enclosure to this letter provides a mark-up of the FSAR pages incorporating revisions in redline/strikeout format. NuScale will include this change as part of a future revision to the NuScale Design Certification Application.

NuScale plans to incorporate FSAR changes associated with the equipment survivability evaluation called for in Reference 1 and the hydrogen and oxygen monitoring questions raised during the April 9, 2019 public teleconference in a forthcoming FSAR markup submittal scheduled for May 23, 2019.

This letter makes no regulatory commitments or revisions to any existing regulatory commitments.

If you have any questions, please feel free to contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com.

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0419-65280 Page 2 of 2 04/19/19 Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Getachew Tesfaye, NRC, OWFN-8H12

Enclosure:

Changes to Final Safety Analysis Report, Chapters 1, Introduction and General Description of the Plant, Chapter 3, Design of Structures, Systems, Components and Equipment, Chapter 12, Radiation Protection and Chapter 15, Transient and Accident Analyses NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0419-65280

Enclosure:

Changes to Final Safety Analysis Report, Chapters 1, Introduction and General Description of the Plant, Chapter 3, Design of Structures, Systems, Components and Equipment, Chapter 12, Radiation Protection and Chapter 15, Transient and Accident Analyses NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Final Safety Analysis Report Introduction Table 1.1-1: Acronyms and Abbreviations (Continued)

Acronym or Description Abbreviation BDBEE beyond design basis external event BDG backup diesel generator BOC beginning of cycle BOL beginning of life BOP balance-of-plant BPDS balance-of-plant drain system BPE bioprocessing equipment BPSS backup power supply system BPVC Boiler and Pressure Vessel Code BRL Ballistic Research Laboratory BRVS battery room ventilation system BTP Branch Technical Position BWR boiling water reactor CAM continuous air monitor CARS condenser air removal system CAS central alarm station CAS compressed air system CCBE common cause basic event CCDF conditional core damage frequency CCDP conditional core damage probability CCF common cause failure CCFL counter current flow limitation CCFP conditional containment failure probability CDE core damage event CDF core damage frequency CDI conceptual design information CDM certified design material CDST core damage source term CEA control element assembly CES containment evacuation system CET containment event tree CEUS central and eastern United States CFD computational fluid dynamics CFDS containment flooding and drain system CFR Code of Federal Regulations CFT containment flange tool CHF critical heat flux CHFR critical heat flux ratio CFWS condensate and feedwater system CHRS containment heat removal system CHWS chilled water system CILRT containment integrated leak rate test CIM civil interface macro CIP clean-in-place CIS containment isolation system CIV containment isolation valve CLRF conditional large release frequency CLRT containment leakage rate testing CMAA Crane Manufacturers Association of America CMS code management software Tier 2 1.1-5 Draft Revision 3

NuScale Final Safety Analysis Report Introduction Table 1.1-1: Acronyms and Abbreviations (Continued)

Acronym or Description Abbreviation DIM display interface module DMA dimethylamine DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio DOE Department of Energy DOT Department of Transportation D-RAP Design Reliability Assurance Program DSRS Design Specific Review Standard DSS digital safety system DSW dry solid waste DTC Doppler temperature coefficient, fuel temperature coefficient, Doppler coefficient DWS demineralized water system EAB exclusion area boundary EAL Emergency Action Level ECCS emergency core cooling system ECL effluent concentration limit EDL equivalent dead load EDMG extensive damage mitigation guidelines EDNS normal DC power system EDSS highly reliable DC power system EDSS-C EDSS-common EDSS-MS EDSS-module-specific EDV engineering design verification EFDS equipment and floor drainage system EFPD effective full-power days EFPY effective full-power years EHVS 13.8 kV and switchyard system EIM equipment interface module ELAP extended loss of AC power ELVS low voltage AC electrical distribution system ELWR evolutionary light water reactor EMC electromagnetic compatibility EMDAP evaluation model development and assessment process EMDM electromagnetic drive mechanism EMI electromagnetic interference EMVS medium voltage AC electrical distribution system EOC end of cycle EOF emergency operations facility EOL end of life EOP emergency operating procedure EPA electrical penetration assembly EPA Environmental Protection Agency EPG emergency procedure guidelines EPRI Electric Power Research Institute EPZ emergency planning zone EQ equipment qualification EQDP equipment qualification data package EQRF equipment qualification record file ERDA Energy Research and Development Administration ERDS emergency response data system Tier 2 1.1-7 Draft Revision 3

NuScale Final Safety Analysis Report Introduction Table 1.1-1: Acronyms and Abbreviations (Continued)

Acronym or Description Abbreviation PDC principal design criteria PDIL power dependent insertion limit PDIT differential pressure indicating transmitter PFT process feed tank PGA peak ground acceleration pHT concentration of H+ ion on a logarithmic scale (temperature dependent)

PID proportional integral derivative PING particulate, iodine, and noble gas PIRT phenomena identification and ranking table PIT pressure indicating transmitter PLC programmable logic controller PLD pool leakage detection PLDD programmable logic design description PLDP Programmable Logic Development Plan PLDS pool leakage detection system PLHGR peak linear heat generation rate PLM priority logic module PLRS programmable logic requirement specification PLS plant lighting system PLVVP Programmable Logic Verification and Validation Plan PMF probable maximum flood PMP probable maximum precipitation PORV power-operated relief valve POS plant operating state POV power-operated valve PPE personnel protective equipment PPS plant protection system PRA probabilistic risk assessment PRV pressure relief valve PSCIV primary system containment isolation valves PSCS pool surge control system PSD power spectra density PSMS power supply monitoring system PSS process sampling system PST phase separator tank PSTN public switched telephone network PTAC performance and test acceptance criteria band PTS pressurized thermal shock PVC polyvinyl chloride PVMS plant-wide video monitoring system PWHT post-weld heat treatment PWR pressurized water reactor PWS potable water system PWSCC primary water stress-corrosion cracking PZR pressurizer QA quality assurance QAP Quality Assurance Program QAPD Quality Assurance Program Description QPD quadrant power difference QD quick disconnect Tier 2 1.1-13 Draft Revision 3

Table 1.9-2: Conformance with Regulatory Guides (Continued)

Tier 2 NuScale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1.97 Criteria for Accident Monitoring 4 Partially Conforms The NuScale design satisfies power supply require- 3.11 Instrumentation for Nuclear ments in Section 6.6 of IEEE Std 497-2002 for Type B Appendix 3C Power Plants and C variables with highly reliable power rather 5.4 than with Class 1E. The portions of RG 1.97 dealing 7.1 with 10 CFR 50.34(f)(2)(xix) are addressed in 7.2 Section 19.2.3.3.8.

8.3 11.5 12.3 14.3 19.2 1.98 Assumptions Used for Evaluating 0 Not Applicable This RG is applicable only to BWR designs. Not Applicable the Potential Radiological Conse-quences of a Radioactive Offgas System Failure in a Boiling Water Reactor 1.99 Radiation Embrittlement of Reac- 2 Conforms None. 5.3 1.9-19 tor Vessel Materials 1.100 Seismic Qualification of Electrical 3 Partially Conforms This RG is applicable except for aspects related to: (1) 3.9 and Active Mechanical Equip- when site-specific spectra exceed the certified 3.10 ment and Functional Qualifica- design spectra (e.g., Position C1.2.1.g); and (2) quali-5.2 tion of Active Mechanical fication of new and replacement equipment in older Equipment for Nuclear Power unresolved safety issue A46 plants (e.g., Position 14.3 Plants C.1.2.2.j). Not applicable to electrical equipment. Appendix 3C Site-specific guidance is the responsibility of the COL applicant. RG 1.100 endorses Conformance with Regulatory Criteria ASME QME-1 2007. NuScale complies with the non-mandatory Appendix QR-B with the following exceptions:

QR-B5200, Identification and Specification of Qualifi-cation Requirements, (g) material activation energy.

QR-B5300 Selection of Qualification Methods for determination and recording of shelf life of Draft Revision 3 nonmetallics.

QR-B5500 Documentation, (h) shelf life preservation requirements.

Appendix 3C describes the exceptions cited.

Table 1.9-2: Conformance with Regulatory Guides (Continued)

Tier 2 NuScale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1.180 Guidelines for Evaluating Electro- 1 Partially Conforms Aspects of this guidance related to the design of SSC 3.11 magnetic and Radio-Frequency to address effects of electromagnetic and radio-fre- 7.2 Interference in Safety-Related quency interference (EMI/RFI) are applicable.

9.5 Instrumentation and Control Sys- Aspects of this guidance related to the design of tems site-specific SSC and installation and testing prac-tices for addressing the effects of EMI/RFI and power surges on safety-related I&C systems are the respon-sibility of the COL applicant or licensee.

1.181 Content of the Updated Final - Not Applicable This guidance governs site-specific reporting activi- Not Applicable Safety Analysis Report in Accor- ties that are the responsibility of the COL applicant dance with 10 CFR 50.71(e) or licensee.

1.183 Alternative Radiological Source 0 Partially Conforms For the NuScale design, the safety analysis shows 6.4 Terms for Evaluating Design Basis that core damage does not occur during any design 9.3 Accidents at Nuclear Power Reac- basis event. Thus, the RG 1.183 guidance is partially 12.2 tors applicable to the NuScale dose consequence analy-sis. The basis and justification for departures from 15.0.2 the RG 1.183 guidance for the limiting dose conse- 15.0.3 1.9-29 quence analysis for NuScale are provided in a Topi- 15.6 cal Report. NuScale uses the alternative source term non-LOCA or transient-specific guidance of RG 1.183 15.7 for Chapter 15 events. 15.10 1.184 Decommissioning of Nuclear 1 Not Applicable This RG governs site-specific decommissioning plan- Not Applicable Power Reactors ning and implementation activities that are the responsibility of the COL applicant or licensee.

1.185 Standard Format and Content for 1 Not Applicable This RG governs site-specific decommissioning plan- Not Applicable Post-Shutdown Decommission- ning activities that are the responsibility of the COL Conformance with Regulatory Criteria ing Activities Report applicant or licensee.

1.186 Guidance and Examples for Iden- - Not Applicable This RG endorses NEI 97-04 Appendix B is the Not Applicable tifying 10 CFR 50.2 Design Bases responsibility of the COL applicant or licensee.

1.187 Guidance for Implementation of - Not Applicable This RG implements change process requirements Not Applicable 10 CFR 50.59, Changes, Tests, and that are the responsibility of the COL applicant or Experiments licensee.

1.188 Standard Format and Content for 1 Not Applicable This RG is applicable to operating reactor licensees Not Applicable Draft Revision 3 Applications to Renew Nuclear seeking to renew their operating licenses.

Power Plant Operating Licenses

Table 1.9-5: Conformance with TMI Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued)

Tier 2 NuScale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 50.34(f)(2)(v) Provide for automatic indication of the bypassed Conforms None. 7.1 and operable status of safety systems (I.D.3) 7.2.4 7.2.13 50.34(f)(2)(vi) Provide the capability of high point venting of Departure The venting of noncondensible gases is unnecessary to 5.4.4 noncondensible gases from the reactor coolant ensure long term core cooling capability.

system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity. (II.B.1) 50.34(f)(2)(vii) Perform radiation and shielding design reviews of Conforms None.The NuScale design supports an exemption from 12.2 spaces around systems that may, as a result of an the portions of 10 CFR 50.34(f)(2)(viii) requiring 12.3.112.4 accident, contain accident source term capability for obtaining and analyzing post-accident 12.419.2 1.9-207 radioactive materials, and design as necessary to samples of the reactor coolant system and permit adequate access (II.B.2) containment without exceeding prescribed radiation dose limits. Therefore, the NuScale design does not contain vital areas, as defined by NUREG-0737, Item II.B.2, other than the main control room and technical support center. Protection of necessary equipment from radiation is reasonably assured through demonstrating equipment survivability.

Conformance with Regulatory Criteria Draft Revision 3

Table 1.9-5: Conformance with TMI Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued)

Tier 2 NuScale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 50.34(f)(2)(xix) Provide instrumentation adequate for Conforms None. 7.1.1 monitoring plant conditions following an 7.1.2 accident that includes core damage (II.F.3) 7.2.13 19.2 50.34(f)(2)(xx) Provide power supplies for pressurizer relief Departure The requirements of 10 CFR 50.34(f)(2)(xx) for power 5.4.5 valves, block valves, and level indicators (II.G.1) supplies for pressurizer relief valves and block valves 7.2.13 are not technically relevant to the NuScale design. The 8.1.4 NuScale design supports an exemption from the 8.3.1 portions of 10 CFR 50.34(f)(2)(xx) related to pressurizer 8.3.2 level indicators.

50.34(f)(2)(xxi) Design auxiliary heat removal systems such that Not Applicable This requirement applies only to BWR designs. Not Applicable necessary automatic and manual actions can be taken to ensure proper functioning when the main feedwater system is not operable (II.K.1.22) 50.34(f)(2)(xxii) Perform a failure modes and effects analysis of Not Applicable This requirement explicitly states its applicability only Not Applicable the integrated control system (ICS) to include to B&W plant designs. This applicability reflects aspects 1.9-212 consideration of failures and effects of input and of the B&W ICS design that were identified following output signals to the ICS (II.K.2.9) the TMI incident as design/reliability deficiencies, and are not pertinent to the NuScale design.

50.34(f)(2)(xxiii) Provide, as part of the reactor protection system, Not Applicable This requirement applies only to B&W plant designs. Not Applicable an anticipatory reactor trip that would be actuated on loss of main feedwater and on turbine trip (II.K.2.10) 50.34(f)(2)(xxiv) Provide the capability to record reactor vessel Not Applicable This requirement applies only to BWR designs. Not Applicable water level in one location on recorders that meet normal post-accident recording requirements Conformance with Regulatory Criteria (II.K.3.23) 50.34(f)(2)(xxv) Provide an onsite Technical Support Center and Partially Conforms None. 13.3 onsite Operational Support Center (III.A.1.2) 50.34(f)(2)(xxvi) Provide for leakage control and detection in the Partially Conforms This requirement is applicable to the DCA to the extent 5.4 design of systems outside containment that it is relevant to the standard plant design. Aspects of 6.3.1 contain (or might contain) accident source term this requirement that are pertinent to testing and 9.3.2 radioactive materials (III.D.1.1) operational programs are the responsibility of the COL 9.3.4 Draft Revision 3 applicant.

50.34(f)(2)(xxvii) Provide for monitoring of in-plant radiation and Conforms None. 11.5 airborne radioactivity (III.D.3.3) 11.6 12.3.4

Tier 2 NuScale Final Safety Analysis Report RAI 02.03.01-4, RAI 19-36 Table 1.9-8: Conformance with SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs" Issue Description Conformance Comments Section Status I.A Use of a Physically-Based Source Term: Incorporation of engineering judgment Conforms None. 15.0.3 and a more realistic source term in design that deviates from the siting 15.10 requirements in 10 CFR 100.

I.B Anticipated Transient without SCRAM (ATWS): Position on the current practices Partially Conforms The NuScale design relies on diversity 15.8 and design features to achieve a high degree of protection against an ATWS. within the module protection system (MPS) to reduce the risk associated with ATWS events.

I.C Mid-Loop Operation: Position on design features necessary to ensure a high Not Applicable Design does not use external loops Not Applicable degree of reliability of RHR systems in PWR. and no drain down condition for refueling.

I.D Station Blackout (SBO): Position on methods to mitigate the effects of a loss of all Not Applicable The relevance of the SECY-90-016 Not Applicable AC power. SBO issue to passive ALWR designs was deferred to and addressed in 1.9-219 Section F of SECY-94-084 and SECY-95-132. The NuScale design conforms to the passive plant guidance these documents.

I.E Fire Protection: Positions on design configuration and features the fire protection Conforms None. Appendix 9A system and other management schemes to ensure safe shutdown of the reactor.

I.F Intersystem LOCA: Position on acceptable design practices and preventative Conforms None. 9.3.4 measures to minimize the probability of an ISLOCA. 19.2.2 I.G Hydrogen Control: Position on acceptable requirements to measure and mitigate Partially Conforms 6.2.5 the effects of hydrogen produced due to a water reaction with zirconium fuel Conformance with Regulatory Criteria cladding.

I.H Core Debris Coolability: Acceptability criteria for cooling area and quenching Conforms None. 19.2 ability regarding corium interaction with concrete.

I.I High-Pressure Core Melt Ejection: Position on acceptable design features to Conforms None. 19.2.3 prevent the event of a high-pressure core melt ejection.

I.J Containment Performance: Position on acceptable conditional containment Conforms None. 19.1 Draft Revision 3 failure probabilities or other analyses to ensure a high degree of protection from 19.2 the containment.

I.K Dedicated Containment Vent Penetration: Position for a dedicated vent Conforms None. 19.2.4 penetration to preclude containment failure resulting from a containment over-pressurization event.

Methodology for Environmental Qualification of Electrical and NuScale Final Safety Analysis Report Mechanical Equipment Design-Basis Accidents (DBAs)

The design basis accidents were reviewed and evaluated to determine which DBAs are addressed in FSAR Chapter 15. Based on this review, the following DBAs are evaluated to determine the mechanical and electrical equipment that requires environmental qualification.

FSAR Section 15.1.5 - steam system piping failure inside and outside of containment.

This covers main steam line breaks (MSLB) inside and outside of containment. For the purpose of environmental qualification, main steam line breaks are considered inside the CNV even though the main steam piping is classified as leak before break (LBB).

FSAR Section 15.2.8 - feedwater system pipe break inside and outside of containment.

This covers feedwater line breaks (FWLB) inside and outside of containment. For the purpose of environmental qualification, feedwater line breaks are considered inside the CNV even though the FW piping is classified as leak before break (LBB).

FSAR Section 15.4.8 - rod ejection accident (REA) reflects a potential break in the RCS pressure boundary. The equipment relied upon to mitigate this accident is the same as that used for the spectrum of small break loss of coolant accidents addressed by FSAR Section 15.6.5. The REA is analyzed as a reactivity event.

FSAR Section 15.6.5 - loss of coolant accidents (LOCA) from spectrum of postulated pipe breaks within the RCS pressure boundary inside and outside of containment.

There are no large break LOCA events for the NuScale design. The small break LOCAs are the result of CVCS pipe rupture events that are postulated inside or outside of containment. The iodine spike design basis source term described in FSAR Section 15.0.3 is used in the EQ program as a bounding surrogate for the radiological consequences of DBEs that result in primary coolant entering the containment.

Note: The core damage event described in FSAR Section 15.10 is a special event that is outside of the scope of the EQ program.

FSAR Section 15.7.4 - radiological consequences of fuel handling accidents. This covers the FHAs within the RXB pool area.

Infrequent Events (IE)

FSAR Section 15.6.2 - radiological consequences of failure of small lines carrying primary coolant outside of containment. Similar to FSAR Section 15.6.5, this covers chemical and volume control systems (CVCS) pipe rupture events that are postulated inside or outside of containment.

Other Design Basis Events FSAR Section 3.6 - high energy line breaks (HELB) outside containment. This covers HELB outside of containment that are not already addressed by FSAR Sections 15.1.5, 15.2.8, or 15.6.5, such as the postulated rupture of the module heatup system (MHS) piping in the gallery areas of the RXB.

Tier 2 3C-12 Draft Revision 3

Methodology for Environmental Qualification of Electrical and NuScale Final Safety Analysis Report Mechanical Equipment FSAR Section 3.6 - moderate energy line breaks (MELB) outside containment.

Normal and Bounding Conditions Containment vessel and reactor building pressure and humidity experienced during the indicated DBE are shown in Table 3C-7. Equipment that is required to perform a design function related to safety, and could potentially be subjected to the design basis environments, is qualified to these conditions for the required operating time.

RPV and containment vessel metal temperatures in the lower (liquid) space with corresponding liquid temperatures for the bounding DBAs are shown on Figure 3C-1.

RPV and containment vessel metal temperatures in the upper (vapor) space with corresponding vapor temperatures for the bounding DBAs are shown on Figure 3C-2.

The average vapor temperatures at the top of module for the bounding DBAs, and assuming a vented bioshield, are shown on Figure 3C-3. The maximum vapor temperatures for elevation 145' in the RXB from the same bounding DBAs are shown on Figure 3C-4.

3C.5.5 Design Basis Event Radiation Doses RAI 03.11-1, RAI 03.11-4 NuScale Topical Report, TR-0915-17565-P (Reference 3C-5) provides the methodology for determining the accident source terms for equipment following design basis events. The limiting event and associated source terms from the design basis accidents discussed above were used to determine total integrated doses for equipment qualification.

The accident conditions integrated doses within the reactor building were determined using the maximum normal core radionuclide inventory. The maximum normal core inventory bounds the equilibrium cycle burnup for the NuScale Power Module reactor and is representative of operating cycle characteristics for environmental qualification purposes. The required dose used for environmental qualification considers the total integrated dose consisting of the normal dose plus the accident dose corresponding to the required post-accident operating time. The normal dose considers gamma and neutron effects, while the accident dose considers the gamma and beta dose that is expected at the equipment location.

Based on the above, the integrated doses following a design basis event are shown in Table 3C-8.

For discussion on gamma and beta radiation effects, refer to Section 3.11.5.

3C.6 Qualification Methods A qualification program plan defines tests, inspections, performance evaluation, acceptance criteria, and required analysis to demonstrate that, when called upon, the qualified equipment can perform its specified design function(s) related to safety for the required post-accident operating time with margin to failure.

Tier 2 3C-13 Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 03.11-1, RAI 03.11-2S1, RAI 03.11-4 Table 3C-8: Limiting Design Basis Accident EQ Radiation Dose Accident Integrated Dose (rads)

Zone Dose 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> Integrated 6.40E02 8.89E03 1.23E04 2.59E04 2.82E04 A

Integrated 2.09E03 2.10E04 2.78E04 6.55E04 8.84E04 Integrated 6.40E02 8.89E03 1.23E04 2.59E04 2.82E04 B

Integrated 2.09E03 2.10E04 2.78E04 6.55E04 8.84E04 Integrated 2.91E05 4.38E06 6.38E06 2.00E07 3.94E07 C

Integrated 8.96E05 9.07E06 1.20E07 2.85E07 3.84E07 Integrated 2.91E05 4.38E06 6.38E06 2.00E07 3.94E07 D

Integrated 8.96E05 9.07E06 1.20E07 2.85E07 3.84E07 Integrated 2.91E05 4.38E06 6.38E06 2.00E07 3.94E07 E

Integrated 8.96E05 9.07E06 1.20E07 2.85E07 3.84E07 Integrated 2.91E05 4.38E06 6.38E06 2.00E07 3.94E07 F

Integrated 8.96E05 9.07E06 1.20E07 2.85E07 3.84E07 Integrated 7.58E01 3.13E03 6.28E03 9.58E04 6.33E05 G

Integrated 3C-29 7.27E03 7.71E04 1.05E05 3.34E05 6.66E05 Integrated 5.50E01 1.69E03 2.99E03 1.56E04 2.48E04 Methodology for Environmental Qualification of Electrical and H

Integrated 7.65E01 2.32E03 4.08E03 2.22E04 3.95E04 Integrated 6.40E00 1.60E02 2.69E02 1.80E03 4.75E03 I

Integrated 1.94E01 5.78E02 1.02E03 7.89E03 2.25E04 J Integrated - - - - -

Integrated 6.17E02 1.24E04 1.71E04 3.95E04 5.39E04 K Integrated - - - - -

Integrated 1.56E-02 3.00E-01 3.73E-01 4.74E-01 4.76E-01 L Integrated - - - - -

Integrated 1.56E-02 3.00E-01 3.73E-01 4.74E-01 4.76E-01 M Integrated - - - - -

Integrated 3.60E01 6.81E02 8.94E02 1.85E03 2.63E03 Mechanical Equipment N Integrated - - - - -

Integrated 2.78E-04 5.70E-03 7.00E-03 7.00E-03 7.00E-03 Draft Revision 3

NuScale Final Safety Analysis Report Radiation Sources The control rod assemblies are irradiated during reactor operations. Because the reactor core operates in an all-rods-out configuration, it is assumed that only the tip of the control rod is irradiated. This portion of the control rod assembly (CRA) consists of Ag-In-Cd neutron absorber. The major input assumptions are listed in Table 12.2-26.

The CRA gamma spectra are listed in Table 12.2-27.

Secondary Source Rod The secondary source is antimony and beryllium (Sb-Be) and is irradiated for nine cycles. Flux is the same as for the in-core instruments (Section 12.2.1.10).

The gamma ray source strengths associated with the secondary source rods are listed in Table 12.2-29 for various times after shutdown.

12.2.1.12 Secondary Coolant System The secondary coolant system is expected to contain minimal radioactivity during normal operations. Primary-to-secondary leaks through the steam generator can introduce primary coolant activity into the secondary system with the resultant contamination level being dependent upon the activity level in the primary coolant and the magnitude of the steam generator leak. Because the condensate polishing system is a full flow system, the condensate polishers were evaluated for the radioactive material that could accumulate on the resins during the period between resin regenerations. Assuming the secondary coolant is at the design basis concentrations (Table 11.1-5), resin decontamination factors consistent with NUREG-0017, and a ten day resin regeneration period, the accumulation of radioactive material is less than 100 mCi.

12.2.1.13 Post-Accident Sources RAI 12.02-3, RAI 12.02-11, RAI 12.03-31, RAI 12.03-32, RAI 12.03-33, RAI 12.03-34, RAI 12.03-35, RAI 12.03-36, RAI 12.03-37, RAI 12.03-39, RAI 12.03-40 Consistent with 10 CFR 50.34(f)(2)(vii), areas that could contain post-accident sources were evaluated for equipment protection and access. The iodine spike design basis source term (the maximum primary coolant activity released from design basis accidents described in Section 15.0.3) is evaluated for equipment qualification (EQ) in and around an NPM. Three volumes associated with the NPM are evaluated for EQ dose consequences: the reactor pressure vessel and containment vessel combined liquid sump volume, the containment vapor volume, and the bioshield envelope volume. The iodine spike design basis source term maximum post-accident activity concentrations used for equipment qualification evaluation are provided on a mass basis in Table 12.2-37. Plateout of activity onto containment surfaces is neglected due to the small containment volume and the lack of surface coatings inside containment. There is also no aerosol removal assumed. Other assumptions for the post-accident EQ source term are listed in Table 12.2-31. The three volumes are evaluated with conservative assumptions, including instantaneous and homogeneous releases into the volume of interest.

Tier 2 12.2-7 Draft Revision 3

NuScale Final Safety Analysis Report Radiation Sources The maximum primary coolant activity released from design basis accidents, that could be released into the bioshield envelope, is assumed to be released instantaneously and homogeneously throughout the bioshield envelope volume. The remaining primary coolant inventory is assumed to be released instantaneously and homogeneously throughout the containment atmosphere, where it shines and leaks into the bioshield envelope at the technical specification leak rate and remains in the envelope's volume for the duration of the event. Table 12.2-37 provides the maximum post-accident activity concentrations in the NPM on a mass basis. These concentrations apply to both the liquid and vapor spaces. Other major assumptions for the post-accident source term are listed in Table 12.2-31. Plateout of activity onto containment surfaces is neglected due to the small containment volume and the lack of surface coatings inside containment. There is also no aerosol removal assumed. The containment air and water volumes are determined based on the reactor vessel being initially full of water and the reactor vessel and containment vessel water levels being in equilibrium. There are three volumes that are evaluated, which includes the post-accident source term.

Table 12.2-34 lists the integrated post-accident source energy deposition versus time for both photons and electrons for these three evaluated volumes. Table 12.2-34 also tabulates the integrated doses for various times post-accident. For additional details on equipment qualification, see Section 3.11 and Appendix 3.C. Consistent with 10 CFR 50.34(f)(2)(vii), areas that could contain core damage post-accident sources were evaluated for equipment protection. Information on equipment protection from a core damage source term is addressed in Section 19.2.

12.2.1.14 Other Contained Sources There are no other identified contained sources that exceed 100 mCi, including HVAC filters. To evaluate the accumulation of radioactive material on the Reactor Building HVAC system HEPA filters, the airborne radioactivity in the Reactor Building due to pool evaporation and primary coolant leaks was deposited on filters assuming a 99 percent particulate efficiency and two years of operation. For the pool evaporation portion, the Reactor Building HVAC system provides a ventilation flow rate equivalent to one air volume change per hour. For the primary coolant leakage portion, the activity that becomes airborne is captured and filtered by the ventilation system. The resultant accumulation of radioactive material is less than 100 mCi.

COL Item 12.2-1: A COL applicant that references the NuScale Power Plant design certification will describe additional site-specific contained radiation sources that exceed 100 millicuries (including sources for instrumentation and radiography) not identified in Section 12.2.1.

12.2.2 Airborne Radioactive Material Sources This section describes the airborne radioactive material sources that form part of the basis for design of ventilation systems and personnel protective measures, and also are considered in personnel dose assessment.

12.2.2.1 Reactor Building Atmosphere RAI 12.02-12S1, RAI 12.02-20 Tier 2 12.2-8 Draft Revision 3

NuScale Final Safety Analysis Report Radiation Sources RAI 12.02-3, RAI 12.02-11 Table 12.2-31: Post-Accident Equipment Qualification Source Term Input Assumptions Parameter Value Containment release delay 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Containment release duration 1.0E-05 hours Containment leak rate 0.2%/day Containment leak rate after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.1%/day Aerosol fraction of non-noble gases released 100%

Bioshield envelope volume 6475 ft3 Primary coolant water density 43.6 lb/ft3 Air density 0.07 lb/ft3 Containment air volume 3635 ft3 Combined water volume 2500 ft3 Tier 2 12.2-64 Draft Revision 3

NuScale Final Safety Analysis Report Radiation Protection Design Features controlled through a portal located in the Annex Building. Radiological areas are posted with signage in compliance with 10 CFR 20.1901 and 20.1902.

RAI 12.03-21, RAI 12.03-22 High radiation areas either are locked or have alarmed barriers. For areas that are not within lockable enclosures or other barriers, the area will be barricaded and posted, and be provided with a visible warning light. Positive control is exercised over each individual entry when access to the area is required, and egress from the area is not impeded.

COL Item 12.3-1: A COL applicant that references the NuScale Power Plant design certification will develop the administrative controls regarding access to high radiation areas per the guidance of Regulatory Guide 8.38.

RAI 12.03-21, RAI 12.03-22 Very high-radiation areas are locked. Positive control is exercised over each individual entry when access to the area is required, and egress from the area is not impeded. Access to very high-radiation areas complies with guidance in RG 8.38.

The locations of very high-radiation areas are listed on Table 12.3-3.

COL Item 12.3-2: A COL applicant that references the NuScale Power Plant design certification will develop the administrative controls regarding access to very high radiation areas per the guidance of Regulatory Guide 8.38.

COL Item 12.3-3: A COL applicant that references the NuScale Power Plant design certification will specify personnel exposure monitoring hardware, specify contamination identification and removal hardware, and establish administrative controls and procedures to control access into and exiting the radiologically controlled area.

12.3.1.3.2 Accident Conditions A radiation and shielding design review has been performed of spaces around systems that may contain accident source term materials, consistent with 10 CFR 50.34(f)(2)(vii). Post-accident access is discussed in Section 12.4.1.8 and equipment qualification is addressed in Section 12.2.1.13 and Section 3.11. A radiation and shielding design review has been performed of spaces around systems that may contain core damage source term materials, consistent with 10 CFR 50.34(f)(2)(vii).

The resultant equipment protection from a core damage source term is addressed in Section 19.2. Area radiation monitors are provided to indicate the post-accident radiation levels, to monitor plant areas during the progression of a postulated accident, and provide local indication to plant personnel prior to area entry.

See Section 7.1 for additional information on post-accident monitoring (PAM) instrumentation.

Tier 2 12.3-8 Draft Revision 3

NuScale Final Safety Analysis Report Dose Assessment Occupational doses are estimated for a single NPM refueling outage and for an entire year, assuming six NPM refueling outages. Table 12.4-7 provides dose estimates for the various refueling activities.

12.4.1.7 Overall Plant Doses The estimated annual personnel doses associated with the activities discussed above are summarized in Table 12.4-1.

Occupational personnel dose estimates are calculated assuming a 12-NPM site and 24-month fuel cycle for NPM operation, which equates to six refueling outages per year.

12.4.1.8 Post-Accident Actions RAI 12.03-1, RAI 12.03-31, RAI 12.03-32, RAI 12.03-33, RAI 12.03-34, RAI 12.03-35, RAI 12.03-36, RAI 12.03-37, RAI 12.03-39, RAI 12.03-40 There are no credited post-accident operator actions to mitigate an event outside of the main control room for design basis events, as described in Chapter 15, therefore there are no vital areas in the NuScale design outside of the main control room and technical support center. The operator dose assessments for the main control room and the technical support center are provided in Section 15.0.3. The NuScale design supports an exemption from the portions of 10 CFR 50.34(f)(2)(viii) related to the design criterion for obtaining and analyzing post-accident samples of the reactor coolant system and containment without exceeding prescribed radiation dose limits.

The underlying purpose of this rule is to ensure the capability to assess the presence and extent of core damage during an accident. The NuScale design meets this purpose by other means, namely by using radiation monitors under the bioshield and core exit temperature indications. These means have the advantage of benefiting public health and safety by avoiding unnecessary operator dose, preventing the spread of contamination, and reducing the potential for radiation leaks and spills. These advantages are in conformance with the principles of ALARA, minimizing the spread of contamination, reducing the generation of radioactive waste, and facilitating eventual decommissioning. However, as described in Section 9.3.2, the NuScale design facilitates the option of post-accident sampling and a sampling contingency plan (COL Item 9.3-2). Refer to Part 7, Chapter 16 for further details.As described in Section 9.3.2, the process sampling system may be used as part of a contingency action to obtain post-accident samples, which would potentially expose the operator to post-accident radiation sources. The primary means to detect and monitor fuel damage uses core exit temperature indication and radiation monitors located under the NPM bioshield. The post-accident primary coolant sample is collected via the normal CVCS sample line flow path to the primary system sample panel located in the CVCS gallery.

To perform primary liquid sampling, operators would enter the RXB at the 100 elevation, descend to the 50 elevation using a stairwell, and traverse to the sample panels in the CVCS gallery and the counting room and hot lab on elevation 50' of the RXB. If the background dose rate in the counting room is too high, operators would use the counting room in the Annex Building. These areas are depicted in Figure 12.3-4a through Figure 12.3-4d. These post-accident radiation zone maps represent a composite of maximum dose rates developed by using the highest dose rate in a particular Reactor Building area resulting from design basis accidents occurring on the module with the highest calculated dose rate. Therefore, the radiation zones depicted Tier 2 12.4-4 Draft Revision 3

NuScale Final Safety Analysis Report Dose Assessment in Figure 12.3-4a through Figure 12.3-4d will not occur simultaneously. For post-accident sampling operator activities, the limiting design basis accident is a break of a small line carrying primary coolant, with a coincident iodine spike. This also assumes that the sampling system has not been rendered inoperable due to the accident. For example, a small line break on a CVCS line upstream of the sample line tap causes the sampling system to be non-functional. Consistent with 10 CFR 50.34(f)(2)(vii), post-accident radiological conditions were evaluated and determined that primary coolant sampling activities expose operators to dose rates up to 70 mrem/hr at the sample panel, with much of the collection activities resulting in dose rates less than 13 mrem/

hr. Post-accident doses in the counting room and hot lab were determined to be less than 2.5 mrem/hr. To perform containment gas sampling, operators would perform the necessary functions from the main control room.

RAI 12.03-1, RAI 12.03-31, RAI 12.03-32, RAI 12.03-33, RAI 12.03-34, RAI 12.03-35, RAI 12.03-36, RAI 12.03-37, RAI 12.03-39, RAI 12.03-40 The operator's exposure to airborne activity is considered as part of the dose evaluation, but is not expected to result in significant doses. This is due to the barrier created by the RXB walls around the reactor pool, which include airtight seals for doors and other penetrations, between the reactor pool area and other areas in the RXB. This barrier minimizes the migration of airborne contamination from the airspace above the pool to other areas of the RXB. The RXB HVAC system uses smoke dampers to minimize the leakage between the pool area and other portions of the building. In addition, the generation of airborne contamination is minimized in the CVCS gallery area through the control of the temperature and pressure of the primary coolant during sampling.

Also, if radiological conditions warrant, respiratory protection equipment can be provided to post-accident sampling personnel.

RAI 12.03-1, RAI 12.03-31, RAI 12.03-32, RAI 12.03-33, RAI 12.03-34, RAI 12.03-35, RAI 12.03-36, RAI 12.03-37, RAI 12.03-39, RAI 12.03-40 The decision to perform post-accident sampling will be determined and initiated by the site staff considering the expected radiological conditions and radiological dose to operating personnel when making the decision to accessing these areas to perform post-accident sampling. A summary of the assumed parameters for the post-accident sampling operator dose evaluation are provided in Table 12.4-8. Post-accident sampling performed consistent with approved procedures prevents radiation exposures to individuals from exceeding 5 rem to the whole body or 50 rem to the extremities, consistent with 10 CFR 50.34(f)(2)(viii).

12.4.1.9 Construction Activities For the construction of an additional NuScale Power Plant adjacent to an existing NuScale Power Plant, the estimated annual radiation exposure to a construction worker is estimated based upon a construction staffing plan over the estimated construction period. It is estimated that the annual dose for a construction worker is 1.64 mrem/year.

COL Item 12.4-1: A COL applicant that references the NuScale Power Plant design certification will estimate doses to construction personnel from a co-located existing operating nuclear power plant that is not a NuScale Power Plant.

RAI 02.03.01-2, RAI 02.03.05-1 Tier 2 12.4-5 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses

5) increase in reactor coolant inventory

6) decrease in reactor coolant inventory

7) radioactive release from a subsystem or component RAI 15.00.03-1 Table 15.0-1 lists the events selected for evaluation in Sections 15.1 through 15.7 and a list of computer codes used for analyzing each event. An additional event, the radiological consequences of the design basis source term (DBST) described in Section 15.0.3.8.6, isSpecial events, such as the beyond-design-basis core damage event (CDE), are also included in Table 15.0-1.

15.0.0.2 Design Basis Event Classification NuScale DBEs are classified by frequency of occurrence, including those events that are expected to occur within the NPM lifetime as well as those that are postulated but not expected to occur during the NPM lifetime. The NuScale DBE spectrum is developed by considering DBEs associated with current generation plants and unique events resulting from NuScale Power Plant design features, including review of PRA initiators.

This approach ensures the design considers a broad spectrum of potential events.

Classification by frequency of occurrence is used to assign the analysis acceptance criteria for the event.

The set of DBEs establishes the design adequacy of the NPM and NuScale Power Plant to limit radiological releases below regulatory guidelines.

15.0.0.2.1 Classification by Event Frequency and Type Design basis event classification by frequency is based on fourthree distinct categories:

• anticipated operational occurrences (AOOs)

• infrequent events (IEs)

• postulated accidents

• special events Events that are expected to occur one or more times during an NPM lifetime are classified as AOOs. Events that are not expected to occur during an NPM lifetime are classified as IEs or postulated accidents or may be conservatively classified as AOOs. In general, events that are not considered to be within the design basis are evaluated in Chapter 19; however, those beyond design basis events (BDBEs) that are explicitly defined by regulation are addressed in this chapter. These events are termed special events. For example, the CDE described in Section 15.10 is a special event.

Special events also encompass defense-in-depth and common cause failures (CCFs) of digital control systems, as described in Branch Technical Position 7-19.

Tier 2 15.0-2 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses MSIBV is not credited for event mitigation based on potential failure of the safety-related MSIBVs and for these reasons the nonsafety-related MSIBV is not identified in Table 15.0-9 as nonsafety-related equipment credited for event mitigation.

RAI 15-17S1 Classification information for the secondary MSIVs, MSIBVs, FWRVs, and the nonsafety-related feedwater check valves are listed in Section 3.2, Table 3.2-1. The secondary MSIVs and MSIBVs are described in Section 10.3.2. The FWRVs and nonsafety-related feedwater check valves are described in Section 10.4.7.

The reactor pool liner, described in Section 9.2.5, is a nonsafety-related component of the reactor pool used as the ultimate heat sink (UHS). Section 9.2.5 describes how the pool liner meets the criteria for event mitigation in that water leakage from the liner is detectable and leakage is a nonconsequential random and independent failure. Therefore, any event that progresses to using DHRS, or convection cooling through the containment vessel to the reactor pool with the use of RVVs and RRVs uses the UHS and the pool liner.

15.0.0.7 Multiple Module Events Chapter 15 DBEs are analyzed for a single NPM. Chapter 21 discusses the suitability of shared components and the design measures taken to ensure these components do not introduce multi-module risks. Section 19.1 discusses consideration of multi-module events.

15.0.1 Radiological Consequence Analyses Using Alternative Source Terms A modified version of the alternative source term methodology is used to evaluate radiological consequences of DBEs and the beyond-design-basis event CDE. The source term methodology and the application of that methodology are described in Section 15.0.3.

15.0.2 Review of Transient and Accident Analysis Methods This section summarizes the principal computer codes used in transient analyses and describes the evaluation model development and assessment process (EMDAP). A roadmap with references to NuScale topical and technical reports required to develop those evaluation models is provided in Table 15.0-10.

Several different licensing methodologies are required to provide the neutronic, thermal-hydraulic, and radiological response of the plant to postulated accidents, IEs, and AOOs.

The NuScale Power Plant licensing methodologies include the computer programs and the calculation framework for a specific transient or accident such as the mathematical models used, assumptions included in the programs, and procedures for treating the program input and output information. The licensing methodology also includes required assumptions about the plant equipment availability, combinations with external events, and other information necessary to specify the calculation procedure and to meet the requirements of the GDC of 10 CFR 50, Appendix A.

Tier 2 15.0-12 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses PIM code is described in Section 4.4.7. The PIM code relies on the published description of the theory and numerical methods of RAMONA, but is not a direct derivative of the coding. The PIM code has been developed independently to suit the geometry and specific needs of the NPM. The main advantage of the RAMONA-type algorithm is the absence or insignificance of numerical damping that affects other time-domain codes and requires extensive studies and adjustments before they can be successfully benchmarked and reliably used. Reference 15.0-10 provides details about the process used to select and qualify the PIM code.

15.0.2.3 Subchannel Analysis VIPRE-01 is a subchannel analysis tool designed for general-purpose thermal-hydraulic analysis under normal operating conditions, operational transients, and events of moderate severity.

VIPRE-01 is used to generate local thermal-hydraulic conditions for CHF tests in developing a CHF correlation. VIPRE-01 provides local thermal-hydraulic conditions throughout the reactor core used in calculating the minimum critical heat flux ratio (MCHFR). VIPRE-01 also provides more realistic boundary conditions, such as the axial profiles for the coolant temperature and wall heat transfer coefficient in a limiting subchannel, for the fuel rod performance analyses. VIPRE-01 qualification is applicable to the NuScale implementation. VIPRE-01 is validated against applicable test data that spans the plant range and establishes the code accuracy and uncertainty.

Reference 15.0-1 provides an applicability assessment of the models, correlations, and features in the VIPRE-01 code for the NPM design.

NuScale-specific CHF correlations have been developed to better represent the NuScale core and fuel assembly design. The NuScale-specific CHF correlations are described in Section 4.4.4 and detailed in Reference 15.0-2. NuScale-specific CHF correlations have been added to the existing suite of VIPRE-01 CHF correlations as an enhancement to VIPRE-01.

15.0.2.4 Radiological Analyses Methodology The computer codes used in calculating DBE, and beyond-design-basis event CDE, doses are described below. Reference 15.0-4 provides additional details on each of the computer codes described below.

15.0.2.4.1 SCALE 6.1, TRITON, and ORIGEN-SCALE SCALE 6.1 modular code package, developed by Oak Ridge National Laboratory, is used for development of reactor core and primary coolant fission product source terms. The TRITON and ORIGEN-ARP analysis sequences of the SCALE 6.1 modular code package, and ORIGEN-S (ORIGEN-SCALE), run as a standalone module, are used to generate radiation source terms for the fuel assemblies and primary coolant. This software has been extensively used in the evaluation of operating large LWRs. The operating environment, nuclear fuel, and structural materials in the NuScale Power Plant design are similar to, or bounded by, those typically found in large PWRs.

Tier 2 15.0-18 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses TRITON is used to generate burnup-dependent cross sections for NuScale fuel assemblies for subsequent use in the ORIGEN-ARP depletion module. The TRITON sequence of the SCALE code package is a multipurpose control module for nuclide transport and depletion, including sensitivity and uncertainty analysis. TRITON can generate problem-dependent and exposure dependent cross sections as well as perform multi-group transport calculations in one-dimensional, two-dimensional, or three-dimensional geometries.

ORIGEN-ARP is a SCALE depletion analysis sequence used to perform point-depletion and decay calculations with the ORIGEN-S module using problem-dependent and burnup-dependent cross sections.

The ORIGEN-S module of SCALE is used to calculate the time-dependent isotopic concentrations of materials in a NuScale fuel assembly by modeling the fission, transmutation, and radioactive decay of fuel isotopes, fission products, and activation products in the assembly. The input isotopic concentrations for those calculations take into account the various chemical and physical processes occurring in the reactor systems and the processing of the liquid, solid, and gaseous waste streams. As a part of the ORIGEN-S decay calculations, time-dependent radiation source terms, (i.e., the activities, neutron spectra, and gamma spectra due to the radioactive isotopes) present in the fuel and waste streams are calculated for use in subsequent dose rate evaluations.

15.0.2.4.2 ARCON96 Onsite and offsite atmospheric dispersion factors for DBEs, and the beyond-design-basis event CDE, are calculated with ARCON96. The program implements the guidance provided in RG 1.194. The code implements:

• a building wake dispersion algorithm

• an assessment of ground level, building vent, elevated, and diffuse source release modes

• hour-by-hour meteorological observations

• sector averaging and directional dependence of dispersion conditions NuScale uses ARCON96 for various time periods at the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ) as well as the control room and technical support center (TSC). Justification for utilizing ARCON96 for offsite locations is provided in Reference 15.0-4.

15.0.2.4.3 RADTRAD RADTRAD is used to estimate radionuclide transport and removal of radionuclides and dose at selected receptors for the various DBEs and the beyond-design-basis event CDE. Given the radionuclide inventory, release fractions, and timing, RADTRAD estimates doses at the EAB and LPZ, and inside the control room and TSC. As material is transported from the point of release, the input model can account for processes that may reduce the quantity of radioactive material.

Material can flow between buildings, from buildings to the environment, or into Tier 2 15.0-19 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses the control room and TSC through filters, piping or other connectors. An accounting of the amount of radioactive material retained in these pathways is maintained. Decay and in-growth of daughter products can be calculated over time as material is transported. Reference 15.0-4 describes the use of RADTRAD for NuScale application.

15.0.2.4.4 MELCOR MELCOR is used to model the progression of severe accidents through modeling the major systems of the reactor plant and their coupled interactions (NUREG/

CR-6119, Rev. 2). Specific use relevant to the application of the DBA source termCDE described in Section 15.10 includes:

• thermal-hydraulic response of the primary coolant system and containment vessel

• core uncovering, fuel heatup, cladding oxidation, fuel degradation, and core material melting and relocation

• aerosol generation

• in-vessel and ex-vessel hydrogen production and transport

• fission product release (aerosol and vapor), transport, and deposition 15.0.2.4.5 STARNAUA This code is an aerosol transport and removal software program that is an enhanced version of NAUAHYGROS and was developed to perform aerosol removal calculations in support of work to develop and apply a more realistic source term for advanced and operating LWRs.

STARNAUA models natural removal of containment aerosols by gravitational settling and diffusiophoresis, and considers the effect of hygroscopicity (growth of hygroscopic aerosols due to steam condensation on the aerosol particles) on aerosol removal. STARNAUA enhancements of NAUAHYGROS include addition of:

• a model for thermophoresis

• a model for spray removal

• the capability to directly input either steam condensation rate or condensation heat transfer rate and total heat transfer rate such as would be provided from an external containment thermal-hydraulics code calculation The NuScale realistic source term methodology, used to support the radiological consequence analysis of the CDE described in Section 15.10, is consistent with existing industry practice used for large passive plant design certification. More detail on the application of STARNAUA is provided in Reference 15.0-4.

Tier 2 15.0-20 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses 15.0.2.4.6 NuScale pHT Code The NuScale pHT code is used to calculate post-accident aqueous molar concentration of hydrogen ions utilizing the methodology described in Reference 15.0-4 to support the radiological consequence analysis of the CDE described in Section 15.10. Calculation of the extent of iodine re-evolution inside containment is dependent on the time dependent pHT. This program takes inputs for:

• initial boron and lithium concentrations

• total core inventory of iodine and cesium

• integrated photon dose to the containment and total dose to the coolant

• initial mass of coolant

• mass of coolant

• temperature of coolant The program then calculates the coolant temperature-dependent pHT as a function of time.

15.0.2.4.7 MCNP6 The MCNP6 is used for evaluating potential shine radiological exposures or doses to operators in the control room following a radiological release event. Both sky-shine and shine from filters are evaluated. MCNP is a general-purpose tool used for neutron, photon, electron, or coupled neutron, photon, and electron transport.

MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori.

The code is well-suited to performing fixed source calculations.

MCNP uses continuous energy cross-section data. For photons, the code accounts for incoherent and coherent scattering, the possibility of fluorescent emission after photoelectric absorption, and absorption in electron-positron pair production.

Electron and positron transport processes account for angular deflection through multiple Coulomb scattering, collisional energy loss with optional straggling, and the production of secondary particles including x-rays, knock-on and Auger electrons, bremsstrahlung, and annihilation gamma rays from positron annihilation at rest. The MCNP code is commercially-grade dedicated under the NuScale NQA-1 program described in Reference 15.0-4.

15.0.3 Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors RAI 15.00.03-1 This section presents the methodology used to perform the calculations associated with the radiological consequences of the DBSTs and the core damage source term (CDST) associated with the beyond-design-basis CDE. and other limiting event types.

Tier 2 15.0-21 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses Table 15.0-11 identifies the list of events analyzed for radiological consequences. Results from the application of this methodology are provided in Table 15.0-12.

15.0.3.1 Introduction RAI 15.00.03-1 This section describes the NuScale conservative methodology for developing accident source terms and performing the corresponding radiological consequence analyses.

Key unique features of the NuScale methodology are the:

• use of ARCON96 to calculate off-site atmospheric dispersion factors

• development of an iodine spike DBST to meet the intent of 10 CFR 52.47(a)(2)(iv)and a beyond-design-basis CDST that are each assessed against the radiological criteria of 10 CFR 52.47(a)(2)(iv). If both the design-basis iodine spike DBST and the beyond-design-basis CDST analyses show acceptable dose results, then 10 CFR 52.47(a)(2)(iv) is met.

• use of the STARNAUA containment aerosol transport code in the range of NuScale containment conditions.

RAI 15.00.03-1 10 CFR 52.47(a)(2)(iv) requires nuclear power reactor design certification applicants to evaluate the consequences of a fission product release into the containment assuming the facility is being operated at the maximum licensed power level and to describe those design features intended to mitigate the radiological consequences of an accident. NuScale follows the approach of the 2012 Nuclear Energy Institute (NEI) position paper on small modular reactor source terms (Reference 15.0-6) by referring to the scenario described in footnote 3 of 10 CFR 52.47(a)(2)(iv) as the maximum hypothetical accident (MHA). Source terms are divided into two principal categories titled "Category 1" and "Category 2.

The Category 1 source terms include deterministic accidents and are analyzed using the guidance of RG 1.183. Exceptions to RG 1.183 are due to differences between the NuScale Power Plant design and large LWRs, as outlined in Reference 15.0-4.

The Category 2 source term consists of the MHA scenario in which significant core damage occurs.

RAI 15.00.03-1 The MHA has historically been linked to a large-break LOCA in large LWRs. The NPM has no large diameter primary coolant system piping; therefore, a large-break LOCA cannot be postulated as the basis for the MHA radiological consequence analysis for NuScale.

Surrogate accident scenarios, denoted as source term design basis accidents, are used to identify appropriate severe accident scenarios to address the MHA for the off-site and control room dose evaluations.

As stated in RG 1.183, "the design basis accidents were not intended to be actual event sequences, but rather, were intended to be surrogates to enable deterministic evaluation of the response of a facility's engineered safety features." The NuScale Tier 2 15.0-22 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses design has DBEs that result in primary coolant entering the containment and the iodine spike DBST described in Section 15.0.3.8.6 is used to bound the radiological consequences of these events. The beyond-design-basis CDE described in Section 15.10, with its associated CDST that is composed of a set of key parameters derived from a spectrum of surrogate accident scenarios, is also postulated. The design-basis iodine spike DBST and the beyond-design-basis CDST are each assessed against the radiological criteria of 10 CFR 52.47(a)(2)(iv). If both the design-basis iodine spike DBST and the beyond-design-basis CDST analyses show acceptable dose results, then 10 CFR 52.47(a)(2)(iv) is met. The analysis of the beyond-design-basis CDST against the acceptance criteria of 10 CFR 52.47(a)(2)(iv) provides reasonable assurance that, even in the extremely unlikely event of a severe accident, the facilitys design features and site characteristics provide adequate protection of the public.

RAI 15.00.03-1 The DBST is composed of a set of key parameters, such as fuel release fractions and timing, derived from a spectrum of source term design basis accidents. The radiological consequence analysis of the MHA using the DBST is discussed in Section 15.0.3.8.6.

However, there are no DBEs that result in significant core damage for the NuScale Power Plant.

Table 15.0-11 identifies design basisthe list of events evaluated for radiological consequences, cross references them to RG 1.183, and identifies the primary source of radiation for the event. Table 15.0-12 provides the iodine spike DBST and the CDST dose results. The results meet acceptance criteria and therefore, 10 CFR 52.47(a)(2)(iv) is met.

15.0.3.2 Methodology Overview The methodology used to perform the Category 1 DBA and the Category 2 MHA events considers:

• atmospheric dispersion

• design basis source term

• containment aerosol generation and removal

• post-accident temperature dependent pH RAI 15.00.03-1 The Category 1DBE radiological consequence analyses follow the guidance of RG 1.183 methodology modified to reflect the difference in the NuScale Power Plant design from large PWRs, as described in Reference 15.0-4. This methodology addresses the submersion and inhalation doses and the direct shine doses from contained or external sources. The key elements of this methodology are:

• Severe accident sequences are modeled using MELCOR.

• Thermal-hydraulic conditions are modeled using NRELAP.

• Source term and dose evaluations are calculated using RADTRAD.

• Meteorological dispersion is calculated using ARCON96.

Tier 2 15.0-23 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses

• Fission product removal rates are calculated using approved methodology.

• The evaluation of post-accident pH on the chemical form of iodine is considered.

Section 15.0.2.4 summarizes the computer codes used for calculating MHA and DBE doses.

15.0.3.3 General Dose Analysis Inputs The following sections summarize the key aspects for calculating DBA dosesthe radiological consequences of the DBEs.

15.0.3.3.1 Core Radionuclide Inventory The isotopic inventories of fuel assemblies are calculated using SCALE 6.1 which is described in Section 15.0.2.4. Isotopic concentrations are based on the detailed geometry of a fuel assembly, rated power plus uncertainty, maximum assembly average exposure, and a range of U-235 enrichments. The isotopic inventory is calculated at a number of time steps in the fuel cycle. Table 11.1-1 provides the maximum core isotopic inventory.

RAI 15.00.03-1 15.0.3.3.2 Primary Coolant Radionuclide Inventory For the radiological consequence analysis, the radioiodine concentrations in the primary coolant system are set at the maximum dose equivalent values permitted by design basis limits. Table 15.0-14 provides the primary coolant radionuclides and nominal inventory assumed in the dose analyses presented in Section 15.0.3.8.

The iodine appearance rates, including the pre-incident appearance rates, are described in Section 15.0.3.8, where used.

RAI 15.00.03-1 15.0.3.3.3 Secondary Coolant Activity Large PWR designs contain a large volume of secondary system water on the "shell" side of the SG. Through primary-to-secondary leakage limits and monitoring by sampling, this water volume contains levels of iodine that are limited operationally.

A sensitivity study was performed in Reference 15.0-4 for the steam generator tube failure (SGTF) and main steam line break (MSLB) events assuming the liquid secondary coolant in the SG was at the primary coolant design basis limit concentration. The sensitivity study demonstrated dose results are not sensitive to the initial secondary side activity. This conclusion is supported by comparing the secondary coolant source terms shown in Table 11.1-5 with the primary coolant source terms shown in Table 15.0-14.

RAI 15.00.03-1 Tier 2 15.0-24 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses 15.0.3.3.4 Source Term Release Fraction and Timing for Dose AnalysisNot Used The DBST is derived using the methodology described in Reference 15.0-4. The intact-containment internal events are considered when developing the DBST. A subset of these events are modeled using MELCOR to provide a representative range of release timing and fractions. Five beyond-design basis Level 1 PRA events were selected in which significant core damage was predicted to occur. Each of the five sequences involves failure of the ECCS with either no valves opening, the RVVs failing to open, or the RRVs failing to open. In each case, the DHRS is assumed available to remove heat. The cases are:

1) A LOCA of RCS injection line with ECCS valves failing to open.

2) A LOCA of RCS injection line with RVVs failing to open.

3) A LOCA of RCS injection line with RRVs failing to open.

4) Loss of DC power with the RVVs failing to open.

5) Loss of DC power with the RRVs failing to open.

The DBST on-site and off-site radiological consequences use event-specific radionuclide groups, release timing, release fractions, and aerosol removal, as described in Section 15.0.3.9. The aerosol removal rate as a function of time, and other key inputs such as atmospheric dispersion factors and isotopic inventories, are input into RADTRAD for calculating the radiological consequences.

RAI 15.00.03-1 15.0.3.3.5 Release Timing and MagnitudeNot Used As with radionuclide groups, design-specific representative results for release timing and magnitude from severe accident evaluations are used for the methodology to reflect current practices and appropriately model the specific event. The approach used to identify the release magnitudes and timing are provided in Reference 15.0-4. Release timing and magnitude for the DBST are listed in Tables 12.2-28 and 12.2-29.

RAI 15.00.03-1 15.0.3.3.6 Aerosol and Elemental Iodine RemovalNot Used Natural deposition phenomena including sedimentation, diffusiophoresis, thermophoresis, and hygroscopicity result in aerosol removal. The NuScale aerosol removal methodology uses the aerosol removal code STARNAUA to track these various deposition phenomena in calculating time-dependent airborne aerosol mass and removal rates. Aerosol removal rates used for the DBST are located in Table 12.2-30. A key assumption of the NuScale aerosol transport methodology is that there is no maximum limit on iodine decontamination factor because removal is facilitated by natural processes as opposed to an active spray system. The NuScale removal rate calculation methodology is based on the calculated time-Tier 2 15.0-25 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses dependent airborne aerosol mass in accordance with Appendix A, Section 3.3 of RG 1.183. NuScale conservatively does not take credit for elemental iodine removal.

Rather, only aerosol removal is credited. A summary of the aerosol transport and removal calculation process is described in Reference 15.0-4.

15.0.3.3.7 Aerosol Resuspension and RevaporizationNot Used Treatment of aerosol resuspension and revaporization is discussed n Reference 15.0-4.

15.0.3.3.8 RADTRAD Modeling Consistent with RG 1.183:

• The RADTRAD decay and daughter product modeling option is used to include progeny from the decay of parent radionuclides that are significant with regard to radiological consequences and the released radioactivity. The calculated total effective dose equivalent (TEDE) is the sum of the committed effective dose equivalent from inhalation and the deep dose equivalent from external exposure from tracked isotopes.

• RADTRAD does not include corrections for depletion of the effluent plume by deposition on the ground.

• RADTRAD determines the maximum two-hour TEDE by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments of successive two-hour periods.

RAI 15.00.03-1 15.0.3.3.9 Post-Accident pHT CodeNot Used The DBST methodology calculates the post-accident temperature-dependent pH.

The pHT code is used to calculate the extent of iodine re-evolution inside containment. During the postulated DBST event, additional acids and bases may enter the coolant and cause a change in pH. The expected overall pH of the coolant is modeled over a time period of 30 days. Section 15.0.2.4 provides a discussion of the NuScale pHT program used to calculate post- accident temperature-dependent pH.

Details about the methodologies used for evaluating post-accident temperature-dependent pH in coolant water following a significant core damage event are presented in Reference 15.0-4. The results of implementing the methodology show that the post-accident temperature-dependent pH inside containment is between 6.0 and 7.0.

15.0.3.3.10 Iodine Re-evolutionNot Used The DBST methodology assumes a negligible amount of iodine re-evolution occurs between temperature-dependent pH values of 6.0 and 7.0 and does not need to be explicitly included in the dose analysis calculation. This position simplifies the Tier 2 15.0-26 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses analysis without an impact to the conservatism of the calculated dose results. The treatment of iodine re-evolution is described in Reference 15.0-4.

15.0.3.3.11 Atmospheric Dispersion Factors (/Q), Breathing Rates, and Occupancy Factors RAI 02.03.04-1 Atmospheric dispersion factor (/Q) inputs to RADTRAD are derived as described in Reference 15.0-4 with assumptions shown in Table 15.0-20 and Table 15.0-21.

Table 2.0-1 provides the accident release /Q values.

Control room and offsite breathing rate and control room occupancy factor inputs to RADTRAD, consistent with RG 1.183, are listed in Table 15.0-13.

15.0.3.3.12 Dose Conversion Factors Consistent with RG 1.183, dose conversion factors from Environmental Protection Agency Federal Guidance Report No. 11 (Reference 15.0-8) and Report No. 12 (Reference 15.0-9) are used for dose analysis.

RAI 15.00.03-1 15.0.3.4 Containment Leakage Containment leakage is described in Reference 15.0-4 and is consistent with the recommendations of RG 1.183. The design-basis containment leak rate is provided in Table 6.5-1., is consistent with the recommendations of RG 1.183, and is listed in Table 12.2-28.

15.0.3.5 Secondary-Side Decontamination The helical coil steam generators of the NuScale Power Plant design are different than that of a large PWR because the primary coolant is on the outside of the tubes. As a result, there is no bulk water volume in which decontamination can easily occur.

Reference 15.0-4 provides the details about the decontamination factor used in the helical coil steam generators as well as the treatment of iodine deposition in the main steam piping and the condenser.

15.0.3.6 Reactor Building Decontamination Factors Reactor Building (RXB) decontamination factors are described in Reference 15.0-4.

15.0.3.7 Receptor Location Considerations RAI 15.00.03-2 Potential on-site radiological receptor locations considered in this evaluation are the control room and TSC; potential off-site locations are the EAB and LPZ. Figure 15.0-3 shows the schematic of the RADTRAD code nodalization used to model these locations for leakage paths from the containment or RXB. Figure 15.0-4 shows the RADTRAD Tier 2 15.0-27 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses

• The control room is habitable during a loss of normal AC power as the CRHS automatically activates after 10 minutes without normal AC power, as described in Section 6.4.3.

• Control room ventilation is designed to minimize in-leakage.

RAI 15.00.03-7

• The control room is designed with a two-door air lock system. Therefore, in-leakage of 5 cfm is assumed for ingress and egress. An additional 147-cfm of in-leakage is also assumed.

RAI 15.00.03-2 The control room ventilation system design modeling assumptions are provided in Table 15.0-15. Details about system operation with CRHS are provided in Section 6.4 and Section 9.4.1.

No credit is taken for the use of personal protective equipment, such as beta radiation resistant protective clothing, eye protection, or self-contained breathing apparatus. No credit is taken for prophylactic drugs such as potassium iodide pills.

Potential shine radiological exposures to operators within the control room following a radiological release event are evaluated. Direct shine, sky-shine and shine from filters are evaluated using MCNP, as described in Section 15.0.2.4.7.

Reference 15.0-4 provides additional details regarding the calculation of shine doses. The 30-day cumulative doses due to either recirculation filter or cloud-shine in the control room are added to the dose results from DBEs provided in Table 15.0-12.

Shine doses are well below the regulatory limit of 5 rem because of the heavy shielding provided by the wall and floors of the Control Building.

RAI 15.00.03-7 15.0.3.7.2 Technical Support Center Design Accident analyses are performed for one emergency mode: that of uninterrupted power supply with continuous filtered airflow to the Technical Support Center (TSC) envelope for the event duration. In the event of immediate loss of power with control room habitability system (CRHS) activation, TSC personnel are evacuated and the TSC functions are transferred to an alternate site-specific location. With loss of power with CRHS activation, the TSC is evacuated since it is not serviced by the CRHS.

RAI 15.00.03-7 The key design features assumed for the technical support center are summarized as follows:

RAI 15.00.03-7, RAI 15.00.03-8

• The nonsafety-related normal TSC ventilation system filters remove 99 percent of iodine under accident conditions.

RAI 15.00.03-7, RAI 15.00.03-8 Tier 2 15.0-29 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses

• The nonsafety-related normal TSC ventilation is isolated by a non safety-related control system once the radioactivity measured at the duct intake reaches the isolation signal setpoint. The setpoint for the radiation monitor to redirect air through the air filtration unit is 10ten-times background. The setpoint for CRHS initiation and CREcontrol room envelope isolation is 10ten-times the expected radiation out of the filtration unit following a DBE, which indicates a failure of the filtration unit to remove sufficient radioactivity. The time between when the radiation concentration reaches the detector setpoint and radiation enters the technical support center (TSC) envelope is assumed to be 30 seconds. Ten times the expected post-accident radiation analytical limits for noble gases, particulate and iodine are shown in Table 15.0-19.

RAI 15.00.03-7

• 10-cfm of in-leakage is assumed for ingress and egress. An additional 56 cfm of in-leakage is also assumed.

RAI 15.00.03-7 The technical support center ventilation system design modeling assumptions are provided in Table 15.0-18.

RAI 15.00.03-7 No credit is taken for the use of personal protective equipment, such as beta radiation resistant protective clothing, eye protection, or self-contained breathing apparatus. No credit is taken for prophylactic drugs such as potassium iodide pills.

RAI 15.00.03-7 Potential shine radiological exposures to operators within the TSC following a radiological release event are evaluated. Direct shine, sky-shine and shine from filters are evaluated using MCNP, as described in Section 15.0.2.4.7.

Reference 15.0-4 provides additional details regarding the calculation of shine doses.

RAI 09.02.05-2 15.0.3.7.3 Reactor Building Pool Boiling Radiological Consequences Without available power for the active cooling systems, the addition of makeup water, or operator action, the sensible and decay heat from the NPMs and spent fuel would heat the pool water and could eventually cause the water in the UHS pools to boil. Table 9.2.5-2 shows that it takes longer than 61 hours7.060185e-4 days <br />0.0169 hours <br />1.008598e-4 weeks <br />2.32105e-5 months <br /> for the pool to reach boiling after a loss of normal AC power event. However, if the pool were to boil, the dose would be less than 0.5 rem TEDE onsite and offsite.

15.0.3.8 Consequence Analyses of Category 1 EventsDesign-Basis Source Terms 15.0.3.8.1 Failure of Small Lines Carrying Primary Coolant Outside Containment Failure of small lines carrying primary coolant outside containment is not an event addressed in RG 1.183. The methodology used for determining dose consequences, including the iodine spiking assumptions for this event, is similar to Tier 2 15.0-30 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses dropped fuel assembly lands horizontally on the top of the weir wall providing the minimum water depth above the dropped assembly. The methodology for determining fuel handling accident radiological consequences is consistent with the guidance provided in Appendix B of RG 1.183.

The inventory of fission products available for release at the time of the accident is dependent on a number of factors, such as the power history of the fuel assembly, the time delay between reactor shutdown and the beginning of fuel handling operations, the volatility of the nuclides, and the number of fuel rods damaged in a fuel assembly handling accident. The activity available for release is based on 102 percent power, bounding core inventory provided in Table 11.1-1, and a 1.4 radial peaking factor with 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> decay from time of reactor shutdown to the beginning of fuel handling operation. Activity is instantaneously released into the pool water from all fuel rods in the dropped assembly.

The following is a summary of the assumptions used from Appendix B of RG 1.183:

• radionuclides considered include xenon, krypton, halogen, cesium, and rubidium

• release fractions are from RG 1.183, Table 3 RAI 15.00.03-5

• depth of water above the damaged fuel of 23 feet is assumed RAI 15.00.03-5

• overall effective decontamination factor of 200 is assumed

• iodine chemical form released from the pool is 57 percent elemental iodine and 43 percent organic iodide

• no reduction or mitigation of noble gas radionuclides released from the fuel is assumed

• radionuclides are released to the environment over a two-hour period There are no single failures assumed for this event. Noble gases and iodines are released from the pool, while the cesiums and rubidiums are particulates and remain in the pool. The activity released from the pool to the RXB is assumed to be instantaneously released to the environment without holdup or mitigation. 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 potential radiological consequences of a fuel handling accident are summarized in Table 15.0-12.

RAI 15.00.03-1 15.0.3.8.6 Radiological Analysis of the Iodine Spike Design-Basis Source TermCategory 2 Maximum Hypothetical Accident Section 15.0.3.1 discusses how a MHA has historically been linked to a large-break LOCA in large LWRs and that, for the NPM, a large-break LOCA cannot physically be postulated as the basis for the MHA radiological consequence analysis. Section 15.6.5 presents the LOCA analysis, which shows that no fuel failures occur. The NuScale design has DBEs that result in primary coolant entering an intact containment and the Tier 2 15.0-35 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses iodine spike DBST is used to bound the radiological consequences of these events. The design-basis iodine spike DBST and the beyond-design-basis CDST described in Section 15.10 are each assessed against the radiological criteria of 10 CFR 52.47(a)(2)(iv). If both the design-basis iodine spike DBST and the beyond-design-basis CDST analyses show acceptable dose results 10 CFR 52.47(a)(2)(iv) is met. However, 10 CFR 52.47 (a)(2)(iv) requires nuclear power reactor design certification applicants to evaluate the consequences of a fission product release into the containment assuming substantial meltdown of the core. Therefore, this section presents source term design basis accidents. A source term design basis accident is a postulated accident scenario, meant as a surrogate to the large break LOCA typically evaluated by LWRs to meet the regulatory intent of addressing the MHA. Five source term design basis accidents derived from the Level 1 PRA were used to establish the DBST described in Section 15.0.3.3.4 in accordance with the methodology of Reference 15.0-4. Parameters associated with the DBST are presented in Table 12.2-28, Table 12.2.29, and Table 12.2-30.

RAI 15.00.03-1 To address 10 CFR 52.47 (a)(2)(iv), the DBST is assumed to occur, resulting in significant core damage. Activity is assumed to be released from the fuel over a specified time period, as described in Reference 15.0-4 and presented in Table 12.2-28, and assumed to homogeneously mix in the containment atmosphere. Removal of aerosol in the containment occurs through natural deposition phenomena. The aerosol removal methodology utilizes the code STARNAUA to determine the time-dependent airborne aerosol mass and removal rates, as described in Reference 15.0-4. Activity is released to the atmosphere from the containment at the design basis leakage rate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of the limit after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Reference 15.0-4 provides the methodology for the radiological consequences of the iodine spike DBST.MHA, based on the guidance provided in Appendix A of RG 1.183.

Assumptions used from Appendix A of RG 1.183 are:

• The chemical form of radioiodine released to the containment atmosphere is 95%

cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. Note that the methodology considers cesium iodide as an aerosol.

CP3-1603, Pending CP3-1655

• The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the containment.

RAI 15.00.03-1 The radioactive source term is calculated from the maximum core inventory provided in Table 11.1-2, multiplied by the release fractions provided in Table 12.2-29. The timing of the release and the radionuclide groups assumed, and the iodine removal mechanisms in the containment are provided in Table 12.2-28 and Table 12.2-30.

RAI 15.00.03-1 RADTRAD is used to determine the dose, as outlined in Section 15.0.3.3.8. The control room model is described in Section 15.0.3.7.1. The potential radiological consequences of the iodine spike DBST are presented in Table 15.0-12.

Tier 2 15.0-36 Draft Revision 3

NuScale Final Safety Analysis Report Transient and Accident Analyses The overcooling return to power analysis demonstrates that DBEs, where a most reactive control rod is assumed stuck out upon reactor trip, can be safely cooled by DHRS, or DHRS transitioning to ECCS cooling, without challenging MCHFR limits. Additionally, return to power scenarios with extended ECCS core cooling are limited by the density reactivity feedback as generated by the boiling in the core such that these scenarios are well bounded by the DHRS transition event due to the relative power levels in the core.

RAI 15-1 The evaluation of an overcooling return to power event demonstrates that design limits are not exceeded and the overcooling return to power event is non-limiting with respect to DBEs.

15.0.7 References 15.0-1 NuScale Power, LLC, Subchannel Analysis Methodology, TR-0915-17564, Rev.

1.

15.0-2 NuScale Power, LLC, NuScale Power Critical Heat Flux Correlations, TR-0116-21012, Rev. 1.

15.0-3 NuScale Power, LLC, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422, Rev. 0.

15.0-4 NuScale Power, LLC, Accident Source Term Methodology, TR-0915-17565, Rev. 23.

15.0-5 NuScale Power, LLC, Non-Loss-of-Coolant Accident Transient Analysis Methodology, TR-0516-49416, Rev. 1.

15.0-6 Nuclear Energy Institute, Small Modular Reactor Source Terms,

[Position Paper] December 27, 2012, Washington, DC.

15.0-7 NuScale Power, LLC, Long-Term Cooling Methodology, TR-0916-51299, Rev. 0.

15.0-8 U.S. Environmental Protection Agency, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020, 1988.

15.0-9 U.S. Environmental Protection Agency, External Exposure to Radionuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, 1993.

15.0-10 NuScale Power, LLC, Evaluation Methodology for Stability Analysis of NuScale Power Module, TR-0516-49417, Rev.0.

15.0-11 NuScale Power, LLC, NuScale Rod Ejection Accident Methodology, TR-0716-50350, Rev. 0.

Tier 2 15.0-46 Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-9, RAI 15-9S1 Table 15.0-1: Design Basis Events Section Type Classification Computer Code Used 15.0 Transient and Accident Analysis 15.0.3 Radiological Consequences of Category 2 Maximum Hypothetical AccidentIodine Spike Design-Basis Source Postulated RADTRAD Term (10 CFR 52.47(a)(2)(iv)) AccidentN/A(6) ORIGEN STARNAUA pHT ARCON96 15.0.6 Return to Power Event - NuScale specific phenomenon N/A(4) NRELAP5 15.1 Increase in Heat Removal by Secondary System 15.1.1 Decrease in Feedwater Temperature AOO NRELAP5 VIPRE-01 15.1.2 Increase in Feedwater Flow AOO NRELAP5 VIPRE-01 15.1.3 Increase in Steam Flow AOO NRELAP5 VIPRE-01 15.0-48 15.1.4 Inadvertent Opening of Steam Generator Relief or Safety Valve AOO NRELAP5 VIPRE-01 15.1.5 Steam Piping Failures Inside and Outside of Containment Postulated Accident NRELAP5 VIPRE-01 RADTRAD ORIGEN ARCON96 15.1.6 Loss of Containment Vacuum/Containment Flooding AOO NRELAP5 VIPRE-01 15.2 Decrease in Heat Removal by the Secondary System Transient and Accident Analyses 15.2.1 Loss of External Load AOO NRELAP5 VIPRE-01 15.2.2 Turbine Trip AOO NRELAP5 VIPRE-01 Draft Revision 3 15.2.3 Loss of Condenser Vacuum AOO NRELAP5 VIPRE-01 15.2.4 Closure of Main Steam Isolation Valve AOO NRELAP5 VIPRE-01 15.2.5 Steam Pressure Regulator Failure (Closed) N/A(1) N/A

Table 15.0-1: Design Basis Events (Continued)

Tier 2 NuScale Final Safety Analysis Report Section Type Classification Computer Code Used 15.6 Decrease in Reactor Coolant Inventory 15.6.1 Inadvertent Opening of Reactor Safety Valve AOO See 15.6.6 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment IE NRELAP5 RADTRAD ORIGEN ARCON96 15.6.3 Steam Generator Tube Failure Postulated Accident RADTRAD NRELAP5 ORIGEN ARCON96 15.6.4 Main Steam Line Failure Outside Containment (BWR) N/A(1) N/A 15.6.5 Loss-of-Coolant Accidents Resulting From a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Postulated Accident NRELAP5 Pressure Boundary 15.6.6 Inadvertent Operation of Emergency Core Cooling System AOO NRELAP5 15.7 Radioactive Release from a Subsystem or Component 15.7.1 Gaseous Waste Management System Leak or Failure N/A(2) N/A 15.0-50 15.7.2 Liquid Waste Management System Leak or Failure N/A(2) N/A 15.7.3 Postulated Radioactive Releases Due to Liquid Containing Tank Failures N/A(2) RADTRAD, ORIGEN, ARCON96 15.7.4 Fuel Handling Accidents Postulated Accident RADTRAD, ORIGEN, ARCON96 15.7.5 Spent Fuel Cask Drop Accident Postulated Accident Not analyzed 15.7.6 NuScale Power Module Drop Accident N/A(3) Not analyzed Transient and Accident Analyses Special Events 15.8 Anticipated Transient Without Scram (10 CFR 50.62) Special Event No analysis required.

15.9 Stability - note that stability is not an event. The NPM is protected from this phenomenon by MPS trips and N/A(4) PIM technical specification initial conditions.

15.10 Core Damage Source Term (10 CFR 52.47(a)(2)(iv)) Special Event RADTRAD Draft Revision 3 ORIGEN STARNAUA pHT ARCON96 MELCOR

Table 15.0-1: Design Basis Events (Continued)

Tier 2 NuScale Final Safety Analysis Report Section Type Classification Computer Code Used 8.4 Station Blackout (10 CFR 50.63) N/A(5) NRELAP5 Notes:

(1) Design feature is not part of NuScale design.

(2) Events are described in Chapter 11.

(3) Module drop is considered a Beyond Design Basis Event.

(4) Event is analyzed to AOO Acceptance Criteria.

(5) Event is included in the loss of non-emergency AC power analysis described in Section 15.2.6.

(6) The iodine spike DBST is not an event, rather it serves as a bounding surrogate for design-basis loss of primary coolant into containment events described in Section 15.6.

15.0-51 Transient and Accident Analyses Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report RAI 15-15 Table 15.0-2: Acceptance Criteria-Thermal Hydraulic and Fuel Classification(5) Fuel Clad(1) RCS Pressure Main Steam Containment Event Progression System Pressure AOO Fuel cladding integrity shall be 110% of system design 110% of system Peak pressure An AOO should not develop into a maintained by ensuring that minimum pressure design pressure design pressure(4) more serious plant condition without DNBR remains above the 95/95 DNBR other faults occurring independently.

limit. Satisfaction of this criterion precludes the possibility of a more serious event during the lifetime of the plant.

IE Fuel cladding integrity will be 120% of system design 120% of system Peak pressure Shall not, by itself, cause a maintained if the minimum DNBR pressure design pressure design pressure(4) consequential loss of required remains above the 95/ 95 DNBR limit. If functions of systems needed to cope the minimum DNBR does not meet with the fault, including those of the these limits, then the fuel is assumed to RCS and the reactor containment have failed. system.

Postulated Fuel cladding integrity will be 120% of system design 120% of system Peak pressure Shall not, by itself, cause a Accidents(2),(3) maintained if the minimum DNBR pressure design pressure design pressure(4) consequential loss of required 15.0-52 remains above the 95/ 95 DNBR limit. If functions of systems needed to cope the minimum DNBR does not meet with the fault, including those of the these limits, then the fuel is assumed to RCS and the reactor containment have failed. system.

Special Event (SBO) Core cooling refer to Section 8.4 N/A N/A N/A Notes:

(1) Minimum critical heat flux ratio (CHFR) is used instead of minimum DNBR, as described in Section 4.4.2.

(2) See Table 15.0-3 for acceptance criteria for the Rod Ejection Accident.

(3) See Table 15.0-4 for acceptance criteria for Loss of Coolant Accidents.

(4) See Section 6.2.1.1 for containment pressure design limits.

Transient and Accident Analyses (5) The iodine spike DBST and core damage event associated CDST do not have thermal hydraulic or fuel acceptance criteria.

Draft Revision 3

Tier 2 NuScale Final Safety Analysis Report Table 15.0-5: Acceptance Criteria-Radiological Event Analysis Release Duration Exclusion Area Boundary Control Room(3)

And Low Population Zone(1)(2) (Rem-TEDE)

(Rem-TEDE)

Loss of Coolant Accident (iodine spike 30 days for all leakage pathways. 25 5 DBST)

Rod Ejection Accident Per Appendix H of RG 1.183, a radiological analysis is 6.3 5 not required as the consequences of this event are bounded by the consequences of other analyzed events.

Steam Generator Tube Failure Affected steam generator: until time to isolation. 25 5 (Fuel Damage or pre-incident spike)

Unaffected steam generator: until reactor shut down 2.5 and depressurized. (Coincident Iodine Spike)

Main Steam Line Break Until reactor shut down and depressurized. 25 5 (Fuel Damage or pre-incident spike) 15.0-55 2.5 (Coincident Iodine Spike)

Fuel Handling Accidents 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.3 5 Small Lines Carrying Primary Coolant Until isolation, if capable, or until reactor shut down 2.5 5 Outside Containment and depressurized.

Feedwater System Pipe Breaks Inside Radiological analysis is not required as the 2.5(4) 5 and Outside Containment radiological consequences of the FWLB event are bounded by the consequences of a steam line break discussed in Section 15.0.3.

Core Damage Event 30 days for leakage pathways. 25 5 Transient and Accident Analyses Notes:

(1) Based on 10 CFR 52.47 (LOCA), RG 1.183, and 10 CFR 20.1301.

(2) Individual at the EAB shall not receive dose limit for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period flowing the onset of release.

(3) Based on 10 CFR 52.47 and is for the duration of the event.

Draft Revision 3 (4) Small fraction (10%) of regulatory dose reference value (25 rem TEDE).

Table 15.0-8: Reactivity Coefficients (Continued)

Tier 2 NuScale Final Safety Analysis Report Section Design Basis Event Power Level Moderator Doppler Coefficient(1)

% HFP (160 MWt) Temperature Coefficient(1) 15.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power 102% See Section 15.4.2 -1.40 pcm/°F 75%

50%

25%

15.4.3 Control Rod Misoperation (System Malfunction or Operator Error) See Section 15.4.3 See Section 15.4.3 See Section 15.4.3 15.4.4 Startup of an Inactive Loop or Recirculation Loop at an Incorrect N/A N/A N/A Temperature 15.4.5 Flow Controller Malfunction Causing an Increase in Core Flow Rate (BWR) N/A N/A N/A 15.4.6 Inadvertent Decrease in Boron Concentration in Reactor Coolant System See 15.4.1 and See 15.4.1 See 15.4.1 and 15.4.2(2) and15.4.2(2) 15.4.2(2) 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper N/A N/A N/A Position 15.4.8 Spectrum of Rod Ejection Accidents See Section 15.4.8 See Section 15.4.8 See Section 15.4.8 15.0-60 15.5 Increase in Reactor Coolant Inventory 15.5.1 Chemical and Volume Control System Malfunction 102% -43.0 pcm/°F -1.40 pcm/°F 15.6 Decrease in Reactor Coolant Inventory (3) 15.6.1 Inadvertent Operation of A Reactor Safety Valve See Section 15.6.6 See Section 15.6.6 See Section 15.6.6 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 102% -43.0 pcm/°F (EOC) -2.25 pcm/°F (EOC) 0.0 pcm/°F (BOC) -1.40 pcm/°F (BOC) 15.6.3 Steam Generator Tube Failure 102% -43.0 pcm/°F (EOC) -2.25 pcm/°F (EOC) 0.0 pcm/°F (BOC) -1.40 pcm/°F (BOC) 15.6.4 Main Steam Line Failure Outside Containment (BWR) N/A N/A N/A Transient and Accident Analyses 15.6.5 Loss-of-Coolant Accidents Resulting From a Spectrum of Postulated Piping 102% 0.0 pcm/°F -1.40 pcm/°F Breaks Within the Reactor Coolant Pressure Boundary 15.6.6 Inadvertent Operation of Emergency core Cooling System 102% 0.0 pcm/°F -1.40 pcm/°F 15.7 Radioactive release from a subsystem or component 15.7.1 Gaseous Waste Management System Leak or Failure N/A N/A N/A Draft Revision 3 15.7.2 Liquid Waste Management System Leak or Failure N/A N/A N/A 15.7.3 Postulated Radioactive Releases Due to Liquid Containing Tank Failures N/A N/A N/A 15.7.4 Fuel Handling Accidents N/A N/A N/A 15.7.5 Spent Fuel Cask Drop Accident N/A N/A N/A

Table 15.0-8: Reactivity Coefficients (Continued)

Tier 2 NuScale Final Safety Analysis Report Section Design Basis Event Power Level Moderator Doppler Coefficient(1)

% HFP (160 MWt) Temperature Coefficient(1) 15.7.6 NuScale Power Module Drop Accident N/A N/A N/A Special Events 15.8 Anticipated Transient Without Scram N/A N/A N/A 15.9 Stability See Section 15.9 See Section 15.9 See Section 15.9 15.10 Core Damage Event N/A N/A N/A Note:

(1) The moderator temperature coefficients and doppler coefficients include calculational uncertainty.

(2) The reactivity insertions possible for an inadvertent decrease in Boron concentration are bounded by the reactivity insertions analyzed in Section 15.4.1 and Section 15.4.2.

(3) The iodine spike DBST, which is a surrogate for the Section 15.6 DBEs that result in primary coolant entering the containment, does not have associated reactivity coefficients.

15.0-61 Transient and Accident Analyses Draft Revision 3

Table 15.0-9: Assumed Single Failures and Credited Nonsafety-Related Systems (Continued)

Tier 2 NuScale Final Safety Analysis Report Section Design Basis Event Assumed Single Failure Credited Nonsafety-rRelated sSystem 15.4 Reactivity and Power Distribution Anomalies 15.4.1 Uncontrolled Control Rod Assembly Withdrawal from a No adverse single failures None Subcritical or Low Power or Startup Condition 15.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power No adverse single failures None 15.4.3 Control Rod Misoperation (System Malfunction or Operator Failure of an ex-core flux detector with respect to power-related None Error) trips 15.4.4 Startup of an Inactive Loop or Recirculation Loop at an N/A N/A Incorrect Temperature 15.4.5 Flow Controller Malfunction Causing an Increase in Core N/A N/A Flow Rate (BWR) 15.4.6 Inadvertent Decrease in Boron Concentration in Reactor No adverse single failures None Coolant System 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an No adverse single failures None Improper Position 15.4.8 Spectrum of Rod Ejection Accidents No adverse single failures None 15.0-63 15.5 Increase in Reactor Coolant Inventory 15.5.2 Chemical and Volume Control System Malfunction No adverse single failures None 15.6 Decrease in Reactor Coolant Inventory (1) 15.6.1 Inadvertent Opening of Reactor Safety Valve See 15.6.6 See 15.6.6 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside No adverse single failures None Containment 15.6.3 Steam Generator Tube Failure Failure of MSIV for faulted steam generator - Radiological. Secondary MSIV No adverse single failures - thermal hydraulic acceptance criteria None 15.6.4 Main Steam Line Failure Outside Containment (BWR) N/A N/A Transient and Accident Analyses 15.6.5 Loss-of-Coolant Accidents Resulting From a Spectrum of No adverse single failures None Postulated Piping Breaks Within the Reactor Coolant Pressure 15.6.6 Inadvertent Operation of Emergency Core Cooling System No adverse single failures None Draft Revision 3 15.7 Radioactive release from a subsystem or component 15.7.1 Gaseous Waste Management System Leak or Failure N/A N/A 15.7.2 Liquid Waste Management System Leak or Failure N/A N/A 15.7.3 Postulated Radioactive Releases Due to Liquid Containing N/A N/A Tank Failures

Table 15.0-9: Assumed Single Failures and Credited Nonsafety-Related Systems (Continued)

Tier 2 NuScale Final Safety Analysis Report Section Design Basis Event Assumed Single Failure Credited Nonsafety-rRelated sSystem 15.7.4 Fuel Handling Accidents N/A N/A 15.7.5 Spent Fuel Cask Drop Accident N/A N/A 15.7.6 NuScale Power Module Drop Accident N/A N/A Special Events 15.8 Anticipated Transient Without Scram N/A N/A 15.9 Stability No adverse single failures None 15.10 Core Damage Event N/A N/A Note:

(1) The iodine spike DBST, which is a surrogate for the Section 15.6 DBEs that result in primary coolant entering the containment, does not have assumed single failures or credited nonsafety-related systems.

15.0-64 Transient and Accident Analyses Draft Revision 3

Table 15.0-10: Referenced Topical and Technical Reports (Continued)

Tier 2 NuScale Final Safety Analysis Report Topical or Technical Report Rev No Description NRC SER Report Reference Describes assumptions, codes, and methodologies used to calculate the radiological consequences of design basis accidents, including the iodine spike design-basis source term, and the beyond-design basis core damage event.

Describes the methodologies associated with performing the beyond-design-basis core damage source term radiological analysis and associated aerosol transport and iodine re-evolution assessment methodologies.Describes the methodology for establishing the NuScale design basis source term (DBST) release timing and magnitude that meets 10 CFR 52.47(a)(2)(iv). Describes the Accident Source Term TR-0915-17565 32 DBST associated aerosol transport and iodine re-evolution assessment Not issued Methodology (Topical) methodologies.

Describes the STARNAUA aerosol modeling to the range of post-accident containment conditions and justifies the assumption that no elemental iodine decontamination factor limit should be applied to natural aerosol removal phenomenon in the NuScale containment.

Describes the use of ARCON96 for establishing offsite atmospheric dispersion.

15.0-66 Discusses how NuScale Power, LLC, meets the NRC requirements for use of VIPRE-01 Safety Evaluation Reports (SERs), the modelling methodology for performing steady state and transient subchannel analyses, and the qualification of the code for application to the NuScale Power Plant design.

Subchannel Analysis TR-0915-17564 1 Explains why the methodology is independent of any one CHF correlation and Not issued Methodology (Topical) may be used for NuScale applications if methodology requirements are satisfied.

Describes methodology for treatment of uncertainties in the NuScale subchannel methodology.

Transient and Accident Analyses Provides the bases for use of critical heat flux (CHF) correlations in VIPRE-01within its range of applicability, along with its associated correlation limit, for NuScale Power Critical the NuScale Power, LLC, Design Certification Application and for the safety Heat Flux Correlations TR-0116-21012 1 Not issued analysis of the NPM with NuFuel-HTP2' fuel. The report describes the tests, test (Topical) facilities, statistical methods, base CHF correlation development, NSPX factor Draft Revision 3 development, and final validation for the development of the CHF correlation.

Tier 2 NuScale Final Safety Analysis Report Table 15.0-11: Summary of Applicable Radiological Events to the NuScale Design Event Dose Thermal Hydraulic Regulatory NEI Primary Source of Radiation Consequence Analysis Section Guide 1.183 Position Paper Analysis Section Appendix (1) Category (2)

Failure of Small Lines Carrying Primary 15.0.3.8.1 15.6.2 N/A 1 Coolant activity (with pre-existing and Coolant Outside Containment coincident iodine spiking)

Steam gGenerator tTube fFailure 15.0.3.8.2 15.6.3 F 1 Coolant activity (with pre-existing and coincident iodine spiking)

Main sSteam lLine bBreak 15.0.3.8.3 15.1.5 E 1 Coolant activity (with pre-existing and coincident iodine spiking)

Rod eEjection aAccident 15.0.3.8.4 15.4.8 H 1 Damaged fuel Fuel hHandling aAccident 15.0.3.8.5 N/A B 1 Damaged fuel Maximum Hypothetical Accident 15.0.3.8.69 N/A N/A 2 Damaged fuelCoolant activity (with (significant core damage)Iodine Spike pre-existing and coincident iodine spiking)

Design-Basis Source Term(3)

Core Damage Event(2) 15.10.2 N/A N/A Damaged fuel 15.0-68 Notes:

(1) Appendices C, D, and G were not included because they are not applicable to the NuScale design.

(2) The CDE is a beyond-design-basis special event.Reference 15.0-6.

(3) The iodine spike DBST is not an event, rather it serves as a bounding surrogate for design-basis loss of primary coolant into containment events described in Section 15.6.

Transient and Accident Analyses Draft Revision 3

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 BasisChapter 15 Analyses Event Location Acceptance Criteria Dose (rem TEDE)

(rem TEDE)

Failure of Small Lines Carrying Primary Coolant EAB 6.3 0.02 Outside Containment LPZ 6.3 0.04 CR 5.0 0.08 Steam Generator Tube Failure EAB 25.0 0.08 (pre-incident iodine spike) LPZ 25.0 0.08 CR 5.0 0.20 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 <0.01 CR 5.0 0.01 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 Iodine Spike Design Basis Source Term(1) EAB 25.0 <0.010.63 (pre-incident iodine spike)(significant core damage) LPZ 25.0 <0.011.37 CR 5.0 <0.012.14 Iodine Spike Design-Basis Source Term(1) EAB 25.0 <0.01 (coincident iodine spike) LPZ 25.0 <0.01 CR 5.0 <0.01 Core Damage Event(2) EAB 25.0 0.63 LPZ 25.0 1.37 CR 5.0 2.14 Notes:

(1) The iodine spike DBST is not an event, rather it serves as a bounding surrogate for design-basis loss of primary coolant into containment events described in Section 15.6.

(2) The CDE is a beyond-design-basis special event.

Tier 2 15.0-69 Draft Revision 3

NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory At this point, the LOCA event transitions to the post-LOCA long term core cooling phase. A gradual cool down and depressurization of the containment and RPV is occurring, as evident in the pressure, temperature, and level response depicted in Figure 15.6-42, Figure 15.6-45, Figure 15.6-47, and Figure 15.6-52 through Figure 15.6-54. Stable ECCS cooling is established and the module remains in a safe condition with liquid level maintained above the core throughout the entire transient duration. Collapsed liquid level has recovered to an equilibrium level of approximately 10 feet above the top of the core by the end of the transient.

During the post-LOCA long-term core cooling phase, the containment and RPV temperatures and pressures continue to decrease, indicating that the decay and residual heat are being removed from the RPV and containment. Sensitivities show that the maximum cooldown scenario consisting of a letdown line break with a reactor pool temperature of 40 degrees F, 20 percent pressurizer level and a 1.2 multiplier on decay heat results in the RCS minimum collapsed level. Boron precipitation does not occur at the time of the minimum collapsed liquid level, based on the core temperature being less than the highest boron precipitation temperatures for the highest boron concentration. Boron precipitation is also evaluated for the minimum RCS temperature during the 72-hour time following the LOCA, indicating that boron precipitation does not occur.

The MPS is credited to protect the NPM in the event of a LOCA. The following MPS signals provide the plant with protection during a LOCA:

• high pressurizer pressure

• high containment pressure

• low pressurizer level

• low pressurizer pressure

• low low pressurizer level

• high containment water level

• low RCS level 15.6.5.4 Radiological Consequences Section 15.0.3 presents the iodine spike design basis source term (DBST) methodology and the radiological consequences of the Category 2 maximum hypothetical accidentiodine spike DBST. The LOCA does not result in fuel failure, therefore the design basis source termiodine spike DBST bounds the source term, and thus the dose consequences, of a LOCA.

15.6.5.5 Conclusions The acceptance criteria for a LOCA, per 10 CFR 50.46(b), are listed in Table 15.0-4. These acceptance criteria, followed, by how the NuScale Power Plant design meets them, are listed below.

1) Peak cladding temperature - The calculated maximum fuel element cladding temperature shall not exceed 2200 degrees F.

Tier 2 15.6-18 Draft Revision 3

NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory open. Pressure and temperature inside the RPV continue a gradual downward trend, as shown in Figure 15.6-57, Figure 15.6-63, and Figure 15.6-64.

After the remaining ECCS valves open and pressure equalizes across the RRVs, liquid coolant from the containment begins to flow into the RPV downcomer region. This establishes a two phase natural circulation loop through the ECCS valves with steam exiting the pressurizer area into containment through the RVVs and liquid returning from the containment to the RPV through the RRVs. Decay heat and residual heat is transferred from the containment to the reactor pool resulting in the pressure and the temperature inside the RPV and containment continuing to decrease.

The transient continues until stable ECCS cooling has been established and RCS pressure and temperature continues to decrease. The module remains in a safe condition with liquid level maintained above the top of the core through the entire transient. The fuel volume average temperature is shown in Figure 15.6-65 and fuel cladding temperature is shown in Figure 15.6-66.

The MPS is credited to protect the module in the event of an inadvertent opening of an RVV by the following MPS signals:

• high containment pressure, and

• high containment water level No operator actions are credited for this event.

The event transitions to long-term cooling, similar to that described in Section 15.6.5.

15.6.6.4 Radiological Consequences Section 15.0.3 provides the radiological consequences for the NuScale infrequent events and postulated accidents. Radiological consequence analyses are not required for AOOs. Section 15.0.3 also presents the design basis source termiodine spike DBST methodology and the radiological consequences of the Category 2 maximum hypothetical accidentiodine spike DBST. The inadvertent opening of an RPV valve does not result in fuel failure, therefore the design basis source termiodine spike DBST bounds the source term, and thus the dose consequences, of this event.

15.6.6.5 Conclusions The acceptance criteria for an AOO are listed in Table 15.0-2. These acceptance criteria, followed, by how the NuScale Power Plant design meets them, are listed below.

Table 15.6-17 provides the results of the limiting scenario of a spurious opening of an RVV.

1) Fuel cladding integrity shall be maintained by ensuring that minimum DNBR remains above the 95/95 DNBR limit. Minimum critical heat flux ratio (MCHFR) is used instead of minimum DNBR, as described in Section 4.4.2.

Tier 2 15.6-25 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event 15.10 Core Damage Event A beyond-design-basis core damage event (CDE), with an associated core damage source term (CDST) composed of a set of key parameters derived from a spectrum of surrogate accident scenarios, is postulated. The beyond-design-basis CDE analysis and the design-basis iodine spike design basis source term (DBST) analysis described in Section 15.0.3 are each assessed against the radiological criteria of 10 CFR 52.47(a)(2)(iv), and if both analyses show acceptable dose results, 10 CFR 52.47(a)(2)(iv) is met. The analysis of the beyond-design-basis CDST against the acceptance criteria of 10 CFR 52.47(a)(2)(iv) provides reasonable assurance that, even in the extremely unlikely event of a severe accident, the facilitys design features and site characteristics provide adequate protection of the public.

The inputs, methods, and assumptions used to derive the CDST and analyze its radiological consequences are discussed in Section 15.10.1. The radiological consequences of the CDST are discussed in Section 15.10.2.

15.10.1 Inputs, Methods, and Assumptions 15.10.1.1 Core Radionuclide Inventory The core radionuclide inventory described in Section 15.0.3, and shown in Table 11.1-1, is assumed for the CDST.

15.10.1.2 Primary Coolant Radionuclide Inventory The primary coolant radionuclide inventory is assumed to be zero and is not considered as a contributor to dose in the CDE radiological consequence analysis, in accordance with the methodology of Reference 15.0-4.

15.10.1.3 Secondary Coolant Activity The secondary coolant activity is assumed to be zero and is not considered as a contributor to dose in the CDE radiological consequence analysis, in accordance with the methodology of Reference 15.0-4.

15.10.1.4 Source Term Release Timing and Magnitude The CDST associated with the CDE is composed of a set of key parameters, such as fuel release fractions and timing, derived from a spectrum of surrogate accident scenarios.

A surrogate accident scenario is a postulated event that results in core damage with subsequent release of appreciable quantities of fission products into an intact containment, that serves as a surrogate to the large break loss-of-coolant accident with a substantial meltdown of the core typically evaluated by light water reactors as the maximum hypothetical accident. Five surrogate accident scenarios derived from intact-containment internal events in the Level 1 probabilistic risk assessment were used to establish the CDST in accordance with the methodology of Reference 15.0-4.

Each of the five surrogate accident sequence cases involves various failure of the emergency core cooling system (ECCS), (i.e., all ECCS valves failing to open, the reactor vent valves (RVVs) failing to open, or the reactor recirculation valves (RRVs) failing to Tier 2 15.10-1 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event open). In each case, the decay heat removal system is assumed available to remove heat. The five surrogate accident scenario cases are summarized as follows:

Case 1: chemical and volume control system (CVCS) injection line break with all ECCS valves failing to open.

Case 2: CVCS injection line break with RVVs failing to open.

Case 3: CVCS injection line break with RRVs failing to open.

Case 4: loss of direct current (DC) power with the RVVs failing to open.

Case 5: loss of DC power with the RRVs failing to open.

Each surrogate accident scenario case is modeled using MELCOR to provide a representative range of release timing and fractions for the development of the CDST.

Section 15.0.2.4 provides a discussion of the MELCOR computer code. The methodology used to identify the release magnitude and timing of the CDST is provided in Reference 15.0-4. Release timing and core inventory release fractions for the CDST are listed in Table 15.10-1 and Table 15.10-2, respectively.

The radioactive source term is calculated by multiplying the maximum core inventory provided in Table 11.1-1 by the release fractions provided in Table 15.10-2. The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the containment.

15.10.1.5 Aerosol and Elemental Iodine Removal Natural deposition phenomena, including sedimentation, diffusiophoresis, thermophoresis, and hygroscopicity, result in aerosol removal. The NuScale aerosol removal methodology uses the aerosol removal code STARNAUA to track these various deposition phenomena in calculating time-dependent airborne aerosol mass and removal rates. Section 15.0.2.4 provides a discussion of the STARNAUA computer code.

Aerosol removal rates used for the CDE radiological consequence analysis are provided in Table 15.10-3. A key assumption of the NuScale aerosol transport methodology is that there is no maximum limit on iodine decontamination factor because removal is facilitated by natural processes, as opposed to an active spray system. The NuScale removal rate calculation methodology is based on the calculated time-dependent airborne aerosol mass in accordance with Appendix A, Section 3.3 of RG 1.183. NuScale conservatively does not take credit for elemental iodine removal. Rather, only aerosol removal is credited. A summary of the aerosol transport and removal calculation process is described in Reference 15.0-4. Treatment of aerosol resuspension and revaporization is discussed in Reference 15.0-4.

15.10.1.6 Chemical Form of Iodine Reference 15.0-4 provides the methodology for the radiological consequences of the CDE, based on the guidance provided in Appendix A of RG 1.183. The chemical form of radioiodine released to the containment atmosphere is 95 percent cesium iodide, Tier 2 15.10-2 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event 4.85 percent elemental iodine, and 0.15 percent organic iodide. Note that the methodology considers cesium iodide as an aerosol.

15.10.1.7 RADTRAD Modeling The RADTRAD modeling techniques described in Section 15.0.3.3.8 are used to analyze the CDST.

15.10.1.8 pHT and Iodine Re-Evolution The CDE radiological consequence methodology calculates the post-accident pHT. The pHT code is used to calculate the extent of iodine re-evolution inside containment.

During the postulated CDE, additional acids and bases may enter the coolant and cause a change in pHT. The expected overall pHT of the coolant is modeled over a period of 30 days. Section 15.0.2 provides a discussion of the NuScale pHT program used to calculate post-accident pHT.

Details about the methodologies used for evaluating post-accident pHT in coolant water following an event that results in significant core damage are presented in Reference 15.0-4. The results of implementing the methodology show the post-accident pHT inside containment is between 6.0 and 7.0.

The CDE radiological consequence methodology assumes a negligible amount of iodine re-evolution occurs between pHT values of 6.0 and 7.0, and does not need to be explicitly included in the dose analysis calculation. This assumption simplifies the analysis without an impact to the conservatism of the calculated dose results. The treatment of iodine re-evolution is described in Reference 15.0-4.

15.10.1.9 Atmospheric Dispersion Factors (/Q), Breathing Rates, and Occupancy Factors The atmospheric dispersion factor (/Q), breathing rate, and occupancy factor inputs to RADTRAD described in Section 15.0.3.3.11 are assumed in the CDE radiological consequence analysis.

15.10.1.10 Dose Conversion Factors The dose conversion factors described in Section 15.0.3.3.12 are assumed in the CDE radiological consequence analysis.

15.10.1.11 Containment Leakage The containment leakage assumptions described in Section 15.0.3.4 are assumed in the CDE radiological consequence analysis. Activity is released to the atmosphere from the containment at the design-basis containment leak rate provided in Table 6.5-1 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50 percent of the design-basis containment leak rate after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Tier 2 15.10-3 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event 15.10.1.12 Secondary-Side Decontamination The secondary-side decontamination assumptions described in Section 15.0.3.5 are assumed in the CDE radiological consequence analysis.

15.10.1.13 Reactor Building Decontamination Factors The Reactor Building decontamination assumptions described in Section 15.0.3.6 are assumed in the CDE radiological consequence analysis.

15.10.1.14 Receptor Locationsiderations The receptor location, control room, technical support center, and Reactor Building pool boiling radiological consequence assumptions described in Section 15.0.3.7 are assumed in the CDE radiological consequence analysis. The control room model is described in Section 15.0.3.7.

15.10.2 Radiological Consequences of the Core Damage Source Term Using the inputs, methods, and assumptions described in Section 15.10.1, the potential radiological consequences of the CDE are calculated and presented in Table 15.0-12. As shown in Table 15.0-12, NuScale meets the radiological acceptance criteria for the CDE.

Because NuScale already meets the radiological acceptance criteria, NuScale elected not to exercise the provisions in Reference 15.0-4 that allow less conservative analysis assumptions for the beyond-design-basis CDE (e.g., 50th percentile /Q values, Reactor Building decontamination factors, median core radionuclide inventory, etc.).

Tier 2 15.10-4 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event Table 15.10-1: Core Damage Source Term Release Timing Parameter Value Delay of radionuclide release into containment 3.80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> Duration of radionuclide release into containment 1.00 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> Tier 2 15.10-5 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event Table 15.10-2: Core Inventory Release Fractions Radionuclide Group Release Fraction into Containment Vessel Noble gases 3.9E-01 Halogens 1.4E-01 Alkali metals 2.0E-01 Alkaline earths 5.3E-03 Tellurium group 1.5E-01 Molybdenum group 4.9E-02 Noble metals 7.9E-04 Lanthanides 2.1E-08 Cerium group 2.1E-08 Tier 2 15.10-6 Draft Revision 3

NuScale Final Safety Analysis Report Core Damage Event Table 15.10-3: Containment Aerosol Removal Rates Time (hours) Removal Rate (hour-1) 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 Tier 2 15.10-7 Draft Revision 3