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Source Term Revision, Revision 0, PM-0118-58201
	ML18019A163
Person / Time
Site:	NuScale
Issue date:	01/23/2018
From:	Becker G, Gardner D, Shaver M; NuScale
To:	; Office of New Reactors
References
	LO-0118-58195 PM-0118-58201-NP, Rev. 0
	Download: ML18019A163 (22)
	v • d • e

NuScale Nonproprietary NuScale Source Term Revision Mark Shaver Radiological Engineering Supervisor Gary Becker Regulatory Affairs Counsel Darrell Gardner Licensing Project Manager January 23, 2018 PM-0118-58201 -NP *. NUSCALE Revision : 0 Copyright 2018 by NuScale Power, LLC. .:* POWER-Template # : 0000-21727-F0 1 R1

Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0000633.

This presentation was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights.

Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

Outline

Purpose

Current DCA Source Terms

Current AST LTR Approach

DCA Source Term Changes

- Source Term Changes Rationale

AST LTR Changes

- AST LTR Iodine Spike MHA

- AST LTR Core Damage MHA

Regulatory Considerations

DCA Impacts

Schedule

Summary 3

Purpose

Inform NRC of forthcoming changes to NuScale Accident Source Term Methodology Topical Report (AST LTR) and the DCA source term associated with the maximum hypothetical accident (MHA)

Note: "MHA" is shorthand for the scenario described in 10 CFR 52 .4 7(a)(2)(iv)

Elicit NRC feedback on

- technical approach

- regu latory interpretations

- schedules 4

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Current DCA Source Terms Various source terms are used in various design-basis analyses. Only the underlined text below will change.

Technica l 0.0028% Failed 0.028% Failed Specification 1% Failed Fuel Contents of 1 Core Damage Fuel Fraction Fuel Fraction Limit Primary Fraction Fuel Assembly Coo lant 5

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Current AST LTR Approach

MHA source term is derived from <1E-10/year PRA core damage sequences

- core damage leads to fiss ion product release into containment 6

DCA Source Term Changes

Incredible source terms (CDF<1 E-6/year) Technical involving core damage Specification Limit Core Darnage will only be used for Prirnary Coolant EPZ and SAMOA 7

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Source Term Changes Rationale

NuScale has elected to make these changes because the current MHA approach is overly conservative given the low probability of internal events core damage (<1 E-9/year)

- existing analyses using the current methodology result in overly conservative design requirements for

equipment qualification

control room habitability design

Additionally, for defense-in-depth, the NuScale containment's primary purpose is to facilitate emergency core cooling, with a secondary function of retaining fission products

- both prevention and mitigation functions 8

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Source Term Changes Rationale Steam escapes Condenses on Re-enters RPV Returns to core RPV CNV wall at RRV from RRV 9

Source Term Changes Rationale

IAEA defense-in-depth report (INSAG-10)

NuScale's containment provides both Level 4 and Level 3 TABLE I. LEVELS OF DEFENCE lN DEPTH Levels of defcncx Objective &sential means in depth Prevention of abnonnal opcralion Conservative design and high and failures quality in conslruction and operation Lcvc12 Control of abno.mtal operation and Corttrot limiting and protection detection of failures systems and other surveillance features Level 3 Control of accidents wid1in the Engineered safety features an.d OC!iign basis accident p rocedmcs Level 4 Control of severe plant conditions, Complementary* measures and incl.,.,ling prevention of accident accwent manageinent progression and mitigation of the consequences of severe accidents 1..c..*et 5 Mil:igsticm of radiological Off-sile emergency te_*pon~e:

consequ:cncxs of significant rclea~~s of radioactive material~

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AST LTR Changes

Logic gate added near beginning of AST LTR

/ >1E-6(year intact

/ containment core damage event?

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AST LTR Iodine Spike MHA

Surrogate MHA event derived from sequences

>1 E-6/year not resulting in core damage

- LOeA in containment initiating event (e.g., eves break in containment) is used, as it is representative of a spectrum of events that result in primary coolant in the containment

Radionuclides in primary coolant at technical specification limits released from core into containment

- assume Iodine spiking

Significantly less technically complex to analyze than the core damage MHA and significantly smaller source term 12 PM-0118-58201-NP

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AST LTR Core Damage MHA

Alternative path of logic gate is essentially the current MHA approach shown in AST LTR Rev. 2 that NRC already reviewed

- two changes to how dose is calculated from the MHA source term in AST LTR Rev. 3

aerosol modeling

leak rate assumptions 13 PM-0118-58201-NP Revision : 0 Copyright 2018 by NuScale Power, LLC .

Regulatory Considerations

10 CFR 52.47(a)(2):

" .. . The following power reactor design characteristics will be taken into consideration by the Commission: ...

(iii) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials; and (iv) ...design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release 3 from the core into the containment. .. The applicant shall perform an evaluation and analysis of the postulated fission product release ... to evaluate the offsite radiological consequences. "

Regulatory Considerations

10 CFR 52.47, Footnote 3 "The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products. "

Regulatory Considerations

Footnote 3 derives from 10 CFR Part 100 siting evaluation

- included phrase: "that would result in potential hazards not exceeded by those from any accident considered credible" and explicit reference to TID-14844 (see current 10 CFR 100.11)

- Part 100 rulemaking to "decouple siting from accident source term and dose calculations" modified to be compatible with "revised accident source terms" (63 FR 65159)

- 10 CFR 50.67 AST rulemaking , "There is no regulatory requirement for a specific source term for reactors to be licensed in the future" (64 FR 71995)

Therefore, iodine spike MHA should not require an exemption from 10 CFR 52.47(a)(2)(iv) 16 PM-0118-58201 -NP *:* NUSCALE Revision : O Copyright 201 8 by Nu Scale Power, LLC . .=* POWE::RN Template # : 0000-21 727-F01 R1

Regulatory Considerations

Containment leakage integrity remains applicable

- GDC 16, Containment design: " ... an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. .. "

Containment's function in accident prevention and mitigation is addressed by other requirements

- 10 CFR 52.47(a)(4 ): assess "adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents."

- 10 CFR 50.46 ECCS acceptance criteria, GDC 34 residual heat removal, GDC 35 emergency core cooling 17 PM-0 118-5820 1-NP '; NUSCALE Revision : 0 Copyright 2018 by NuScale Power, LLC . .**PoweR-r emplate # : 0000-21727-F01 R1

Regulatory Considerations

Consistent with other source term requirements

- GDC 19, Control room: "A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident."

- 10 CFR 52.79(a)(1 )(vi): Comparable requirement for COL applications, using actual site characteristics. COL would be expected to use same approach 18 PM-0118-58201-NP *. NUSCALE Revision : 0 Copyright 2018 by NuScale Power, LLC . ~.* POWER-Temp1ate # : 0000-21727-F01 R1

DCA Impacts

Referenced AST LTR Rev. 3 methodology change

- option of core damage or iodine spike MHA source term based on whether >1 E-6/year core damage event exists

- change to aerosol modeling

- change to leak rate assumptions

FSAR sections will use iodine spike MHA source term

- 3.11 and 3.C Environmental Qualification Methodology

- 12.2 Radiation Sources

Schedule

RAI 9224 response

- July 2018

AST LTR Rev. 3 submittal

- July 2018

DCA update unilateral letter

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Summary

Core damage MHA source term is overly conservative as a design basis for the NuScale Power Module

Iodine spike MHA source term consistent with 10 CFR 52.47(a)(2)(iv) and related requirements

RAI 9224 response, AST LTR Rev. 3, and associated DCA markups to be submitted July 2018 21 PM-0118-58201-NP :* *: NUSCALE Revision : 0 Copyright 2018 by NuScale Power, LLC. :. ,;* POWE:R-Temp1ate#: 0000-21727-F01 R1

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+44 (0) 2079 321700 http://www.nuscalepower.com 22 PM-0 11 8-58201-NP *. NUSCALE Revision : O Copyright 20 18 by NuScale Power, LLC . ** POWER-Temp1ate #: 0000-2 1727-F0 1 R1

ML18019A163

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