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LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation: NuScale Source Term Methodology Application, PM-1119-67928, Revision 0
	ML19324C836
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Site:	NuScale
Issue date:	11/13/2019
From:	Rad Z; NuScale
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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	L0-1119-67933
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L0-1119-67933 November 13, 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 Presentation Materials Entitled "ACRS Subcommittee Presentation: NuScale Source Term Methodology Application,*

PM-1119-67928, Revision 0 The purpose of this submittal is to provide presentation materials to the NRG for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on November 20, 2019. The materials support NuScala's presentation of the "Accident Source Term Methodology* topical report The enclosure to this letter is the non proprietary version of the presentation titled

ACRS Subcommittee Presentation* NuScale Source Term Methodology Application,* PM-1119-67928, Revision 0 This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

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

Sincerely,

~

Director, Regulatory Affairs NuScale Power, LLC Distribution:

Robert Taylor, NRC, OWFN-8H12 Michael Snodderly, NRG, OWFN-8H12 Samuel Lee, NRC, O\\NFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Getachew Tesfaye, NRG, OWFN-8H12

Enclosure:

"ACRS Subcommittee Presentation: NuScale Source Term Methodology Application,"

PM-1119-67928, Revision 0 NuScale Power, LLC 1100 NE Crcle Blvd, Surte 200 Cavalhs, Oregon 97330 Office 541 360--0500 Fax 541 207 3928 Wiffl nusraepower com

L0-1119-67933

Enclosure:

"ACRS Subcommittee Presentation-NuScale Source Term Methodology Application," PM-1119-67928, Revision 0 NuScale Power, LLC 1100 NE Crrc!e Blvd, Surte 200 Corvallis, Oregon 97330 Office 541.360--0500 Fax 541 207 3928 www nuscaJaoower com

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2 PM-1119-67928 ReVJSlon. 0 Presenters Mark Shaver Radiological Engineering Supervisor Paul Guinn Radiological Safety Analyst Carrie Fosaaen Licensing Manager Jim Osborn Licensing Engineer Gary Becker Regulatory Affairs Council Copyright 2019 by NuScale Power, LLC.

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Agenda

Source-term-related open items

Accident source terms applications

Other Topics 3

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Acronyms 4

PM-111 ~7928 Revision 0 Term Definition

!_AS_~---*

_][ accide~t sour~ term

__ ]

--- -- - ----- ---- - - -- -- ----- - 1*- --

BTP

,: branch technical position

[ CDE ______________ Jl.9ore damage ev_e_nt _______==-c--_------J Ir

. COST

core damage source term

[CR ___ -___ ------~

- ][ co~-trol ro~~

__________ 1

--DB_A __ ---- - - -- ---- --- - -

'.~-des~-~asis a~~i~~i~~t ___ - - - --,

[ OBFFF ________________ _][ design basis_ failed fuel fraction j r

OBST

i design basis source term

~a-=c:_ ___ ---=--=~-~----~7-I[ envir~n~ental __ qualification __ ]

ESF

--1: engi~eere~i safuty f~atu~e - - -- -- 1

[~F-~~~--_

--=*----=-=------- __.JI fail-ed fu~I-fraction

________ -__ ]

I MHA

-- - -- r~aximum hyp~th~ti~~I ~~~id~~t:

l PAM_. -~- ~ ___.. _. _" _ ~~"~J post-accident man itoring__ _ __ J

PAS

1 post-accident sampling

[ PSCT ______ -__ --

--__ J[p~oi surge_~o~trol t~nk -~----J

.-TIO ____ - ------- ------ --- --- ------(t~t~I i~t~g~~t~~i ci~~~-- -- ------ ---

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Source-Term-Related Open Items

FSAR Chapter 2

- Open Item 02.03.04-1: staff evaluation to determine if TR-0915-17565 is acceptable for calculating accident offsite x/Q values

FSAR Chapter 3

- RAI 8837, multiple questions: staff request for clarification of TIO calculation methodology for DCA Part 2, Appendix 3C, Table 3C-8

FSAR Chapter 11

- RAI 9161, Question 11.01-1: staff evaluation of DBFFF as application in source terms for radiation shielding, ventilation systems, and radiation zoning

- RAI 9253, Question 11.01-2: staff request for inclusion of COL Item 11.2-3: evaluation of PSCT for BTP 11-6 5

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Source-Term-Related Open Items

FSAR Chapter 12

- Multiple items

FSAR Chapter 13

- RAI 9825, Question 13.03-1: staff evaluation of process sampling system

FSAR Chapter 15

- Open Item 15.0.2-6: staff review of the use of ARCON96, STARNAUA, and pHr as part of NuScale methodology (described in TR-0915-17565) for performing OBA radiological consequence analyses

FSAR Chapter 19

- Open Item 19.2.4-1: Possible inadequate description of equipment survivability in Ch.19; addressed by Ch. 19 revision and RAI 9705 6

PM-1119-67928 ReVISIOn 0 Copynght 2019 by NuScale Power, LLC.

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Source Term Overview Bounding Fuel lsotopics 66 ppm FFF Normal PCA 10x FFF !

660 ppm FFF Design Basis PCA = TS

""'.1, -

.~ -

- -- r ----

Single Assembly Activity Content

\\.

FSAR Ch.15 FHA Dose

--+I

\\

FSAR Ch. 15 :

REA Dose 1

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I CDST Release Fractions FSAR Ch.15 CDE Dose

~ FSARCh.19 Equip. Surv.

Dose TR-0915-17565 Content TS PCA + SOOx Iodine Spike f-----+, FSAR Ch.15 Iodine Spike DBST Dose

~ FSAR Ch. 3 Envr. Qual.

Dose i.........+

FSAR Ch.15 DBA Dose

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7 PM-1119-67928 Rev1s1on O Copynght 2019 by NuScale Power, LLC.

FSAR Ch.11 Normal Effluent FSAR Ch.12 Shielding 1% Failed Fuel FSAR Ch.12 Gaseous Tank Failure l

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Source Term Overview Single Assembly Activity Content r

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FSAR Ch.15 FHA Dose

....,.,..--------.... \\

FSAR Ch. 15 :

8 PM-1119-67928 ReVISlon* 0 REA Dose 1

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I Bounding Fuel lsotopics CDST Release Fractions 66 ppm FFF Normal PCA 10x FFF !

660 ppm FFF Design Basis PCA = TS TS PCP. + SOOx Iodine Spike FSARCh.11 Normal Effluent FSARCh.12 FSARCh.12 Gaseous Shielding Tank Failure r

1% Failed Fuel r - - - - - - - - ** -* - - - - - - - - - - - - - - - - - - - - I

--l+r f----+

FSAR ch. 15 Evaluate radiological 1

I FSAR Ch. 15 Iodine Spike consequences for I

CDE Dose DBST Dose,,

acceptability

- -- - - - - - - *~ - - --- - - --- - - - - - - --- - -

L-..+

FSAR Ch.19 f----+

FSAR Ch. 3 Equip. Surv.

Envr. Qual.

Dose Dose

,I

,I r

'I L--..+

FSAR Ch.15 DBA Dose

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Source Term Overview Single Assembly Activity Content FSAR Ch.15 FHA Dose

' ' ' ' ' ' ' L_-..r,--------,,

FSAR Ch. 15 :

9 PM-1119-67928 Revision: 0 I

REA Dose 1

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Bounding Fuel lsotopics CDST Release Fractions

__.r FSAR Ch. 15 CDE Dose 66 ppm FFF Normal PCA 10x FFF !

660 ppm FFF Design Basis PCA = TS i

TS PCA+ SOOx Iodine Spike f--+

FSAR Ch. 15 Iodine Spike DBST Dose

'I FSAR Ch.11 Normal Effluent

~

'I FSAR Ch.12

~

FSAR Ch.12 Gaseous Shielding Tank Failure

~

r 1% Failed Fuel r - - - - - - - - -

- I r

---I+

FSAR Ch. 19 f--+

FSAR Ch. 3 I

Equip. Surv.

Envr. Qual.

I Dose Dose Assure equipment functionality r

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Chapter 2 AST Application

In general, NuScale's site parameters are consistent with past applicant precedents and the EPRI ALWR URD.

Notable differences are

- Much smaller site footprint

Less atmospheric dispersion

Atmospheric dispersion (X/Q) methodology based on ARCON96

- NuScale site boundary (-140m) vs traditional LWR (-800-6000m)

- ARCON96 used in control room X/Q analyses is closer to NuScale distances and empirically proven to produce more accurate results than PAVAN at shorter distances

- Methodology described in AST L TR 10 PM-1119-67928 ReVJS1on 0 Copynght 2019 by NuScale Power, LLC.

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Chapter 3 Normal EQ Dose

FSAR Appendix 3C describes environmental qualification (EQ) program for qualifying equipment

- Normal operation dose for EQ derived from direct gamma emitted by design basis source term (-6-7 failed rods/core)

- Integrated dose for conservative 60-year equipment life

Environmental Qualification (EQ) program includes equipment in 10 CFR 50.49 scope: safety-related electric equipment and certain PAM equipment specified in RG 1.97 11 PM-1119-67928 RevJSIOn 0 Copyright 2019 by NuScale Power, LLC.

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Chapter 3 Accident EQ Dose

Accident EQ dose for FSAR Appendix 3C derived from both gamma and beta emitters from design basis source term (-6-7 failed rods/core+ iodine spike)

AST L TR Rev. 4 expanded scope to provide accident EQ dose methodology

- Iodine spike OBST is a design basis event and thus addressed by EQ per 10 CFR 50.49

- COST is a beyond design basis event, and thus beyond the scope of EQ 12 PM-1119-67928 ReVISIOn 0

Per SECY-90-016: stringent safety-related requirements, including 10 CFR 50.49, were not "commensurate with the importance of the safety functions to be performed" during severe accident miUgation.

Equipment survivability applied instead.

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Chapter 11 Source Terms

Two source term models are developed for both primary and secondary coolants:

- Design Basis and Normal Effluent ("Realistic") coolant source terms have three components:

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Water activation products

>> Calculated from first principles

>> The same concentration for both Normal Effluent and Design Basis

Corrosion activation products

>> Utilized ANSI 18.1-1999, adjusted to NuScale plant parameters

>> The same concentration for both Normal Effluent and Design Basis

>> The only component that strictly used the regulatory guidance provided

Fission products

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Chapter 11 Source Terms (cont.)

Normal Effluent ("Realistic") source term fission products

<1 failed rod per core (66 ppm) failure rate is assumed

Supported by industry experience from large PWRs, which shows much improvement since the 1970s

>> 90-95%, of US LWRs are zero-defect since 2010

Industry data (1987-2010) shows that most failures (90%) are due to grid-to-rod fretting (77°/o) and debris (13%)

>> NuScale uses natural circulation, which mitigates these mechanisms

>> Technical Report TR-1116-52065, Rev. 1

Design Basis source term fission products 14 PM-1119-67928 ReV1st0n: 0

7 failed rods per core (660 ppm) failure rate is assumed

>> 1 Ox normal effluent source term

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Chapter 12 Application

12.2 Source Terms

- Chapter 11 Design Basis Source Terms (660 ppm) used for normal operations design and shielding

- AST Design Basis Accident Iodine Spike source term for equipment qualification evaluations

12.3 Radiation Protection Features of the design accounting for Design Basis Source Terms (660 ppm)

12.4 Dose Assessments are informed by Design Basis Source Terms (660 ppm) 15 PM-1119-67928 R0VJS1on 0 Copyright 2019 by NuScale Power, LLC.

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Chapter 15 Design Basis Events

Small line break outside containment (FSAR § 15.0.3.8.1)

- Iodine-spiked primary source

Steam generator tube failure (FSAR § 15.0.3.8.2)

- Iodine-spiked primary source

Main steam line break (FSAR § 15.0.3.8.3)

- Iodine-spiked primary source

Rod ejection accident (FSAR § 15.0.3.8.4)

- Damaged fuel source

Fuel handling accident (FSAR § 15.0.3.8.5)

- Damaged fuel source

Iodine spike OBST (FSAR § 15.0.3.8.6)

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Chapter 15 DBE Dose Results I

Event Iodine Spike Design Basis Source Term (pre-incident iodine spike)

Iodine Spike Design-Basis Source Term (coincident iodine spike)

Main Steam Line Break (pre-incident iodine spike)

Main Steam Line Break (coincident iodine spike)

Steam generator tube failure (pre-incident iodine spike)

Steam generator tube failure (coincident iodine spike)

Primary coolant line break Fuel handling accident Location EAB LPZ CR EAB LPZ CR EAB LPZ CR EAB LPZ CR EAB LPZ CR EAB LPZ CR EAB LPZ CR EAB LPZ CR Acceptance Criteria rem TEDE 25.0 25.0 5.0 25.0 25.0 5.0 25.0 25.0 5.0 2.5 2.5 5.0 25.0 25.0 5.0 2.5 2.5 5.0 6.3 6.3 5.0 6.3 6.3 5.0 17 PM-1119-67928 Revision: 0 Copyright 2019 by NuScale Power, LLC.

Dose rem TEDE

<0.01

<0.01 0.01

<0.01

<0.01 0.08 0.08 0.20

<0.01

<0.01 0.02 0.04 0.08 0.55 0.55 0.89 WNUSCALE" f'owe1 lo

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Chapter 15 Core Damage Event

Core damage event:

- A special event (beyond design basis) with radionuclides from core damage released into an intact containment

- Postulated to enable deterministic evaluation of the response of the facility and site to the maximum hypothetical accident (i.e. a "substantial meltdown" event)

Five surrogate accident scenarios derived from intact-containment internal events in the Level 1 PRA were used to establish the COST

- The minimum onset time for fission product release from the gap, the release duration associated with minimum release onset time, and the median value of the release fractions determined from the spectrum of surrogate accident scenarios are used as the COST.

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Chapter 15 CDST Dose Results Event Location Acceptance Criteria (rem TEDE) Dose (rem TEDE)

EAB 25.0 LPZ 25.0 -~-~---..---

0.63 1.37 2.14 19 PM-1119-67928 CR

5.0 Revision

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Chapter 19 Application

Functionality of equipment that is necessary for mitigating a severe accident is, commensurate with the importance of the safety functions to be performed, reasonably assured by demonstrating equipment survivability

- The core damage source term (COST) is considered in the equipment survivability evaluation to demonstrate necessary equipment is available in a severe accident for its required functional duration

Following a severe accident, containment integrity and post-accident monitoring must be maintained

- Post-accident monitoring is not relied upon for mitigating severe accidents, but is intended only to provide information on severe accident conditions.

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21 Other Topics

1. PAS exemption request

- NuScale requested exemption from 10 CFR 50.34(f)(2)(viii) based on alternate means to assess core damage.

2. Application of GDCs to beyond design-basis accidents

- In general, NuScale maintains that 10 CFR 50 Appendix A does not apply to severe, beyond design-basis accidents, unless specifically invoked (e.g., GDC 19 via NUREG-0737, Item 11.8.2-10 CFR 50.34(f)(2)(vii)).

3. Radiological consequence contribution from potential leaks in non-safety hydrogen monitoring lines in 10 CFR 52.47(a)(2)(iv) analysis.

- NRC RAI 9690 Question No. 01.05-40

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1. Post-Accident Sampling Exemption Request

One of several TMl-related requirements that expressly considers a core melt source term.

PAS capability is not needed because NuScale design ensures the capability to assess core damage by other means.

- Under the Bioshield radiation monitors

- Core exit temperature indicators

Advantages

- Source term remains contained within module

- Reduced opportunity for leaks and spills

- Reduced operator doses

Exemption request is found in DCA Part 7, Chapter 16 22 PM-1119-67928 ReVJS1on* 0 Copyright 2019 by NuScale Power, LLC.

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2. Applicability of GDC 19

The GDCs define and establish acceptance criteria for design basis events for LWRs.

- 68 FR 54123, "Combustible Gas Control in Containment:

"The postulated accidents used in the development of [the GDCs] are normally design-basis accidents. The NRC believes it is not appropriate to address severe accident design requirements in the General Design Criteria."

- For Large LWRs, the "design basis LOCA" radiological consequences assessment (FSAR 15.6.5) is based on a core damage event. Thus, the control room dose limits of GDC 19 apply to this assessment.

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2. Applicability of GDC 19 (cont.)

TMI Action Item 11.B.2 (10 CFR 50.34(f)(2)(vii))

- Required a design review to ensure adequate shielding for operator access and component protection for "degraded core" accidents "beyond the design basis."

- Assured the design and licensing basis of then-operating plants was in-line with current guidance. The RG 1.3 and 1.4 source terms and the operator dose limits of GDC 19 were prescribed.

- Thus, the operator access requirements of 11.B.2 are redundant to GDC 19 under current guidance for Large LWRs.

NuScale's approach classifies the COE as a beyond design basis event

- GDC 19 does not apply

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3.

Hydrogen Monitoring System Leak

NRC issued RAI 9690 requesting that NuScale postulate a leak from the hydrogen monitoring system and analyze the resultant radiological consequences.

NuScale believes accounting for such leakage in the COST analysis is unnecessary to reasonably assure adequate protection

- Hydrogen monitoring capability is provided only for severe accidents and is not germane to any OBA.

- If the system leaks excessively, operators will isolate the leak, but this would be an unplanned and unexpected post-accident activity, and therefore does not require a separate dose analysis.

- Such potential leakage contributors are not included in guidance or past applications, apparently due to low risk.

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3.

Hydrogen Monitoring System Leak

NuScale followed established guidance provided in RG 1.183, Appendix A to evaluate offsite doses following COE.

NRC guidance excludes these known potential leakage pathways from the design basis accident radiological consequence analysis.

- RG 1.183 and SRP 15.6.5 include only containment and ESF system leakage for PWRs.

- TMI Item 11.B.2: "Leakage of systems located outside of containment need not be considered for the shielding review; leakage from those systems is "treated under Item 111.D.1.1."

- Item 111.D.1.1: requires a Leakage Control Program to "minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency."

3.

Hydrogen Monitoring System Leak

NuScale's COST evaluation is a beyond design basis event analysis.

NuScale does not believe potential leakage represents a significant safety risk.

Therefore, the existing guidance is adequate for NuScale to provide reasonable assurance that the worker and public are protected.

However, NRC staff have stated that they cannot reach a finding on the issue, and therefore intend to exclude the hydrogen monitoring leakage from issue resolution in the NuScale DC rulemaking.

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Questions?

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Portland Office 6650 SW Redwood Lane, Suite 210 Portland, OR 97224 971.371.1592 Corvallis Office 1100 NE Circle Blvd., Suite 200 Corvallis, OR 97330 541. 360. 0500 Rockville Office 11333 Woodglen Ave., Suite 205 Rockville, MD 20852 301. 770.0472 Charlotte Office 2815 Coliseum Centre Drive, Suite 230 Charlotte, NC 28217 980.349.4804 Richland Office 1933 Jadwin Ave., Suite 130 Richland, WA 99354 541. 360. 0500 Arlington Office 2300 Clarendon Blvd., Suite 1110 Arlington, VA 22201 London Office 1st Floor Portland House Bressenden Place London SW1E 5BH United Kingdom

+44 (0) 2079 321700 http://www. nuscalepower. com

'}I Twitter: @NuScale_Power 29 PM-1119-67928 ReV1s1on 0 Copynght 2019 by NuScale Power, LLC.

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