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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

ACRS Subcommittee Presentation I

I i

' NuScale Source Term l

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Presenters Mark Shaver Radiological Engineering Supervisor Paul Guinn Radiological Safety Analyst Carrie Fosaaen Licensing Manager Jim Osborn Licensing Engineer Gary Becker Regulatory Affairs Council 2

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Agenda

Source-term-related open items

Accident source terms applications

Other Topics 3

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Acronyms Term Definition

!_AS_~---* ______ _][ accide~t sour~ term __ ]

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

BTP ,: branch technical position

[ CDE ______________ Jl.9ore damage ev_e_nt_______ Ir

==-c--_------J

.' 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

.JI

[~F-~~~--_--=*----=-=------- __ fail-ed fu~I- fraction ________-__ ]

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

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

PAS : post-accident sampling 1

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

.-TIO____ - -- ----- -- ---- --- --- -- ----(t~t~I i~t~g~~t~~i ci~~~-- -- --- --- ---

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1, - - - - - -

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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 FSAR Ch.11 PCA Normal Effluent 10x FFF ! .

660 ppm FFF Design FSAR Ch.12 FSAR Ch.12 Basis PCA = TS Gaseous Shielding

- - - ""'.1, - - - - - - - - - - .~ - - - - - - - - -- r- - - -

, Tank Failure Single Assembly Activity Content CDST Release Fractions TS PCA + SOOx Iodine Spike , 1% Failed Fuel l

f-----+,

FSAR Ch.15 FSAR Ch.15 FSAR Ch.15 Iodine Spike FHA Dose CDE Dose DBST Dose

\. ,

--+I

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

~ FSARCh.19 ~ FSAR Ch. 3

FSAR Ch. 15 :

1 Equip. Surv. Envr. Qual.

I REA Dose I

\ I Dose Dose i.........+

FSAR Ch.15 TR-0915-17565 Content DBA Dose

,J L------------------ -----------

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

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Source Term Overview Bounding Fuel lsotopics 66 ppm FFF Normal -- FSARCh.11 PCA Normal Effluent 10x FFF ! '-

660 ppm FFF Design - FSARCh.12 FSARCh.12 Basis PCA = TS Gaseous

- Shielding Tank Failure Single Assembly Activity Content CDST Release Fractions TS PCP. + SOOx Iodine Spike 1% Failed Fuel r

r - - - - - - - - ** -* - - - - - - - - - - - - - - - - - - - - I

.~ :

r FSAR Ch.15

--l+r 15

" f----+ FSAR ch. 15 Evaluate radiological 1 I FSAR Ch. Iodine Spike consequences for I FHA Dose CDE Dose

' : '- ,, DBST Dose ,, acceptability

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

- - - - - - "'- - - *~ - - - - - - - - - - - - - - - - - - - - -

L-..+ FSAR Ch.19 f----+ FSAR Ch. 3

-- : FSAR Ch. 15 :

1 Equip. Surv. Envr. Qual.

I* REA Dose

\

_________ .,I I Dose

,I Dose

,I r 'I L--..+

FSAR Ch.15 DBA Dose -

,I 8

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Source Term Overview Bounding Fuel lsotopics 66 ppm FFF Normal - FSAR Ch.11

'I PCA Normal Effluent 10x FFF ! ~

660 ppm FFF Design -

~

FSAR Ch.12

'I FSAR Ch.12 Basis PCA = TS

- Shielding Gaseous Tank Failure Single Assembly CDST Release i

TS PCA+ SOOx r

~

Activity Content Fractions Iodine Spike 1% Failed Fuel

,----. __.r f--+ FSAR Ch. 15

FSAR Ch.15 FSAR Ch. 15 Iodine Spike

FHA Dose CDE Dose

' DBST Dose L_-..r,--------,,

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

- - - - - - - - - - - - - - I

---I+ FSAR Ch. 19 f--+ FSAR Ch. 3 Assure

FSAR Ch. 15 :

I I Equip. Surv. Envr. Qual. equipment

REA Dose 1 I I I Dose Dose functionality r

+ FSAR Ch. 15 DBA Dose 9

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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 LTR 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 Appendi x 3C derived from both gamma and beta emitters from design basis source term (-6-7 failed rods/cor e+ iodine spike)

AST LTR 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

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:

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

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

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

>> 1Ox 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 Assessm ents 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)

Chapter 15 DBE Dose Results Acceptance Criteria Dose Event Location rem TEDE rem TEDE EAB 25.0 <0.01 Iodine Spike Design Basis Source Term LPZ 25.0 <0.01 (pre-incident iodine spike)

CR 5.0 <0.01 EAB 25.0 <0.01 Iodine Spike Design-Basis Source Term LPZ 25.0 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 25.0 <0.01 Main Steam Line Break LPZ 25.0 <0.01 (pre-incident iodine spike)

CR 5.0 0.01 I

EAB 2.5 <0.01 Main Steam Line Break LPZ 2.5 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 25.0 0.08 Steam generator tube failure LPZ 25.0 0.08 (pre-incident iodine spike)

CR 5.0 0.20 EAB 2.5 <0.01 Steam generator tube failure LPZ 2.5 <0.01 (coincident iodine spike)

CR 5.0 <0.01 EAB 6.3 0.02 Primary coolant line break LPZ 6.3 0.04 CR 5.0 0.08 EAB 6.3 0.55 Fuel handling accident LPZ 6.3 0.55 CR 5.0 0.89 17 PM-1119-67928 WNUSCALE "

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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.

18 PM-1 119-67928

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

EAB 25.0 0.63 LPZ 25.0 1.37

-~-~---..---

CR 5.0 2.14 19 PM-1119-67928

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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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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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'}I Twitter: @NuScale_Power NUSCALE~

Power for all humankind 29 PM-1119-67928 ReV1s1on 0 Copynght 2019 by NuScale Power, LLC.

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