ML19324C839

From kanterella
Jump to navigation Jump to search
LLC Submittal of Presentation Materials Entitled ACRS Subcommittee Presentation NuScale Topical Report -Accident Source Term Methodology, PM-1119-67927, Revision 0
ML19324C839
Person / Time
Site: NuScale
Issue date: 11/13/2019
From: Rad Z
NuScale
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LO-1119-67932 PM-1119-67927, Rev 0
Download: ML19324C839 (27)
v • d • e


Text

L0-1119-67932 November 13, 2019 Docket No.52-048 U.S. Nuclear Regulatory Comm1ss1on ATIN: Document Control Desk One Whrte Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Presentation Materials Entitled "ACRS Subcommittee Presentation* NuScale Topical Report -Accident Source Term Methodology," PM-1119-67927, Revision 0 The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on November 20, 2019. The materials support NuScale's presentation of the "Accident Source Term Methodology" topical report.

The enclosure to this letter is the nonproprietary version of the presentation titled

  • ACRS Subcommittee Presentation: NuScale Topical Report -Accident Source Term Methodology," PM-1119-67927, Revision 0.

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

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

Sincerely,

~~

Director, Regulatory Affairs NuScale Power, LLC Distnbution:

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

Enclosure:

"ACRS Subcommittee Presentation: NuScale Topical Report -Accident Source Term Methodology," PM-1119-67927, Revision 0 NuScale Power, LLC 1100 NE Crrcie Blvd, Surte 200 CoMtlhs, Oregoo 97330 Office 541 360-0500 Fax 541 207 3928 W'fNI nuscalepower com

Enclosure:

"ACRS Subcommittee Presentation NuScale Topical Report-Accident Source Term Methodology,"

PM-1119-67927, ReVJSion O NuScale Power, LLC 1100 NE Circle Blvd, Sulle 200 Corvalis, Oregon 97330 Office 541 360-0500 Fax 5412073928

'{NH nuscalepower com L0-1119-67932

ACRS Subcommittee Presentation I

I :. N uScale Topical Report I.

i.

Accident Source Term I

1 Methodology PM-1119-67927 ReV1s1on. 0 November 20, 2019 Copyright 2019 by NuScale Power, LLC.

M

~~g:_~-~~ -

Template# 0000-21727-f01 R5

2 PM-1119-67927 ReVJSlon. 0 Presenters Paul Guinn Radiological Safety Analyst Mark Shaver Radiological Engineering Supervisor Carrie Fosaaen Licensing Manager Jim Osborn Licensing Engineer Gary Becker Regulatory Affairs Council Copyright 2019 by NuScale Power, LLC.

M ti!:',~-~~-~s

  • Template#' 0000-21727-F01 R5

Agenda

  • NuScale AST Methodology Overview
  • Closed Session 3

PM-1119-67927 Revision: O Copyright 2019 by NuScale Power, LLC.

w

!'!.!l~.f ~.~~.

Template# 0000-21727-F01 R5

Acronyms 4

PM-1119-67927 ReVJS1on* 0 Term Definition

[~_?T

____, _______ ~--=-~- ___ _j~~-~i~~~~-~o~!~~ _ter~c--~-==--=-~c-J i

COE I: core damage event

~

[ C~S!_ ___ _ _ __________ -=~ __ _j§~:_d_a_~~~~-~o-~r~_ter~------

__ _J

CNV
containment vessel

[osA _ -- -_ --- ________ -_]~sign basis accident - -____ ]

, OBST

, design basis source term

_,1_

[~_B_

______ __ _ _____ _j[_~~~~-u-~i~~ -~~~a-~~un~a_r}' ______ __j

EQ

.: environmental qualification I

~[_F_H_A_~--~---*-C.. ____ *-_-_-_--_-_ --=--__ -_

-"--=-* _-

__ -_-J~u-el 7 ha-~dling--accid~nt =---~ -=~,-]

!i low population zone LMSLB -

____ -____. ___ ][ main steam_line break_

~ -----------

---,,---~--------------~------------

NEI

' Nuclear Energy Institute I

u~-_P-~-A--_-__ -__ -_--_-___ -_-_--_--__ -_-_ ----~~

__ J~_ri~~~~~ol~-~~-~c~i_v~~ _,_

0 7 J

, REA

rod ejection accident

[_ S~~ _________________________ J~~~~o~-~-~~~cc~~~~t_ ~~~a~i~-- ___ ]

, SGTF I* steam generator tube failure I

Copyright 2019 by NuScale Power, LLC.

M

!!!:l~.f-~1~-

Template #" 0000-21727-F01 R5

NuScale AST Methodology

  • Unique radiological consequence analysis methods within TR-0915-17565 include:

- Atmospheric dispersion

- Core damage source term

- Containment aerosol transport and removal

- Post-accident pHr *

- Iodine spike design basis source term (OBST)

- EQ dose within CNV and bioshield envelope

  • Industry standard radiological consequence analysis methods within TR-0915-17565 include:

- Design basis accident source terms

  • REA, FHA, MSLB, SGTF, and small line break accident methodologies 5

PM-1119-67927 Revision 0 Copynght 2019 by NuScale Power, LLC N ~!J.~.f,~b§

  • Template#. 0000-21727-F01 R5

NuScale AST Methodology

  • Application of core damage event (COE) limited by classifying it as a beyond design basis event

- COE used for control room dose evaluation and 10 CFR 52.47(a)(2)(iv) offsite dose evaluation

- COE not used for EQ

- COE used for equipment survivability evaluation

  • Unique design basis event "iodine spike source term"

- Surrogate for LOCA in containment without fuel damage events 6

PM-1119-67927 Revision 0

  • Assumes 100°/o of radionuclides in primary (plus an iodine spike) are in containment
  • Used in conjunction with COE to evaluate MHA radiological consequences for acceptability
  • Used in EQ Copyright 2019 by NuScale Power, LLC.

W

~!J~.f.!1.~.

Template#. 0000-21727-F01 R5

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

660 ppm FFF Design Basis PCA = TS

~----

Single Assembly CDST Release TS PCA+ SOOx Activity Content Fractions Iodine Spike

'I f---+

--+ FSARCh.15 FSAR Ch.15 FSAR Ch.15 Iodine Spike FHA Dose CDE Dose DBST Dose

\\..

~

~

I I

L. - _.,

\\

--+r

'I I

I L.....-+

FSAR Ch.19 FSAR Ch. 3 I

FSAR Ch.15 I

I Equip. Surv.

Envr. Qual.

I I

REA Dose I

I I

I I

Dose Dose

\\ _________,,,,

,)

I I

I

--+ FSAR Ch.15 TR-0915-17565 Content DBA Dose L--------------~--------------

7 PM-1119-67927 ReVJSion: 0 Copyright 2019 by NuScale Power, LLC.

r

~

FSAR Ch.11 Normal Effluent FSAR Ch.12 Shielding

'I

'I

\\..

FSAR Ch.12 Gaseous Tank Failure i

1% Failed Fuel M

~!,l.~fA~s" Template#". 0000-21727-F01 R5

8 Software Utilized

  • General application software:

SCALE 6.1, TRITON, and ORIGEN-S - used to calculate the time-dependent isotopic source term for all evaluated radiological events ARCON96 - used to calculate onsite and offsite atmospheric dispersion factors all evaluated radiological events RAOTRAO - used to estimate radionuclide transport and removal for all evaluated radiological events

  • OBA application software:

NRELAP5 - used to provide event-specific thermal-hydraulic data for design basis events

  • COE application software:

MELCOR - used to model the progression of severe accidents for the COE STARNAUA-used to perform aerosol removal calculations for the COE NuScale pHLCode - used to calculate post-accident aqueous molar concentration of hydrogen ions for iodine re-evolution evaluation in the COE MCNP6 - used for evaluating potential shine radiological exposures or doses to operators in the control room following a radiological release event PM-111 ~7927 Revision* 0 Copyright 2019 by NuScale Power, LLC.

N ~Y.~.f.!b~.

Template#' 0000-21727-F01 R5

Atmospheric Dispersion

  • TR-0915-17565 position: ARCON96 (RG 1.194) methodology used for the calculation of offsite atmospheric dispersion factors
  • ARCON96 and PAVAN compared to demonstrate conservative application of ARCON96 for NuScale site distances
  • Unique NARCON atmospheric dispersion model created to apply ARCON96 model results which incorporate RG 1.145 (PAVAN methodology) modeling conservatisms 9

PM-1119-67927 Revision. 0 Copyright 2019 by NuScale Power, LLC.

M

!!!"'~-~~1g.

Template#' 0000-21727-F01 R5

Core Damage Source*Term

  • Methodology based on state-of-the-art (post-RG 1.183)

SAND2011-0128 severe accident modeling and the approach of the 2012 Nuclear Energy Institute (NEI) position paper on SMR source terms

  • TR-0915-17565 positions:

- COE treated as beyond-design-basis event for the NuScale design, but evaluated in concert with iodine spike OBST to constitute a bifurcated maximum hypothetical accident

- COST derived from range of five surrogate accident scenarios 10 PM-1119-67927 Rev1s1on 0 taken from Level 1 PRA intact-containment internal events

  • SAN02011-0128 radionuclide groups for the COST release groups
  • COST release fractions taken as medians from SAS spectrum
  • COST release timing from scenario of quickest core damage onset Copyright 2019 by NuScale Power, LLC.

M

~!:',~.f~b§ -

Template#* 0000-21727-F01 R5

CDST Evaluation

  • SAND2011-0128-based representative (median) release fractions from the spectrum of surrogate accident scenarios used as the COST release fractions 11 PM-1119-67927 Revision: 0 1

2 3

4 5

6 7

~

9 SAND2011-0128 Radionuclide Groups Chemical Group Name Elements in Group Noble Gases Kr, Xe Halogens Br, I Alkali Metals Rb, Cs Tellurium Group Se, Sb, Te Alkaline Earths Sr, Ba Molybdenum Group Mo, Nb, Tc Noble Metals Ru, Rh, Pd, Co Lantbanides La, Nd,Eu,.Pm., Pr~ Sm, Y, Cm,Am Cerium Group Ce, Pu, Np, Zr Copyright 2019 by NuScale Power, LLC.

W

!i!J.~.~-~br Template#: 0000-21727-F01 RS

COST Evaluation (continued) omparison of release timing and magnitudes of example surrogate accident scenario cases Description Onset of gap release (hr)

Duration of gap plus early in-vessel release (hr)

Fraction of initial core Noble Gases Halogens Alkali Metals Alkaline Earths inventory Tellurium Group released into Molybdenum containment Group 17.6 12.0 0.39 0.21 0.25 5.9E-3 0.22 6.4E-2 3.8 8.1 1.0 9.0 0.19 0.41 3.5E-2 0.16 5.9E-2 0.22 2.8E-3 6.7E... 3 3.8E-2 0.16 1.3E-2 5.3E-2 21.3 8.1 30 sec 30 sec 1.3 14.0 9.0 1.8 5.63 0.19 0.48 0.39 1

0.872 1.9E-2 0.14 0.14 0.4 0.307 3.1 E-2 0.20 0.3 0.235 2.4E-3 5.3E-3 D.02 0.0054 2.3E-2 0.15 0.15 0.05 0.267 5.8E-3 4.9E-2 4.9E-2 0.0025 0.1 Noble Metals 1.2E-3 1.2E-4 1.5E-3 4.9E-5 7.9E-4 7.9E-4 0.0025 0.006 12 PM-1119-67927 Revision: 0 Lanthanides 3.3E-8 2.6E-9 3.1 E-8 1.1 E-9 2.1 E-8 0.0002 1.1 E-7 Cerium Group 3.3E-8 2.6E-9 3.1E-8 1.1E-9 2.1E-8 2.1E-8 0.0005 1.1E-7 Copyright 2019 by NuScale Power, LLC.

M

~~.~f.~!;r Template#: 0000-21727-F01 R5

RADTRAD Model Nodalizatlion 13 PM-1119-67927 Revision: 0 (C1)

Containment (P1)

Leakage EAB, X/Q #1 (C4)

Environment (P4)

Exhaust (P3 Leakage XJQ#4 (C2)

Control Room (P2) Normal (P5)

Emergency (C3)

Emergency Air (P6)

IIE----' Recirculation (P11)

(C6)

Pre-Filter CRVS (P12)

(C7)

Post-Filter CRVS HVAC XJQ #3..__ ___

(P7) Normal (PB)

Leakage XJQ#4 (P9)

Exhaust (C5)

TSC Filter (P10)

Recirculation Copyright 2019 by NuScale Power, LLC.

W

~!:'.~E~.~.r Template#: 0000-21727-F01 R5

Aerosol Transport and Removal

  • TR-0915-.17565 positions:

- STARNAUA is appropriate for modeling natural removal of containment aerosols for the NuScale design

- Utilizing thermal-hydraulic data associated with the surrogate accident scenario with the minimum time to core damage is appropriate for use in STARNAUA

- No maximum limit on iodine decontamination factor for natural removal of containment aerosols

  • Methodology credits sedimentation, diffusiophoresis, and thermophoresis removal mechanisms
  • Accident-specific natural deposition coefficient outputs of STARNAUA provided to RADTRAD dose transport model 14 PM-1119-67927 RevJS1on 0 Copynght 2019 by NuScale Power, LLC.

w

~!:'.~f-~.L.~.

Template#: 0000-21727-F01 RS

STARNAUA Removal Rate Error

  • Error identified when evaluating RAI 9224 (12/2017)

- Time-dependent aerosol removal rate values appeared nonphysical

  • Immediate actions:

- Condition extent evaluation initiated (12/11/2017)

- Notified NRC staff of identification (-12/15/2017)

- Error discovery letter transmitted to code vendor ( 12/21/2017)

  • Subsequent actions:

- Completed extent evaluation to identify removal rate output issue as only impact

- Developed alternative output postprocessing workaround

- Rededicated code 15 PM-1119-67927 Revision* 0 Copyright 2019 by NuScale Power, LLC.

M

~.!J.~.f ~!c.r Template# 0000-21727-F01 R5

STARNAUA Error Extent

  • Limited to incorrect post-processed aerosol removal (A) values; internal NAUA subroutines not affected
  • STARNAUA built as an extension of NAUAHYGROS 1.0

- Thermophoretic deposition and spray removal models added

- Changed output plot file generation to include calculated 'A values

  • NAUA-based codes historically demonstrated to predict aerosol mass concentration behavior reasonably well
  • STARNAUA confirmed to predict aerosol mass concentration well by multiple experiment benchmarks

- LACE 4, LACE 6, ABCOVE 5, ABCOVE 7 included

  • Removal calculation validated by manual calculation 16 PM-1119-67927 ReVJS1on 0 Copyright 2019 by NuScale Power, LLC.

w

~!:'.~.f ~b.§"

Template#. 0000-21727-F01 R5

17 Post-Accident pH1

  • TR-0915-17565 position: For pHT values of 6.0 or greater, the amount of iodine re-evolution that could occur between pHT values of 6.0 and 7.0 is negligible and not included in the dose calculation
  • Considers acids and bases expected in post-accident environment, including:

- lithium hydroxide, cesium hydroxide, hydrochloric acid, nitric acid, hydriodic acid, and boric acid

  • Given pHT, amount of iodine re-evolution estimated using NUREG/CR-5950 methods and shown to be negligible

(<1 °/o) with respect to other iodine modeling conservatisms PM-111 ~7927 ReVJS1on. 0 Copyright 2019 by NuScale Power, LLC M

~!:Jg:,~h.r Template# 0000-21727-F01 R5

Iodine Spike DBST

  • Iodine spike design basis source term:

- Bounding surrogate source term for any design-basis event involving primary coolant loss inside containment; postulated to enable deterministic evaluation of the response of a facility's engineered safety features.

18 PM-1119-67927 R8VISIOn 0

  • Assumes 100% of radionuclide inventory within 100°/o of primary coolant volume instantaneously, homogenously mixed release
  • Primary coolant inventory contains radionuclide concentrations at tech spec limits, plus iodine spike
  • Conservative evaluation-dependent leakage treatment
  • Additional conservative treatments available for discussion in closed session Copyright 2019 by NuScale Power, LLC.

N !!!l~.f~b.~

  • Template#* 0000-21727-F01 R5

Environmental Qualification Dose

  • TR-0915-17565 position: methodology described in TR-0915-17565, Appendix B is appropriate for calculating environmental qualification doses in the containment

-vessel (CNV) and bioshield envelope regions

  • Surrogate source term bounding for all primary coolant loss design basis events

- Assumes 1 OOo/o of radionuclide inventory within 100% of primary coolant volume instantaneously released inside containment

- Primary coolant inventory contains radionuclide concentrations at tech spec limits, plus iodine spike

- Multiple additional conservative treatments summarized in TR-0915-17565, Appendix 8 19 PM-1119-67927 Revision* 0 Copynght 2019 by NuScale Power, LLC.

W

~Y.?.~~br Template#" D000-21727-F01 R5

Other Positions

  • TR-0915-17565 positions:

- Utilizing the iodine spiking assumptions of RG 1.183 is appropriate

- Utilizing the iodine decontamination factor assumptions of RG 1.183 for the fuel handling *accident is appropriate

- Secondary coolant source negligible with regards to primary coolant source

- Containment shine to the environment is negligible for the NuScale design M

!i!:Jg:_~1~ -

20 PM-1119-67927 ReVJS1on. 0 Copyright 2019 by NuScale Power, LLC.

Template# 0000-21727-F01 RS

Design Basis Accidents

  • Small line break outside containment

- Follows SRP § 15.6.2

- Follows RG 1.183_ App. F

- Follows RG 1.183 App. E

  • Rod ejection accident

- Follows RG 1.183 App. H

  • Fuel handling accident

- Follows RG 1.183 App. B 21 PM-1119-6?927 Rev1s1on 0 Copynght 2019 by NuScale Power, LLC.

N ~!Jg:.~br Template#' 0000-21727-F01 R5

Portland Office 6650 SW Redwood Lane, Suite 210 Portland, OR 97224 971.371.1592 CoNallis Office 1100 NE Circle Blvd., Suite 200 Corvallis, OR 97330 541. 360. 0500 Rockville Office 11333 Woodglen Ave., Buite 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 SW1 E 5BH United Kingdom

+44 (0) 2079 321700 http://www.nuscalepower.com Twitter: @NuScale_Power 22 PM-1119-87927 Revision 0 Copyright 2019 by NuScale Power, LLC.

NUSCALE' Power for all humankind w

~!:l.~.~~!g Template#" 0000-21727-F01 R5

Backup Slides 23 PM-1119-87927 Revision 0 Copynght 2019 by NuScale Power, LLC w

~.!J.~.~~!,g Template#* 0000-21727-F01 R5

Overview of Different Boundaries

  • Restricted Area Boundary

- 10 CFR 50 Appendix I normal releases

  • Low Population Zone (LPZ) and Exclusion Area Boundary (EAB)

- 1 O CFR 50.34(a)(1) and 52.17(a)(1) accident releases

  • Emergency Planning Zone (EPZ)

- Independent of DCA 24 PM-1119-67927 ReVJS1on 0 Copyright 2019 by NuScale Power, LLC.

w

~!J.~.fl~.~g.

Template# 0000-21727-F01 R5

EAB, LPZ 25 PM-1 11 9-67927 Revision: 0 i

Figure 2.3-1: Limiting Analytical Distance to EAB and LPZ Outer Boundary

)

1..__ ____

__ b ooooooo 0000000 II 11

-_ --~ r 11 1 1 I.

Turbine 1 1 r-Build ing

, !D I

11 I I /

I I r- -*--*-*- -*--1 1

1 Radwaste Control I

1 1 I Ii I I

t 1 1 I I I r

Non-Limiting f

! I Build ing Building

~*

11

' ~~-

...... -:,.--.--~ R-e~a-ct-o~r~

__,_, d

.iiiE 1

Bui lding

--..._ Non-Limiting...,_,---->...........,_~~-

~

,_Analytical X/Q Distance

)

1011 1 I I !

t f I,'~

Limiting L

Analytical X/Q L ___ _

Turbine Building Ana lytk a I X/Q f

Distancl! _ ___ ~

1 I 11 I

___ -*-__ --------~-~

j e:JI OIOIOII Distance VkLJ L.J

( 4 0 0 feet )

~

=r""'""";==j==,,,......,.....'j"""'=~

~ -----

~

0000000

--------..i,..... -="=-----..

0 0 0 0 0 0 0

-~ ~~~~==='===='=~

AB, LPZ

.._I NUSCALe **

  • ~

f'owe r l o* u l1 h1i rn o r t l 11 d Copyright 2019 by NuScale Power, LLC.

Template#: 0000-21727-F01 R5

Retrieved from "https://kanterella.com/w/index.php?title=ML19324C839&oldid=4765564"