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Enclosure 1: NuScale Accident Source Terms Methodology Options PM-0818-61166, Revision 0
	ML18222A193
Person / Time
Site:	NuScale
Issue date:	08/08/2018
From:	; NuScale
To:	; Office of New Reactors
References
	LO-0818-61301, PM-0818-61166, Rev 0
	Download: ML18222A193 (28)
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L0-0818-61301 :

"NuScale Accident Source Terms Methodology Options" PM-0818-61166, Revision O NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541 .360-0500 Fax 541.207.3928 WWW nuscalepower Com

NuScale Nonproprietary NuScale Accident Source Term Methodology Options Gary Becker Regulatory Affairs Counsel Sarah Bristol Engineering Supervisor August 9, 2018 PM-0818-61166 Revision: O Copyright 2018 by NuScale Power, LLC.

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Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Numbers DE-NE0000633 and DE-NE0008742.

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.

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Outline

Purpose

Background

Overview of accident source term methodology options

- Options to revise characterization or implementation of the maximum hypothetical accident (MHA) analysis within the NuScale licensing basis

- Five total options, binned into two general groups

Summary 3

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Purpose

Continue discussion of NuScale's proposed accident source terms methodology

Present additional options to address MHA to resolve post-accident sampling and environmental qualification issues

Receive NRC feedback on various options presented 4

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Background

Original NuScale Accident Source Term Methodology Topical Report (AST LTR) led to the use of a maximum hypothetical accident (MHA) that was overly conservative

- Deterministically assumed significant core damage, which becomes design basis for certain aspects of plant design

- Post-accident sampling (PAS) doses and environmental qualification (EQ) of in-containment instruments began to drive possible design changes without identifiable safety benefit

- Notes:

"MHA" is shorthand for the scenario described in 10 CFR 52.4 7(a)(2)(iv), assumed for evaluating public doses at site boundary

NuScale relies on the normal process sampling system to serve as the post-accident sampling system (PASS) 5 PM-0818-6 1166 Revision: O Copyright 2018 by NuScale Power, LLC.

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Background

NuScale proposed a revised methodology to develop MHA in "Accident Source Terms Regulatory Framework White Paper" (WP-0318-58980)

- Credibility of core damage event informs selection of MHA

- MHA and deterministic event sequences together comprise design basis radiological releases

NuScale refined that proposal based on NRC feedback in previous public meetings and added a similar approach - Engineering Analysis (Option 1A below)

NuScale developed other ways to resolve licensing challenges associated with core damage MHA 6

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Regulatory Framework

Offsite dose analysis, 10 CFR 52.47(a)(2)(iv)

- Applicant is required to "assume a fission product release from the core into the containment," to determine that resulting offsite doses are within acceptable limits

Analysis considers leakage from an intact containment, effects of fission product mitigation systems, and postulated site parameters

This assumed fission product release is the "maximum hypothetical accident"

>> " ... a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events ... "

>> Original footnote further states "result[s] in potential hazards not exceeded by those from any accident considered credible"

- Therefore, MHA is historically rooted in worst "credible" accident 7

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Regulatory History

MHAwas originally defined by TID-14844 in 1962

- Generalized hypothetical scenario for site evaluation purposes

Derived from accidents deemed credible for very early reactor designs

Posited event was "arbitrary, " with "loose" tie to a major LOCA; nonmechanistic assumptions based on qualitative notion of credibility

- ACRS acknowledged "for certain reactors ... judgment will indicate that the generalized accident is too severe"

In 1963 AEC accepted a siting analysis reducing the assumed core melt from 100°/o to 6°/o based on crediting emergency core cooling system (ECCS) performance

General Design Criteria, and later 10 CFR 50.46, rendered core melt LOCA beyond the ECCS design basis

But "substantial core melt" persists as normal practice (RG 1.183) 8 PM-0818-61166 Revision: O Copyright 2018 by NuScale Power, LLC.

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Regulatory History

MHA as design basis

- Original role of MHA was to define an exclusion area and low population zone - i.e., separation from public as defense in depth

- Larger reactors could not meet dose limits without crediting engineered safety features (ESFs) to mitigate doses

- MHA became design basis for those ESFs and supporting SSCs (e.g. , control room habitability, instrumentation), despite being beyond design basis for ECCS

- 1996 Part 100 rulemaking moved requirement to Part 50 9

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Regulatory History

NuScale believes it is appropriate to reconsider core damage MHA assumption as design basis

- 1996 Part 100 rulemaking: "events having the very low likelihood of about 10-6 per reactor year or lower have been regarded in past licensing actions to be 'incredible', and as such, have not been required to be incorporated into the design basis of the plant"

- 1999 Alternative Source Terms rulemaking: "there is no regulatory requirement for a specific source term for reactors to be licensed in the future" 10 PM-0818-61166 Revision: O Copyright 2018 by NuScale Power, LLC.

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Overview of Options

There are two categories of options and several possible technical/regulatory variations within each group

- Group 1 options: Evaluation of design basis accidents, defense in depth ,

and risk insights to determine MHA source term

- Group 2 options: Limit application of core damage MHA in the design basis

Under all options:

- Doses to public are shown acceptable for spectrum of postulated events

- Emergency preparedness remains last layer of defense in depth

- No significant differences in design

- Compliance with 10 CFR 50.34(f)(2)(xix), core damage accident monitoring, will be addressed through the approach described in WP-0318-58980 (applying Equipment Survivability policy) 11 PM-08 18-6 11 66 Revision: O Copyright 2018 by NuScale Power, LLC.

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Option Group 1 Overview

Evaluation of design basis accidents and defense in depth to determine MHA source term

- Option 1A: Chapter 15 engineering analysis

- Option 1B: White Paper methodology with expanded risk insights discussion

Both options:

- Address MHA consistently and holistically in NuScale design basis

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Option 1A: Chapter 15 Engineering Analysis

Use iodine spike as MHA source term if:

- Chapter 15 design basis accidents do not result in fuel failure ,

AND

- design incorporates defense in depth , AND

- Chapter 19 confirms likelihood of severe accidents is very small

Otherwise, use core damage MHA source term 13 PM-0818-6 11 66 Revision: O Copyright 2018 by NuScale Power, LLC.

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Option 1A- Evaluate DBAs

NuScale Chapter 15 design basis accidents do not result in fuel failure

- The design eliminates many postulated accidents (large break LOCAs)

- The design increases reliability (natural circulation versus active systems, no reliance on electric power or nonsafety support systems, no operator actions credited)

- AOOs, infrequent events, postulated accidents, and special events use critical heat flux (CHF) as acceptance criteria :

CHF more restrictive than fuel failure (LWR)

- Chapter 15 accidents do not result in core uncovery; maintaining primary inventory ensures core cooling 14 PM-0818-61166 Revision : O Copyright 2018 by NuScale Power, LLC.

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Option 1A - Evaluate DID

NuScale design incorporates layers of defense in depth; defense in depth also accounts for uncertainties:

Prevention:

Simple, passive design eliminates numerous systems and components of principal concern to large light water reactors (LLWRs)- reduced operational challenges/transients

- Containment:

Redundant, diverse, passive safety systems require no AC power or operator actions

Small core size (5°/o LWR) with low power density and high ratio of RCS to core power - increased core coolability.

Additional redundant design features support core cooling (CVCS, CFDS).

Protection:

Reactor pressure vessel (RPV) & containment vessel (CNV) design minimizes challenges to containment integrity (no penetrations below fuel, vessel-in-vessel design , partially submerged in UHS, no concrete).

Seismic Category I reactor building.

Overall robust plant design - minimize challenges and survive hazards.

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Option 1A - Confirm Severe Accident Risk is Very Small

NuScale Chapter 19 confirms risk from severe accidents is very small:

- Very low NuScale PRA core damage frequency (CDF) provides significant margin to NRC's safety goal

- Very low NuScale PRA large release frequency (LRF) provides significant margin to NRC's safety goal and minimizes dose to workers and the public 16 PM-0818-61166 Revision : o Copyright 2018 by NuScale Power, LLC.

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Option 1B: White Paper Framework

Utilize both deterministic analysis and risk insights as proposed in "Accident Source Terms Regulatory Framework White Paper" (WP-0318-58980)

- Option 1B has been expanded to address Staff concerns, which are addressed in subsequent slides:

risk-informed versus risk-based

PRA acceptability

PRA uncertainty

COL items and finality issues

external events 17 PM-0818-6 11 66 Revision: O Copyright 2018 by NuScale Power, LLC.

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Option 1B - Risk-Informed

A NuScale Power Module (NPM) is evaluated against the five key principles of RG 1.174 in a risk-informed decision making process to determine the appropriate MHA source term

1. Current regulations are met, or are addressed through a requested exemption or rule change

2. Safety margins are maintained

3. Risk results are low and meet the Safety Goal Policy Statement and subsidiary objectives

4. Performance measurement strategies are employed to confirm the selection of the MHA source term.

5. Sufficient defense in depth exists (see next slide)

Interdisciplinary team (Engineering , Licensing, PRA) makes final recommendation:

- MHA iodine spike source term if principles and DID guidelines are met

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Option 1B - Defense in Depth

As shown below, the proposed framework includes both quantitative and qualitative elements of likelihood and consequences to evaluate plant capabilities and programmatic elements DID Layer DID Guideline Prevention Maintain frequency of all design basis events <10-2 per module critical year Maintain frequency of all beyond design basis events <1 Q-4 per module critical Containment year and individual sequences <1 Q-6 per module critical year1 Maintain large release frequency of all beyond design basis events <1 o-6 per Protection module critical year2 Ensure no single design or operational feature is exclusively relied upon to Overall ensure any of the layers of defense Notes:

1. The 104 per year threshold for core damage is the subsidiary objective derived to provide margin to NRC's safety goal. The 10.s per year threshold is consistent with past licensing actions (61 FR 65176) and the State-of-the-Art Reactor Consequence Analysis for core damage sequences that did not lead to an early containment failure or containment bypass.

2. The 10*6 large release threshold is consistent with the performance guideline established by NRC to determine whether a level of safety ascribed by a reactor is consistent with the safety goals.it also supports compliance with regulations on dose to workers and the public.

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Option 18 - Uncertainty & Scope

Consistent with the guidelines in NUREG-1855, uncertainties incorporated into the NPM PRA and also addressed through the use of "mean" risk values; sources of uncertainty not resolved via quantification are addressed by other methods such as sensitivity studies, safety margins, and defense in depth

The scope of events for MHA will include intact-containment core damage sequences from the PRA for all modes of operation (i.e., full power and low power and shutdown),

including external events

- The use of intact-containment beyond design basis events is consistent with 10 CFR 52.4 7(a)(2)(iv) in which accidents are hypothesized as releases into containment

In addition the scope of events will include consideration of potential multi-module core damage sequences 20 PM-0818-61166 Revision: O Copyright 2018 by NuScale Power, LLC.

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Option 1 B - PRA Acceptability & Monitoring

The DC/COL-ISG-028 guidance addresses how DC and COL applicants can use the ASME/ANS Standard, as endorsed by RG 1.200, for determining PRA technical adequacy

- Noted exceptions to meeting Capability Category II

generic versus plant-specific operating experience and data, and design and operational expectations versus plant-specific operating procedures and walkdowns

>> The ISG includes enhancements and new requirements to capture the documentation of assumptions, uncertainties, and their impacts due to the status of the design, site, operational, and maintenance information and data

- DC and COL PRAs should include self-assessments and external review

PRA updates will include review of MHA source term selection:

- When the PRA is performed for the COLA, site-specific information will be available, and the PRA will be updated to account for site-specific design information and any design changes

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Option Group 2 Overview

Assume core damage MHA, but limit application of MHA in the design basis

- Option 2A: Analyze MHA only for offsite dose analysis only

- Option 28: Exclude MHA from design basis of PAM and PAS

- Option 2C: Revise PAM and PAS capabilities under MHA

All options:

- Analyze a core damage MHA for offsite dose consequence evaluation

- Address MHA challenges narrowly; does not fully account for "enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials," per 10 CFR 52.47(a)(2)(iii)

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Option 2A

Analyze core damage MHA for offsite dose consequence purposes only

- Approach:

MHA offsite consequences analyzed separately from other Chapter 15 events, does not constitute design basis event

MHA analysis satisfies 10 CFR 52.4 7(a)(2)(iv), remaining Chapter 15 events constitute design basis

EQ, CNV leakage, and control room dose depend only on "Category 1" source terms (no new "Iodine Spike MHA" event)

- Considerations:

May require exemption from historical implementation of 10 CFR 52.47(a)(2)(iv) to define MHA as a BOBE

Does not resolve PAS challenges, additional exemption likely still required (see Option 2C) 23 PM-0818-611 66 Revision: O Copyright 2018 by NuScale Power, LLC.

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Option 28

Exclude MHA from design basis of PAS and EQ

- Approach:

Exemption from 10 CFR 50.49 to exclude MHA as an applicable OBA

Exemption from 10 CFR 50.34(f)(2)(viii) PASS rule that specifies MHA source term for sampling capability basis

Core damage MHA remains a Chapter 15 analysis associated with design basis of CNV leakage and control room dose

PAS capability demonstrated for non-core damage events only

- Considerations:

Selective, inconsistent application of core damage MHA

Uncertain basis for non-core damage capability PAS and PAM 24 PM-0818-6 11 66 Revision: O Copyright 2018 by NuScale Power, LLC.

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Option 2C

Revise PAM and PAS capabilities under MHA

-Approach

Revise EQ basis (e.g., FSAR Table 3.11-1) to reduce qualification duration for PAM

Instrumentation qualified to same dose levels as typical LWRs. PAM instrumentation qualified for duration actually required by operators for core damage event

Full exemption from 10 CFR 50.34(f)(2)(viii) PAS rule

- Considerations

PAM approach conforms with regulations and guidance

Departure from precedent of longer PAM survival durations justified by NuScale's lack of required operator actions

PAS would be limited to contingency means for sampling, only 25 PM-0818-61 166 Revision: O Copyright 2018 by NuScale Power, LLC .

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Conclusion

NuScale continues to believe it is appropriate to revisit the assumption of core damage for MHA

NuScale presented 3 additional options to more narrowly address design challenges associated with the core damage MHA 26 PM-0818-61166 Revision: O Copyright 2018 by NuScale Power, LLC.

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ML18222A193

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