ML19316B738

CFR 50.69 License Amendment Request, Pre-Submittal Meeting, November 12, 2019, Presentation by PSEG Nuclear LLC

ML19316B738
Person / Time
Site:	Hope Creek
Issue date:	11/12/2019
From:	Public Service Enterprise Group
To:	Office of Nuclear Reactor Regulation
Kim J
References
Download: ML19316B738 (8)

Text

Hope Creek 10 CFR 50.69 License Amendment Request Pre-Submittal Meeting November 12, 2019 1

Agenda Introduction/Opening Remarks Hope Creek Submittal Overview PRA Technical Adequacy Deviations from NEI 00-04

• Passive Categorization

• Seismic Justification Schedule Closing Remarks 2

PRA Technical Adequacy - Current PRA Models Model - Internal Events (With Internal Flooding)

Probabilistic Risk Assessment (PRA) model technical adequacy previously evaluated by Staff

• Inverter allowed outage time Facts & Observations (F&O) closure peer review performed 2017 in accordance with NEI 05-04 Appendix X.

Peer Review closed all F&Os meeting closure requirements.

Upgrades verified to have been peer reviewed.

Changes to as-built, as-operated plant reviewed periodically to determine if model impacts require off-cycle update.

Sensitivity studies in accordance with NEI 00-04 for areas such as Human Reliability Analysis (HRA) and Common Cause Failures (CCF).

Sensitivity study for FLEX to determine FLEX human event impacts.

3

PRA Technical Adequacy - Current PRA Models Model - Fire PRA Fire Probabilistic Risk Assessment (PRA) model technical adequacy previously evaluated by Staff

• Inverter allowed outage time Facts & Observations (F&O) closure peer review performed September 2018 with NRC observation in accordance with NEI 05-04 Appendix X.

Peer Review closed all F&Os meeting closure requirements.

Upgrades verified to have been peer reviewed.

Changes to as-built, as-operated plant reviewed periodically to determine if model impacts require off-cycle update.

Sensitivity studies in accordance with NEI 00-04 for areas such as Human Reliability Analysis (HRA) and Common Cause Failures (CCF).

Sensitivity study for FLEX to determine FLEX human event impacts.

4

Deviations from NEI 00-04 Passive Categorization Passive components and passive function of active components evaluated using Arkansas Nuclear One Risk-Informed Repair/Replacement Activities (RI-RRA) method.

Previously approved by NRC for Vogtle 10 CFR 50.69 application.

ASME Code Class 1 SSCs with pressure retaining function and supports will be assigned as high safety significant (HSS) for passive categorization.

• Resulting HSS for risk-informed safety classification cannot be changed by integrated decision-making panel (IDP).

5

Deviations from NEI 00-04 Seismic Hazard Assessment EPRI Technical Report 3002012988 for seismic considerations in categorization process.

Basis for meeting Tier 1 site criteria provided.

System seismic insights to be provided to IDP.

Lessons learned from lead Tier 1 plant incorporated.

6

Schedule NEI Coordinating Committee review and comment complete.

LAR submittal target by end of November 2019.

7

Closing Remarks 8