ML19316B738
| 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
Pre-Submittal Meeting November 12, 2019 Hope Creek 10 CFR 50.69 License Amendment Request 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