ML111640160

Summary of Meeting with STP Nuclear Operating Company Regarding Risk-Informed GSI-191, Assessment of Debris Accumulation on Pressurized-Water Reactor (PWR) Sump Performance; Resolution Approach for South Texas, Units 1 and 2 (TAC ME5358-ME5

ML111640160
Person / Time
Site:	South Texas
Issue date:	06/21/2011
From:	Balwant Singal Plant Licensing Branch IV
To:
Singal, Balwant, 415-3016, NRR/DORL/LPL4
References
TAC ME5358, TAC ME5359
Download: ML111640160 (8)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 21, 2011 LICENSEE: STP NUCLEAR OPERATING COMPANY FACILITY: SOUTH TEXAS PROJECT, UNITS 1 AND 2

SUMMARY

OF JUNE 2, 2011, PRE-LICENSING MEETING WITH STP NUCLEAR OPERATING COMPANY TO DISCUSS THE PROPOSED RISK INFORMED APPROACH TO THE RESOLUTION OF GSI-191, "ASSESSMENT OF DEBRIS ACCUMULATION ON PWR SUMP PERFORMANCE" (TAC NOS. ME5358 AND ME5359)

On June 2, 2011, a public meeting was held between the U.S. Nuclear Regulatory Commission (NRC), and representatives of STP Nuclear Operating Company (STPNOC, the licensee), at NRC Headquarters, Two White Flint North, 11545 Rockville Pike, Rockville, MD. The meeting notice and agenda, dated May 16, 2011, is located at Agencywide Documents Access and Management System (ADAMS) Accession No. IVIL111470652. The purpose of the meeting was to discuss the proposed risk-informed approach to the resolution of GSI-191, "Assessment of Debris Accumulation on PWR [pressurized-water reactor] Sump Performance." South Texas Project (STP) is the lead plant and STPNOC plans to submit a License Amendment Request (LAR) by May/June 2012. The licensee previously met the NRC staff on February 22, 2011, and provided an overall view of the proposed approach. The summary for the meeting on February 22, 2011, dated AprilS, 2011, is located at ADAMS Accession No. ML110770005.

STPNOC and the NRC staff decided to have a series of meetings with the NRC staff to discuss the following individual key sub-topics in more detail.

• Risk Informed (RI) GSI-191 Closure Plan, Conceptual Flow Diagram, and Licensing Strategy

• Containment Building Computer-Aided Design (CAD) Model

• Probabilistic Risk Assessment (PRA) Modeling

• Loss-of-Coolant Accident (LOCA) Initiating Event Frequencies and Uncertainties

• Thermal Hydraulics and Downstream Effects

• Jet Plume Formation, Zone of Influence (ZOI), and Debris Generation

• Debris Transport and Sump Performance

• Uncertainty Quantification and In-Vessel Effects

• Additional sub-topics may be added as needed basis

-2 The following topics were discussed during the meeting on June 2, 2011:

1. Risk Informed (RI) GSI-191 Closure Plan, Conceptual Flow Diagram, and Licensing Strategy

2. Containment Building Computer-Aided Design (CAD) Model

3. Probabilistic Risk Assessment (PRA) Modeling

4. Thermal Hydraulics and Downstream Effects

5. Loss-of-Coolant Accident (LOCA) Initiating Event Frequencies and Uncertainties The licensee's presentation slides are located at ADAMS Accession No. ML11157A01 O. A list of meeting attendees is provided in the Enclosure to this meeting summary.

Meeting Summary Risk Informed (RI) GSI-191 Closure Plan, Conceptual Flow Diagram, and Licensing Strategy The licensee presented the following:

• Flow chart representing overall process for defining inputs, developing models, performing the analysis, and analyzing and validating the results and analysis.

• Integrated closure plan showing the activities to be performed in 2011. The licensee plans to submit the LAR in May/June 2012.

• Graphical representation of the flow of data inputs and various calculation models. The NRC staff suggested that the licensee ensure that the model reflects how the uncertainties are being propagated.

• Licensing strategy, including the potential need for an exemption. The licensee debated if an exemption from the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Section 50.46 will be required. However, the NRC staff was of the opinion that an exemption will be required. The licensee took an action to discuss it internally and there will be more discussion during the future meetings.

Containment Building Computer-Aided Design (CAD) Model The licensee presented the main characteristics of the CAD model being prepared in support of the effort and explained the capabilities of the model. The NRC staff wanted to know the licensee plans to prepare separate CAD models for Units 1 and 2. The licensee indicated that configuration differences between Units 1 and 2 are so insignificant, that a separate model is not required and the model is being prepared based on Unit 1 design. The NRC staff also asked a number of questions to understand how the piping welds, hanger locations, and insulation is being incorporated in to the model.

- 3 PRA Modeling The licensee presented an overview of status and direction of the PRA model in support of the proposed effort. The licensee presented the following key features of the new PRA model:

• Develop expanded LOCA event trees

• The current model does not explicitly address sump plugging phenomena, but the new model will depict detailed representation of the plugging phenomena

• The following logic changes will be incorporated in to the model:

)0- Breakout sump blockage logic

)0- Add logic to represent in-core "downstream" phenomena

)0- Add logic to represent potential operational strategies

)0- Differentiate LOCAs based on potential impacts due to sump blockage

)0- The results of the detailed analysis and their insights to be incorporated and reflected by the PRA model Thermal Hydraulics and Downstream Effects The licensee explained that simulations of the reactor coolant system (RCS) thermo-hydraulic response by use of RELAP5-3D will be performed to determine:

• The main thermo-hydraulic parameters which are expected to affect the debris generation and transport inside the containment.

• The results of analysis will provide boundary conditions for the jet model development and estimate the effects of the debris deposition in the sump screen on the system response during accident requiring long-term cooling.

• Analysis will be performed for different break locations and sizes.

• Sensitivity analysis will also be performed for range of conditions requiring recirculation. Software DAKOTA coupled with RELAP5 will be used to perform the sensitivity analysis and uncertainty quantification.

The licensee indicated that the debris transport and deposition through downcomer, lower plenum, and reactor core will be investigated to determine the effects of the debris on the thermo-hydraulic response of the system (core and fuel temperature distribution) for downstream effects.

The licensee indicated that Computational Fluid Dynamics (CFD) codes will be used to predict the debris transport and deposition in the core and RELAP5-3D will be used to perform thermal hydraulics simulations of the reactor system NRC staff was concerned with the use of CFD Code and suggested that results of the analytical results provided by the use of CFD Code should be validated against the results of actual testing. The licensee was receptive to the suggestion. The NRC staff also suggested that the

-4 CFD model should have capability to evaluate the impact of non-conforming conditions discovered later. The licensee took an action to address the staff's suggestion.

Loss-of-Coolant Accident (LOCA) Initiating Event Frequencies and Uncertainties The licensee defined the scope, technical approach, key inputs and outputs, interface with other GSI-191 tasks, and issues and strategies for resolution for this project.

The following key process features to determine LOCA initiating event frequencies and uncertainties were described:

• Incorporate insights from previous work on LOCA frequencies.

• Characterize LOCA initiating events and their frequencies with respect to specific component, materials, and dimensions, specific locations, range of break sizes, degradation mechanisms and mitigation effectiveness and other break characteristics.

• Quantify uncertainties (augmented by sensitivity studies).

• Utilize passive component reliability methods and data from RHSI technology.

• Utilize PIPExp database to help resolve uncertainties in failure rates.

• Utilize information from NUREG-1829 and NUREG/CR-5750 in optimal manner.

• Consider probabilistic fracture mechanics evaluations.

• Use of Bayes' method for uncertainty treatment developed during EPRI RI-ISI program.

• Use of Markov model to evaluate the impact of changes to inspection on pipe failure rates.

• Provide estimates of LOCA initiating event frequencies and uncertainties for PRA model (RISKMAN).

• Provide location specific conditional probability vs. break size information for debris formation/thermal hydraulics.

• Address technical issues such as impact of aging on LOCA frequencies, reconcile differences between STP LOCA results vs. NUREG-1829 and NUREG/CR-5750.

Action Items The NRC and STPNOC staff agreed on the following actions before the next meeting.

NRC Staff Actions Provide feedback to STPNOC so that the agenda focus for the next meeting can be directed toward the resolution of issues. The NRC staff plans to provide input in the area of LOCA initiating event frequencies and uncertainties.

- 5 STPNOC Actions

• Produce an example component LOCA frequency calculation for the NRC staff's review.

• The CFD needs to be revised to show that the LOCA frequencies have a path to the PRA as well as CASA Grande Physics models. STP will update and enhance the slide prior to the next teleconference.

• Provide feedback to the NRC staff regarding STPNOC's approach towards deciding whether the chemical effects will be handled as a separate technical area and/or a separate top event in the PRA.

• STPNOC will compare the CFD model being used for downstream effects against the PWR Owner's Group testing.

No Public Meeting Feedback Forms were received for this meeting.

Please direct any inquiries to me at (301) 415-3016, or Balwant.Singal@nrc.gov.

Sincerely,

~L pD£ Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosure:

List of Attendees cc w/encl: Distribution via Listserv

LIST OF ATTENDEES JUNE 2. 2011, PRE-LICENSING MEETING RISK-INFORMED APPROACH TO RESOLUTION OF GSI-191 ISSUE STP NUCLEAR OPERATING COMPANY SOUTH TEXAS PROJECT UNITS 1 AND 2 DOCKET NOS. 50-498 AND 50-499 NAME TITLE ORGANIZATION Tim Sande** Principle Mechanical Engineer Alion Science and Technology

• Dave Midlik Reg. Affairs Coordinator Southern Nuclear Stewart Bailey Branch Chief NRC/NRR/DSS/SSIB Steve Smith Reactor Systems Engineer NRC/NRR/DSS/SSIB Stephen Dismore Senior Reliability Risk Analyst NRC/NRR/DRAIAPLA Paul Klein Senior Materials Engineer NRC/NRR/DCIICSGB Nan Gilles Technical Advisor for Reactors, Commissioner NRC/OCM i George Apostolakis i Dan O' Neal Reliability Risk Engineer NRC/NRR/DRAIAPLA

! John Burke Senior Engineer NRC/RES/DE/MEEB Donnie Harrison Branch Chief NRC/NRR/DRAIAPLA Jana Bergman - Scientech Balwant K. Singal Senior Project Manager NRC/NRR/DORL Ravi Grover Project manager NRC/NRR/DORL Charles Bowman General Manager, Nuclear Safety Assurance STPNOC I Steve Blossom Project Manager, Special Projects STPNOC Bruce Letellier** Technical Staff Los Alamos National Lab Jamie Paul Licensing STPNOC I Ernie Kee Risk Management STPNOC Karl Fleming** President KNF Consulting Services Rick Grantom Manager Risk Project STPNOC

! Yassin Hassa** Professor Texas A&M University i Rodlfo Vaghetto Phd Student Texas A&M University John Tsao Senior Materials Engineer NRC/NRR/DCIICPNB I I

David H. Johnson Vice President ABS Consulting Ralph Architzel Senior Reactor Engineer NRC/NRR/DSS/SSIB Blake Purnell Project Manager NRC/NRR/DPR/PGCB I Ervin Geiger Reactor System Engineer NRC/NRR/DSS/SSIB Bruce Lin Mechanical Engineer NRC/RES/DE/MEEB I John Lehning Reactor Engineer NRC/NRR/DSS/SNPB Michael Golay President MGI Enclosure

-2 NAME TITLE ORGANIZATION Robert Tregoning" Senior Technical Advisor NRC/RESIDE Jeff Weyhmiller* Engineer III Palisades Brian Brogan* Senior Staff Engineer Palisades Bill Beckius* Project Engineer Palisades Paul Stevenson* - Westinghouse Electric

• Attended the meeting via teleconference

• • Represented STPNOC Abbreviations:

NRC - U.S. Nuclear Regulatory Commission NRR - Office of Nuclear Reactor Regulation SSIB - Safety Issues Resolution Branch DORL - Division of Operating Reactor Licensing DSS - Division of Safety Systems STPNOC - STP Nuclear Operating Company ORA - Division of Risk Assessment APLA - PRA Licensing Branch DCI- Division of Component Integrity CSGB - Steam Generator Tube Integrity and Chemical Engineering Branch CPNB - Piping and NDE Branch DPR - Division of Policy and Rulemaking PGCB - Generic Communications and Power Uprate Branch RES - Office of Nuclear Regulatory Research DE - Division of Engineering MEEB - Mechanical and Electrical Engineering Branch OCM - The Commission

ML111640160 OFFICE INRRlLPL4/PM NRRlLPL4/LA NRRlDSS/SSIB/BC NRRlDRAlAPLAlBC NRRlLPL4/BC NRRlLPL4/PM BSingal NAME BSingal JBurkhardt SBailey DHarrison MMarkley (LWilkins for)

DATE 6/13/11 6/13/11 6/14/11 6/14/11 6/20/11 !f1l11