Attachment_1_Task_Order_No_1_Statement_of_Work.pdf

TASK ORDER NO. 1

STATEMENT OF WORK

1. PROJECT TITLE

Consequential Steam Generator Tube Rupture (C-SGTR) PRA

2. BACKGROUND

In 2009 the NRC closed the Steam Generator Action Plan (SGAP) (ADAMS ML100340994), based in part on the resolution of a substantial number of technical issues associated with the development of improved methods for assessing the risk associated with steam generator (SG) tubes under accident conditions. The NRC’s Office of Nuclear Reactor Regulation (NRR) subsequently recognized that additional research was needed to further enhance the analysis methods, tools, and expertise developed as a part of the NRC’s SGAP. NRR then requested that the NRC’s Office of Nuclear Regulatory Research (RES) develop additional products to facilitate the development and review of future risk assessments involving consequential steam generator tube rupture (C-SGTR). These products were to build upon research conducted for resolution of the SGAP and were to include additional thermal-hydraulic and structural analyses; characterization of SG flaw distributions; and improved probabilistic risk assessment (PRA) modeling.

To obtain technical assistance for this research project, the NRC awarded a contract Titled, ““Simplified Method for Assessing the Risk Associated with Consequential Steam Generator Tube Rupture Events,” documented a simplified methodology for quantitative assessment of the risk associated with consequential steam generator tube rupture during a severe accident after the onset of core damage, and during a Design Basis Accident (DBA) event before the onset of core damage. Key aspects of this work included:

Estimating the probabilities of large early releases and containment bypass that is the main focus for severe accidents; and

Screening probabilities for both core damage and containment bypass that are addressed for DBA events.

The contractor performed a number of innovative technical research tasks to support this work, including the following:

Developing an approach for probabilistically characterizing SG flaw distributions obtained from nuclear power plant operating experience (e.g., nondestructive examination of SG tubes) and applied the approach to representative data samples to develop a set of flaw-density probability density functions.

Selecting the accident scenarios of interest for C-SGTR events based on his extensive Probabilistic Risk Assessment (PRA) experience.

Characterizing the likelihood of design basis and severe-accident induced SG tube failures using a state-of-the-art approach for integrating PRA results, thermal-hydraulic analysis, and structural analysis for the reactor coolant system.

o The principal investigator developed the logic for calculator software capable of probabilistically quantifying the potential for severe accident induced SG tube failure given a representative SG flaw distribution and accident sequence.

In order to estimate offsite radiological releases associated with C-SGTR events, the principal investigator developed a PRA bridge tree model that allowed analysts to determine the probabilities of various release categories as part of a Level-2 (post core damage) PRA analysis.

o Using this model, the principal investigator developed an approach for estimating the large early release frequency (LERF) for C-SGTR events.

The subject matter in the C-SGTR report was expanded into a broad-scope draft NUREG report titled “Consequential SGTR Analysis for Westinghouse and Combustion Engineering Plants with Thermally-Treated Alloy 600 and 690 Steam Generator Tubes.”

The draft NUREG report included the following:

Accident sequence definitions and selection of pressure-induced and thermally-induced C-SGTR scenarios of interest for both Westinghouse and Combustion Engineering plant designs

Development of a simplified large early release frequency (LERF) model

Characterization and estimation of SG tube flaw distributions

Illustrative examples on the use of SG flaw distributions

The draft NUREG report provides conclusions and insights based on the steam generator geometry and high-temperature behavior of reactor coolant system (RCS) piping and SG tubes associated with two different PWR designs. State-of-the-art modeling of thermal-hydraulic behavior during severe accident conditions and high-temperature response of component systems has been employed in the study.

Assessment of potential fission product release potential due to C-SGTR following a severe accident sequence based on a LERF PRA model is also presented.

3. SCOPE

The scope of Task Order 1 is to provide technical assistance to ensure that the draft NUREG report is finalized as a NUREG report and issued in a timely manner, and that NRC offices are instructed on how to use the C-SGTR PRA methodology. The objective of this task is to expand and update Probabilistic Risk Assessment (PRA) methods used in nuclear regulatory activities, in compliance with the 1995 Nuclear Regulatory Commission (NRC) policy statement for implementing risk-informed regulation.

4. SPECIFIC TASKS

The contractor shall provide technical assistance to ensure that:

The draft NUREG report is finalized as a NUREG report and issued in a timely manner

NRC offices are instructed how to use the C-SGTR PRA methodology in risk-informed decisions involving steam generators that can include, but are not limited to:

o Rulemaking o Licensing o Oversight o Incident Response

This task will require extensive working knowledge of severe accident induced steam generator tube rupture modeling (for example, demonstrated knowledge of modeling described in NUREG-1570 and later NRC efforts undertaken as part of the Steam Generator Action Plan), severe accident behavior of nuclear power plant systems, modeling of high temperature behavior of nuclear power plant components including steam generator tubes, and estimation of radiation released as a consequence of severe accident induced C-SGTRs. In addition, the contractor is required to have extensive working knowledge with the Consequential Steam Generator Tube Rupture (C-SGTR) Calculator Performance: (1) As outlined in the diagram shown under Level-1 PRA of the Base Contract; and (2) Contained in publicly available briefing slides that were presented at the ACRS Meeting of the Subcommittees on Metallurgy & Reactor Fuels and PRA on April 07, 2015.

5. DELIVERABLES AND DELIVERY SCHEDULE

A Monthly Letter Status Report (MLSR) shall be provided by the contractor by the 20th of each month. This report will be used by the Government to assess the adequacy of the resources proposed by the contractor to accomplish the work contained in this contract and provide status of contractor progress in achieving tasks and producing deliverables. The report shall include contract/order summary information, work completed during the specified period, milestone schedule information, problem resolution, travel plans, and staff hour summary. The MLSR shall be delivered via email in PDF format.

6. LEVEL OF EFFORT

The estimated Level of Effort for Task Order 1 is 700 Hours.

7. PLACE OF PERFORMANCE

It is expected that the work will be performed at the contractor’s site.

8. TRAVEL

Estimated travel to NRC Headquarters in Rockville, Maryland is shown in the table below.

Meeting Title Estimated Date Number of Contractor Personnel to Attend Meeting

Kickoff Meeting At Start of Task 1

Progress Meeting No. 1 3 Months After Start of Task 1

Progress Meeting No. 2 6 Months After Start of Task 1

Draft Report Presentation 8 Months After Start of Task 1

Final Report Presentation 12 Months After Start of Task

Contingency Meeting No. 1 Unspecified 1

Contingency Meeting No. 2 Unspecified 1

9. PERIOD OF PERFORMANCE

12 Months from effective date task order.

10. SECURITY

The work to be done under this contract is UNCLASSIFIED.