[Federal Register Volume 71, Number 156 (Monday, August 14, 2006)]
[Notices]
[Pages 46522-46525]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-13236]


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NUCLEAR REGULATORY COMMISSION


Pressurized Thermal Shock; Reports on the Technical Basis and 
Public Workshop

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of availability; notice of workshop.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is making available 
reports documenting the technical basis for a proposed revision of the 
NRC's pressurized thermal shock regulations. The NRC will also be 
conducting a two-day public workshop on this topic. The workshop is 
open to the public and all interested parties may attend.

DATES: The NRC is not soliciting comments at this time; however, NRC 
will request formal public comments when a notice of proposed 
rulemaking is published in the Federal Register. The public workshop 
will be: September 7, 2006, from 8:30 a.m.-12 p.m., Room T10-A1, and 
from 1 p.m.-4:45 p.m., Room T9-A1; September 8, 2006, from 9:30 a.m.-
3:45 p.m., Room T9-A1. If you plan to attend the workshop you are 
encouraged to preregister in order to facilitate security check-in on 
the day of the meeting.

ADDRESSES: Documents related to the proposed technical basis can be 
accessed electronically at http://www.nrc.gov/reading-rm/adams/web-based.html. From this site, you can access ADAMS, which provides text 
and image files of the NRC's publicly available documents. If you do 
not have access to ADAMS or if you experience problems accessing 
documents in ADAMS, contact the NRC's Public Document Room (PDR) 
reference staff by telephone at 1-800-397-4209 or 301-415-4737, or by 
e-mail to [email protected]. These documents may also be viewed on public 
computers located in the NRC's Public Document Room, Room O1-F21, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852. 
The PDR reproduction contractor will provide hard copies of the 
documents for a fee.

FOR FURTHER INFORMATION CONTACT: Dr. Mark T. Kirk, Office of Nuclear 
Regulatory Research, Component Integrity Branch, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
6015, facsimile 301-415-5074; e-mail [email protected].

SUPPLEMENTARY INFORMATION: During the operation of a nuclear power 
plant, the walls of the reactor pressure vessel (RPV) are exposed to 
neutron radiation, resulting in localized embrittlement of the vessel 
steel and weld materials in the core area. If an embrittled RPV had an 
existing flaw of critical size and certain postulated severe system 
transients were to occur, the flaw could very rapidly propagate through 
the vessel, resulting in a through-wall crack and challenging the 
integrity of the RPV. The postulated severe transients of concern, 
known as pressurized thermal shock (PTS) events, are characterized by a 
rapid cooling (i.e., thermal shock) of the internal RPV surface in 
combination with repressurization of the RPV. The coincident occurrence 
of critical-size flaws, embrittled vessel steel and weld material, and 
a severe PTS transient is a very low-probability event. Additionally, 
only a few currently operating pressurized-water reactors are projected 
to closely approach the current statutory limit on the level of 
embrittlement, as set forth in Title 10, Section 50.61, of the Code of 
Federal Regulations (10 CFR 50.61), ``Fracture Toughness Requirements 
for Protection Against Pressurized Thermal Shock Events,'' during their 
planned operational life.
    Advancements in our understanding and knowledge of materials 
behavior, our ability to realistically model plant systems and 
operational characteristics, and our ability to better evaluate PTS 
transients to estimate loads on vessel walls led NRC to conclude that 
the earlier analysis, conducted in the course of developing the PTS 
Rule in the 1980s, contained significant conservatism. Consistent with 
the NRC's Strategic Plan and the strategy to use realistically 
conservative, safety-focused research programs to resolve safety-
related issues, the NRC's Office of Nuclear Regulatory Research (RES) 
undertook a project in 1999 to develop a technical basis to support a 
risk-informed revision of 10 CFR 50.61. Two central features of the 
research approach include a focus on the use of realistic input values 
and models and an explicit treatment of uncertainties (using currently 
available uncertainty analysis tools and techniques). This approach 
improved significantly upon that employed to establish the 
embrittlement limits in 10 CFR 50.61, which originally included 
intentional and unquantified conservatism in many aspects of the 
analysis and treated uncertainties implicitly by incorporating them 
into the models.
    In early 2005, RES completed a series of reports that describe in 
detail the results of the project initiated in 1999. The information in 
these reports demonstrates that even through the

[[Page 46523]]

period of license extension, the likelihood of vessel failure 
attributable to PTS is extremely low ([ap]10-8/year). These 
results provide evidence that the statutory embrittlement limit 
established in 10 CFR 50.61 can be modified significantly to reduce 
unnecessary conservatism without affecting safety. This is possible 
because the operating reactor fleet has little probability of exceeding 
the limits on the frequency of reactor vessel failure, as they relate 
to NRC guidelines on core damage frequency and large early release 
frequency during either the currently licensed lifetime or the period 
of license extension.
    In early 2005, the RES reports were discussed with the NRC's 
Advisory Committee on Reactor Safeguards (ACRS) during a series of 
public meetings. Following these meetings, the ACRS sent letters to the 
NRC expressing the view that RES had developed a sound technical basis 
for a risk-informed revision of 10 CFR 50.61. More recently (June--
October 2005) staff from the NRC's Office of Nuclear Reactor Regulation 
(NRR) reviewed this technical basis and found it acceptable to begin 
the rulemaking process contingent upon the following three 
prerequisites:
    (1) Commission approval of the rulemaking plan, and dedication of 
resources
    (2) Successful resolution of outstanding technical issues 
identified in the existing technical basis
    (3) Making the technical basis documents available to the public
    This notice addresses prerequisite number 3.

Public Availability of Reports

    The following table provides the document titles and Agencywide 
Documents Access and Management System (ADAMS) accession numbers for 
all of the reports that, collectively, comprise the proposed technical 
basis for risk-informed revision of 10 CFR 50.61. The NRC staff 
recommends that persons interested in obtaining an overview of the 
technical basis and the recommended revisions to 10 CFR 50.61 focus 
their attention first on ADAMS Accession ML061580318. 
Interested persons can find more detailed information in the other 
supporting documents.

----------------------------------------------------------------------------------------------------------------
                 Topical area                  ADAMS ML                  Author & title
----------------------------------------------------------------------------------------------------------------
Previous Results.............................        ML030090626  Kirk, M.T., ``Technical Basis for Revision of
                                                                   the Pressurized Thermal Shock (PTS) Screening
                                                                   Criteria in the PTS Rule (10 CFR 50.61),''
                                                                   December 2002.
Current Results Summary......................        ML061580318  Kirk, M.T., et al., ``Technical Basis for
                                                                   Revision of the Pressurized Thermal Shock
                                                                   (PTS) Screening Limit in the PTS Rule (10 CFR
                                                                   50.61): Summary Report,'' NUREG-1806, Vol. 1.
Probabilistic Risk Assessment & Human Factors        ML992710066  Sui, N., ``Uncertainty Analysis and
 Analysis.                                                         Pressurized Thermal Shock: An Opinion,''
                                                                   September 3, 1999.
                                                     ML061580379  Whitehead, D.W., and A.M. Kolaczkowski, ``PRA
                                                                   Procedures and Uncertainty for PTS
                                                                   Analysis,'' NUREG/CR-6859.
                                                     ML042880452  Kolaczkowski, A.M., et al., ``Oconee
                                                                   Pressurized Thermal Shock (PTS) Probabilistic
                                                                   Risk Assessment (PRA),'' September 28, 2004.
                                                     ML042880454  Whitehead, D.W., et al., ``Beaver Valley
                                                                   Pressurized Thermal Shock (PTS) Probabilistic
                                                                   Risk Assessment (PRA),'' September 28, 2004.
                                                     ML042880473  Whitehead, D.W., et al., ``Palisades
                                                                   Pressurized Thermal Shock (PTS) Probabilistic
                                                                   Risk Assessment (PRA),'' October 6, 2004.
                                                     ML042880482  Whitehead, D.W., et al., ``Generalization of
                                                                   Plant-Specific Pressurized Thermal Shock
                                                                   (PTS) Risk Results to Additional Plants,''
                                                                   December 14, 2004.
                                                     ML042880476  Kolaczkowski, A.M. et al., ``Estimates of
                                                                   External Events Contribution to Pressurized
                                                                   Thermal Shock (PTS) Risk,'' October 1, 2004.
Thermal Hydraulics...........................        ML050390012  Bessette, D.E., ``Thermal-Hydraulic Evaluation
                                                                   of Pressurized Thermal Shock,'' NUREG-1809.
                                                     ML043570429  Reyes, J.N., et al., ``Scaling Analysis for
                                                                   the OSU APEX-CE Integral System Test
                                                                   Facility,'' NUREG/CR-6731.
                                                     ML043570405  Reyes, J.N., et al., ``Final Report for the
                                                                   OSU APEX-CE Integral System Test Facility
                                                                   Test Results,'' NUREG/CR-6856.
                                                     ML043570394  Fletcher, C.D., D.A. Prelewicz, and W.C.
                                                                   Arcieri, ``RELAP5/MOD3.2.2 Gamma Assessment
                                                                   for Pressurized Thermal Shock Applications,''
                                                                   NUREG/CR-6857.
                                                     ML061100488  Chang, Y.H.J., A. Mosleh, and K. Almenas,
                                                                   ``Thermal-Hydraulic Uncertainty Analysis in
                                                                   Pressurized Thermal Shock Risk Assessment:
                                                                   Methodology and Implementation on Oconee-1,
                                                                   Beaver Valley, and Palisades Nuclear Power
                                                                   Plants,'' NUREG/CR-6899.
                                                     ML043570385  Arcieri, W.C., R.M. Beaton, C.D. Fletcher, and
                                                                   D.E. Bessette, ``RELAP5 Thermal-Hydraulic
                                                                   Analysis To Support PTS Evaluations for the
                                                                   Oconee-1, Beaver Valley-1, and Palisades
                                                                   Nuclear Power Plants,'' NUREG/CR-6858.
                                                     ML061170401  Arcieri, W.C., C.D. Fletcher, and D.E.
                                                                   Bessette, ``RELAP5/MOD3.2.2 Gamma Results for
                                                                   the Palisades 1D Downcomer Sensitivity
                                                                   Study,'' August 31, 2004.
                                                     ML042880480  Junge, M., ``PTS Consistency Effort,'' October
                                                                   6, 2004.
Probabilistic Fracture Mechanics.............        ML061580343  Kirk, M.T., et al., ``Probabilistic Fracture
                                                                   Mechanics: Models, Parameters, and
                                                                   Uncertainty Treatment Used in FAVOR Version
                                                                   04.1,'' NUREG-1807.
                                                     ML051790410  Simonen, F.A., S.R. Doctor, G.J. Schuster, and
                                                                   P.G. Heasler, ``A Generalized Procedure for
                                                                   Generating Flaw-Related Inputs for the FAVOR
                                                                   Code,'' NUREG/CR-6817, Rev. 1.

[[Page 46524]]

 
                                                     ML061580369  Williams, P.T., T.L. Dickson, and S. Yin,
                                                                   ``Fracture Analysis of Vessels--Oak Ridge,
                                                                   FAVOR v04.1: Computer Code: Theory and
                                                                   Implementation of Algorithms, Methods, and
                                                                   Correlations,'' NUREG/CR-6854.
                                                     ML061580375  Dickson, T.L., P.T. Williams, and S. Yin,
                                                                   ``Fracture Analysis of Vessels--Oak Ridge,
                                                                   FAVOR v04.1, Computer Code: User's Guide,''
                                                                   NUREG/CR-6855.
                                                     ML061580358  Malik, S.N.M., ``FAVOR Code Versions 2.4 and
                                                                   3.1: Verification and Validation Summary
                                                                   Report,'' NUREG-1795.
                                                     ML042960391  Dickson, T.L., and S. Yin, ``Electronic
                                                                   Archival of the Results of Pressurized
                                                                   Thermal Shock Analyses for Beaver Valley,
                                                                   Oconee, and Palisades Reactor Pressure
                                                                   Vessels Generated with the 04.1 Version of
                                                                   FAVOR,'' October 15, 2004.
                                                     ML061580349  Kirk, M.T., et al., ``Sensitivity Studies of
                                                                   the Probabilistic Fracture Mechanics Model
                                                                   Used in FAVOR,'' NUREG-1808.
----------------------------------------------------------------------------------------------------------------

Public Workshop

    The NRC will conduct a public workshop on September 7-8, 2006, at 
NRC Headquarters, 11545 Rockville Pike, Rockville, Maryland. The 
purpose of this workshop is to inform the public of the reports 
detailed in the preceding section of this notice. A preliminary agenda 
for the workshop follows. If you plan to attend this meeting you are 
urged to contact Dr. Mark Kirk via e-mail to [email protected] at least 3 
business days before the meeting date so that your name can be included 
on the list of attendees and so you can be advised of any revisions to 
the agenda. You are strongly encouraged to communicate via e-mail, as 
this will facilitate the most efficient response to your inquiry.

Preliminary Agenda

                                                               Thursday, September 7, 2006
                                             [8:30 a.m.-12 p.m., Room T10-A1; 1 p.m.-4:45 p.m., Room T9-A1]
--------------------------------------------------------------------------------------------------------------------------------------------------------
             Start time                     Stop time            Duration [min]                 Topic                      Presenter/moderator
--------------------------------------------------------------------------------------------------------------------------------------------------------
8:30...............................  9.....................  30....................  Background of PTS Project   Kirk
                                                                                      (Overview, Objectives,
                                                                                      Reviews Performed to
                                                                                      Date).
9..................................  9:30..................  30....................  Status and Plan for         Mencinsky
                                                                                      Rulemaking.
9:30...............................  9:45..................  15....................  Overview of Reports.......  Kirk
9:45...............................  10:15.................  30....................  Questions from Public       Hardies
                                                                                      Regarding Reports and
                                                                                      Regulatory Process.
10:15..............................  10:30.................  30....................  Break.....................  .......................................
10:30..............................  11....................  30....................  Modeling Approach:          Kirk
                                                                                      Overview.
11.................................  11:30.................  30....................  Modeling Approach: Risk     Kolaczkowski
                                                                                      Assessment and Human
                                                                                      Factors.
11:30..............................  12....................  30....................  Modeling Approach: Thermal- Bessette
                                                                                      Hydraulics.
12.................................  1.....................  60....................  Lunch.....................  .......................................
1..................................  1:30..................  30....................  Modeling Approach:          Kirk
                                                                                      Fracture Mechanics and
                                                                                      Material Embrittlement.
1:30...............................  2.....................  30....................  Questions from the Public   Hardies
                                                                                      Regarding Modeling
                                                                                      Approach.
2..................................  3:30..................  90....................  Summary of Results from     Kirk
                                                                                      Baseline Analysis of
                                                                                      Three Plants.
3:30...............................  3:45..................  15....................  Break.....................  .......................................
3:45...............................  4:45..................  60....................  Questions from Public       Hardies
                                                                                      Regarding Baseline
                                                                                      Analysis.
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                                                                Friday, September 8, 2006
                                                            [9:30 a.m.-3:45 p.m., Room T9-A1]
--------------------------------------------------------------------------------------------------------------------------------------------------------
             Start time                     Stop time            Duration [min]                 Topic                      Presenter/Moderator
--------------------------------------------------------------------------------------------------------------------------------------------------------
9:30...............................  10:30.................  60....................  Summary of Study            Kirk
                                                                                      Generalizing the Results
                                                                                      to All Domestic PWRs.
10:30..............................  11....................  30....................  Questions from the Public   Hardies
                                                                                      Regarding Generalization.
11.................................  11:30.................  30....................  Proposed Allowable Through- Siu
                                                                                      Wall Cracking Frequency
                                                                                      Limit.
11:30..............................  11:45.................  15....................  Questions from the Public   Hardies
                                                                                      Regarding Through-Wall
                                                                                      Cracking Frequency Limit.
11:45..............................  1.....................  75....................  Lunch.....................  .......................................
1..................................  1:30..................  30....................  Proposed Material           Kirk
                                                                                      Embrittlement-Based
                                                                                      Reference Temperature
                                                                                      Limits for Use in a
                                                                                      Revised Version of 10 CFR
                                                                                      50.61.
1:30...............................  2.....................  30....................  Questions from Public       Hardies
                                                                                      Regarding Reference
                                                                                      Temperature Limits.
2..................................  3.....................  60....................  General Questions from      Hardies
                                                                                      Public.
3..................................  3:15..................  15....................  Break.....................  .......................................
3:15...............................  3:45..................  30....................  Summary...................  Hardies
--------------------------------------------------------------------------------------------------------------------------------------------------------



[[Page 46525]]

    Dated at Rockville, Maryland, this 4th day of August, 2006.

    For the U.S. Nuclear Regulatory Commission,
James T. Wiggins,
Deputy Director, Office of Nuclear Regulatory Research.
 [FR Doc. E6-13236 Filed 8-11-06; 8:45 am]
BILLING CODE 7590-01-P