2200°F is alive, but it has never been well!

I have "parked the following documents here. Sorting and discussion will follow at a later date.May 21, 2010
MEMORANDUM TO: Robert D. Carlson, Chief
Financial Analysis Branch
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
FROM: Richard F. Dudley, Senior Project Manager /RA/
Financial Analysis Branch
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
SUBJECT: FORTHCOMING PUBLIC MEETING TO DISCUSS
NRC STAFF RESOLUTION OF PUBLIC COMMENTS ON
SUPPLEMENTAL PROPOSED RULE ON RISK-INFORMED CHANGES
TO LOSS-OF-COOLANT ACCIDENT TECHNICAL REQUIREMENTS
(10 CFR 50.46a) AND DRAFT FINAL RULE LANGUAGE
DATE & TIME: Friday, June 4, 2010
8:30 a.m. – 12:30 p.m.
LOCATION: U.S. Nuclear Regulatory Commission
Two White Flint North, Room T-2B3
11545 Rockville Pike
Rockville, Maryland 20852
PURPOSE: To discuss (1) NRC responses to major comments received on the
10 CFR 50.46a Supplemental Proposed Rule published on August 10,
2009 (74 FR 40006) and (2) NRC’s associated draft final rule
language.
Draft final rule language is available for review in preparation for this
meeting at ADAMS Accession Number ML101250271 and on
Regulations.gov under Docket ID NRC-2004-0006. Note that the NRC is
not soliciting or accepting formal public comments at this time. This
preliminary rule language has not been fully vetted through internal
NRC reviews, and therefore may be subject to significant revision before
formal publication.
MEETING CONTACT: Richard F. Dudley, NRR
301-415-1116
Richard.Dudley@nrc.gov
- 2 -
CATEGORY 3: * This is a Category 3 Meeting: The public is invited to participate in this
meeting by providing comments and asking questions throughout the
meeting.
The NRC provides reasonable accommodations to individuals with
disabilities where appropriate. If you need a reasonable accommodation
to participate in this meeting (e.g., sign language), or need this meeting
notice or other information from the meeting in another format, please
notify the NRC meeting contact by May 28, 2010, so that arrangements
can be made.
Teleconferencing: Interested members of the public unable to attend the
meeting may participate by telephone via a toll-free teleconference. For
details, please call the NRC meeting contact listed or the toll-free number,
1-800-368-5642, and ask the operator for a connection to the meeting
contact. Those interested in participating in this meeting by
teleconference should call or email the meeting contact as soon as
possible, but no later than May 28, 2010.
PARTICIPANTS: Participants from the NRC include members of the Office of Nuclear
Reactor Regulation (NRR), the Office of New Reactors (NRO), and the
Office of Nuclear Regulatory Research (RES).
NRC Industry
T. Collins, NRR B. Bradley, NEI
S. Dinsmore, NRR V. Anderson, NEI
R. Landry, NRO
Docket ID: NRC-2004-0006; RIN 3150-AH29
Enclosure:
Meeting Agenda
* Commission’s Policy Statement on “Enhancing Public Participation in NRC Meetings”
(67 FR 36920, May 28, 2002)
- 2 -
CATEGORY 3: * This is a Category 3 Meeting: The public is invited to participate in this
meeting by providing comments and asking questions throughout the
meeting.
The NRC provides reasonable accommodations to individuals with
disabilities where appropriate. If you need a reasonable accommodation
to participate in this meeting (e.g., sign language), or need this meeting
notice or other information from the meeting in another format, please
notify the NRC meeting contact by May 28, 2010, so that arrangements
can be made.
Teleconferencing: Interested members of the public unable to attend the
meeting may participate by telephone via a toll-free teleconference. For
details, please call the NRC meeting contact listed or the toll-free number,
1-800-368-5642, and ask the operator for a connection to the meeting
contact. Those interested in participating in this meeting by
teleconference should call or email the meeting contact as soon as
possible, but no later than May 28, 2010.
PARTICIPANTS: Participants from the NRC include members of the Office of Nuclear
Reactor Regulation (NRR), the Office of New Reactors (NRO), and the
Office of Nuclear Regulatory Research (RES).
NRC Industry
T. Collins, NRR B. Bradley, NEI
S. Dinsmore, NRR V. Anderson, NEI
R. Landry, NRO
Docket ID: NRC-2004-0006; RIN 3150-AH29
Enclosure:
Meeting Agenda
* Commission’s Policy Statement on “Enhancing Public Participation in NRC Meetings”
(67 FR 36920, May 28, 2002)
DISTRIBUTION:
See next page
ADAMS Accession No: ML101400381
OFFICE NRR/PFAB: PM NRR/PFAB: BC
NAME RDudley RCarlson
DATE 5/21 /10 5/21 /10
OFFICIAL RECORD COPY
Memo to Robert D. Carlson from Richard F. Dudley dated May , 2010
SUBJECT: FORTHCOMING CATEGORY 3 PUBLIC MEETING TO DISCUSS NRC STAFF
RESOLUTION OF PUBLIC COMMENTS ON SUPPLEMENTAL PROPOSED RULE ON RISKINFORMED
CHANGES TO LOSS-OF-COOLANT ACCIDENT TECHNICAL REQUIREMENTS (10
CFR 50.46a) AND DRAFT FINAL RULE LANGUAGE
DISTRIBUTION:
Public
PFAB r/f
PMNS
RidsNrrOd
RidsNrrDpr
RCarlson
RDudley
Enclosure
Public Meeting Agenda
10 CFR 50.46a – Risk-Informed ECCS Requirements
June 4, 2010 (8:30 am to 12:30 pm)
Two White Flint North Room T08 A01
Rockville, Md
Friday, June 4, 2010
Welcoming Remarks, Purpose and Meeting Conduct 8:30 am – 8:40 am
R. Dudley
Introductory Remarks 8:40 am – 8:50 am
Summary of Public Comment Resolution 8:50 am – 9:00 am
R. Dudley
Discussion of Public Comments and Rule Changes Related to 9:00 am – 10:30 am
Risk Assessment S. Dinsmore
Break
Response to Stakeholder Questions 10:45am – 11:15 am
R. Dudley/all
Additional Opportunity for Stakeholder Discussion 11:15 am – 12:15 pm
NRC Closing Remarks and Adjournment 12:15 pm – 12:30 pm

The telecom bridge no. for this meeting is 888-455-9746; pass code = 19442.
Richard Dudley 415-1115


1.
(80)
50.46a Draft Final Rule Language, "Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements".
ML101250271
2010-05-11
32

§ 50.46a DRAFT FINAL RULE LANGUAGE

Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements

(ADAMS Accession no. ML101250271)

3. In § 50.46, paragraph (a) is amended by adding an introductory paragraph and revising paragraph (a)(1)(i) to read as follows:

§ 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power plants.

(1)(i) The ECCS system must be designed so that its calculated cooling performance following postulated LOCAs conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated LOCAs of different sizes, locations,
and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must
include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II Required Documentation, sets forth the documentation requirements for each evaluation model. This section does not apply to a nuclear power reactor facility for which the certifications required under § 50.82(a)(1) have been submitted.

(1) ECCS evaluation f 1 or LOCAs involving breaks at or below the TBS. ECCS cooling performance at or below the TBS must be calculated in accordance with an evaluation model that meets the requirements of either section I to Appendix K of this part, or the following requirements, and must demonstrate that the acceptance criteria in paragraph (e)(3) of this section are satisfied. The evaluation model must be used for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs involving breaks at or below the TBS are analyzed. The evaluation
model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (e)(3) of this section, there is a high level of probability that the criteria would not be exceeded.

(3) Acceptance criteria for LOCAs involving breaks at or below the TBS. The following acceptance criteria must be used in determining the acceptability of ECCS cooling performance:

(i) Peak cladding temperature. The calculated maximum fuel element cladding temperature must not exceed 2200°F.

(ii) Maximum cladding oxidation. The calculated total oxidation of the cladding must not at any location exceed 0.17 times the total cladding thickness before oxidation. As used in this paragraph, total oxidation means the total thickness of cladding metal that would be locally converted to oxide if all the oxygen absorbed by and reacted with the cladding locally were converted to stoichiometric zirconium dioxide. If cladding rupture is calculated to occur, the inside surfaces of the cladding must be included in the oxidation, beginning at the calculated time of rupture. Cladding thickness before oxidation means the radial distance from inside to
outside the cladding, after any calculated rupture or swelling has occurred but before significant oxidation. Where the calculated conditions of transient pressure and temperature lead to a prediction of cladding swelling, with or without cladding rupture, the unoxidized cladding thickness must be defined as the cladding cross-sectional area, taken at a horizontal plane at
the elevation of the rupture, if it occurs, or at the elevation of the 1 highest cladding temperature if no rupture is calculated to occur, divided by the average circumference at that elevation. For ruptured cladding the circumference does not include the rupture opening.

(iii) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam must not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(iv) Coolable geometry. Calculated changes in core geometry must be such that the core remains amenable to cooling.

(v) Long term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature must be maintained at an acceptably low value and decay heat must be removed for the extended period of time required by the long-lived radioactivity remaining in the core.


APPENDIX A TO PART 50 -GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS

CRITERIA

Criterion 35--Emergency core cooling. A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.