[Federal Register Volume 72, Number 191 (Wednesday, October 3, 2007)]
[Proposed Rules]
[Pages 56275-56287]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 07-4887]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 72, No. 191 / Wednesday, October 3, 2007 /
Proposed Rules
[[Page 56275]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AI01
Alternate Fracture Toughness Requirements for Protection Against
Pressurized Thermal Shock Events
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend
its regulations to provide updated fracture toughness requirements for
protection against pressurized thermal shock (PTS) events for
pressurized water reactor (PWR) pressure vessels. The proposed rule
would provide new PTS requirements based on updated analysis methods.
This action is desirable because the existing requirements are based on
unnecessarily conservative probabilistic fracture mechanics analyses.
This action would reduce regulatory burden for licensees, specifically
those licensees that expect to exceed the existing requirements before
the expiration of their licenses, while maintaining adequate safety.
These new requirements would be voluntarily utilized by any PWR
licensee as an alternative to complying with the existing requirements.
DATES: Submit comments by December 17, 2007. Submit comments specific
to the information collection aspects of this rule by November 2, 2007.
Comments received after these dates will be considered if it is
practical to do so, but assurance of consideration cannot be given to
comments received after these dates.
ADDRESSES: You may submit comments by any one of the following methods.
Please include the following number ``RIN 3150-AI01'' in the subject
line of your comments. Comments on rulemakings submitted in writing or
in electronic form will be made available for public inspection.
Because your comment will not be edited to remove any identifying or
contact information, the NRC cautions you against including any
information in your submission that you do not want to be publicly
disclosed.
Submit comments via the Federal e-Rulemaking Portal http://www.regulations.gov. Mail comments to: Secretary, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and
Adjudications Staff.
E-mail comments to: [email protected]. If you do not receive a reply e-
mail confirming that we have received your comments, contact us
directly at (301) 415-1966. Address questions about our rulemaking Web
site to Carol Gallagher (301) 415-5905; E-mail [email protected].
Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland
20852, between 7:30 a.m. and 4:15 p.m. Federal workdays (telephone
(301) 415-1966).
Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at
(301) 415-1101.
You may submit comments on the information collections by the
methods indicated in the Paperwork Reduction Act Statement.
Publicly available documents related to this rulemaking may be
viewed electronically on the public computers located at the NRC's
Public Document Room (PDR), O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, MD 20852-2738. The PDR reproduction
contractor will copy documents for a fee.
Publicly available documents created or received at the NRC after
November 1, 1999, are available electronically at the NRC's Electronic
Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this
site, the public can gain entry into the NRC's Agencywide Document
Access and Management System (ADAMS), which provides text and image
files of NRC's public documents. If you do not have access to ADAMS or
if there are problems in accessing the documents located in ADAMS,
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737, or
by e-mail to [email protected].
FOR FURTHER INFORMATION CONTACT: Mr. George Tartal, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone (301) 415-0016; e-mail: [email protected], or Mr.
Barry Elliot, Office of Nuclear Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 415-
2709; e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Background
II. Section-by-Section Analysis
III. Agreement State Compatibility
IV. Availability of Documents
V. Plain Language
VI. Voluntary Consensus Standards
VII. Finding of No Significant Environmental Impact: Environmental
Assessment
VIII.Paperwork Reduction Act Statement
IX. Regulatory Analysis
X. Regulatory Flexibility Certification
XI. Backfit Analysis
I. Background
Pressurized thermal shock events are system transients in a
pressurized water reactor (PWR) in which severe overcooling occurs
coincident with high pressure. The thermal stresses caused by rapid
cooling of the reactor vessel inside surface combine with the stresses
caused by high pressure. The aggregate effect of these stresses is an
increase in the potential for fracture if a preexisting flaw is present
in a material susceptible to brittle failure. The ferritic, low alloy
steel of the reactor vessel beltline adjacent to the core where neutron
radiation gradually embrittles the material over the lifetime of the
plant may be such a material.
The toughness of ferritic reactor vessel materials is characterized
by a ``reference temperature for nil ductility transition''
(RTNDT). RTNDT is referred to as a ductile-to-
brittle transition temperature. At temperatures below RTNDT
fracture occurs very rapidly, by cleavage, a behavior referred to as
``brittle.'' As temperatures increase above RTNDT,
progressively larger amounts of deformation occur before rapid cleavage
fracture occurs. Eventually, at temperatures above approximately
RTNDT + 60 [deg]F, there is no longer adequate stress
intensification to promote cleavage and fracture occurs by the slower
mechanism of micro-void initiation, growth, and coalescence into the
crack, a behavior referred to as ``ductile.''
At normal operating temperature, ferritic reactor vessel materials
are usually tough. However, neutron
[[Page 56276]]
radiation embrittles the material over time, causing a shift in
RTNDT to higher temperatures. Correlations based on test
results for unirradiated and irradiated specimens have been developed
to calculate the shift in RTNDT as a function of neutron
fluence (the integrated neutron flux over a specified time of plant
operation) for various material compositions. The value of RTNDT
at a given time in a reactor vessel's life is used in fracture
mechanics calculations to determine the probability that assumed pre-
existing flaws would propagate when the reactor vessel is stressed.
The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, adopted on
July 23, 1985 (50 FR 29937), establishes screening criteria below which
the potential for a reactor vessel to fail due to a PTS event is deemed
to be acceptably low. The screening criteria effectively define a
limiting level of embrittlement beyond which operation cannot continue
without further plant-specific evaluation. Regulatory Guide (RG) 1.154,
``Format and Content of Plant-Specific Pressurized Thermal Shock
Analysis Reports for Pressurized Water Reactors,'' indicates that
reactor vessels that exceed the screening criteria in the rule may
continue to operate provided they can demonstrate a mean through-wall
crack frequency (TWCF) from PTS-related events of no greater than 5 x
10-6 per reactor year.
Any reactor vessel with materials predicted to exceed the screening
criteria in 10 CFR 50.61 may not continue to operate without
implementation of compensatory actions or additional plant-specific
analyses unless the licensee receives an exemption from the
requirements of the rule. Acceptable compensatory actions are neutron
flux reduction, other plant modifications to reduce PTS event
probability or severity, and reactor vessel annealing, which are
addressed in 10 CFR 50.61(b)(3), (b)(4), and (b)(7); and 10 CFR 50.66,
respectively.
No currently operating PWR reactor vessel is projected to exceed
the 10 CFR 50.61 screening criteria before the expiration of its 40
year operating license. However, several PWR reactor vessels are
approaching the screening criteria, while others are likely to exceed
the screening criteria during their first license renewal periods.
Technical Basis for the Proposed Amendment
The NRC's Office of Nuclear Regulatory Research (RES) has completed
a research program to update the PTS regulations. The results of this
research program conclude that the risk of through-wall cracking due to
a PTS event is much lower than previously estimated. This finding
indicates that the screening criteria in 10 CFR 50.61 are unnecessarily
conservative and may impose an unnecessary burden on some licensees.
Therefore, the NRC is proposing a new rule, 10 CFR 50.61a, which would
provide alternative screening criteria and corresponding embrittlement
correlations based on the updated technical basis. The updated
embrittlement correlation is the projected increase in the Charpy V-
notch 30 ft-lb transition temperature for reactor vessel materials
resulting from neutron radiation and is calculated using equations 5
through 7 of the proposed rule. The proposed rule would be voluntary
for all holders of a PWR operating license under 10 CFR part 50 or a
combined license under 10 CFR part 52, although it is intended for
licensees with reactor vessels that cannot demonstrate compliance with
the more restrictive criteria in 10 CFR 50.61. The requirements of 10
CFR 50.61 would continue to apply to licensees who choose not to
implement 10 CFR 50.61a.
The following two reports provide the technical basis for this
rulemaking: (1) NUREG-1806, ``Technical Basis for Revision of the
Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR
50.61): Summary Report,'' and (2) NUREG-1874, ``Recommended Screening
Limits for Pressurized Thermal Shock (PTS).'' These reports summarize
and reference several additional reports on the same topic. The updated
technical basis indicates that, after 60 years of operation, the risk
of reactor vessel failure due to a PTS event is much lower than
previously estimated. The updated analyses were based on information
from three currently operating PWRs. Because the severity of the risk-
significant transient classes (i.e., primary side pipe breaks, stuck
open valves on the primary side that may later re-close) is controlled
by factors that are common to PWRs in general, the NRC concludes that
the TWCF results and resultant RT-based screening criteria developed
from their analysis of three plants can be applied with confidence to
the entire fleet of operating PWRs. This conclusion is based on an
understanding of characteristics of the dominant transients that drive
their risk significance and on an evaluation of a larger population of
high embrittlement PWRs. This evaluation revealed no design,
operational, training, or procedural factors that could credibly
increase either the severity of these transients or the frequency of
their occurrence in the general PWR population above the severity/
frequency characteristic of the three plants that were modeled in
detail.
The current guidance provided by Regulatory Guide 1.174, Revision
1, ``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis,''
for large early release frequency (LERF) was used to relate the PTS
screening criteria in 10 CFR 50.61a to an acceptable yearly limit of 1
x 10-6 per reactor year on reactor vessel TWCF. Although
many post-through-wall cracking accident progressions are expected to
lead only to core damage (which suggests a 1 x 10-5 events
per year limit on TWCF per Regulatory Guide 1.174), uncertainties in
the accident progression analysis led to the recommendation of adopting
the more conservative TWCF limit of 1 x 10-6 per reactor
year based on LERF.
The updated technical basis uses many different models and
parameters to estimate the yearly probability that a PWR will develop a
through-wall crack as a consequence of PTS loading. One of these models
is a revised embrittlement correlation that uses information on the
chemical composition and neutron exposure of low alloy steels in the
reactor vessel's beltline region to estimate the resistance to fracture
of these materials. Although the general trends of the embrittlement
models in 10 CFR 50.61 and the proposed rule are similar, the form of
the revised embrittlement correlation differs substantially from the
correlation in the existing 10 CFR 50.61. The correlation in 10 CFR
50.61a has been updated to more accurately represent the substantial
amount of reactor vessel surveillance data that has accumulated since
the embrittlement correlation was last revised during the 1980s.
This proposed rule would differ from the current rule in that it
would contain a requirement for licensees who choose to follow its
requirements to analyze the results from the American Society of
Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)
Section XI in service inspection volumetric examinations. This
requirement would be provided in paragraph (e) of the proposed rule.
The examinations and analyses would confirm that the flaw density and
size in the licensee's reactor vessel beltline are bounded by the flaw
density and size utilized in the technical basis. The technical basis
was developed using a flaw density, spatial distribution, and size
distribution determined from a small amount of experimental data, as
[[Page 56277]]
well as from physical models and expert elicitation. The experimental
data included 22,210 cubic inches of weld metal, 3845 cubic inches of
plate, and 1650 cubic inches of clad. The experimental data were
obtained from samples removed from reactor vessel materials from
cancelled plants (Shoreham and the Pressure Vessel Research Users
Facility (PVRUF) vessel). The NRC considers that the analysis of the
ASME Code inservice inspection volumetric examination is needed to
confirm that the flaw density and size distributions in the reactor
vessel to which the proposed rule may be applied are consistent with
those in the technical basis because the experimental data was obtained
from a limited number of reactor vessels.
Paragraph (g)(6)(ii)(c) of 10 CFR 50.55a requires licensees to
implement Supplements 4 and 6 in Appendix VIII to ASME BPV Code Section
XI after November 22, 2000. Supplement 4 contains qualification
requirements for the reactor vessel inservice inspection volume from
the clad-to-base metal interface to the inner 1.0 inch or 10 percent of
the vessel thickness, whichever is larger. Supplement 6 contains
qualification requirements for reactor vessel weld volumes other than
those near the clad-to-base metal interface.
The performance of inspectors who have gone through the Supplement
4 qualification process has been documented in a paper by Becker
(Becker, L., ``Reactor Pressure Vessel Inspection Reliability,''
Proceeding of the Joint EC-IAEA Technical Meeting on the Improvement in
In-Service Inspection Effectiveness, Petten, the Netherlands, November
2002). Analysis of the results reported in this paper indicates that an
inspector using a Supplement 4 qualification procedure would have an 80
percent probability of detecting a flaw with a through-wall extent of
0.1 inch and would have an approximately 99 percent probability of
detecting a flaw with a through-wall extent of 0.3 inch. Therefore,
there is an 80 percent or greater probability of detecting a flaw that
contributes to crack initiation from PTS events in reactor vessels with
embrittlement conditions characteristic of 1 x 10-6 per
reactor-year TWCF when they are inspected using ASME BPV Code Section
XI, Appendix VIII, Supplement 4 requirements.
The true flaw density for flaws with a through wall extent of
between 0.1 and 0.3 inch can be inferred from the ASME Code examination
results and the probability of detection. The proposed rule would
require licensees to determine if:
(1) The indication density and size within the weld and base metal
inservice inspection volume from the clad-to-base metal interface to
the inner 1.0 inch or 10 percent of the vessel thickness are within the
flaw density and size distributions that were used in the technical
basis represented in Tables 2 and 3 in the proposed rule;
(2) Any indications within the weld and base metal inservice
inspection volume from the clad-to-base metal interface to the inner
1.0 inch or 10 percent of the vessel thickness are larger than the
sizes in Tables 2 and 3;
(3) Any indications between the clad-to-base metal interface and
three-eights of the vessel thickness exceed the size allowable in ASME
BPV Code Section XI, Table IWB-3510-1; or
(4) Any linear indications that penetrate through the clad into the
welds or the adjacent base metal.
The technical basis for the proposed rule concludes that flaws as
small as 0.1 inch deep contribute to TWCF and that nearly all of the
contributions come from flaws in the range below 1 inch deep for
reactor vessels with embrittlement characteristics of TWCF equal to 1 x
10-6 per reactor year. The peak contribution comes from
flaws between 0.1 and 0.2 inch deep, because that is the range that has
the maximum combined effect from the number of flaws, which is
decreasing with flaw size, and their susceptibility to brittle
fracture, which is increasing with flaw size. For weld flaws that
exceed the sizes in the table, the risk analysis indicates that a
single flaw can be expected to contribute a significant fraction of the
1 x 10-6/reactor-year limit on TWCF. Therefore, if a flaw of
that size is found in a reactor vessel, it is important to more
accurately assess if its size and location with respect to the local
level of embrittlement challenge the regulatory limit.
The technical basis for the proposed rule indicates that flaws
buried deeper than 1 inch from the inner surface of the reactor vessel
are not as susceptible to brittle fracture as similar size flaws
located closer to the inner surface. Therefore, the proposed rule would
not require the comparison of the density of such flaws, but still
would require large flaws, if discovered, to be evaluated for
contributions to TWCF if they are within the inner three-eights of the
vessel thickness. This requirement would be provided in paragraph
(e)(4)(iv) of the proposed rule. The limitation for flaw acceptance,
specified in ASME Code Section XI Table IWB-3510-1, approximately
corresponds to the threshold for flaw sizes that can make a significant
contribution to TWCF if present in reactor vessel material at this
depth. Therefore, this proposed rule would require these flaws to be
evaluated for contribution to TWCF in addition to the other evaluations
for such flaws that are prescribed in the ASME Code.
The numerical values in Tables 2 and 3 of the proposed rule would
represent the number of flaws in each size range that were derived from
the technical basis. Table 2 for the weld flaws is limited to flaw
sizes that are frequent enough to be expected to occur in most plants.
Similarly, Table 3 for the plate and forging flaws stops at the maximum
flaw size that was modeled for these materials in the technical basis.
If one or more larger flaws are found in a reactor vessel, they must be
evaluated to ensure that they are not causing the TWCF for that reactor
vessel to exceed the regulatory limit.
Surface cracks that penetrate through the stainless steel clad into
the welds or the adjacent base metal were not included in the technical
basis because these types of flaws have not been observed in the
beltline of an operating PWR reactor vessel. However, flaws of this
type were observed in the Quad Cities Unit 2 reactor vessel head in
1990 (NUREG-1796, ``Safety Evaluation Report related to the License
Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad
Cities Nuclear Power Station, Units 1 and 2''). The observed cracks had
a maximum depth into the base metal of approximately 6 mm (0.24 inch)
and penetrated through the stainless steel clad. Quad Cities Units 2
and 3 are boiling water reactors which are not susceptible to PTS
events and hence are not subject to 10 CFR 50.61. The cracking at Quad
Cities Unit 2 was attributed to intergranular stress corrosion cracking
(IGSCC) of the stainless steel cladding, which has not been observed in
PWR reactor vessels, and hot cracking of the low alloy steel metal
base. If these cracks were in the beltline region of a PWR, they would
be a significant contributor to TWCF because of their size and
location. The proposed rule would require licensees to determine if
cracks of this type exist in the beltline weld region at each ASME Code
Section XI ultrasonic examination. This requirement would be provided
in paragraph (e)(2) of the proposed rule.
Development of Tables 2 and 3 Flaw Density and Size Screening Criteria
The ASME Code specifies that the dimension of flaws detected by
nondestructive examination be
[[Page 56278]]
expressed to the nearest 0.05 inch for indications less than 1 inch.
Hence, the examination results from the ASME Code volumetric
examination will be reported in multiples of 0.05 inch with a range of
0.025 inch. Therefore, Tables 2 and 3 in the proposed rule
describe the flaw density in multiples of 0.05 inch with a size range
of 0.025 inch.
The ASME Code standard for reporting flaw sizes did not match the
size increments in the technical basis. Therefore, the NRC staff
developed a procedure to distribute the flaws used in the technical
basis into ASME Code-sized ranges. This is explained in greater detail
in the NRC staff document ``Development of Flaw Size Distribution
Tables for Draft Proposed Title 10 of the Code of Federal Regulations
(10 CFR) 50.61a'' (refer to ADAMS accession number ML070950392).
The values in Tables 2 and 3 of the proposed rule exceed the values
for those size ranges that were developed from the laboratory analyses
of the two reactor vessels. It was decided to allow licensees to use
the Table 2 and 3 values instead of the values that would come from the
laboratory results because it is still conservative to model all of the
flaws as if they were the largest size for each of the ASME Code size
ranges. In effect, some of the conservatism that was in the original
risk modeling is being made available to licensees for demonstrating
that the results of an individual plant's ASME Code examinations are
consistent with the underlying technical basis.
Rulemaking Initiation
In SECY-06-0124, dated May 26, 2006, the NRC staff presented a
rulemaking plan to the Commission to amend fracture toughness
requirements for PWRs. In this SECY paper, the NRC staff proposed four
options for rulemaking. The NRC staff recommended Option 3, which would
allow licensees to voluntarily implement the less restrictive screening
limits based on the updated technical basis and insert the updated
embrittlement correlation into 10 CFR 50.61 to maintain regulatory
consistency and implement the best state-of-the-art embrittlement
correlation in both 10 CFR 50.61 and 10 CFR 50.61a. This recommendation
was based on providing the necessary relief to licensees that would
otherwise expend considerable resources to justify continued plant
operation beyond the screening criteria in 10 CFR 50.61 (via
compensatory actions, plant-specific analyses, annealing or exemption),
while also requiring all licensees to recalculate their embrittlement
metric to ensure that all plants' analyses are consistent.
In a Staff Requirements Memorandum (SRM) dated June 30, 2006, the
Commission approved the initiation of the rulemaking as specified in
Option 2 of the rulemaking plan. This option would require licensees to
continue to meet the requirements of 10 CFR 50.61, which provides
adequate protection against PTS events, without implementing the
updated embrittlement correlation. For licensees whose reactor vessels
do not meet the requirements of 10 CFR 50.61, Option 2 would allow
licensees to voluntarily implement 10 CFR 50.61a which utilizes the
less restrictive screening limits based on the updated technical basis
as well as the updated embrittlement correlation. Accordingly, the
proposed rule provides for a voluntary alternative to the current set
of PTS requirements for any PWR licensee. The NRC considered requiring
new plants to use the best available embrittlement correlation (i.e.,
the embrittlement correlation developed for the new rule). The NRC
believes that such a requirement was not necessary to provide adequate
protection of public health and safety. The NRC believes that imposing
the existing 10 CFR 50.61, without modification, on new reactors would
ensure that adequate protection concerns would be met. The NRC believes
that the proposed rule's requirements should be a voluntary alternative
available to new plants, if needed.
In implementing the rulemaking plan, the proposed rule would
provide a new section, 10 CFR 50.61a, for the new set of fracture
toughness requirements. The NRC decided that providing a new section
containing the updated screening criteria and updated embrittlement
correlations would be appropriate because the Commission directed the
NRC staff to prepare a rulemaking which would allow current PWR
licensees to implement the new requirements of 10 CFR 50.61a or
continue to comply with the current requirements of 10 CFR 50.61.
Alternatively, the NRC could have revised 10 CFR 50.61 to include the
new requirements, which could be implemented as an alternative to the
current requirements. However, providing two sets of requirements
within the same regulatory section was considered confusing and/or
ambiguous as to which requirements apply to which licensees. The
proposed rule would provide a voluntary alternative to the current
rule, which further prompted the NRC to keep the current, mandatory
requirements separate from the new, voluntarily-implemented
requirements. As a result, the proposed new rule would retain the
current requirements in 10 CFR 50.61 for PWR licensees choosing not to
implement the less restrictive screening limits, and would present new
requirements in 10 CFR 50.61a as a voluntary relaxation for any PWR
licensee.
II. Section-by-Section Analysis
Section 50.61--Fracture Toughness Requirements for Protection Against
Pressurized Thermal Shock Events
Section 50.61 contains the current requirements for pressurized
thermal shock screening limits and embrittlement correlations.
Paragraph (b) of this section would be modified to reference the
proposed new section, Sec. 50.61a, as a voluntary alternative to
compliance with the requirements of Sec. 50.61. No changes are made to
the current pressurized thermal shock screening criteria, embrittlement
correlations, or any other related requirements in this section.
Section 50.61a--Alternate Fracture Toughness Requirements for
Protection Against Pressurized Thermal Shock Events
Proposed new Sec. 50.61a would contain pressurized thermal shock
screening limits based on updated probabilistic fracture mechanics
analyses. This new section would provide similar requirements to that
of Sec. 50.61, fracture toughness requirements for protection against
pressurized thermal shock events for pressurized water nuclear power
reactors. However, Sec. 50.61a would differ extensively in how the
licensee determines the resistance to fractures initiating from
different flaws at different locations in the vessel beltline, as well
as in the fracture toughness screening criteria. The proposed rule
would require quantifying PTS reference temperatures (RTMAX-
X) for flaws along axial weld fusion lines, plates, forgings, and
circumferential weld fusion lines, and comparing the quantified value
against the RTMAX-X screening criteria. Although comparing
quantified values to the screening criteria is also required by the
current Sec. 50.61, the proposed Sec. 50.61a would provide screening
criteria that vary depending on material product form and vessel wall
thickness. Further, the embrittlement correlation and the method of
calculation of RTMAX-X values in Sec. 50.61a would differ
significantly from that in Sec. 50.61 as described in the technical
basis for this rule. The new embrittlement correlation was developed
using multivariable
[[Page 56279]]
surface-fitting techniques based on pattern recognition, understanding
of mechanisms, and engineering judgement. The embrittlement database
used for this analysis was derived primarily from the Power Reactor
Embrittlement Data Base (PR-EDB) developed at Oak Ridge National
Laboratory. The updated RTMAX-X estimation procedures
provide a more realistic (compared to the existing regulation) method
for estimating the fracture toughness of reactor vessel materials over
the lifetime of the plant.
Paragraph (a) would contain definitions for terms used in Sec.
50.61a. It would also provide that terms defined in Sec. 50.61 also
have the same meaning in Sec. 50.61a unless otherwise noted.
Paragraph (b) would describe the applicability of Sec. 50.61a to
PWRs as an alternative to the requirements of Sec. 50.61. The
requirements of this section would provide a voluntarily-implemented
alternative to the current requirements of Sec. 50.61 for any current
PWR licensee or future holder of a PWR operating license or combined
license.
Paragraph (c) would set forth the requirements governing NRC
approval of a licensee's use of Sec. 50.61a. The licensee would make
the formal request to the NRC via a license amendment, and only upon
approval of the license amendment by the NRC would a licensee be
permitted to implement Sec. 50.61a. In the licensee's amendment
request, the required information would include (a) calculating the
values of RTMAX-X values as required by paragraph (c)(1),
(b) examining and assessing flaws discovered by ASME Code inspections
as required by paragraph (c)(2), and (c) comparing the RTMAX-X
values against the applicable screening criteria as required by
paragraph (c)(3). In doing so, the licensee would also be required to
utilize paragraphs (e)(1) through (e)(3), paragraph (f), and paragraph
(g) in order to perform the necessary calculations, comparisons,
examinations, assessments, and analyses.
Paragraph (d) would define the requirements for subsequent
examinations and flaw assessments after initial approval to use Sec.
50.61a has been obtained under the requirements of paragraph (c). It
would also define the required compensatory measures or analyses to be
taken if a licensee determines that the screening criteria will be
exceeded. Paragraph (d)(1) would define the requirements for subsequent
RTMAX-X assessments consistent with the requirements of
paragraphs (c)(1) and (c)(3). Paragraph (d)(2) would define the
requirements for subsequent examination and flaw assessments utilizing
the requirements of paragraphs (e)(1), (e)(1)(i), (e)(1)(ii), (e)(2),
and (e)(3). Paragraphs (d)(3) through (d)(7) would define the
requirements for implementing compensatory measures or plant-specific
analyses should the value of RTMAX-X be projected to exceed
the PTS screening criteria in Table 1 of this section.
Paragraph (e) would define the requirements for verifying that the
PTS screening criteria in Sec. 50.61a are applicable to a particular
reactor vessel. The proposed rule would require that verification be
based on an analysis of test results from ultrasonic examination of the
reactor vessel beltline materials required by Section XI of the ASME
Code.
Paragraph (e)(1) would establish cumulative limits on flaw density
and size within the ASME Code, Section XI, Appendix VIII, Supplement 4
inspection volume, which corresponds to a depth of approximately one
inch from the clad-to-base metal interface. The allowable number of
flaws provided in Tables 2 and 3 are cumulative values. If flaws exist
in larger increments, the allowable number of flaws is the value in
Table 2 or 3 for that increment minus the total number of flaws in all
larger increments. Flaws in this inspection volume contribute
approximately 97-99 percent to the TWCF at the screening limit.
Paragraph (e)(1)(i) would describe the flaw density limits for
welds.
Paragraph (e)(1)(ii) would describe the flaw density limits for
plates and forgings.
Paragraph (e)(1)(iii) would describe the specific ultrasonic
examination and neutron fluence information to be submitted to the NRC.
The NRC would utilize this information to evaluate whether plant-
specific information gathered in accordance with this rule suggests
that the NRC staff should generically re-examine the technical basis
for the rule.
Paragraph (e)(2) would require that licensees verify that no clad-
base metal interface flaws within the ASME Code, Section XI, Appendix
VIII, Supplement 4 inspection volume open to the vessel inside surface.
These types of flaws could have a substantial effect on the TWCF.
Paragraph (e)(3) would establish limits on flaw density and size
beyond the ASME Code, Section XI, Appendix VIII, Supplement 4
inspection volume to three-eights of the reactor vessel thickness from
the interior surface. Flaws in this inspection volume contribute
approximately 1-3 percent to the TWCF at the screening criteria. Flaws
exceeding this limit could affect the TWCF. Flaws greater than three-
eights of the reactor vessel thickness from the interior surface do not
contribute to the TWCF at the screening limit.
Paragraph (e)(4) would establish requirements to be met if flaws
exceed the limits in (e)(1) and (e)(3) or open to the inside surface of
the reactor vessel. This section requires an analysis to demonstrate
the reactor vessel would have a TWCF of less than 1 x 10-6
per reactor-year. The analysis could be a complete, plant-specific,
probabilistic fracture mechanics analysis or could be a simplified
analysis of flaw size, location and embrittlement to demonstrate that
the actual flaws in the reactor vessel are not in locations that would
cause the TWCF to be greater than 1 x 10-6 per reactor-year.
This paragraph would be required to be implemented if the requirements
of (e)(1) through (e)(3) are not satisfied.
Paragraph (e)(5) would describe the critical parameters to be
addressed if flaws exceed the limits in (e)(1) and (e)(3) or if the
flaws would open to the inside surface of the reactor vessel. This
paragraph would be required to be implemented if the requirements of
(e)(1) through (e)(3) are not satisfied.
Paragraph (f) would define the process for calculating RTMAX-X
values. These values would be based on the vessel's copper, manganese,
phosphorus, and nickel weight percentages, reactor cold leg
temperature, and neutron flux and fluence values, as well as the
unirradiated RTNDT of the product form in question.
Paragraph (g) would provide the necessary equations and variables
required by paragraph (f) of this section.
Table 1 would provide the PTS screening criteria for comparison
with the licensee's calculated RTMAX-X values. Tables 2 and
3 would provide values to be used in paragraph (e) of this section.
Tables 4 and 5 would provide values to be used in paragraph (f) of this
section.
III. Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement States Programs,'' approved by the Commission on June 20,
1997, and published in the Federal Register (62 FR 46517; September 3,
1997), this rule is classified as compatibility category ``NRC.''
Agreement State Compatibility is not required for Category ``NRC''
regulations. The NRC program elements in this category are those that
relate directly to areas of regulation reserved to the NRC by the
[[Page 56280]]
Atomic Energy Act or the provisions of Title 10 of the Code of Federal
Regulations (10 CFR). Although an Agreement State may not adopt program
elements reserved to NRC, it may wish to inform its licensees of
certain requirements via a mechanism that is consistent with the
particular State's administrative procedure laws, but does not confer
regulatory authority on the State.
IV. Availability of Documents
The following table lists documents relating to this rulemaking
which are available to the public and how they may be obtained.
Public Document Room (PDR). The NRC's Public Document Room is
located at the NRC's headquarters at 11555 Rockville Pike, Rockville,
MD 20852.
NRC's Electronic Reading Room (ERR). The NRC's electronic reading
room is located at http://www.nrc.gov/reading-rm.html.
----------------------------------------------------------------------------------------------------------------
Document PDR Web ERR (ADAMS)
----------------------------------------------------------------------------------------------------------------
Regulatory Analysis.......................... X X ML070570383
OMB Supporting Statement..................... X X ML070570446
SECY-06-0124, May 26, 2006, Rulemaking Plan X ............ ML060530624
Request for Commission Approval.
SRM-SECY-06-0124, June 30, 2006, Staff X ............ ML061810148
Requirements--Commission Approval of
Rulemaking Plan.
NUREG-1796, ``Safety Evaluation Report X ............ ML043060581
Related to the License Renewal of the
Dresden Nuclear Power Station, Units 2 and 3
and Quad Cities Nuclear Power Station, Units
1 and 2''.
NUREG-1806, ``Technical Basis for Revision of X ............ ML061580318
the Pressurized Thermal Shock (PTS)
Screening Limits in the PTS Rule (10 CFR
50.61): Summary Report''.
NUREG-1874, ``Recommended Screening Limits X ............ ML070860156
for Pressurized Thermal Shock (PTS)''.
Regulatory Guide 1.154, ``Format and Content X ............ ML003740028
of Plant-Specific Pressurized Thermal Shock
Analysis Reports for Pressurized Water
Reactors''.
Regulatory Guide 1.174, ``An Approach for X ............ ML023240437
Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes
to the Licensing Basis''.
Memorandum from Elliot to Mitchell, dated X ............ ML070950392
April 3, 2007, ``Development of Flaw Size
Distribution Tables for Draft Proposed Title
10 of the Code of Federal Regulations (10
CFR) 50.61a''.
----------------------------------------------------------------------------------------------------------------
V. Plain Language
The Presidential memorandum dated June 1, 1998, entitled ``Plain
Language in Government Writing'' directed that the Government's writing
be in plain language. This memorandum was published on June 10, 1998
(63 FR 31883). The NRC requests comments on the proposed rule
specifically with respect to the clarity and effectiveness of the
language used. Comments should be sent to the address listed under the
ADDRESSES caption of the preamble of this document.
VI. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995, Pub.
L. 104-113, requires that Federal agencies use technical standards that
are developed or adopted by voluntary consensus standards bodies unless
using such a standard is inconsistent with applicable law or is
otherwise impractical.
The NRC considered using American Society for Testing and Materials
(ASTM) standard E-900, ``Standard Guide for Predicting Radiation-
Induced Temperature Transition Shift in Reactor Vessel Materials. This
standard contains a different embrittlement correlation than that of
this proposed rule. However, the correlation developed by RES has been
more recently calibrated to available data. As a result, ASTM standard
E-900 is not a practical candidate for application in the technical
basis for the proposed rule because it does not represent the broad
range of conditions necessary to justify a revision to the regulations.
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code requirements are utilized as part of the volumetric
examination analysis requirements of the proposed rule. ASTM Standard
Practice E 185, ``Standard Practice for Conducting Surveillance Tests
for Light-Water Cooled Nuclear Power Reactor Vessels'' is incorporated
by reference in 10 CFR 50 Appendix H and utilized to determine 30-foot-
pound transition temperatures. These standards were selected for use in
the proposed rule based on their use in other regulations within Part
50 and their applicability to the subject of the desired requirements.
The NRC will consider using other voluntary consensus standards if
appropriate standards are identified.
VII. Finding of No Significant Environmental Impact: Environmental
Assessment
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR part 51, that this rule, if adopted, would not be a
major Federal action significantly affecting the quality of the human
environment and, therefore, an environmental impact statement is not
required. The basis for this determination is as follows:
Environmental Impacts of the Action
This environmental assessment focuses on those aspects of Sec.
50.61a where there is a potential for an environmental impact. The NRC
has concluded that there will be no significant radiological
environmental impacts associated with implementation of the rule
requirements for the following reasons:
(1) Section 50.61a would maintain the same functional requirements
for the facility as the existing PTS rule in Sec. 50.61 as a voluntary
alternative to the existing rule. This proposed rule would establish
screening criteria, limiting levels of embrittlement beyond which
operation cannot continue without further plant-specific evaluation or
modifications, as well as require calculation of the maximum
embrittlement predicted at the end of the licensed period of operation.
The screening criteria provide reasonable assurance that licensees
operating below (predicted embrittlement less than) the screening
criteria could endure a pressurized thermal shock event without
fracture of vessel materials, thus assuring integrity of the reactor
pressure vessel.
(2) The new rule is risk-informed and in accordance with the NRC's
1995 PRA policy statement and risk-informed regulation guidance.
Sufficient safety margins are maintained to ensure that any potential
increases in core damage frequency (CDF) and large early release
frequency (LERF) resulting from
[[Page 56281]]
implementation of Sec. 50.61a are negligible.
The action will not significantly increase the probability or
consequences of accidents, result in changes being made in the types of
any effluents that may be released off site, or result in a significant
increase in occupational or public radiation exposure. Therefore, there
are no significant radiological environmental impacts associated with
this action.
With regard to potential nonradiological impacts, implementation of
the rule requirements has no impact on the facility other than to
provide a more realistic method of calculating PWR vessel fracture
toughness with associated limits. Nonradiological plant effluents are
not affected and there are no other environmental impacts. Therefore,
the NRC concludes that there are no significant environmental impacts
associated with the action.
Alternatives to the Action
As an alternative to the rulemaking described above, the NRC
considered not taking the action (i.e., the ``no-action'' alternative).
Not adopting the more realistic and less conservative regulation would
result in no change in environmental impacts for current PWRs or those
that would be expected for future PWRs under 10 CFR 50.61.
Agencies and Persons Consulted
The NRC staff developed the proposed rule and this environmental
assessment. Under the NRC's stated policy, a copy of this environmental
assessment will be provided to the state liaison officials as part of
the publication of the proposed rule for public comment.
Conclusion
On the basis of this environmental assessment, the NRC concludes
that the action would not have a significant effect on the quality of
the human environment. Accordingly, the NRC has determined not to
prepare an environmental impact statement for the action.
The determination of this environmental assessment is that no
significant offsite impact to the public from this action would occur.
However, the general public should note that the NRC is seeking public
participation. Comments on any aspect of the environmental assessment
may be submitted to the NRC as indicated under the ADDRESSES heading.
The NRC has sent a copy of this proposed rule to every State
Liaison Officer and requested their comments on the environmental
assessment.
VIII. Paperwork Reduction Act Statement
This proposed rule would contain new or amended information
collection requirements that are subject to the Paperwork Reduction Act
of 1995 (44 U.S.C. 3501, et seq.). This proposed rule has been
submitted to the Office of Management and Budget for review and
approval of the information collection requirements.
Type of submission, new or revision: Revision.
The title of the information collection: 10 CFR part 50,
``Alternate Fracture Toughness Requirements for Protection against
Pressurized Thermal Shock Events (10 CFR 60.61 and 50.61a)'' proposed
rule.
The form number if applicable: Not applicable.
How often the collection is required: Collections would be
initially required for PWR licensees utilizing the requirements of 10
CFR 50.61a as a voluntary alternative to the requirements of 10 CFR
50.61. Collections would also be required, after voluntary
implementation of the new Sec. 50.61a, when any change is made to the
design or operation of the facility that affects the calculated
RTMAX-X value. Collections would also be required during the
scheduled periodic ultrasonic examination of beltline welds.
Who will be required or asked to report: Any PWR licensee
voluntarily utilizing the requirements of 10 CFR 50.61a in lieu of the
requirements of 10 CFR 50.61 would be subject to all of the proposed
requirements in this rulemaking.
An estimate of the number of annual responses: 2.
The estimated number of annual respondents: 1.
An estimate of the total number of hours needed annually to
complete the requirement or request: 264 hours (24 hours annually for
recordkeeping plus 240 hours annually for reporting).
Abstract: The NRC is proposing to amend its regulations to provide
updated fracture toughness requirements for protection against
pressurized thermal shock (PTS) events for pressurized water reactor
(PWR) pressure vessels. The proposed rule would provide new PTS
requirements based on updated analysis methods. This action is
necessary because the existing requirements are based on unnecessarily
conservative probabilistic fracture mechanics analyses. This action
would reduce regulatory burden for licensees, specifically those
licensees that expect to exceed the existing requirements before the
expiration of their licenses. These new requirements would be
voluntarily utilized by any PWR licensee as an alternative to complying
with the existing requirements.
The U.S. Nuclear Regulatory Commission is seeking public comment on
the potential impact of the information collections contained in this
proposed rule and on the following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
2. Estimate of burden?
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
4. How can the burden of the information collection be minimized,
including the use of automated collection techniques?
A copy of the OMB clearance package may be viewed free of charge at
the NRC Public Document Room, One White Flint North, 11555 Rockville
Pike, Room O-1 F21, Rockville, MD 20852. The OMB clearance package and
rule are available at the NRC worldwide Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html for 60 days after the
signature date of this notice.
Send comments on any aspect of these proposed information
collections, including suggestions for reducing the burden and on the
above issues, by November 2, 2007 to the Records and FOIA/Privacy
Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by Internet electronic mail to
[email protected] and to the Desk Officer, Office of Information and
Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and
Budget, Washington, DC 20503. Comments received after this date will be
considered if it is practical to do so, but assurance of consideration
cannot be given to comments received after this date. You may also
comment by telephone at (202) 395-3087.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
[[Page 56282]]
IX. Regulatory Analysis
The Commission has prepared a draft regulatory analysis on this
proposed regulation. The analysis examines the costs and benefits of
the alternatives considered by the Commission. The Commission requests
public comments on this draft regulatory analysis. Availability of the
regulatory analysis is provided in Section IV. Comments on the draft
regulatory analysis may be submitted to the NRC as indicated under the
ADDRESSES heading of this document.
In addition, the Commission also requests public comments on the
cost and benefit of requiring PWR licensees to revise their vessel
analyses if the updated embrittlement correlation were imposed in 10
CFR 50.61. This would differ from the proposed rule, which leaves the
technical content of 10 CFR 50.61 unchanged.
X. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the
Commission certifies that this rule would not, if promulgated, have a
significant economic impact on a substantial number of small entities.
This proposed rule would affect only the licensing and operation of
nuclear power plants. The companies that own these plants do not fall
within the scope of the definition of ``small entities'' set forth in
the Regulatory Flexibility Act or the size standards established by the
NRC (10 CFR 2.810).
XI. Backfit Analysis
The NRC has determined that the requirements in this proposed rule
do not constitute backfitting as defined in 10 CFR 50.109(a)(1).
Therefore, a backfit analysis has not been prepared for this proposed
rule.
The requirements of the current PTS rule, 10 CFR 50.61, would
continue to apply to all PWR licensees, and would not change as a
result of this proposed rule. The requirements of the proposed PTS
rule, 10 CFR 50.61a, would not be required, but could be voluntarily
utilized, by any PWR licensee. Licensees choosing to implement the
proposed PTS rule would be required to comply with its requirements as
a voluntary alternative to complying with the requirements of the
current PTS rule. Because the proposed PTS rule would not be mandatory
for any PWR licensee, but rather could be voluntarily implemented by
any PWR licensee, the NRC finds that this amendment would not
constitute backfitting.
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing to
adopt the following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note). Section 50.7 also
issued under Pub. L. 95-601, sec. 10, 92 Stat. 2951 (42 U.S.C.
5841). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955,
as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-190, 83
Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 50.103
also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C.
2138).
Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec.
185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and
Appendix Q also issued under sec. 102, Pub. L. 91-190, 83 Stat. 853
(42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec.
204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and
50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C.
2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42
U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68
Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under
sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
2. In Sec. 50.61, paragraph (b)(1) is revised to read as follows:
Sec. 50.61 Fracture toughness requirements for protection against
pressurized thermal shock events.
* * * * *
(b) Requirements. (1) For each pressurized water nuclear power
reactor for which an operating license has been issued under this part
or a combined license issued under Part 52 of this chapter, other than
a nuclear power reactor facility for which the certifications required
under Sec. 50.82(a)(1) have been submitted, the licensee shall have
projected values of RTPTS or RTMAX-X, accepted by
the NRC, for each reactor vessel beltline material for the EOL fluence
of the material in accordance with this section or Sec. 50.61a. For a
licensee choosing to comply with this section, the assessment of
RTPTS must use the calculation procedures given in paragraph
(c)(1) of this section, except as provided in paragraphs (c)(2) and
(c)(3) of this section. The assessment must specify the bases for the
projected value of RTPTS for each vessel beltline material,
including the assumptions regarding core loading patterns, and must
specify the copper and nickel contents and the fluence value used in
the calculation for each beltline material. This assessment must be
updated whenever there is a significant \2\ change in projected values
of RTPTS, or upon request for a change in the expiration
date for operation of the facility.
* * * * *
3. Section 50.61a is added to read as follows:
Sec. 50.61a Alternate fracture toughness requirements for protection
against pressurized thermal shock events.
(a) Definitions. Terms in this section have the same meaning as
those set forth in 10 CFR 50.61(a), with the exception of the term
``ASME Code''.
(1) ASME Code means the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code, Section III, Division I, ``Rules for
the Construction of Nuclear Power Plant Components,'' and Section XI,
Division I, ``Rules for Inservice Inspection of Nuclear Power Plant
Components,'' edition and addenda and any limitations and modifications
thereof as specified in Sec. 50.55a.
(2) RTMAX AW means the material property which characterizes the
reactor vessel's resistance to fracture initiating from flaws found
along axial weld fusion lines. RTMAX-AW is determined under
the provisions of paragraph (f) of this section and has units of
[deg]F.
(3) RTMAX PL means the material property which characterizes the
reactor vessel's resistance to fracture initiating from flaws found in
plates in regions that are not associated with welds found in plates.
RTMAX-PL is determined under the provisions of paragraph (f)
of this section and has units of [deg]F.
(4) RTMAX FO means the material property which characterizes the
reactor vessel's resistance to fracture initiating from flaws in
forgings that are not associated with welds found in forgings.
RTMAX-FO is determined under the provisions of paragraph (f)
of this section and has units of [deg]F.
[[Page 56283]]
(5) RTMAX CW means the material property which characterizes the
reactor vessel's resistance to fracture initiating from flaws found
along the circumferential weld fusion lines. RTMAX-CW is
determined under the provisions of paragraph (f) of this section and
has units of [deg]F.
(6) RTMAX X means any or all of the material properties RTMAX-
AW, RTMAX-PL, RTMAX-FO, or RTMAX-CW
for a particular reactor vessel.
(7) [phis]t means fast neutron fluence for neutrons with energies
greater than 1.0 MeV. [phis]t is determined under the provisions of
paragraph (g) of this section and has units of n/cm\2\.
(8) [phis] means average neutron flux. [phis] is determined under
the provisions of paragraph (g) of this section and has units of n/
cm\2\/sec.
(9) [Delta]T30 means the shift in the Charpy V-notch transition
temperature produced by irradiation defined at the 30 ft-lb energy
level. The [Delta]T30 value is determined under the provisions of
paragraph (g) of this section and has units of [deg]F.
(10) Surveillance data means any data that demonstrates the
embrittlement trends for the beltline materials, including, but not
limited to, data from test reactors or surveillance programs at other
plants with or without a surveillance program integrated under 10 CFR
part 50, Appendix H.
(11) TC means cold leg temperature under normal full power
operating conditions, as a time-weighted average from the start of full
power operation through the end of licensed operation. TC
has units of [deg]F.
(b) Applicability. Each holder of an operating license under this
part or holder of a combined license under part 52 of this chapter of a
pressurized water nuclear power reactor may utilize the requirements of
this section as an alternative to the requirements of 10 CFR 50.61.
(c) Request for Approval. Prior to implementation of this section,
each licensee shall submit a request for approval in the form of a
license amendment together with the documentation required by
paragraphs (c)(1), (c)(2), and (c)(3) of this section for review and
approval to the Director, Office of Nuclear Reactor Regulation
(Director). The information required by paragraphs (c)(1), (c)(2), and
(c)(3) of this section must be submitted for review and approval by the
Director at least three years before the limiting RTPTS
value calculated under 10 CFR 50.61 is projected to exceed the PTS
screening criteria in 10 CFR 50.61 for plants licensed under 10 CFR
part 50 or 10 CFR part 52.
(1) Each licensee shall have projected values of RTMAX-X
for each reactor vessel beltline material for the EOL fluence of the
material. The assessment of RTMAX-X values must use the
calculation procedures given in paragraphs (f) and (g) of this section,
except as provided in paragraphs (f)(6) and (f)(7) of this section. The
assessment must specify the bases for the projected value of
RTMAX-X for each reactor vessel beltline material, including
the assumptions regarding future plant operation (e.g., core loading
patterns, projected capacity factors, etc.); the copper (Cu),
phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor
cold leg temperature (TC); and the neutron flux and fluence
values used in the calculation for each beltline material.
(2) Each licensee shall perform an examination and an assessment of
flaws in the reactor vessel beltline as required by paragraph (e) of
this section. The licensee shall verify that the requirements of
paragraphs (e)(1) through (e)(3) have been met and submit all
documented indications and the neutron fluence map required by
paragraph (e)(1)(iii) to the Director in its application to utilize 10
CFR 50.61a. If analyses performed under paragraph (e)(4) of this
section are used to justify continued operation of the facility,
approval by the Director is required prior to implementation.
(3) Each licensee shall compare the projected RTMAX-X
values for plates, forgings, axial welds, and circumferential welds to
the PTS screening criteria for the purpose of evaluating a reactor
vessel's susceptibility to fracture due to a PTS event. If any of the
projected RTMAX-X values are greater than the PTS screening
criteria in Table 1 of this section, then the licensee may propose the
compensatory actions or plant-specific analyses as required in
paragraphs (d)(3) through (d)(7) of this section, as applicable, to
justify operation beyond the PTS screening criteria in Table 1 of this
section.
(d) Subsequent Requirements. Licensees who have been approved to
utilize 10 CFR 50.61a under the requirements of paragraph (c) of this
section shall comply with the requirements of this paragraph.
(1) Whenever there is a significant change in projected values of
RTMAX-X, such that the previous value, the current value, or
both values, exceed the screening criteria prior to the expiration of
the plant operating license; or upon the licensee's request for a
change in the expiration date for operation of the facility; a re-
assessment of RTMAX-X values documented consistent with the
requirements of paragraph (c)(1) and (c)(3) of this section must be
submitted for review and approval to the Director. If the Director does
not approve the assessment of RTMAX-X values, then the
licensee shall perform the actions required in paragraphs (d)(3)
through (d)(7) of this section, as necessary, prior to operation beyond
the PTS screening criteria in Table 1 of this section.
(2) Licensees shall determine the impact of the subsequent flaw
assessments required by paragraphs (e)(1)(i), (e)(1)(ii), (e)(2), and
(e)(3) of this section and shall submit the assessment for review and
approval to the Director within 120 days after completing a volumetric
examination of reactor vessel beltline materials as required by Section
XI of the ASME Code. If a licensee is required to implement paragraphs
(e)(4) and (e)(5) of this section, a re-analysis in accordance with
paragraphs (e)(4) and (e)(5) of this section is required within one
year of the subsequent ASME Code inspection.
(3) If the value of RTMAX-X is projected to exceed the
PTS screening criteria, then the licensee shall implement those flux
reduction programs that are reasonably practicable to avoid exceeding
the PTS screening criteria. The schedule for implementation of flux
reduction measures may take into account the schedule for review and
anticipated approval by the Director of detailed plant-specific
analyses which demonstrate acceptable risk with RTMAX-X
values above the PTS screening criteria due to plant modifications, new
information, or new analysis techniques.
(4) If the analysis required by paragraph (d)(3) of this section
indicates that no reasonably practicable flux reduction program will
prevent the RTMAX-X value for one or more reactor vessel
beltline materials from exceeding the PTS screening criteria, then the
licensee shall perform a safety analysis to determine what, if any,
modifications to equipment, systems, and operation are necessary to
prevent the potential for an unacceptably high probability of failure
of the reactor vessel as a result of postulated PTS events if continued
operation beyond the PTS screening criteria is to be allowed. In the
analysis, the licensee may determine the properties of the reactor
vessel materials based on available information, research results and
plant surveillance data, and may use probabilistic fracture mechanics
techniques. This analysis must be submitted to the Director at least
three years before RTMAX-X is
[[Page 56284]]
projected to exceed the PTS screening criteria.
(5) After consideration of the licensee's analyses, including
effects of proposed corrective actions, if any, submitted under
paragraphs (d)(3) and (d)(4) of this section, the Director may, on a
case-by-case basis, approve operation of the facility with RTMAX-X
values in excess of the PTS screening criteria. The Director will
consider factors significantly affecting the potential for failure of
the reactor vessel in reaching a decision.
(6) If the Director concludes, under paragraph (d)(5) of this
section, that operation of the facility with RTMAX-X values
in excess of the PTS screening criteria cannot be approved on the basis
of the licensee's analyses submitted under paragraphs (d)(3) and (d)(4)
of this section, then the licensee shall request a license amendment,
and receive approval by the Director, prior to any operation beyond the
PTS screening criteria. The request must be based on modifications to
equipment, systems, and operation of the facility in addition to those
previously proposed in the submitted analyses that would reduce the
potential for failure of the reactor vessel due to PTS events, or on
further analyses based on new information or improved methodology.
(7) If the limiting RTMAX-X value of the facility is
projected to exceed the PTS screening criteria and the requirements of
paragraphs (d)(3) through (d)(6) of this section cannot be satisfied,
the reactor vessel beltline may be given a thermal annealing treatment
under the requirements of Sec. 50.66 to recover the fracture toughness
of the material. The reactor vessel may be used only for that service
period within which the predicted fracture toughness of the reactor
vessel beltline materials satisfy the requirements of paragraphs (d)(1)
through (d)(6) of this section, with RTMAX-X values
accounting for the effects of annealing and subsequent irradiation.
(e) Examination and Flaw Assessment Requirements. The volumetric
examinations results evaluated under paragraphs (e)(1), (e)(2), and
(e)(3) of this section must be acquired using procedures, equipment and
personnel that have been qualified under the ASME Code, Section XI,
Appendix VIII, Supplement 4 and Supplement 6.
(1) The licensee shall verify that the indication density and size
distributions within the ASME Code, Section XI, Appendix VIII,
Supplement 4 inspection volume \1\ are within the flaw density and size
distributions in Tables 2 and 3 of this section based on the test
results from the volumetric examination. The allowable number of flaws
specified in Tables 2 and 3 of this section represent a cumulative flaw
size distribution for each ASME flaw size increment. The allowable
number of flaws for a particular ASME flaw size increment represents
the maximum total number of flaws in that and all larger ASME flaw size
increments. The licensee shall also demonstrate that no flaw exceeds
the size limitations specified in Tables 2 and 3 of this section.
---------------------------------------------------------------------------
\1\ The ASME Code, Section XI, Appendix VIII, Supplement 4 weld
volume is the weld volume from the clad-to-base metal interface to
the inner 1.0 inch or 10 percent of the vessel thickness, whichever
is greater.
---------------------------------------------------------------------------
(i) The licensee shall determine the allowable number of weld flaws
for the reactor vessel beltline by multiplying the values in Table 2 of
this section by the total length of the reactor vessel beltline welds
that were volumetrically inspected and dividing by 1000 inches of weld
length.
(ii) The licensee shall determine the allowable number of plate or
forging flaws for their reactor vessel beltline by multiplying the
values in Table 3 of this section by the total plate or forging surface
area that was volumetrically inspected in the beltline plates or
forgings and dividing by 1000 square inches.
(iii) For each indication detected in the ASME Code, Section XI,
Appendix VIII, Supplement 4 inspection volume, the licensee shall
document the dimensions of the indication, including depth and length,
the orientation of the indication relative to the axial direction, and
the location within the reactor vessel, including its azimuthal and
axial positions and its depth embedded from the clad-to-base metal
interface. The licensee shall also document a neutron fluence map,
projected to the date of license expiration, for the reactor vessel
beltline clad-to-base metal interface and indexed in a manner that
allows the determination of the neutron fluence at the location of the
detected indications.
(2) The licensee shall identify, as part of the examination
required by paragraph (c)(2) of this section and any subsequent ASME
Code, Section XI ultrasonic examination of the beltline welds, any
indications within the ASME Code, Section XI, Appendix VIII, Supplement
4 inspection volume that are located at the clad-to-base metal
interface. The licensee shall verify that such indications do not open
to the vessel inside surface using a qualified surface or visual
examination.
(3) The licensee shall verify, as part of the examination required
by paragraph (c)(2) of this section and any subsequent ASME Code,
Section XI ultrasonic examination of the beltline welds, all
indications between the clad-to-base metal interface and three-eights
of the reactor vessel thickness from the interior surface are within
the allowable values in ASME Code, Section XI, Table IWB-3510-1.
(4) The licensee shall perform analyses to demonstrate that the
reactor vessel will have a through-wall crack frequency (TWCF) of less
than 1x10-6 per reactor-year if the ASME Code, Section XI
volumetric examination required by paragraph (c)(2) or (d)(2) of this
section indicates any of the following:
(i) The indication density and size in the ASME Code, Section XI,
Appendix VIII, Supplement 4 inspection volume is not within the flaw
density and size limitations specified in Tables 2 and 3 of this
section;
(ii) Any indication in the ASME Code, Section XI, Appendix VIII,
Supplement 4 inspection volume that is larger \2\ than the sizes in
Tables 2 and 3 of this section;
---------------------------------------------------------------------------
\2\ Table 2 for the weld flaws is limited to flaw sizes that are
expected to occur and were modeled from the technical basis
supporting this rule. Similarly, Table 3 for the plate and forging
flaws stops at the maximum flaw size modeled for these materials in
the technical basis supporting this rule.
---------------------------------------------------------------------------
(iii) There are linear indications that penetrate through the clad
into the low alloy steel reactor vessel shell; or
(iv) Any indications between the clad-to-base metal interface and
three-eights \3\ of the vessel thickness exceed the size allowable in
ASME Code, Section XI, Table IWB-3510-1.
---------------------------------------------------------------------------
\3\ Because flaws greater than three-eights of the vessel wall
thickness from the inside surface do not contribute to TWCF, flaws
greater than three-eights of the vessel wall thickness from the
inside surface need not be analyzed for their contribution to PTS.
---------------------------------------------------------------------------
(5) The analyses required by paragraph (e)(4) of this section must
address the effects on TWCF of the known sizes and locations of all
indications detected by the ASME Code, Section XI, Appendix VIII,
Supplement 4 and Supplement 6 ultrasonic examination out to three-
eights of the vessel thickness from the inner surface, and may also
take into account other reactor vessel-specific information, including
fracture toughness information.
(f) Calculation of RTMAX-X values. Each licensee shall
calculate RTMAX-X values for each reactor vessel beltline
material using [phi]t. [phi]t must be calculated using an NRC-approved
methodology.
(1) The values of RTMAX-AW, RTMAX-PL,
RTMAX-FO, and RTMAX-CW must be
[[Page 56285]]
determined using Equations 1 through 4 of this section.
(2) The values of [Delta]T30 must be determined using
Equations 5 through 7 of this section, unless the conditions specified
in paragraph (f)(6)(iv) of this section are met, for each axial weld
fusion line, plate, and circumferential weld fusion line. The
[Delta]T30 value for each axial weld fusion line calculated
as specified by Equation 1 of this section must be calculated for the
maximum fluence ([phi]FL) occurring along a particular axial
weld fusion line. The [Delta]T30 value for each plate
calculated as specified by Equation 1 of this section must be
calculated for tFL occurring along a particular axial weld fusion line.
The [Delta]T30 value for each plate or forging calculated as
specified by Equations 2 and 3 of this section are calculated for the
maximum fluence ([phi]tMAX) occurring at the clad-to-base
metal interface of each plate or forging. In Equation 4, the
[phi]tFL value used for calculating the plate, forging, and
circumferential weld RTMAX-CW value is the maximum [phi]
occurring for each material along the circumferential weld fusion line.
(3) The values of Cu, Mn, P, and Ni in Equations 6 and 7 of this
section must represent the best estimate values for the material weight
percentages. For a plate or forging, the best estimate value is
normally the mean of the measured values for that plate or forging. For
a weld, the best estimate value is normally the mean of the measured
values for a weld deposit made using the same weld wire heat number as
the critical vessel weld. If these values are not available, either the
upper limiting values given in the material specifications to which the
vessel material was fabricated, or conservative estimates (mean plus
one standard deviation) based on generic data \4\ as shown in Table 4
of this section for P and Mn, must be used.
---------------------------------------------------------------------------
\4\ Data from reactor vessels fabricated to the same material
specification in the same shop as the vessel in question and in the
same time period is an example of ``generic data.''
---------------------------------------------------------------------------
(4) The values of RTNDT(u) must be evaluated according
to the procedures in the ASME Code, Section III, paragraph NB-2331. If
any other method is used for this evaluation, the licensee shall submit
the proposed method for review and approval by the Director along with
the calculation of RTMAX-X values required in paragraph
(c)(1) of this section.
(i) If a measured value of RTNDT(u) is not available, a
generic mean value of RTNDT(u) for the class \5\ of material
must be used if there are sufficient test results to establish a mean.
---------------------------------------------------------------------------
\5\ The class of material for estimating RTNDT(u) must be
determined by the type of welding flux (Linde 80, or other) for
welds or by the material specification for base metal.
---------------------------------------------------------------------------
(ii) The following generic mean values of RTNDT(u) must
be used unless justification for different values is provided: 0 [deg]F
for welds made with Linde 80 weld flux; and -56 [deg]F for welds made
with Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes.
(5) The value of Tc in Equation 6 of this section must
represent the weighted time average of the reactor cold leg temperature
under normal operating full power conditions from the beginning of full
power operation through the end of licensed operation.
(6) The licensee shall verify that an appropriate RTMAX-X
value has been calculated for each reactor vessel beltline material.
The licensee shall consider plant-specific information that could
affect the use of Equations 5 though 7 of this section for the
determination of a material's [Delta]T30 value.
(i) The licensee shall evaluate the results from a plant-specific
or integrated surveillance program if the surveillance data has been
deemed consistent as judged by the following criteria:
(A) The surveillance material must be a heat-specific match for one
or more of the materials for which RTMAX-X is being
calculated. The 30-foot-pound transition temperature must be determined
as specified by the requirements of 10 CFR 50 Appendix H.
(B) If three or more surveillance data points exist for a specific
material, the surveillance data must be evaluated for consistency with
the model in Equations 5, 6, and 7 as specified by paragraph (f)(6)(ii)
of this section. If fewer than three surveillance data points exist for
a specific material, then Equations 5, 6, and 7 of this section must be
used without performing the consistency check.
(ii) The licensee shall estimate the mean deviation from the model
(Equations 5, 6 and 7 of this section) for the specific data set (i.e.,
a group of surveillance data points representative of a given
material). The mean deviation from the model for a given data set must
be calculated using Equations 8 and 9 of this section. The mean
deviation for the data set must be compared to the maximum heat-average
residual given in Table 5 or Equation 10 of this section and based on
the material group into which the surveillance material falls and the
number of available data points. The licensee shall determine, based on
this comparison, if the surveillance data show a significantly
different trend than the model predicts. The surveillance data analysis
must follow the criteria in paragraphs (f)(6)(iii) through (f)(6)(iv)
of this section. For surveillance data sets with greater than 8 shift
points, the maximum credible heat-average residual must be calculated
using Equation 10 of this section. The value of [sigma] used in
Equation 10 of this section must comply with Table 5 of this section.
(iii) If the mean deviation from the model for the data set is
equal to or less than the value in Table 5 or the value using Equation
10 of this section, then the [Delta]T30 value must be
determined using Equations 5, 6, and 7 of this section.
(iv) If the mean deviation from the model for the data set is
greater than the value in Table 5 or the value using Equation 10 of
this section, the [Delta]T30 value must be determined using
the surveillance data. If the mean deviation from the model for the
data set is outside the limits specified in Equation 10 of this section
or in Table 5 of this section, the licensee shall review the data base
for that heat in detail, including all parameters used in Equations 4,
5, and 6 of this section and the data used to determine the baseline
Charpy V-notch curve for the material in an unirradiated condition. The
licensee shall submit an evaluation of the surveillance data and its
[Delta]T30 and RTMAX-X values for review and
approval by the Director no later than one year after the surveillance
capsule is withdrawn from the reactor vessel.
(7) The licensee shall report any information that significantly
improves the accuracy of the RTMAX-X value to the Director.
Any value of RTMAX-X that has been modified as specified in
paragraph (f)(6)(iv) of this section is subject to the approval of the
Director when used as provided in this section.
(g) Equations and variables used in this section.
Equation 1: RTMAX-AW = MAX {[RTNDT(u)-plate +
[Delta]T30-plate([phi]tFL)],
[RTNDT(u)-axialweld +
[Delta]T30-axialweld([phi]tFL)]{time}
Equation 2: RTMAX-PL = RTNDT(u)-plate +
[Delta]T30-plate([phi]tMAX)
Equation 3: RTMAX-FO = RTNDT(u)-forging +
[Delta]T30-forging([phi]tMAX)
Equation 4: RTMAX-CW = MAX {[RTNDT(u)-plate +
[Delta]T30-plate([phi]tMAX)],
[RTNDT(u)-circweld +
[Delta]T30-circweld([phi]tMAX)],
[RTNDT(u)-forging +
[Delta]T30-forging([phi]tMAX)]{time}
Equation 5: [Delta]T30 = MD + CRP
Equation 6: MD = A [middot] (1 - 0.001718 [middot] TC)
[middot] (1 + 6.13 [middot] P [middot] Mn2.471) [middot]
[phi]te0.5
Equation 7: CRP = B [middot] (1 + 3.77 [middot] Ni1.191)
[middot] f(Cue,P) [middot]
g(Cue,Ni,[phi]te) VVVVVVV
[[Page 56286]]
Where:
P [wt-%] = phosphorus content
Mn [wt-%] = manganese content
Ni [wt-%] = nickel content
Cu [wt-%] = copper content
A = 1.140 x 10-7 for forgings
= 1.561 x 10-7 for plates
= 1.417 x 10-7 for welds
B = 102.3 for forgings
= 102.5 for plates in non-Combustion Engineering manufactured
vessels
= 135.2 for plates in Combustion Engineering vessels
= 155.0 for welds
[phi]te = [phi]t for [phi] greater than or equal to 4.39 x
1010 n/cm2/sec
= [phi]t [middot] (4.39 x 1010/[phi])0.2595
for [phi] less than 4.39 x 1010 n/cm2/sec
Where:
[phi] [n/cm2/sec] = average neutron flux
t [sec] = time that the reactor has been in full power operation
[phi]t [n/cm2] = [phi] [middot] t
f(Cue,P) = 0 for Cu <= 0.072
= [Cue - 0.072]0.668 for Cu > 0.072 and P
<= 0.008
= [Cue - 0.072 + 1.359 [middot] (P-
0.008)]0.668 for Cu > 0.072 and P > 0.008
and Cue = 0 for Cu <= 0.072
= MIN (Cu, maximum Cue) for Cu > 0.072
and maximum Cue = 0.243 for Linde 80 welds
= 0.301 for all other materials
g(Cue,Ni,[phi]te) = 0.5 + 0.5 [middot]
tanh{[log10([phi]te) + 1.1390 [middot]
Cue - 0.448 [middot] Ni - 18.120] / 0.629{time}
Equation 8: Residual [reg] = measured [Delta]T30 - predicted
[Delta]T30 (by Equations 5, 6, and 7)
Equation 9: Mean deviation for a data set of n data points =
[GRAPHIC] [TIFF OMITTED] TP03OC07.013
Equation 10: Maximum credible heat-average residual = 3[sigma]/
n0.5
Where:
n = number of surveillance shift data points (sample size) in the
specific data set
[sigma] = standard deviation of the residuals about the model for a
relevant material group given in Table 5.
Table 1.--PTS Screening Criteria
----------------------------------------------------------------------------------------------------------------
RT MAX-X limits [[deg]F] for different vessel wall
thicknesses \6\ (TWALL)
Product form and RT MAX-Values --------------------------------------------------------
9.5 in. < TWALL 10.5 in. < TWALL
TWALL <= 9.5 in. <= 10.5 in. <= 11.5 in.
----------------------------------------------------------------------------------------------------------------
Axial Weld RTMAX-AW.................................... 269 230 222
Plate RTMAX-PL......................................... 356 305 293
Forging without underclad cracks RTMAX-FO.............. 356 305 293
Axial Weld and Plate RTMAX-AW + RTMAX-PL............... 538 476 445
Circumferential Weld RTMAX-CW \7\...................... 312 277 269
Forging with underclad cracks RTMAX-FO................. 246 241 239
----------------------------------------------------------------------------------------------------------------
Table 2.--Allowable Number of Flaws in Welds
----------------------------------------------------------------------------------------------------------------
Allowable number of cumulative
flaws per 1000 inches of weld
ASME section XI flaw size per IWA-3200 Range of through-wall extent length in the ASME section XI
(TWE) of flaw (in.) appendix VIII supplement 4
inspection volume
----------------------------------------------------------------------------------------------------------------
0.05........................................ 0.025 <= TWE < 0.075.......... Unlimited
0.10........................................ 0.075 <= TWE < 0.125.......... 166.70
0.15........................................ 0.125 <= TWE < 0.175.......... 90.80
0.20........................................ 0.175 <= TWE < 0.225.......... 22.82
0.25........................................ 0.225 <= TWE < 0.275.......... 8.66
0.30........................................ 0.275 <= TWE < 0.325.......... 4.01
0.35........................................ 0.325 <= TWE < 0.375.......... 3.01
0.40........................................ 0.375 <= TWE < 0.425.......... 1.49
0.45........................................ 0.425 <= TWE < 0.475.......... 1.00
----------------------------------------------------------------------------------------------------------------
Table 3.--Allowable Number of Flaws in Plates or Forging
----------------------------------------------------------------------------------------------------------------
Allowable number of cumulative
flaws per 1000 square inches of
Range of through-wall extent inside diameter surface area in
ASME section XI flaw size per IWA-3200 (TWE) of flaw (in.) forgings or plates in the ASME
section XI appendix VIII
supplement 4 inspection volume \8\
----------------------------------------------------------------------------------------------------------------
0.05........................................ 0.025 <= TWE < 0.075.......... Unlimited
0.10........................................ 0.075 <= TWE < 0.125.......... 8.049
0.15........................................ 0.125 <= TWE < 0.175.......... 3.146
0.20........................................ 0.175 <= TWE < 0.225.......... 0.853
0.25........................................ 0.225 <= TWE < 0.275.......... 0.293
0.30........................................ 0.275 <= TWE < 0.325.......... 0.0756
0.35........................................ 0.325 <= TWE < 0.375.......... 0.0144
----------------------------------------------------------------------------------------------------------------
Table 4.--Conservative Estimates for Chemical Element Weight Percentages
------------------------------------------------------------------------
Materials P Mn
------------------------------------------------------------------------
Plates.......................... 0.014 1.45
Forgings........................ 0.016 1.11
[[Page 56287]]
Welds........................... 0.019 1.63
------------------------------------------------------------------------
Table 5.--Maximum Heat-Average Residual [[deg]F] for Relevant Material Groups by Number of Available Data Points
----------------------------------------------------------------------------------------------------------------
Number of available data points
Material group [sigma] -----------------------------------------------------
[[deg]F] 3 4 5 6 7 8
----------------------------------------------------------------------------------------------------------------
Welds, for Cu > 0.072........................... 26.4 45.7 39.6 35.4 32.3 29.9 28.0
Plates, for Cu > 0.072.......................... 21.2 36.7 31.8 28.4 26.0 24.0 22.5
Forgings, for Cu > 0.072........................ 19.6 33.9 29.4 26.3 24.0 22.2 20.8
Weld, Plate or Forging, for Cu <= 0.072......... 18.6 32.2 27.9 25.0 22.8 21.1 19.7
----------------------------------------------------------------------------------------------------------------
Dated at Rockville, Maryland, this 27th day of September 2007.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
---------------------------------------------------------------------------
\6\ Wall thickness is the beltline wall thickness including the
clad thickness.
\7\ RTPTS limits contributes 1 x 10-8 per
reactor year to the reactor vessel TWCF.
\8\ Excluding underclad cracks in forgings.
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[FR Doc. 07-4887 Filed 10-2-07; 8:45 am]
BILLING CODE 7590-01-P