Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 46557-46569 [E8-18429]
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Proposed Rules
Federal Register
Vol. 73, No. 155
Monday, August 11, 2008
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.
NUCLEAR REGULATORY
COMMISSION
10 CFR Part 50
RIN 3150–AI01
[NRC–2007–0008]
Alternate Fracture Toughness
Requirements for Protection Against
Pressurized Thermal Shock Events
Nuclear Regulatory
Commission.
ACTION: Supplemental Proposed Rule.
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AGENCY:
SUMMARY: The Nuclear Regulatory
Commission (NRC) is considering the
adoption of provisions regarding
applicability of the rule and new
provisions regarding procedures to
perform surveillance data checks related
to the updated fracture toughness
requirements for protection against
pressurized thermal shock (PTS) events
for pressurized water reactor (PWR)
pressure vessels. The NRC is
considering these provisions as an
alternative to the provisions previously
noticed for public comment on October
3, 2007 (72 FR 56275).
DATES: Submit comments on this
proposed rule by September 10, 2008.
Submit comments on the information
collection aspects on this proposed rule
by September 10, 2008.
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 submitted in
writing or in electronic form will be
made available for public inspection.
Because your comments 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.
Federal e Rulemaking Portal: Go to
https://www.regulations.gov and search
for documents filed under Docket ID
NRC–2007–0008. Address questions
about NRC dockets to Carol Gallagher
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(301) 415–5905; e-mail
Carol.Gallager@nrc.gov.
Mail comments to: Secretary, U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001, ATTN:
Rulemakings and Adjudications Staff.
E-mail comments to:
Rulemaking.Comments@nrc.gov. If you
do not receive a reply e-mail confirming
that we have received your comments,
contact us directly at (301) 415–1966.
Hand deliver comments to: 11555
Rockville Pike, Rockville, Maryland
20852, between 7:30 am and 4:15 pm
during Federal workdays. (Telephone
(301) 415–1966).
Fax comments to: Secretary, U.S.
Nuclear Regulatory Commission at (301)
415–1101.
You can access publicly available
documents related to this document
using the following methods:
NRC’s Public Document Room (PDR):
The public may examine publicly
available documents at the NRC’s PDR,
Public File Area O–F21, One White
Flint North, 11555 Rockville Pike,
Rockville, Maryland. The PDR
reproduction contractor will copy
documents for a fee.
NRC’s Agencywide Document Access
and Management System (ADAMS):
Publicly available documents created or
received at the NRC are available
electronically at the NRC’s Electronic
Reading Room at https://www.nrc.gov/
reading-rm/adams.html. From this page,
the public can gain entry into 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’s
PDR reference staff at 1–800–397–4209,
or (301) 415–4737, or by e-mail to
PDR.Resource@nrc.gov.
Ms.
Veronica M. Rodriguez, Office of
Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001; telephone
(301) 415–3703; e-mail:
Veronica.Rodriguez@nrc.gov, Mr. Barry
Elliot, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001; telephone (301) 415–2709; e-mail:
Barry.Elliot@nrc.gov, or Mr. Mark Kirk,
Office of Nuclear Regulatory Research,
U.S. Nuclear Regulatory Commission,
Washington, DC 20555–0001; telephone
FOR FURTHER INFORMATION CONTACT:
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(301) 415–6015; e-mail:
Mark.Kirk@nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Introduction
II. Background
III. Discussion
IV. Responses to Comments on the Proposed
Rule
V. Section-by-Section Analysis
VI. Specific Request for Comments
VII. Availability of Documents
VIII. Plain Language
IX. Voluntary Consensus Standards
X. Finding of No Significant Environmental
Impact: Availability
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Act Certification
XIV. Backfit Analysis
I. Introduction
The NRC published a proposed rule
on alternate fracture toughness
requirements for protection against
Pressurized Thermal Shock (PTS) for
public comments in the Federal
Register on October 3, 2007 (72 FR
56275). This rule provides 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
utilized by any Pressurized Water
Reactor (PWR) licensee as an alternative
to complying with the existing
requirements.
During the development of the PTS
final rule, the NRC determined that
several changes to the proposed rule
language may be needed to adequately
address issues raised in stakeholder’s
comments. The NRC also determined, in
response to a stakeholder comment, that
the characteristics of advanced PWR
designs were not considered in the
technical analysis made for the
proposed rule. The NRC does not have
assurance that reactors that commence
commercial power operation after the
effective date of this rule will have
operating characteristics and materials
of fabrication similar to those evaluated
as part of the technical basis for the
proposed rule. Therefore, the NRC has
concluded that it would be prudent to
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limit the applicability and the use of
§ 50.61a to currently-operating plants
only, and proposes to modify the
applicability provisions of the proposed
rule accordingly.
Also, several stakeholders questioned
the accuracy and validity of the generic
embrittlement curves in the proposed
rule. The NRC wants to ensure that the
predicted values from the proposed
embrittlement trend curves provide an
adequate basis for implementation of
the rule. Therefore, the NRC has
continued to work on statistical
procedures to identify deviations from
generic embrittlement trends, such as
those described in § 50.61a(f)(6) of the
proposed rule. Based on this work, the
NRC is considering enhancing the
procedure described in paragraph
§ 50.61a(f)(6) to, among other things,
detect signs from the plant- and heatspecific surveillance data of
embrittlement trends that are not
reflected by Equations 5, 6 and 7 of the
rule that may emerge at high fluences.
Because these proposed modifications
may not represent a logical outgrowth
from the October 2007 proposed rule’s
provisions, the NRC concludes that
obtaining stakeholder feedback on the
proposed alternative provisions through
the use of a supplemental proposed rule
is appropriate. As discussed in Section
VI of this notice, the NRC will consider
comments on §§ 50.61a(b); (f)(6)(i)
through (f)(6)(vi); Equations 10, 11, and
12 in § 50.61a(g); and Tables 5, 6, and
7 of this supplemental proposed rule.
The NRC is also requesting comments
on whether there should be additional
language added to § 50.61a(e) to allow
licensees to account for the effects of
sizing errors. This supplemental
proposed rule does not reflect other
modifications or editorial and
conforming changes that the NRC is
considering to incorporate in the final
rule as a result of the public comments
on the October 2007 proposed rule.
II. Background
PTS events are system transients in a
PWR in which severe overcooling
occurs coincident with high pressure.
The thermal stresses are caused by rapid
cooling of the reactor vessel inside
surface, which combine with the
stresses caused by high pressure. The
aggregate effect of these stresses is an
increase in the potential for fracture if
a pre-existing 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, can be susceptible to brittle
fracture.
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The PTS rule, described in § 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 § 50.61 may
continue to operate provided they can
demonstrate a mean through-wall crack
frequency (TWCF) from PTS-related
events of no greater than 5 × 10¥6 per
reactor year.
Any reactor vessel with materials
predicted to exceed the screening
criteria in § 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, plant modifications to reduce
PTS event probability or severity, and
reactor vessel annealing, which are
addressed in §§ 50.61(b)(3), (b)(4), and
(b)(7); and § 50.66, ‘‘Requirements for
Thermal Annealing of the Reactor
Pressure Vessel.’’
Currently, no operating PWR reactor
vessel is projected to exceed the § 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.
The NRC’s Office of Nuclear
Regulatory Research (RES) developed a
technical basis that supports updating
the PTS regulations. This technical basis
concluded that the risk of through-wall
cracking due to a PTS event is much
lower than previously estimated. This
finding indicated that the screening
criteria in § 50.61 are unnecessarily
conservative and may impose an
unnecessary burden on some licensees.
Therefore, the NRC created a new rule,
§ 50.61a, which provides alternate
screening criteria and corresponding
embrittlement correlations based on the
updated technical basis. 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, in a Staff Requirements
Memorandum (SRM) dated June 30,
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2006, to prepare a rulemaking which
would allow current PWR licensees to
implement the new requirements of
§ 50.61a or continue to comply with the
current requirements of § 50.61.
Alternatively, the NRC could have
revised § 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 NRC published the proposed
rulemaking on the alternate fracture
toughness requirements for protection
against PTS for public comment in the
Federal Register on October 3, 2007 (72
FR 56275). The proposed rule provided
an alternative to the current rule, which
a licensee may choose to adopt. This
prompted the NRC to keep the current
requirements separate from the new
alternative requirements. As a result, the
proposed rule retained the current
requirements in § 50.61 for PWR
licensees choosing not to implement the
less restrictive screening limits, and
presented new requirements in § 50.61a
as an alternative relaxation for PWR
licensees.
III. Discussion
The NRC published a proposed new
rule, § 50.61a (October 3, 2007, 72 FR
56275), that would provide new PTS
requirements based on updated analysis
methods because the existing
requirements are based on unnecessarily
conservative probabilistic fracture
mechanics analyses. Stakeholders’
comments raised concerns related to the
applicability of the rule and the
accuracy and validity of the generic
embrittlement curves. The NRC
reconsidered the technical and
regulatory issues in these areas and is
considering adopting the modified
provisions regarding the applicability of
the rule and new provisions regarding
procedures to perform surveillance data
checks described in this supplemental
proposed rule. The NRC will consider
comments on §§ 50.61a(b), (f)(6)(i)
through (f)(6)(vi); Equations 10, 11 and
12 in § 50.61a(g); and Tables 5, 6, and
7 of this supplemental proposed rule.
As described in Section VI of this
notice, the NRC is also requesting
comments on whether there should be
additional language added to § 50.61a(e)
to allow licensees to account for the
effects of sizing errors. The NRC will
consider the October 2007 proposed
rule, the supplemental proposed rule,
and the comments received in response
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to both, when deciding whether to
adopt a final PTS rule.
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Applicability of the Proposed Rule,
§ 50.61a(b)
The supplemental proposed rule
differs from the proposed rule and from
§ 50.61 in that it proposes to limit the
use of § 50.61a to currently operating
plants only. It cannot be demonstrated,
a priori, that reactors which commence
commercial power operation after the
effective date of this rule will have
operating characteristics, in particular
identified PTS event sequences and
thermal-hydraulic responses, which are
consistent with the reactors which were
evaluated as part of the technical basis
for this rule. Other factors, including
materials of fabrication and welding
methods, could also vary. Hence, the
use of § 50.61a would be limited to
currently operating PWR facilities
which are known to have characteristics
consistent with those assumed in the
technical basis. The NRC also proposes
to allow the holder of the operating
license for Watts Bar Unit 2 to adopt the
requirements in § 50.61a as this facility
has operating characteristics consistent
with those assumed in the technical
basis. The NRC recognizes that licensees
for reactors who commence commercial
power operation after the effective date
of this rule may, under the provisions of
§ 50.12, seek an exemption from
§ 50.61a(b) to apply this rule if a plantspecific basis analyzing their operating
characteristics, materials of fabrications,
and welding methods is provided.
Surveillance Data, § 50.61a(f)
Section 50.61a(f) of the proposed rule
defines the process for calculating the
values for the material properties (i.e. ,
RTMAX–X) for a particular reactor vessel.
These values would be based on the
vessel material’s copper, manganese,
phosphorus, and nickel weight
percentages, reactor cold leg
temperature, and fast neutron flux and
fluence values, as well as the
unirradiated nil-ductility transition
reference temperature (i.e., RTNDT).
Section 50.61a(f) of the proposed rule
included a procedure by which the
RTMAX–X values, which are predicted for
plant-specific materials using a generic
temperature shift (i.e., DT30)
embrittlement trend curve, are
compared with heat-specific
surveillance data that are collected as
part of 10 CFR Part 50, Appendix H
surveillance programs. The purpose of
this comparison is to assess how well
the surveillance data are represented by
the generic embrittlement trend curve. If
the surveillance data are close
(closeness is assessed statistically) to the
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generic embrittlement trend curve, then
the predictions of this embrittlement
trend curve are used. This is expected
to normally be the case. However, if the
heat-specific surveillance data deviate
significantly, and non-conservatively,
from the predictions of the generic
embrittlement trend curve, this
indicates that alternative methods (i.e.,
other than, or in addition to, the generic
embrittlement trend curve) may be
needed to reliably predict the
temperature shift trends, and to estimate
RTMAX–X, for the conditions being
assessed. However, alternative methods
for temperature shift prediction are not
prescribed by § 50.61a(f) of the proposed
rule.
Although standard and accepted
procedures exist to assess the statistical
significance of the differences between
heat-specific surveillance data and the
generic embrittlement trend curve,
similarly standard and acceptable
procedures are not available to assess
the practical importance of such
differences. The practical importance of
statistically significant deviations is best
assessed by licensees on a case-by-case
basis, which would be submitted for the
review of the Director of NRR, as
prescribed by § 50.61a(f).
The method described in the
proposed rulemaking to compare the
heat-specific surveillance data collected
as part of 10 CFR part 50, Appendix H
surveillance programs to the generic
temperature shift embrittlement trend
curve included a single statistical test.
This statistical test was set forth by
Equations 9 and 10, and Table 5. This
test determined if, on average, the
temperature shift from the surveillance
data was significantly higher than the
temperature shift of the generic
embrittlement trend curve. The NRC has
determined that, while necessary, this
single test is not sufficient to ensure that
the temperature shift predicted by the
embrittlement trend curve well
represents the heat-specific surveillance
data. Specifically, this single statistical
test cannot determine if the temperature
shift from the surveillance data shows a
more rapid increase after significant
radiation exposure than the progression
predicted by the generic embrittlement
trend curve. To address this potential
deficiency, which could be particularly
important during a plant’s period of
extended operation, the NRC added two
more statistical tests in this
supplemental proposed rulemaking,
which are expressed by Equations 11
and 12 and by Tables 6 and 7. Together,
these two additional tests determine if
the surveillance data from a particular
heat show a more rapid increase after
significant radiation exposure than the
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progression predicted by the generic
embrittlement trend curve.
The NRC documented the technical
basis for the proposed alternative in the
following reports: (1) ‘‘Statistical
Procedures for Assessing Surveillance
Data for 10 CFR Part 50.61a,’’ (ADAMS
Accession No. ML081290654), and (2)
‘‘A Physically Based Correlation of
Irradiation Induced Transition
Temperature Shifts for RPV Steel,’’
(ADAMS Accession No. ML081000630).
IV. Responses to Comments on the
Proposed Rule
The NRC received 5 comment letters
on the proposed 10 CFR 50.61a rule
published on October 3, 2007 (72 FR
56275). The following paragraphs
discuss those comments which are
directly associated with the
supplemental proposed rule’s
provisions on the applicability of the
rule and surveillance data procedures.
The remainder of the comments and the
NRC responses will be provided in the
Federal Register notice for the final
rule.
Comments on the Applicability of the
Proposed Rule
Comment: The commenters stated
that the rule, as written, is only
applicable to the existing fleet of PWRs.
The characteristics of advanced PWR
designs were not considered in the
analysis. The commenters suggested
adding a statement to state that this rule
is applicable to the current PWR fleet
and not the new plant designs.
[PWROG–5, EPRI–5]
Response: The NRC agrees with the
comment that this rule is only
applicable to the existing fleet of PWRs.
The NRC cannot be assured that reactors
that commence commercial power
operation after the effective date of this
rule will have operating characteristics,
in particular identified PTS event
sequences and thermal-hydraulic
responses, which are consistent with the
reactors that were evaluated as part of
the technical basis for § 50.61a. Other
factors, including materials of
fabrication and welding methods, could
also vary. Therefore, the NRC agrees
with the commenters that it would be
prudent to restrict the use of § 50.61a to
current plants. As a result of this
comment, the NRC proposes to modify
§ 50.61a(b) and the statement of
considerations of the rule to reflect this
position to limit the use of the rule to
currently operating plants.
Comments on Surveillance Data
Comment: The commenters stated
that there is little added value in the
requirement to assess the surveillance
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data as a part of this rule because
variability in data has already been
accounted for in the derivation of the
embrittlement correlation.
The commenters also stated that there
is no viable methodology for adjusting
the projected DT30 for the vessel based
on the surveillance data. Any effort to
make this adjustment is likely to
introduce additional error into the
prediction. Note that the embrittlement
correlation described in the basis for the
revised PTS rule (i.e., NUREG–1874)
was derived using all of the currently
available industry-wide surveillance
data.
In the event that the surveillance data
does not match the DT30 value predicted
by the embrittlement correlation, the
best estimate value for the pressure
vessel material is derived using the
embrittlement correlation. The likely
source of the discrepancy is an error in
the characterization of the surveillance
material or of the irradiation
environment. Therefore, unless the
discrepancy can be resolved, obtaining
the DT30 prediction based on the best
estimate chemical composition for the
heat of the material is more reliable than
a prediction based on a single set of
surveillance measurements.
The commenters suggested removing
the requirement to assess surveillance
data, including Table 5, of this rule.
[PWROG–4, EPRI–4, NEI–2]
Response: The NRC does not agree
with the proposed change. The NRC
believes that there is added value in the
requirement to assess surveillance data.
Although variability has been accounted
for in the derivation of the
embrittlement correlation, it is the
NRC’s view that the surveillance
assessment required in § 50.61a(f)(6) is
needed to determine if the
embrittlement for a specific heat of
material in a reactor vessel is consistent
with the embrittlement predicted by the
embrittlement correlation.
The commenters also assert that there
is no viable methodology for adjusting
the projected DT30 for the vessel based
on the surveillance data, and that any
adjustment is likely to introduce
additional error into the prediction. The
NRC believes that although there is no
single methodology for adjusting the
projected DT30 for the vessel based on
the surveillance data, it is possible, on
a case-specific basis, to justify
adjustments to the generic DT30
prediction. For this reason the rule does
not specify a method for adjusting the
DT30 value based on surveillance data,
but rather requires the licensee to
propose a case-specific DT30 adjustment
procedure for review and approval from
the Director. Although the commenters
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assert that it is possible that error could
be introduced, it is the NRC view that
appropriate plant-specific adjustments
based upon available surveillance data
may be necessary to project reactor
pressure vessel embrittlement for the
purpose of this rule.
As the result of these public
comments, the NRC has continued to
work on statistical procedures to
identify deviations from generic
embrittlement trends, such as those
described in § 50.61a(f)(6) of the
proposed rule. Based on this work, the
NRC is considering further enhancing
the procedure described in paragraph
(f)(6) to, among other things, detect
signs from the plant- and heat-specific
surveillance data that may emerge at
high fluences of embrittlement trends
that are not reflected by Equations 5, 6,
and 7. The empirical basis for the NRC’s
concern regarding the potential for unmodeled high fluence effects is
described in documents located at
ADAMS Accession Nos. ML081120253,
ML081120289, ML081120365,
ML081120380, and ML081120600. The
technical basis for the enhanced
surveillance assessment procedure is
described in the document located at
ADAMS Accession No. ML081290654.
V. Section-by-Section Analysis
The following section-by-section
analysis only discusses the
modifications in the provisions related
to the applicability of the rule and
surveillance data procedures that the
NRC is considering as an alternative in
this supplemental proposed rule. The
NRC is only seeking comments on these
alternative provisions. This
supplemental proposed rule does not
reflect other modifications or editorial
and conforming changes that the NRC is
considering to incorporate as a result of
the public comments on the proposed
rule that were not discussed in this
notice as they will be provided in the
Federal Register notice for the final
rule.
would evaluate the surveillance for
consistency with the embrittlement
model by following the procedures
specified by §§ 50.61a(f)(6)(ii), (f)(6)(iii),
and (f)(6)(iv) of the supplemental
proposed rule.
Proposed § 50.61a(f)(6)(ii)
The proposed language for
§ 50.61a(f)(6)(ii) would establish the
requirements to perform an estimate of
the mean deviation of the data set from
the embrittlement model. The mean
deviation for the data set would be
compared to values given in Table 5 or
Equation 10 of this section. The NRC
proposes to modify this paragraph to
state that the surveillance data analysis
would follow the criteria in
§§ 50.61a(f)(6)(v) and (f)(6)(vi) of the
supplemental proposed rule.
Proposed § 50.61a(f)(6)(iii)
The NRC proposes to modify
§ 50.61a(f)(6)(iii) to establish the
requirements to estimate the slope of the
embrittlement model residuals (i.e., the
difference between the measured and
predicted value for a specific data
point). The licensee would estimate the
slope using Equation 11 and compare
this value to the maximum permissible
value in Table 6, both from the
supplemental proposed rule. This
surveillance data analysis would follow
the criteria in §§ 50.61a(f)(6)(v) and
(f)(6)(vi) of the supplemental proposed
rule.
Proposed § 50.61a(f)(6)(iv)
The NRC proposes to modify
§ 50.61a(f)(6)(iv) to establish the
requirements to estimate an outlier
deviation from the embrittlement model
for the specific data set using Equations
8 and 12. The licensee would compare
the normalized residuals to the
allowable values in Table 7 of the
supplemental proposed rule. This
surveillance data analysis would follow
the criteria in §§ 50.61a(f)(6)(v) and
(f)(6)(vi) of the supplemental proposed
rule.
Proposed § 50.61a(b)
The proposed language for § 50.61a(b)
would establish the applicability of the
rule. The NRC proposes to modify this
paragraph to limit the use of this rule to
currently-operating plants only.
Proposed § 50.61a(f)(6)(v)
The NRC proposes to add paragraph
(f)(6)(v) to establish the criteria to be
satisfied in order to calculate the DT30
shift values.
Proposed § 50.61a(f)(6)(i)
The proposed language for
§ 50.61a(f)(6)(i) would establish the
requirements to perform data checks to
determine if the surveillance data show
a significantly different trend than what
the embrittlement model in this rule
predicts. The NRC proposes to modify
§ 50.61a(f)(6)(i)(B) to state that licensees
Proposed § 50.61a(f)(6)(vi)
The NRC proposes to add paragraph
(f)(6)(vi) to establish the actions to be
taken by a licensee if the criteria in
paragraph (f)(6)(v) of this section are not
met. The licensee would need to submit
an evaluation of the surveillance data
and propose values for DT30,
considering their plant-specific
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surveillance data, for the review and
approval by the Director. The licensee
would need to submit an evaluation of
each surveillance capsule removed from
the vessel after the submittal of the
initial application for review and
approval by the Director no later than 2
years after the capsule is withdrawn
from the vessel.
Proposed § 50.61a(g)
The proposed language for § 50.61a(g)
would provide the necessary equations
and variables required by the proposed
changes in § 50.61a(f)(6). The NRC
proposes to modify Equation 10 to
account for 1 percent of significance
level. Equations 11 and 12 would be
added to provide the means for
estimating the slope and the outlier
deviation from the embrittlement
model.
Proposed Tables 5, 6, and 7
Tables 5, 6, and 7 would provide
values to be used in the proposed
changes in § 50.61a(f)(6). The NRC
proposes to modify Table 5 to account
for the use of a 1 percent of significance
level. Tables 6 and 7 would be added to
provide the threshold values for the
slope and the outlier deviation tests.
VI. Specific Request for Comments
The NRC seeks comments on
§§ 50.61a(b), (f)(6)(i) through (f)(6)(vi);
Equations 10, 11, and 12 in § 50.61a(g),
and Tables 5, 6, and 7 of the
supplemental proposed rule. The NRC
is not seeking comments on any other
provisions of the proposed § 50.61a
which remain unchanged from the
October 2007 proposed rule. In
addition, the NRC also requests
comments on the following question:
Adjustments of the Inservice Inspection
Volumetric Examination and Flaw
Assessments
The flaw sizes in Tables 2 and 3 are
selected so that reactor vessels with flaw
sizes less than or equal to those in the
tables will have a TWCF less than or
equal to 1 × 10¥6 per reactor year at the
maximum permissible embrittlement.
The NRC recognizes that the flaw sizes
in these tables represent actual flaw
dimensions while the results from the
ASME Code examinations are estimated
dimensions. The available information
indicates that, for most flaw sizes in
Tables 2 and 3, qualified inspectors will
oversize flaws. Comparing oversized
flaws to the size and density
distributions in Tables 2 and 3 is
conservative and acceptable, but not
necessary. Therefore, NRC is
considering to permit flaw sizes to be
adjusted to account for the effects of
sizing error before comparing the
estimated size and density distribution
to the acceptable size and density
distributions in Tables 2 and 3. This
would be accomplished by requiring
licensees to base the methodology to
account for the effects of sizing error on
statistical data collected from ASME
Code inspector qualification tests. An
acceptable method would include a
demonstration, that accounting for the
effects of sizing error, is unlikely to
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VII. Availability of Documents
The NRC is making the documents
identified below available to interested
persons through one or more of the
following methods, as indicated.
Public Document Room (PDR). The
NRC Public Document Room is located
at 11555 Rockville Pike, Rockville,
Maryland 20852.
Regulations.gov (Web). These
documents may be viewed and
downloaded electronically through the
Federal eRulemaking Portal https://
www.regulations.gov, Docket number
NRC–2007–0008.
NRC’s Electronic Reading Room
(ERR). The NRC’s public electronic
reading room is located at https://
www.nrc.gov/reading-rm.html.
PDR
Federal Register Notice—Proposed Rule: Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (RIN 3150–AI01), 72 FR 56275, October 3,
2007 .................................................................................................................................................
Letter from Thomas P. Harrall, Jr., dated December 17, 2007, ‘‘Comments on Proposed Rule 10
CFR 50, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, RIN 3150–AI01’’ [Identified as Duke] ...............................................................
Letter from Jack Spanner, dated December 17, 2007, ‘‘10 CFR 50.55a Proposed Rulemaking
Comments RIN 3150–AI01’’ [Identified as EPRI] ............................................................................
Letter from James H. Riley, dated December 17, 2007, ‘‘Proposed Rulemaking—Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (RIN
3150–AI01), 72 FR 56275, October 3, 2007 [Identified as NEI] .....................................................
Letter from Melvin L. Arey, dated December 17, 2007, ‘‘Transmittal of PWROG Comments on the
NRC Proposed Rule on Alternate Fracture Toughness Requirements for Protection Against
Pressurized Thermal Shock Events’’, RIN 3150–AI01, PA-MSC–0232 [Identified as PWROG] ....
Letter from T. Moser, dated December 17, 2007, ‘‘Strategic Teaming and Resource Sharing
(STARS) Comments on RIN 3150–AI01, Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events 72 FR 56275 (October 3,2007) [Identified as
STARS] ............................................................................................................................................
‘‘Statistical Procedures for Assessing Surveillance Data for 10 CFR Part 50.61a’’ ...........................
‘‘A Physically Based Correlation of Irradiation Induced Transition Temperature Shifts for RPV
Steel’’ ................................................................................................................................................
Supplemental Regulatory Analysis ......................................................................................................
Supplemental OMB Supporting Statement .........................................................................................
Memo from J. Uhle, dated May 15, 2008, ‘‘Embrittlement Trend Curve Development for Reactor
Pressure Vessel Materials’’ ..............................................................................................................
Draft ‘‘Technical Basis for Revision of Regulatory Guide 1.99: NRC Guidance on Methods to Estimate the Effects of Radiation Embrittlement on the Charpy V-Notch Impact Toughness of Reactor Vessel Materials’’ ........................................................................................................................
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result in accepting actual flaw size
distribution that cause the TWCF to
exceed the acceptance criteria.
Adjusting flaw sizes to account for
sizing error can change an unacceptable
examination result into an acceptable
result; further, collecting, evaluating,
and using data from ASME Code
inspector qualification tests will require
extensive engineering judgment.
Therefore, the methodology would have
to be reviewed and approved by the
Director of the NRC’s Office of Nuclear
Reactor Regulation (NRR) to ensure that
the risk associated with PTS is
acceptable. The NRC requests specific
comments on whether there should be
additional language added to 10 CFR
50.61a(e) to allow licensees to account
for the effects of sizing errors.
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‘‘Comparison of the Predictions of RM–9 to the IVAR and RADAMO Databases’’ ...........................
Memo from M. Erickson Kirk, dated December 12, 2007, ‘‘New Data from Boiling Water Reactor
Vessel Integrity Program (BWRVIP) Integrated Surveillance Project (ISP)’’ ..................................
‘‘Further Evaluation of High Fluence Data’’ .........................................................................................
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VIII. Plain Language
The Presidential memorandum ‘‘Plain
Language in Government Writing’’
published in June 10, 1998 (63 FR
31883), directed that the Government’s
documents be in clear and accessible
language. 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 NRC as explained in the
ADDRESSES heading of this notice.
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IX. Voluntary Consensus Standards
The National Technology Transfer
and Advancement Act of 1995, Public
Law 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 determined that there is
only one technical standard developed
that could be utilized for characterizing
the embrittlement correlations. That
standard is the 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 supplemental proposed rule.
However, the correlation developed by
the NRC 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 supplemental
proposed rule because it does not
represent the broad range of conditions
necessary to justify a revision to the
regulations.
The ASME Code requirements are
utilized as part of the volumetric
examination analysis requirements of
the supplemental 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 Part 50,
Appendix H and utilized to determine
30-foot-pound transition temperatures.
These standards were selected for use in
the supplemental proposed rule based
on their use in other regulations within
10 CFR Part 50 and their applicability
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to the subject of the desired
requirements.
The NRC will consider using a
voluntary consensus standard in the
final rule if an appropriate standard is
identified in the public comment period
for this supplemental proposed rule.
X. Finding of No Significant
Environmental Impact: Availability
The Commission has determined
under the National Environmental
Policy Act of 1969, as amended, and the
Commission’s regulations in 10 CFR
Part 51, Subpart A, 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 determination of this
environmental assessment is that there
will be no significant offsite impact to
the public from this action. This
determination was made as part of the
proposed rulemaking issued on October
3, 2007 (72 FR 56275), and remains
applicable to this supplemental
proposed rulemaking.
XI. Paperwork Reduction Act
Statement
This supplemental 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 supplemental 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 50.61 and 50.61a)’’
supplemental 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 an
alternative to the requirements of 10
CFR 50.61. Collections would also be
required, after implementation of the
new 10 CFR 50.61a, when any change
is made to the design or operation of the
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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: Licensees of currently operating
PWRs 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: 363 hours (253
hours annually for record keeping plus
110 hours annually for reporting).
Abstract: The NRC is proposing to
amend its regulations to provide
updated fracture toughness
requirements for protection against PTS
events for PWR pressure vessels. The
supplemental 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 is
expected to 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 utilized by
licensees of currently operating PWRs as
an alternative to complying with the
existing requirements.
The NRC is seeking public comment
on the potential impact of the
information collections contained in
this supplemental 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?
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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–1F21, Rockville, MD 20852. The
OMB clearance package and rule are
available at the NRC worldwide Web
site: https://www.nrc.gov/public-involve/
doc-comment/omb/. The
document will be available on the NRC
home page site 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
September 10, 2008. 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.
Comments submitted in writing or in
electronic form will be made available
for public inspection. Because your
comments 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. Comments submitted should
reference Docket No. NRC–2007–0008.
Comments can be submitted in
electronic form via the Federal eRulemaking Portal at https://
www.regulations.gov by search for
Docket No. NRC–2007–0008. Comments
can be mailed to NRC Clearance Officer,
Russell Nichols (T–5F52), U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001. Questions about the
information collection requirements
may be directed to the NRC Clearance
Officer, Russell Nichols (T–5 F52), U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001, by
telephone at (301) 415–6874, or by email to INFOCOLLECTS.Resource@
nrc.gov. Comments can be mailed to the
Desk Officer, Office of Information and
Regulatory Affairs, NEOB–10202,
(3150–0011), Office of Management and
Budget, Washington, DC 20503, or by email to Nathan_J._Frey@omb.eop.gov, or
by telephone at (202) 395–7345.
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.
XII. Regulatory Analysis
The NRC has issued a supplemental
regulatory analysis for this
supplemental proposed rulemaking. The
analysis examines the costs and benefits
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of the alternatives considered by the
NRC. The NRC requests public
comments on this supplemental draft
regulatory analysis. Availability of the
supplemental regulatory analysis is
provided in Section VII of this notice.
Comments on the supplemental draft
regulatory analysis may be submitted to
the NRC as indicated under the
ADDRESSES heading of this notice.
XIII. Regulatory Flexibility Act
Certification
In accordance with the Regulatory
Flexibility Act (5 U.S.C. 605(b)), the
NRC certifies that this rule would not,
if promulgated, have a significant
economic impact on a substantial
number of small entities. This
supplemental proposed rule would
affect only the licensing and operation
of currently operating 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).
XIV. Backfit Analysis
The NRC has determined that the
requirements in this supplemental
proposed rule would 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
supplemental proposed rule. The
requirements of the proposed PTS rule,
including those in the supplemental
proposed rule, would not be required,
but could be utilized by PWR licensees
with currently operating plants.
Licensees choosing to implement the
proposed PTS rule would be required to
comply with its requirements as an
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, the NRC finds that this
amendment would not constitute
backfitting.
List of Subjects for 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
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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); sec.
651(e), Pub. L. 109–58, 119 Stat. 806–810 (42
U.S.C. 2014, 2021, 2021b, 2111).
Section 50.7 also issued under Pub. L. 95–
601, sec. 10, 92 Stat. 2951 as amended by
Pub. L. 102–486, sec. 2902, 106 Stat. 3123 (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. Section 50.8(b) is revised to read as
follows:
§ 50.8 Information collection
requirements: OMB approval.
*
*
*
*
*
(b) The approved information
collection requirements contained in
this part appear in §§ 50.30, 50.33,
50.34, 50.34a, 50.35, 50.36, 50.36a,
50.36b, 50.44, 50.46, 50.47, 50.48, 50.49,
50.54, 50.55, 50.55a, 50.59, 50.60, 50.61,
50.61a, 50.62, 50.63, 50.64, 50.65, 50.66,
50.68, 50.69, 50.70, 50.71, 50.72, 50.74,
50.75, 50.80, 50.82, 50.90, 50.91, 50.120,
and appendices A, B, E, G, H, I, J, K, M,
N,O, Q, R, and S to this part.
*
*
*
*
*
3. Section 50.61a is added to read as
follows:
§ 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
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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
§ 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 °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 °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 °F.
(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 °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) jt means fast neutron fluence for
neutrons with energies greater than 1.0
MeV. jt is determined under the
provisions of paragraph (g) of this
section and has units of n/cm2.
(8) j means average neutron flux. j is
determined under the provisions of
paragraph (g) of this section and has
units of n/cm2/sec.
(9) ∆T30 means the shift in the Charpy
V-notch transition temperature
produced by irradiation defined at the
30 ft-lb energy level. The DT30 value is
determined under the provisions of
paragraph (g) of this section and has
units of °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
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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 °F.
(b) Applicability. Each licensee of a
pressurized water nuclear power
reactor, whose original operating license
was issued prior to [EFFECTIVE DATE
OF FINAL RULE], and the holder of any
operating license issued under this part
or part 54 for the Watts Bar Unit 2
facility, 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 this part.
(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
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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 plantspecific 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 reassessment 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 reanalysis 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
E:\FR\FM\11AUP1.SGM
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rwilkins on PROD1PC63 with PROPOSALS
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules
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
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
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46565
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 § 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.
(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-eighths 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 1 × 10¥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
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.
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
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Continued
11AUP1
46566
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules
rwilkins on PROD1PC63 with PROPOSALS
the sizes in Tables 2 and 3 of this
section;
(iii) There are linear indications that
penetrate through the clad into the low
alloy steel reactor vessel shell; or
(iv) Any indications between the cladto-base metal interface and threeeighths 3 of the vessel thickness exceed
the size allowable in ASME Code,
Section XI, Table IWB–3510–1.
(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-eighths 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 jt. jt must be calculated
using an NRC-approved methodology.
(1) The values of RTMAX–AW,
RTMAX–PL, RTMAX–FO, and RTMAX–CW
must be determined using Equations 1
through 4 of this section.
(2) The values of DT30 must be
determined using Equations 5 through 7
of this section, unless the conditions
specified in paragraph (f)(6)(vi) of this
section are met, for each axial weld
fusion line, plate, and circumferential
weld fusion line. The DT30 value for
each axial weld fusion line calculated as
specified by Equation 1 of this section
must be calculated for the maximum
fluence (jtFL) occurring along a
particular axial weld fusion line. The
DT30 value for each plate calculated as
specified by Equation 1 of this section
must be calculated for jtFL occurring
along a particular axial weld fusion line.
The DT30 value for each plate or forging
calculated as specified by Equations 2
and 3 of this section are calculated for
the maximum fluence (jtMAX) occurring
at the clad-to-base metal interface of
each plate or forging. In Equation 4, the
jtFL value used for calculating the plate,
forging, and circumferential weld
RTMAX–CW value is the maximum jt
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 specification 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) 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.
(ii) The following generic mean values
of RTNDT(U) must be used unless
justification for different values is
provided: 0 °F for welds made with
Linde 80 weld flux; and ¥56 °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 DT30 value.
(i) The licensee shall evaluate the
results from a plant-specific or
integrated surveillance program if the
surveillance data satisfy the criteria
described in paragraphs (f)(6)(i)(A) and
(f)(6)(i)(B) of this section:
stops at the maximum flaw size modeled for these
materials in the technical basis supporting this rule.
3 Because flaws greater than three-eighths of the
vessel wall thickness from the inside surface do not
contribute to TWCF, flaws greater than threeeighths of the vessel wall thickness from the inside
surface need not be analyzed for their contribution
to PTS.
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.’’
5 The class of material for estimating RT
NDT(U)
must be determined by the type of welding flux
(Linde 80, or other) for welds or by the material
specification for base metal.
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(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 Part 50,
Appendix H.
(B) If three or more surveillance data
points measured at three or more
different neutron fluences exist for a
specific material, the licensee shall
determine if the surveillance data show
a significantly different trend than the
embrittlement model predicts. This
must be achieved by evaluating the
surveillance data for consistency with
the embrittlement model by following
the procedures specified by paragraphs
(f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of this
section. If fewer than three surveillance
data points exist for a specific material,
then the embrittlement model must be
used without performing the
consistency check.
(ii) The licensee shall estimate the
mean deviation from the embrittlement
model for the specific data set (i.e. , a
group of surveillance data points
representative of a given material). The
mean deviation from the embrittlement
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 derived using Equation 10 of
this section. The maximum heat-average
residual is based on the material group
into which the surveillance material
falls and the number of surveillance
data points. The surveillance data
analysis must use the criteria in
paragraphs (f)(6)(v) and (f)(6)(vi) 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 s used in
Equation 10 of this section must be
obtained from Table 5 of this section.
(iii) The licensee shall estimate the
slope of the embrittlement model
residuals (estimated using Equation 8)
plotted as a function of the base 10
logarithm of neutron fluence for the
specific data set. The licensee shall
estimate the T-statistic for this slope
(TSURV) using Equation 11 and compare
this value to the maximum permissible
T-statistic (TMAX) in Table 6. The
surveillance data analysis must follow
the criteria in paragraphs (f)(6)(v) and
(f)(6)(vi) of this section. For surveillance
data sets with greater than 15 shift
points, the TMAX value must be
calculated using Student’s T
distribution with a significance level (a)
of 1 percent for a one-tailed test.
E:\FR\FM\11AUP1.SGM
11AUP1
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Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules
permissible T-statistic (TMAX) in Table
6; and
(C) The largest normalized residual
value is equal to or less than the
appropriate allowable value from the
third column in Table 7 and the second
largest normalized residual value is
equal to or less than the appropriate
allowable value from the second column
in Table 7.
(vi) If any of the criteria described in
paragraph (f)(6)(v) of this section are not
satisfied, 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 shall, on the basis
of this review, propose DT30 and
RTMAX–X values, considering their
plant-specific surveillance data, to be
used for evaluation relative to the
(iv) The licensee shall estimate the
two largest positive deviations (i.e. ,
outliers) from the embrittlement model
for the specific data set using Equations
8 and 12. The licensee shall compare
the largest normalized residual (r*) to
the appropriate allowable value from
the third column in Table 7 and the
second largest normalized residual to
the appropriate allowable value from
the second column in Table 7. The
surveillance data analysis must follow
the criteria in paragraphs (f)(6)(v) and
(f)(6)(vi) of this section.
(v) The DT30 value must be
determined using Equations 5, 6, and 7
of this section if all three of the
following criteria are satisfied:
(A) The mean deviation from the
embrittlement model for the data set is
equal to or less than the value in Table
5 or the value derived using Equation 10
of this section;
(B) The T-statistic for the slope
(TSURV) estimated using Equation 11 is
equal to or less than the maximum
acceptance criteria of this rule. These
evaluations shall be submitted for the
review and approval by the Director at
the time of the initial application. For
each surveillance capsule removed from
the reactor vessel after the submittal of
the initial application, the licensee shall
perform the analyses required by
paragraph (f)(6) of this section. The
analyses must be submitted for the
review and approval by the Director in
the form of a license amendment, and
must be submitted no later than two
years after the capsule is withdrawn
from the 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 + ∆T30 - plate ( ϕt FL ) , RTNDT ( u ) - axial weld + ∆T30 - axial weld ( ϕt FL )
Equation 2: RTMAX-PL = RTNDT ( u ) - plate + ∆T30 - plate ( ϕt MAX )
rwilkins on PROD1PC63 with PROPOSALS
Where:
P [wt-%] = phosphorus content
Mn [wt-%] = manganese content
Ni [wt-%] = nickel content
Cu [wt-%] = copper content
A = 1.140 × 10¥7 for forgings
= 1.561 × 10¥7 for plates
= 1.417 × 10¥7 for welds
B = 102.3 for forgings
= 102.5 for plates in non-Combustion
Engineering manufactured vessels
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17:37 Aug 08, 2008
Jkt 214001
= 135.2 for plates in Combustion
Engineering vessels
= 155.0 for welds
jte = j for j ≥ 4.39 × 1010 n/cm2/sec
= jt × (4.39 × 1010 / j)0.2595 for j < 4.39
× 1010 n/cm2/sec
Where:
j[n/cm2/sec] = average neutron flux
t[sec] = time that the reactor has been in full
power operation
jt[n/cm2] = j × t
f(Cue,P) = 0 for Cu ≤ 0.072
PO 00000
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Sfmt 4702
= [Cue¥0.072]0.668 for Cu > 0.072 and P ≤
0.008
= [Cue¥0.072 + 1.359 × (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,jte) = 0.5 + (0.5 × tanh {[log10(jte)
+ (1.1390 × Cue)¥(0.448 × Ni)¥18.120]
/ 0.629})
E:\FR\FM\11AUP1.SGM
11AUP1
EP11AU08.021
EP11AU08.020
Equation 7: CRP = B × (1 + 3.77 × Ni1.191 ) × f ( Cu e , P ) × g ( Cu e , Ni, ϕt e )
EP11AU08.019
Equation 6: MD = A × (1 − 0.001718 × TC ) × (1 + 6.13 × P × Mn 2.471 ) × ϕt 0.5
e
EP11AU08.018
Equation 5: ∆T30 = MD + CRP
EP11AU08.017
}
EP11AU08.016
{
Equation 4: RTMAX-CW = MAX RTNDT ( u ) - plate + ∆T30 - plate ( ϕt MAX ) , RTNDT ( u ) - circweld + ∆T30 - circweld ( ϕt MAX ) , RTNDT ( u ) - forging + ∆T30 - forging ( ϕt MAX )
EP11AU08.022
Equation 3: RTMAX-FO = RTNDT ( u ) - forging + ∆T30 - forging ( ϕt MAX )
46568
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules
Equation 8: Residual (r) = measured ∆T30 − predicted ∆T30 (by Equations 5, 6 and 7)
n
Equation 9: Mean deviation for a data set of n data points = (1/n ) × ∑ ri
i =1
Equation 10: Maximum credible heat-average residual = 2.33σ/n 0.5
σ
Where:
n = number of surveillance shift data points
(sample size) in the specific data set
s = standard deviation of the residuals about
the model for a relevant material group
given in Table 5.
Equation 11: TSURV =
Where:
m = the slope of a plot of all of the r values
(estimated using Equation 8) versus the
base 10 logarithm of the neutron fluence
for each r value. The slope shall be
estimated using the method of least
squares.
se(m) = the least squares estimate of the
standard-error associated with the
estimated slope value m.
m
se(m)
Equation 12: r* =
r
σ
Where:
r is defined using Equation 8 and s is given
in Table 5.
TABLE 1—PTS SCREENING CRITERIA
RTMAX-X limits [°F] for different vessel
wall thicknesses 6 (TWALL)
Product form and RTMAX-X values
TWALL ≤
9.5in.
Axial Weld, RTMAX-AW .............................................................................................................................
Plate, RTMAX-PL .......................................................................................................................................
Forging without underclad cracks, RTMAX-FO ..........................................................................................
Axial Weld and Plate, RTMAX-AW + RTMAX-PL ........................................................................................
Circumferential Weld, RTMAX-CW7 ...........................................................................................................
Forging with underclad cracks, RTMAX-FO ...............................................................................................
6 Wall
9.5in. <
TWALL ≤
10.5in.
269
356
356
538
312
246
230
305
305
476
277
241
10.5in. <
TWALL ≤
11.5in.
222
293
293
445
269
239
thickness is the beltline wall thickness including the clad thickness.
limits contributes 1 × 10¥8 per reactor year to the ractor vessel TWCF.
7 RT
PTS
TABLE 2—ALLOWABLE NUMBER OF FLAWS IN WELDS
TWE
TWE
TWE
TWE
TWE
TWE
TWE
TWE
TWE
<
<
<
<
<
<
<
<
<
0.075
0.125
0.175
0.225
0.275
0.325
0.375
0.425
0.475
....................................................................................
....................................................................................
....................................................................................
....................................................................................
....................................................................................
....................................................................................
....................................................................................
....................................................................................
....................................................................................
Unlimited.
166.70.
90.80.
22.82.
8.66.
4.01.
3.01.
1.49.
1.00.
rwilkins on PROD1PC63 with PROPOSALS
TABLE 3—ALLOWABLE NUMBER OF FLAWS IN PLATES OR FORGING
Allowable number of cumulative flaws per 1000
square inches of inside diameter surface area in forgings or plates in the ASME
section XI Appendix VIII
supplement 4 inspection
volume 8
ASME section XI flaw size per IWA–
3200
Range of Through-Wall Extent (TWE) of flaw [in.]
0.05 ....................................................
0.10 ....................................................
0.15 ....................................................
0.025 ≤ TWE < 0.075 ....................................................................................
0.075 ≤ TWE < 0.125 ....................................................................................
0.125 ≤ TWE < 0.175 ....................................................................................
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E:\FR\FM\11AUP1.SGM
11AUP1
Unlimited
8.049
3.146
EP11AU08.024
≤
≤
≤
≤
≤
≤
≤
≤
≤
EP11AU08.023
0.025
0.075
0.125
0.175
0.225
0.275
0.325
0.375
0.425
EP11AU08.031
....................................................
....................................................
....................................................
....................................................
....................................................
....................................................
....................................................
....................................................
....................................................
EP11AU08.030
0.05
0.10
0.15
0.20
0.25
0.30
0.35
0.40
0.45
Range of Through-Wall Extent (TWE) of flaw [in.]
EP11AU08.029
ASME section XI flaw size per IWA–
3200
Allowable number of cumulative flaws per 1000 inches
of weld length in the ASME
section XI Appendix VIII
supplement 4 inspection
volume
46569
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules
TABLE 3—ALLOWABLE NUMBER OF FLAWS IN PLATES OR FORGING—Continued
ASME section XI flaw size per IWA–
3200
0.20
0.25
0.30
0.35
Range of Through-Wall Extent (TWE) of flaw [in.]
....................................................
....................................................
....................................................
....................................................
8 Excluding
Allowable number of cumulative flaws per 1000
square inches of inside diameter surface area in forgings or plates in the ASME
section XI Appendix VIII
supplement 4 inspection
volume 8
0.175
0.225
0.275
0.325
≤
≤
≤
≤
TWE
TWE
TWE
TWE
<
<
<
<
0.225
0.275
0.325
0.375
....................................................................................
....................................................................................
....................................................................................
....................................................................................
0.853
0.293
0.0756
0.0144
underclad cracks in forgings.
TABLE 4—CONSERVATIVE ESTIMATES
FOR CHEMICAL ELEMENT WEIGHT
PERCENTAGES
Materials
P
Plates ................
TABLE 4—CONSERVATIVE ESTIMATES
FOR CHEMICAL ELEMENT WEIGHT
PERCENTAGES—Continued
Mn
0.014
Materials
1.45
P
Forgings ............
Materials
Mn
0.016
TABLE 4—CONSERVATIVE ESTIMATES
FOR CHEMICAL ELEMENT WEIGHT
PERCENTAGES—Continued
1.11
P
Welds ................
Mn
0.019
1.63
TABLE 5—MAXIMUM HEAT-AVERAGE RESIDUAL [°F] FOR RELEVANT MATERIAL GROUPS BY NUMBER OF AVAILABLE DATA
POINTS
[Significance level = 1%]
Number of available data points
s [°F]
Material group
3
Welds, for Cu > 0.072 ......................................................................................
Plates, for Cu > 0.072 ......................................................................................
Forgings, for Cu > 0.072 ..................................................................................
Weld, Plate or Forging, for Cu ≤ 0.072 ...........................................................
TABLE 6—TMAX VALUES FOR THE
SLOPE DEVIATION TEST
3 ................................................
4 ................................................
5 ................................................
6 ................................................
7 ................................................
8 ................................................
9 ................................................
10 ..............................................
11 ..............................................
12 ..............................................
14 ..............................................
15 ..............................................
TMAX
31.82
6.96
4.54
3.75
3.36
3.14
3.00
2.90
2.82
2.76
2.68
2.65
rwilkins on PROD1PC63 with PROPOSALS
TABLE 7—THRESHOLD VALUES FOR
THE OUTLIER DEVIATION TEST (SIGNIFICANCE LEVEL = 1%)
Number of available data points
(n)
Second
largest allowable normalized residual value
(r*)
Largest allowable normalized residual value
(r*)
3 ........................
4 ........................
5 ........................
1.55
1.73
1.84
2.71
2.81
2.88
VerDate Aug<31>2005
17:37 Aug 08, 2008
Jkt 214001
5
6
7
8
35.5
28.5
26.4
25.0
30.8
24.7
22.8
21.7
27.5
22.1
20.4
19.4
25.1
20.2
18.6
17.7
23.2
18.7
17.3
16.4
21.7
17.5
16.1
15.3
TABLE 7—THRESHOLD VALUES FOR
DEPARTMENT OF TRANSPORTATION
THE OUTLIER DEVIATION TEST (SIGNIFICANCE LEVEL = 1%)—Contin- Federal Aviation Administration
ued
14 CFR Part 39
[Significance level = 1%]
Number of available data points
(n)
26.4
21.2
19.6
18.6
4
Number of available data points
(n)
Second
largest allowable normalized residual value
(r*)
Largest allowable normalized residual value
(r*)
1.93
2.00
2.05
2.11
2.16
2.19
2.23
2.26
2.29
2.32
2.93
2.98
3.02
3.06
3.09
3.12
3.14
3.17
3.19
3.21
6 ........................
7 ........................
8 ........................
9 ........................
10 ......................
11 ......................
12 ......................
13 ......................
14 ......................
15 ......................
Dated at Rockville, Maryland, this 24th day
of July 2008.
For the Nuclear Regulatory Commission.
R.W. Borchardt,
Executive Director for Operations.
[FR Doc. E8–18429 Filed 8–8–08; 8:45 am]
BILLING CODE 7590–01–P
PO 00000
Frm 00013
Fmt 4702
Sfmt 4702
[Docket No. FAA–2008–0857; Directorate
Identifier 2007–NM–317–AD]
RIN 2120–AA64
Airworthiness Directives; Dornier
Model 328–300 Airplanes
Federal Aviation
Administration (FAA), Department of
Transportation (DOT).
ACTION: Notice of proposed rulemaking
(NPRM).
AGENCY:
SUMMARY: The FAA proposes to
supersede an existing airworthiness
directive (AD) that applies to all AvCraft
Dornier Model 328–300 airplanes. The
existing AD currently requires
modifying the electrical wiring of the
fuel pumps; installing insulation at the
flow control and shut-off valves, and
other components of the environmental
control system; installing markings at
fuel wiring harnesses; replacing the
wiring harness of the auxiliary fuel
system with a new wiring harness; and
installing insulated couplings in the fuel
E:\FR\FM\11AUP1.SGM
11AUP1
Agencies
[Federal Register Volume 73, Number 155 (Monday, August 11, 2008)]
[Proposed Rules]
[Pages 46557-46569]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E8-18429]
========================================================================
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.
========================================================================
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 /
Proposed Rules
[[Page 46557]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AI01
[NRC-2007-0008]
Alternate Fracture Toughness Requirements for Protection Against
Pressurized Thermal Shock Events
AGENCY: Nuclear Regulatory Commission.
ACTION: Supplemental Proposed Rule.
-----------------------------------------------------------------------
SUMMARY: The Nuclear Regulatory Commission (NRC) is considering the
adoption of provisions regarding applicability of the rule and new
provisions regarding procedures to perform surveillance data checks
related to the updated fracture toughness requirements for protection
against pressurized thermal shock (PTS) events for pressurized water
reactor (PWR) pressure vessels. The NRC is considering these provisions
as an alternative to the provisions previously noticed for public
comment on October 3, 2007 (72 FR 56275).
DATES: Submit comments on this proposed rule by September 10, 2008.
Submit comments on the information collection aspects on this proposed
rule by September 10, 2008.
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 submitted in writing or in electronic form
will be made available for public inspection. Because your comments
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.
Federal e Rulemaking Portal: Go to https://www.regulations.gov and
search for documents filed under Docket ID NRC-2007-0008. Address
questions about NRC dockets to Carol Gallagher (301) 415-5905; e-mail
Carol.Gallager@nrc.gov.
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
E-mail comments to: Rulemaking.Comments@nrc.gov. If you do not
receive a reply e-mail confirming that we have received your comments,
contact us directly at (301) 415-1966.
Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland
20852, between 7:30 am and 4:15 pm during Federal workdays. (Telephone
(301) 415-1966).
Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at
(301) 415-1101.
You can access publicly available documents related to this
document using the following methods:
NRC's Public Document Room (PDR): The public may examine publicly
available documents at the NRC's PDR, Public File Area O-F21, One White
Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR
reproduction contractor will copy documents for a fee.
NRC's Agencywide Document Access and Management System (ADAMS):
Publicly available documents created or received at the NRC are
available electronically at the NRC's Electronic Reading Room at http:/
/www.nrc.gov/reading-rm/adams.html. From this page, the public can gain
entry into 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's PDR
reference staff at 1-800-397-4209, or (301) 415-4737, or by e-mail to
PDR.Resource@nrc.gov.
FOR FURTHER INFORMATION CONTACT: Ms. Veronica M. Rodriguez, Office of
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone (301) 415-3703; e-mail:
Veronica.Rodriguez@nrc.gov, Mr. Barry Elliot, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; telephone (301) 415-2709; e-mail: Barry.Elliot@nrc.gov, or Mr.
Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 415-
6015; e-mail: Mark.Kirk@nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Introduction
II. Background
III. Discussion
IV. Responses to Comments on the Proposed Rule
V. Section-by-Section Analysis
VI. Specific Request for Comments
VII. Availability of Documents
VIII. Plain Language
IX. Voluntary Consensus Standards
X. Finding of No Significant Environmental Impact: Availability
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Act Certification
XIV. Backfit Analysis
I. Introduction
The NRC published a proposed rule on alternate fracture toughness
requirements for protection against Pressurized Thermal Shock (PTS) for
public comments in the Federal Register on October 3, 2007 (72 FR
56275). This rule provides 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 utilized
by any Pressurized Water Reactor (PWR) licensee as an alternative to
complying with the existing requirements.
During the development of the PTS final rule, the NRC determined
that several changes to the proposed rule language may be needed to
adequately address issues raised in stakeholder's comments. The NRC
also determined, in response to a stakeholder comment, that the
characteristics of advanced PWR designs were not considered in the
technical analysis made for the proposed rule. The NRC does not have
assurance that reactors that commence commercial power operation after
the effective date of this rule will have operating characteristics and
materials of fabrication similar to those evaluated as part of the
technical basis for the proposed rule. Therefore, the NRC has concluded
that it would be prudent to
[[Page 46558]]
limit the applicability and the use of Sec. 50.61a to currently-
operating plants only, and proposes to modify the applicability
provisions of the proposed rule accordingly.
Also, several stakeholders questioned the accuracy and validity of
the generic embrittlement curves in the proposed rule. The NRC wants to
ensure that the predicted values from the proposed embrittlement trend
curves provide an adequate basis for implementation of the rule.
Therefore, the NRC has continued to work on statistical procedures to
identify deviations from generic embrittlement trends, such as those
described in Sec. 50.61a(f)(6) of the proposed rule. Based on this
work, the NRC is considering enhancing the procedure described in
paragraph Sec. 50.61a(f)(6) to, among other things, detect signs from
the plant- and heat-specific surveillance data of embrittlement trends
that are not reflected by Equations 5, 6 and 7 of the rule that may
emerge at high fluences.
Because these proposed modifications may not represent a logical
outgrowth from the October 2007 proposed rule's provisions, the NRC
concludes that obtaining stakeholder feedback on the proposed
alternative provisions through the use of a supplemental proposed rule
is appropriate. As discussed in Section VI of this notice, the NRC will
consider comments on Sec. Sec. 50.61a(b); (f)(6)(i) through
(f)(6)(vi); Equations 10, 11, and 12 in Sec. 50.61a(g); and Tables 5,
6, and 7 of this supplemental proposed rule. The NRC is also requesting
comments on whether there should be additional language added to Sec.
50.61a(e) to allow licensees to account for the effects of sizing
errors. This supplemental proposed rule does not reflect other
modifications or editorial and conforming changes that the NRC is
considering to incorporate in the final rule as a result of the public
comments on the October 2007 proposed rule.
II. Background
PTS events are system transients in a PWR in which severe
overcooling occurs coincident with high pressure. The thermal stresses
are caused by rapid cooling of the reactor vessel inside surface, which
combine with the stresses caused by high pressure. The aggregate effect
of these stresses is an increase in the potential for fracture if a
pre-existing 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, can be susceptible to brittle
fracture.
The PTS rule, described in Sec. 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 Sec. 50.61 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 Sec. 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, plant modifications to reduce PTS event probability or
severity, and reactor vessel annealing, which are addressed in
Sec. Sec. 50.61(b)(3), (b)(4), and (b)(7); and Sec. 50.66,
``Requirements for Thermal Annealing of the Reactor Pressure Vessel.''
Currently, no operating PWR reactor vessel is projected to exceed
the Sec. 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.
The NRC's Office of Nuclear Regulatory Research (RES) developed a
technical basis that supports updating the PTS regulations. This
technical basis concluded that the risk of through-wall cracking due to
a PTS event is much lower than previously estimated. This finding
indicated that the screening criteria in Sec. 50.61 are unnecessarily
conservative and may impose an unnecessary burden on some licensees.
Therefore, the NRC created a new rule, Sec. 50.61a, which provides
alternate screening criteria and corresponding embrittlement
correlations based on the updated technical basis. 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, in a Staff Requirements Memorandum
(SRM) dated June 30, 2006, to prepare a rulemaking which would allow
current PWR licensees to implement the new requirements of Sec. 50.61a
or continue to comply with the current requirements of Sec. 50.61.
Alternatively, the NRC could have revised Sec. 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 NRC published the proposed rulemaking on the alternate fracture
toughness requirements for protection against PTS for public comment in
the Federal Register on October 3, 2007 (72 FR 56275). The proposed
rule provided an alternative to the current rule, which a licensee may
choose to adopt. This prompted the NRC to keep the current requirements
separate from the new alternative requirements. As a result, the
proposed rule retained the current requirements in Sec. 50.61 for PWR
licensees choosing not to implement the less restrictive screening
limits, and presented new requirements in Sec. 50.61a as an
alternative relaxation for PWR licensees.
III. Discussion
The NRC published a proposed new rule, Sec. 50.61a (October 3,
2007, 72 FR 56275), that would provide new PTS requirements based on
updated analysis methods because the existing requirements are based on
unnecessarily conservative probabilistic fracture mechanics analyses.
Stakeholders' comments raised concerns related to the applicability of
the rule and the accuracy and validity of the generic embrittlement
curves. The NRC reconsidered the technical and regulatory issues in
these areas and is considering adopting the modified provisions
regarding the applicability of the rule and new provisions regarding
procedures to perform surveillance data checks described in this
supplemental proposed rule. The NRC will consider comments on
Sec. Sec. 50.61a(b), (f)(6)(i) through (f)(6)(vi); Equations 10, 11
and 12 in Sec. 50.61a(g); and Tables 5, 6, and 7 of this supplemental
proposed rule. As described in Section VI of this notice, the NRC is
also requesting comments on whether there should be additional language
added to Sec. 50.61a(e) to allow licensees to account for the effects
of sizing errors. The NRC will consider the October 2007 proposed rule,
the supplemental proposed rule, and the comments received in response
[[Page 46559]]
to both, when deciding whether to adopt a final PTS rule.
Applicability of the Proposed Rule, Sec. 50.61a(b)
The supplemental proposed rule differs from the proposed rule and
from Sec. 50.61 in that it proposes to limit the use of Sec. 50.61a
to currently operating plants only. It cannot be demonstrated, a
priori, that reactors which commence commercial power operation after
the effective date of this rule will have operating characteristics, in
particular identified PTS event sequences and thermal-hydraulic
responses, which are consistent with the reactors which were evaluated
as part of the technical basis for this rule. Other factors, including
materials of fabrication and welding methods, could also vary. Hence,
the use of Sec. 50.61a would be limited to currently operating PWR
facilities which are known to have characteristics consistent with
those assumed in the technical basis. The NRC also proposes to allow
the holder of the operating license for Watts Bar Unit 2 to adopt the
requirements in Sec. 50.61a as this facility has operating
characteristics consistent with those assumed in the technical basis.
The NRC recognizes that licensees for reactors who commence commercial
power operation after the effective date of this rule may, under the
provisions of Sec. 50.12, seek an exemption from Sec. 50.61a(b) to
apply this rule if a plant-specific basis analyzing their operating
characteristics, materials of fabrications, and welding methods is
provided.
Surveillance Data, Sec. 50.61a(f)
Section 50.61a(f) of the proposed rule defines the process for
calculating the values for the material properties (i.e. ,
RTMAX-X) for a particular reactor vessel. These values would
be based on the vessel material's copper, manganese, phosphorus, and
nickel weight percentages, reactor cold leg temperature, and fast
neutron flux and fluence values, as well as the unirradiated nil-
ductility transition reference temperature (i.e., RTNDT).
Section 50.61a(f) of the proposed rule included a procedure by
which the RTMAX-X values, which are predicted for plant-
specific materials using a generic temperature shift (i.e.,
[Delta]T30) embrittlement trend curve, are compared with
heat-specific surveillance data that are collected as part of 10 CFR
Part 50, Appendix H surveillance programs. The purpose of this
comparison is to assess how well the surveillance data are represented
by the generic embrittlement trend curve. If the surveillance data are
close (closeness is assessed statistically) to the generic
embrittlement trend curve, then the predictions of this embrittlement
trend curve are used. This is expected to normally be the case.
However, if the heat-specific surveillance data deviate significantly,
and non-conservatively, from the predictions of the generic
embrittlement trend curve, this indicates that alternative methods
(i.e., other than, or in addition to, the generic embrittlement trend
curve) may be needed to reliably predict the temperature shift trends,
and to estimate RTMAX-X, for the conditions being assessed.
However, alternative methods for temperature shift prediction are not
prescribed by Sec. 50.61a(f) of the proposed rule.
Although standard and accepted procedures exist to assess the
statistical significance of the differences between heat-specific
surveillance data and the generic embrittlement trend curve, similarly
standard and acceptable procedures are not available to assess the
practical importance of such differences. The practical importance of
statistically significant deviations is best assessed by licensees on a
case-by-case basis, which would be submitted for the review of the
Director of NRR, as prescribed by Sec. 50.61a(f).
The method described in the proposed rulemaking to compare the
heat-specific surveillance data collected as part of 10 CFR part 50,
Appendix H surveillance programs to the generic temperature shift
embrittlement trend curve included a single statistical test. This
statistical test was set forth by Equations 9 and 10, and Table 5. This
test determined if, on average, the temperature shift from the
surveillance data was significantly higher than the temperature shift
of the generic embrittlement trend curve. The NRC has determined that,
while necessary, this single test is not sufficient to ensure that the
temperature shift predicted by the embrittlement trend curve well
represents the heat-specific surveillance data. Specifically, this
single statistical test cannot determine if the temperature shift from
the surveillance data shows a more rapid increase after significant
radiation exposure than the progression predicted by the generic
embrittlement trend curve. To address this potential deficiency, which
could be particularly important during a plant's period of extended
operation, the NRC added two more statistical tests in this
supplemental proposed rulemaking, which are expressed by Equations 11
and 12 and by Tables 6 and 7. Together, these two additional tests
determine if the surveillance data from a particular heat show a more
rapid increase after significant radiation exposure than the
progression predicted by the generic embrittlement trend curve.
The NRC documented the technical basis for the proposed alternative
in the following reports: (1) ``Statistical Procedures for Assessing
Surveillance Data for 10 CFR Part 50.61a,'' (ADAMS Accession No.
ML081290654), and (2) ``A Physically Based Correlation of Irradiation
Induced Transition Temperature Shifts for RPV Steel,'' (ADAMS Accession
No. ML081000630).
IV. Responses to Comments on the Proposed Rule
The NRC received 5 comment letters on the proposed 10 CFR 50.61a
rule published on October 3, 2007 (72 FR 56275). The following
paragraphs discuss those comments which are directly associated with
the supplemental proposed rule's provisions on the applicability of the
rule and surveillance data procedures. The remainder of the comments
and the NRC responses will be provided in the Federal Register notice
for the final rule.
Comments on the Applicability of the Proposed Rule
Comment: The commenters stated that the rule, as written, is only
applicable to the existing fleet of PWRs. The characteristics of
advanced PWR designs were not considered in the analysis. The
commenters suggested adding a statement to state that this rule is
applicable to the current PWR fleet and not the new plant designs.
[PWROG-5, EPRI-5]
Response: The NRC agrees with the comment that this rule is only
applicable to the existing fleet of PWRs. The NRC cannot be assured
that reactors that commence commercial power operation after the
effective date of this rule will have operating characteristics, in
particular identified PTS event sequences and thermal-hydraulic
responses, which are consistent with the reactors that were evaluated
as part of the technical basis for Sec. 50.61a. Other factors,
including materials of fabrication and welding methods, could also
vary. Therefore, the NRC agrees with the commenters that it would be
prudent to restrict the use of Sec. 50.61a to current plants. As a
result of this comment, the NRC proposes to modify Sec. 50.61a(b) and
the statement of considerations of the rule to reflect this position to
limit the use of the rule to currently operating plants.
Comments on Surveillance Data
Comment: The commenters stated that there is little added value in
the requirement to assess the surveillance
[[Page 46560]]
data as a part of this rule because variability in data has already
been accounted for in the derivation of the embrittlement correlation.
The commenters also stated that there is no viable methodology for
adjusting the projected [Delta]T30 for the vessel based on
the surveillance data. Any effort to make this adjustment is likely to
introduce additional error into the prediction. Note that the
embrittlement correlation described in the basis for the revised PTS
rule (i.e., NUREG-1874) was derived using all of the currently
available industry-wide surveillance data.
In the event that the surveillance data does not match the
[Delta]T30 value predicted by the embrittlement correlation,
the best estimate value for the pressure vessel material is derived
using the embrittlement correlation. The likely source of the
discrepancy is an error in the characterization of the surveillance
material or of the irradiation environment. Therefore, unless the
discrepancy can be resolved, obtaining the [Delta]T30
prediction based on the best estimate chemical composition for the heat
of the material is more reliable than a prediction based on a single
set of surveillance measurements.
The commenters suggested removing the requirement to assess
surveillance data, including Table 5, of this rule. [PWROG-4, EPRI-4,
NEI-2]
Response: The NRC does not agree with the proposed change. The NRC
believes that there is added value in the requirement to assess
surveillance data. Although variability has been accounted for in the
derivation of the embrittlement correlation, it is the NRC's view that
the surveillance assessment required in Sec. 50.61a(f)(6) is needed to
determine if the embrittlement for a specific heat of material in a
reactor vessel is consistent with the embrittlement predicted by the
embrittlement correlation.
The commenters also assert that there is no viable methodology for
adjusting the projected [Delta]T30 for the vessel based on
the surveillance data, and that any adjustment is likely to introduce
additional error into the prediction. The NRC believes that although
there is no single methodology for adjusting the projected
[Delta]T30 for the vessel based on the surveillance data, it
is possible, on a case-specific basis, to justify adjustments to the
generic [Delta]T30 prediction. For this reason the rule does
not specify a method for adjusting the [Delta]T30 value
based on surveillance data, but rather requires the licensee to propose
a case-specific [Delta]T30 adjustment procedure for review
and approval from the Director. Although the commenters assert that it
is possible that error could be introduced, it is the NRC view that
appropriate plant-specific adjustments based upon available
surveillance data may be necessary to project reactor pressure vessel
embrittlement for the purpose of this rule.
As the result of these public comments, the NRC has continued to
work on statistical procedures to identify deviations from generic
embrittlement trends, such as those described in Sec. 50.61a(f)(6) of
the proposed rule. Based on this work, the NRC is considering further
enhancing the procedure described in paragraph (f)(6) to, among other
things, detect signs from the plant- and heat-specific surveillance
data that may emerge at high fluences of embrittlement trends that are
not reflected by Equations 5, 6, and 7. The empirical basis for the
NRC's concern regarding the potential for un-modeled high fluence
effects is described in documents located at ADAMS Accession Nos.
ML081120253, ML081120289, ML081120365, ML081120380, and ML081120600.
The technical basis for the enhanced surveillance assessment procedure
is described in the document located at ADAMS Accession No.
ML081290654.
V. Section-by-Section Analysis
The following section-by-section analysis only discusses the
modifications in the provisions related to the applicability of the
rule and surveillance data procedures that the NRC is considering as an
alternative in this supplemental proposed rule. The NRC is only seeking
comments on these alternative provisions. This supplemental proposed
rule does not reflect other modifications or editorial and conforming
changes that the NRC is considering to incorporate as a result of the
public comments on the proposed rule that were not discussed in this
notice as they will be provided in the Federal Register notice for the
final rule.
Proposed Sec. 50.61a(b)
The proposed language for Sec. 50.61a(b) would establish the
applicability of the rule. The NRC proposes to modify this paragraph to
limit the use of this rule to currently-operating plants only.
Proposed Sec. 50.61a(f)(6)(i)
The proposed language for Sec. 50.61a(f)(6)(i) would establish the
requirements to perform data checks to determine if the surveillance
data show a significantly different trend than what the embrittlement
model in this rule predicts. The NRC proposes to modify Sec.
50.61a(f)(6)(i)(B) to state that licensees would evaluate the
surveillance for consistency with the embrittlement model by following
the procedures specified by Sec. Sec. 50.61a(f)(6)(ii), (f)(6)(iii),
and (f)(6)(iv) of the supplemental proposed rule.
Proposed Sec. 50.61a(f)(6)(ii)
The proposed language for Sec. 50.61a(f)(6)(ii) would establish
the requirements to perform an estimate of the mean deviation of the
data set from the embrittlement model. The mean deviation for the data
set would be compared to values given in Table 5 or Equation 10 of this
section. The NRC proposes to modify this paragraph to state that the
surveillance data analysis would follow the criteria in Sec. Sec.
50.61a(f)(6)(v) and (f)(6)(vi) of the supplemental proposed rule.
Proposed Sec. 50.61a(f)(6)(iii)
The NRC proposes to modify Sec. 50.61a(f)(6)(iii) to establish the
requirements to estimate the slope of the embrittlement model residuals
(i.e., the difference between the measured and predicted value for a
specific data point). The licensee would estimate the slope using
Equation 11 and compare this value to the maximum permissible value in
Table 6, both from the supplemental proposed rule. This surveillance
data analysis would follow the criteria in Sec. Sec. 50.61a(f)(6)(v)
and (f)(6)(vi) of the supplemental proposed rule.
Proposed Sec. 50.61a(f)(6)(iv)
The NRC proposes to modify Sec. 50.61a(f)(6)(iv) to establish the
requirements to estimate an outlier deviation from the embrittlement
model for the specific data set using Equations 8 and 12. The licensee
would compare the normalized residuals to the allowable values in Table
7 of the supplemental proposed rule. This surveillance data analysis
would follow the criteria in Sec. Sec. 50.61a(f)(6)(v) and (f)(6)(vi)
of the supplemental proposed rule.
Proposed Sec. 50.61a(f)(6)(v)
The NRC proposes to add paragraph (f)(6)(v) to establish the
criteria to be satisfied in order to calculate the
[Delta]T30 shift values.
Proposed Sec. 50.61a(f)(6)(vi)
The NRC proposes to add paragraph (f)(6)(vi) to establish the
actions to be taken by a licensee if the criteria in paragraph
(f)(6)(v) of this section are not met. The licensee would need to
submit an evaluation of the surveillance data and propose values for
[Delta]T30, considering their plant-specific
[[Page 46561]]
surveillance data, for the review and approval by the Director. The
licensee would need to submit an evaluation of each surveillance
capsule removed from the vessel after the submittal of the initial
application for review and approval by the Director no later than 2
years after the capsule is withdrawn from the vessel.
Proposed Sec. 50.61a(g)
The proposed language for Sec. 50.61a(g) would provide the
necessary equations and variables required by the proposed changes in
Sec. 50.61a(f)(6). The NRC proposes to modify Equation 10 to account
for 1 percent of significance level. Equations 11 and 12 would be added
to provide the means for estimating the slope and the outlier deviation
from the embrittlement model.
Proposed Tables 5, 6, and 7
Tables 5, 6, and 7 would provide values to be used in the proposed
changes in Sec. 50.61a(f)(6). The NRC proposes to modify Table 5 to
account for the use of a 1 percent of significance level. Tables 6 and
7 would be added to provide the threshold values for the slope and the
outlier deviation tests.
VI. Specific Request for Comments
The NRC seeks comments on Sec. Sec. 50.61a(b), (f)(6)(i) through
(f)(6)(vi); Equations 10, 11, and 12 in Sec. 50.61a(g), and Tables 5,
6, and 7 of the supplemental proposed rule. The NRC is not seeking
comments on any other provisions of the proposed Sec. 50.61a which
remain unchanged from the October 2007 proposed rule. In addition, the
NRC also requests comments on the following question:
Adjustments of the Inservice Inspection Volumetric Examination and Flaw
Assessments
The flaw sizes in Tables 2 and 3 are selected so that reactor
vessels with flaw sizes less than or equal to those in the tables will
have a TWCF less than or equal to 1 x 10-6 per reactor year
at the maximum permissible embrittlement. The NRC recognizes that the
flaw sizes in these tables represent actual flaw dimensions while the
results from the ASME Code examinations are estimated dimensions. The
available information indicates that, for most flaw sizes in Tables 2
and 3, qualified inspectors will oversize flaws. Comparing oversized
flaws to the size and density distributions in Tables 2 and 3 is
conservative and acceptable, but not necessary. Therefore, NRC is
considering to permit flaw sizes to be adjusted to account for the
effects of sizing error before comparing the estimated size and density
distribution to the acceptable size and density distributions in Tables
2 and 3. This would be accomplished by requiring licensees to base the
methodology to account for the effects of sizing error on statistical
data collected from ASME Code inspector qualification tests. An
acceptable method would include a demonstration, that accounting for
the effects of sizing error, is unlikely to result in accepting actual
flaw size distribution that cause the TWCF to exceed the acceptance
criteria. Adjusting flaw sizes to account for sizing error can change
an unacceptable examination result into an acceptable result; further,
collecting, evaluating, and using data from ASME Code inspector
qualification tests will require extensive engineering judgment.
Therefore, the methodology would have to be reviewed and approved by
the Director of the NRC's Office of Nuclear Reactor Regulation (NRR) to
ensure that the risk associated with PTS is acceptable. The NRC
requests specific comments on whether there should be additional
language added to 10 CFR 50.61a(e) to allow licensees to account for
the effects of sizing errors.
VII. Availability of Documents
The NRC is making the documents identified below available to
interested persons through one or more of the following methods, as
indicated.
Public Document Room (PDR). The NRC Public Document Room is located
at 11555 Rockville Pike, Rockville, Maryland 20852.
Regulations.gov (Web). These documents may be viewed and downloaded
electronically through the Federal eRulemaking Portal https://
www.regulations.gov, Docket number NRC-2007-0008.
NRC's Electronic Reading Room (ERR). The NRC's public electronic
reading room is located at https://www.nrc.gov/reading-rm.html.
------------------------------------------------------------------------
Document PDR Web ERR (ADAMS)
------------------------------------------------------------------------
Federal Register Notice-- X NRC-2007-0008 ML072750659
Proposed Rule: Alternate
Fracture Toughness
Requirements for Protection
Against Pressurized Thermal
Shock Events (RIN 3150-
AI01), 72 FR 56275, October
3, 2007....................
Letter from Thomas P. X NRC-2007-0008 ML073521542
Harrall, Jr., dated
December 17, 2007,
``Comments on Proposed Rule
10 CFR 50, Alternate
Fracture Toughness
Requirements for Protection
Against Pressurized Thermal
Shock Events, RIN 3150-
AI01'' [Identified as Duke]
Letter from Jack Spanner, X NRC-2007-0008 ML073521545
dated December 17, 2007,
``10 CFR 50.55a Proposed
Rulemaking Comments RIN
3150-AI01'' [Identified as
EPRI]......................
Letter from James H. Riley, X NRC-2007-0008 ML073521543
dated December 17, 2007,
``Proposed Rulemaking--
Alternate Fracture
Toughness Requirements for
Protection Against
Pressurized Thermal Shock
Events (RIN 3150-AI01), 72
FR 56275, October 3, 2007
[Identified as NEI]........
Letter from Melvin L. Arey, X NRC-2007-0008 ML073521547
dated December 17, 2007,
``Transmittal of PWROG
Comments on the NRC
Proposed Rule on Alternate
Fracture Toughness
Requirements for Protection
Against Pressurized Thermal
Shock Events'', RIN 3150-
AI01, PA-MSC-0232
[Identified as PWROG]......
Letter from T. Moser, dated X NRC-2007-0008 ML073610558
December 17, 2007,
``Strategic Teaming and
Resource Sharing (STARS)
Comments on RIN 3150-AI01,
Alternate Fracture
Toughness Requirements for
Protection against
Pressurized Thermal Shock
Events 72 FR 56275 (October
3,2007) [Identified as
STARS].....................
``Statistical Procedures for X ................ ML081290654
Assessing Surveillance Data
for 10 CFR Part 50.61a''...
``A Physically Based X ................ ML081000630
Correlation of Irradiation
Induced Transition
Temperature Shifts for RPV
Steel''....................
Supplemental Regulatory X NRC-2007-0008 ML081440673
Analysis...................
Supplemental OMB Supporting X NRC-2007-0008 ML081440736
Statement..................
Memo from J. Uhle, dated May X ................ ML081120253
15, 2008, ``Embrittlement
Trend Curve Development for
Reactor Pressure Vessel
Materials''................
Draft ``Technical Basis for X ................ ML081120289
Revision of Regulatory
Guide 1.99: NRC Guidance on
Methods to Estimate the
Effects of Radiation
Embrittlement on the Charpy
V-Notch Impact Toughness of
Reactor Vessel Materials''.
[[Page 46562]]
``Comparison of the X ................ ML081120365
Predictions of RM-9 to the
IVAR and RADAMO Databases''
Memo from M. Erickson Kirk, X ................ ML081120380
dated December 12, 2007,
``New Data from Boiling
Water Reactor Vessel
Integrity Program (BWRVIP)
Integrated Surveillance
Project (ISP)''............
``Further Evaluation of High X ................ ML081120600
Fluence Data''.............
------------------------------------------------------------------------
VIII. Plain Language
The Presidential memorandum ``Plain Language in Government
Writing'' published in June 10, 1998 (63 FR 31883), directed that the
Government's documents be in clear and accessible language. 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 NRC as explained in the ADDRESSES heading of this notice.
IX. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 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 determined that there is only one technical standard
developed that could be utilized for characterizing the embrittlement
correlations. That standard is the 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 supplemental proposed rule. However, the
correlation developed by the NRC 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 supplemental
proposed rule because it does not represent the broad range of
conditions necessary to justify a revision to the regulations.
The ASME Code requirements are utilized as part of the volumetric
examination analysis requirements of the supplemental 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 Part 50, Appendix H
and utilized to determine 30-foot-pound transition temperatures. These
standards were selected for use in the supplemental proposed rule based
on their use in other regulations within 10 CFR Part 50 and their
applicability to the subject of the desired requirements.
The NRC will consider using a voluntary consensus standard in the
final rule if an appropriate standard is identified in the public
comment period for this supplemental proposed rule.
X. Finding of No Significant Environmental Impact: Availability
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in 10
CFR Part 51, Subpart A, 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 determination of this environmental assessment is that there
will be no significant offsite impact to the public from this action.
This determination was made as part of the proposed rulemaking issued
on October 3, 2007 (72 FR 56275), and remains applicable to this
supplemental proposed rulemaking.
XI. Paperwork Reduction Act Statement
This supplemental 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 supplemental
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 50.61 and 50.61a)''
supplemental 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 an alternative to the requirements of 10 CFR 50.61.
Collections would also be required, after implementation of the new 10
CFR 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: Licensees of currently
operating PWRs 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: 363 hours (253 hours annually for
record keeping plus 110 hours annually for reporting).
Abstract: The NRC is proposing to amend its regulations to provide
updated fracture toughness requirements for protection against PTS
events for PWR pressure vessels. The supplemental 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 is expected to 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 utilized by licensees of currently operating PWRs
as an alternative to complying with the existing requirements.
The NRC is seeking public comment on the potential impact of the
information collections contained in this supplemental 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?
[[Page 46563]]
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-1F21, Rockville, MD 20852. The OMB clearance package and
rule are available at the NRC worldwide Web site: https://www.nrc.gov/
public-involve/doc-comment/omb/. The document will be
available on the NRC home page site 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 September 10, 2008. 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.
Comments submitted in writing or in electronic form will be made
available for public inspection. Because your comments 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. Comments submitted should
reference Docket No. NRC-2007-0008. Comments can be submitted in
electronic form via the Federal e-Rulemaking Portal at https://
www.regulations.gov by search for Docket No. NRC-2007-0008. Comments
can be mailed to NRC Clearance Officer, Russell Nichols (T-5F52), U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001. Questions
about the information collection requirements may be directed to the
NRC Clearance Officer, Russell Nichols (T-5 F52), U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, by telephone at (301)
415-6874, or by e-mail to INFOCOLLECTS.Resource@nrc.gov. Comments can
be mailed to the Desk Officer, Office of Information and Regulatory
Affairs, NEOB-10202, (3150-0011), Office of Management and Budget,
Washington, DC 20503, or by e-mail to Nathan_J._Frey@omb.eop.gov, or
by telephone at (202) 395-7345.
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.
XII. Regulatory Analysis
The NRC has issued a supplemental regulatory analysis for this
supplemental proposed rulemaking. The analysis examines the costs and
benefits of the alternatives considered by the NRC. The NRC requests
public comments on this supplemental draft regulatory analysis.
Availability of the supplemental regulatory analysis is provided in
Section VII of this notice. Comments on the supplemental draft
regulatory analysis may be submitted to the NRC as indicated under the
ADDRESSES heading of this notice.
XIII. Regulatory Flexibility Act Certification
In accordance with the Regulatory Flexibility Act (5 U.S.C.
605(b)), the NRC certifies that this rule would not, if promulgated,
have a significant economic impact on a substantial number of small
entities. This supplemental proposed rule would affect only the
licensing and operation of currently operating 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).
XIV. Backfit Analysis
The NRC has determined that the requirements in this supplemental
proposed rule would 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 supplemental proposed rule. The requirements of the proposed
PTS rule, including those in the supplemental proposed rule, would not
be required, but could be utilized by PWR licensees with currently
operating plants. Licensees choosing to implement the proposed PTS rule
would be required to comply with its requirements as an 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, the NRC finds that this
amendment would not constitute backfitting.
List of Subjects for 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); sec. 651(e), Pub.
L. 109-58, 119 Stat. 806-810 (42 U.S.C. 2014, 2021, 2021b, 2111).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (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. Section 50.8(b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 50.30, 50.33, 50.34, 50.34a, 50.35,
50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55,
50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64, 50.65, 50.66,
50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90,
50.91, 50.120, and appendices A, B, E, G, H, I, J, K, M, N,O, Q, R, and
S to this part.
* * * * *
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
[[Page 46564]]
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.
(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 licensee of a pressurized water nuclear
power reactor, whose original operating license was issued prior to
[EFFECTIVE DATE OF FINAL RULE], and the holder of any operating license
issued under this part or part 54 for the Watts Bar Unit 2 facility,
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 this part.
(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
reassessment 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 reanalysis 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
[[Page 46565]]
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 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.
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\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.
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(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