[Federal Register Volume 73, Number 155 (Monday, August 11, 2008)]
[Proposed Rules]
[Pages 46557-46569]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-18429]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 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.
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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 http://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
[email protected].
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
E-mail comments to: [email protected]. If you do not
receive a reply e-mail confirming that we have received your comments,
contact us directly at (301) 415-1966.
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
[email protected].
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:
[email protected], Mr. Barry Elliot, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; telephone (301) 415-2709; e-mail: [email protected], or Mr.
Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 415-
6015; e-mail: [email protected].
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 http://www.regulations.gov, Docket number NRC-2007-0008.
NRC's Electronic Reading Room (ERR). The NRC's public electronic
reading room is located at http://www.nrc.gov/reading-rm.html.
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Document PDR Web ERR (ADAMS)
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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: http://www.nrc.gov/public-involve/doc-comment/omb/index.html. 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 http://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 [email protected]. 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 [email protected], 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.
---------------------------------------------------------------------------
\1\ The ASME Code, Section XI, Appendix VIII, Supplement 4 weld
volume is the weld volume from the clad-to-base metal interface to
the inner 1.0 inch or 10 percent of the vessel thickness, whichever
is greater.
---------------------------------------------------------------------------
(i) The licensee shall determine the allowable number of weld flaws
for the reactor vessel beltline by multiplying the values in Table 2 of
this section by the total length of the reactor vessel beltline welds
that were volumetrically inspected and dividing by 1000 inches of weld
length.
(ii) The licensee shall determine the allowable number of plate or
forging flaws for their reactor vessel beltline by multiplying the
values in Table 3 of this section by the total plate or forging surface
area that was volumetrically inspected in the beltline plates or
forgings and dividing by 1000 square inches.
(iii) For each indication detected in the ASME Code, Section XI,
Appendix VIII, Supplement 4 inspection volume, the licensee shall
document the dimensions of the indication, including depth and length,
the orientation of the indication relative to the axial direction, and
the location within the reactor vessel, including its azimuthal and
axial positions and its depth embedded from the clad-to-base metal
interface. The licensee shall also document a neutron fluence map,
projected to the date of license expiration, for the reactor vessel
beltline clad-to-base metal interface and indexed in a manner that
allows the determination of the neutron fluence at the location of the
detected indications.
(2) The licensee shall identify, as part of the examination
required by paragraph (c)(2) of this section and any subsequent ASME
Code, Section XI ultrasonic examination of the beltline welds, any
indications within the ASME Code, Section XI, Appendix VIII, Supplement
4 inspection volume that are located at the clad-to-base metal
interface. The licensee shall verify that such indications do not open
to the vessel inside surface using a qualified surface or visual
examination.
(3) The licensee shall verify, as part of the examination required
by paragraph (c)(2) of this section and any subsequent ASME Code,
Section XI ultrasonic examination of the beltline welds, all
indications between the clad-to-base metal interface and three-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 x 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
[[Page 46566]]
the sizes in Tables 2 and 3 of this section;
---------------------------------------------------------------------------
\2\ Table 2 for the weld flaws is limited to flaw sizes that are
expected to occur and were modeled from the technical basis
supporting this rule. Similarly, Table 3 for the plate and forging
flaws stops at the maximum flaw size modeled for these materials in
the technical basis supporting this rule.
---------------------------------------------------------------------------
(iii) There are linear indications that penetrate through the clad
into the low alloy steel reactor vessel shell; or
(iv) Any indications between the clad-to-base metal interface and
three-eighths \3\ of the vessel thickness exceed the size allowable in
ASME Code, Section XI, Table IWB-3510-1.
---------------------------------------------------------------------------
\3\ Because flaws greater than three-eighths of the vessel wall
thickness from the inside surface do not contribute to TWCF, flaws
greater than three-eighths of the vessel wall thickness from the
inside surface need not be analyzed for their contribution to PTS.
---------------------------------------------------------------------------
(5) The analyses required by paragraph (e)(4) of this section must
address the effects on TWCF of the known sizes and locations of all
indications detected by the ASME Code, Section XI, Appendix VIII,
Supplement 4 and Supplement 6 ultrasonic examination out to three-
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 [phis]t.
[phis]t 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 [Delta]T30 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 [Delta]T30 value for
each axial weld fusion line calculated as specified by Equation 1 of
this section must be calculated for the maximum fluence ([phis]tFL)
occurring along a particular axial weld fusion line. The [Delta]T30
value for each plate calculated as specified by Equation 1 of this
section must be calculated for [phis]tFL occurring along a particular
axial weld fusion line. The [Delta]T30 value for each plate or forging
calculated as specified by Equations 2 and 3 of this section are
calculated for the maximum fluence ([phis]tMAX) occurring at the clad-
to-base metal interface of each plate or forging. In Equation 4, the
[phis]tFL value used for calculating the plate, forging, and
circumferential weld RTMAX CW value is the maximum [phis]t 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\ Data from reactor vessels fabricated to the same material
specification in the same shop as the vessel in question and in the
same time period is an example of ``generic data.''
---------------------------------------------------------------------------
(4) The values of RTNDT(U) must be evaluated according to the
procedures in the ASME Code, Section III, paragraph NB-2331. If any
other method is used for this evaluation, the licensee shall submit the
proposed method for review and approval by the Director along with the
calculation of RTMAX X values required in paragraph (c)(1) of this
section.
(i) If a measured value of RTNDT(U) is not available, a generic
mean value of RTNDT(U) for the class \5\ of material must be used if
there are sufficient test results to establish a mean.
---------------------------------------------------------------------------
\5\ The class of material for estimating RTNDT(U)
must be determined by the type of welding flux (Linde 80, or other)
for welds or by the material specification for base metal.
---------------------------------------------------------------------------
(ii) The following generic mean values of RTNDT(U) must be used
unless justification for different values is provided: 0 [deg]F for
welds made with Linde 80 weld flux; and -56 [deg]F for welds made with
Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes.
(5) The value of Tc in Equation 6 of this section must represent
the weighted time average of the reactor cold leg temperature under
normal operating full power conditions from the beginning of full power
operation through the end of licensed operation.
(6) The licensee shall verify that an appropriate RTMAX X value has
been calculated for each reactor vessel beltline material. The licensee
shall consider plant-specific information that could affect the use of
Equations 5 though 7 of this section for the determination of a
material's [Delta]T30 value.
(i) The licensee shall evaluate the results from a plant-specific
or integrated surveillance program if the surveillance data satisfy the
criteria described in paragraphs (f)(6)(i)(A) and (f)(6)(i)(B) of this
section:
(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 [sigma] 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 ([alpha]) of 1 percent for a one-tailed test.
[[Page 46567]]
(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 [Delta]T30 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 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 [Delta]T30 and RTMAX X values,
considering their plant-specific surveillance data, to be used for
evaluation relative to the 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.
[GRAPHIC] [TIFF OMITTED] TP11AU08.016
[GRAPHIC] [TIFF OMITTED] TP11AU08.017
[GRAPHIC] [TIFF OMITTED] TP11AU08.018
[GRAPHIC] [TIFF OMITTED] TP11AU08.019
[GRAPHIC] [TIFF OMITTED] TP11AU08.020
[GRAPHIC] [TIFF OMITTED] TP11AU08.021
[GRAPHIC] [TIFF OMITTED] TP11AU08.022
Where:
P [wt-%] = phosphorus content
Mn [wt-%] = manganese content
Ni [wt-%] = nickel content
Cu [wt-%] = copper content
A = 1.140 x 10-7 for forgings
= 1.561 x 10-7 for plates
= 1.417 x 10-7 for welds
B = 102.3 for forgings
= 102.5 for plates in non-Combustion Engineering manufactured
vessels
= 135.2 for plates in Combustion Engineering vessels
= 155.0 for welds
[phgr]te = [phgr] for [phgr] >= 4.39 x 1010 n/
cm2/sec
= [phgr]t x (4.39 x 1010 / [phgr])0.2595
for [phgr] < 4.39 x 1010 n/cm2/sec
Where:
[phgr][n/cm2/sec] = average neutron flux
t[sec] = time that the reactor has been in full power operation
[phgr]t[n/cm2] = [phgr] x t
f(Cue,P) = 0 for Cu <= 0.072
= [Cue-0.072]0.668 for Cu > 0.072 and P <=
0.008
= [Cue-0.072 + 1.359 x (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,[phgr]te) = 0.5 + (0.5 x tanh
{[log10([phgr]te) + (1.1390 x Cue)-
(0.448 x Ni)-18.120] / 0.629{time} )
[[Page 46568]]
[GRAPHIC] [TIFF OMITTED] TP11AU08.029
[GRAPHIC] [TIFF OMITTED] TP11AU08.030
[GRAPHIC] [TIFF OMITTED] TP11AU08.031
Where:
n = number of surveillance shift data points (sample size) in the
specific data set
[sigma] = standard deviation of the residuals about the model for a
relevant material group given in Table 5.
[GRAPHIC] [TIFF OMITTED] TP11AU08.023
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.
[GRAPHIC] [TIFF OMITTED] TP11AU08.024
Where:
r is defined using Equation 8 and [sigma] is given in Table 5.
Table 1--PTS Screening Criteria
------------------------------------------------------------------------
RTMAX-X limits [[deg]F] for different
vessel wall thicknesses \6\ (TWALL)
--------------------------------------
Product form and RTMAX-X values 9.5in. < 10.5in. <
TWALL <= TWALL <= TWALL <=
9.5in. 10.5in. 11.5in.
------------------------------------------------------------------------
Axial Weld, RTMAX-AW............. 269 230 222
Plate, RTMAX-PL.................. 356 305 293
Forging without underclad cracks, 356 305 293
RTMAX-FO........................
Axial Weld and Plate, RTMAX-AW + 538 476 445
RTMAX-PL........................
Circumferential Weld, RTMAX-CW\7\ 312 277 269
Forging with underclad cracks, 246 241 239
RTMAX-FO........................
------------------------------------------------------------------------
\6\ Wall thickness is the beltline wall thickness including the clad
thickness.
\7\ RTPTS limits contributes 1 x 10-8 per reactor year to the ractor
vessel TWCF.
Table 2--Allowable Number of Flaws in Welds
----------------------------------------------------------------------------------------------------------------
Allowable number of cumulative flaws per
ASME section XI flaw size per IWA- Range of Through-Wall Extent 1000 inches of weld length in the ASME
3200 (TWE) of flaw [in.] section XI Appendix VIII supplement 4
inspection volume
----------------------------------------------------------------------------------------------------------------
0.05................................. 0.025 <= TWE < 0.075......... Unlimited.
0.10................................. 0.075 <= TWE < 0.125......... 166.70.
0.15................................. 0.125 <= TWE < 0.175......... 90.80.
0.20................................. 0.175 <= TWE < 0.225......... 22.82.
0.25................................. 0.225 <= TWE < 0.275......... 8.66.
0.30................................. 0.275 <= TWE < 0.325......... 4.01.
0.35................................. 0.325 <= TWE < 0.375......... 3.01.
0.40................................. 0.375 <= TWE < 0.425......... 1.49.
0.45................................. 0.425 <= TWE < 0.475......... 1.00.
----------------------------------------------------------------------------------------------------------------
Table 3--Allowable Number of Flaws in Plates or Forging
----------------------------------------------------------------------------------------------------------------
Allowable number of cumulative flaws per
1000 square inches of inside diameter
ASME section XI flaw size per IWA- Range of Through-Wall Extent surface area in forgings or plates in the
3200 (TWE) of flaw [in.] ASME section XI Appendix VIII supplement 4
inspection volume \8\
----------------------------------------------------------------------------------------------------------------
0.05................................. 0.025 <= TWE < 0.075......... Unlimited
0.10................................. 0.075 <= TWE < 0.125......... 8.049
0.15................................. 0.125 <= TWE < 0.175......... 3.146
[[Page 46569]]
0.20................................. 0.175 <= TWE < 0.225......... 0.853
0.25................................. 0.225 <= TWE < 0.275......... 0.293
0.30................................. 0.275 <= TWE < 0.325......... 0.0756
0.35................................. 0.325 <= TWE < 0.375......... 0.0144
----------------------------------------------------------------------------------------------------------------
\8\ Excluding underclad cracks in forgings.
Table 4--Conservative Estimates for Chemical Element Weight Percentages
------------------------------------------------------------------------
Materials P Mn
------------------------------------------------------------------------
Plates........................................ 0.014 1.45
Forgings...................................... 0.016 1.11
Welds......................................... 0.019 1.63
------------------------------------------------------------------------
Table 5--Maximum Heat-Average Residual [[deg]F] for Relevant Material Groups by Number of Available Data Points
[Significance level = 1%]
----------------------------------------------------------------------------------------------------------------
Number of available data points
Material group [sigma] -----------------------------------------------------
[[deg]F] 3 4 5 6 7 8
----------------------------------------------------------------------------------------------------------------
Welds, for Cu > 0.072........................... 26.4 35.5 30.8 27.5 25.1 23.2 21.7
Plates, for Cu > 0.072.......................... 21.2 28.5 24.7 22.1 20.2 18.7 17.5
Forgings, for Cu > 0.072........................ 19.6 26.4 22.8 20.4 18.6 17.3 16.1
Weld, Plate or Forging, for Cu <= 0.072......... 18.6 25.0 21.7 19.4 17.7 16.4 15.3
----------------------------------------------------------------------------------------------------------------
Table 6--TMAX Values for the Slope Deviation Test
[Significance level = 1%]
------------------------------------------------------------------------
Number of available data points (n) TMAX
------------------------------------------------------------------------
3.......................................................... 31.82
4.......................................................... 6.96
5.......................................................... 4.54
6.......................................................... 3.75
7.......................................................... 3.36
8.......................................................... 3.14
9.......................................................... 3.00
10......................................................... 2.90
11......................................................... 2.82
12......................................................... 2.76
14......................................................... 2.68
15......................................................... 2.65
------------------------------------------------------------------------
Table 7--Threshold Values for the Outlier Deviation Test (Significance
Level = 1%)
------------------------------------------------------------------------
Second
largest Largest
allowable allowable
Number of available data points (n) normalized normalized
residual residual
value (r*) value (r*)
------------------------------------------------------------------------
3............................................. 1.55 2.71
4............................................. 1.73 2.81
5............................................. 1.84 2.88
6............................................. 1.93 2.93
7............................................. 2.00 2.98
8............................................. 2.05 3.02
9............................................. 2.11 3.06
10............................................ 2.16 3.09
11............................................ 2.19 3.12
12............................................ 2.23 3.14
13............................................ 2.26 3.17
14............................................ 2.29 3.19
15............................................ 2.32 3.21
------------------------------------------------------------------------
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