[Federal Register Volume 75, Number 1 (Monday, January 4, 2010)]
[Rules and Regulations]
[Pages 13-29]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-31146]
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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: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to provide alternate fracture toughness requirements for
protection against pressurized thermal shock (PTS) events for
pressurized water reactor (PWR) pressure vessels. This final rule
provides alternate 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 reduces regulatory burden for those PWR licensees who
expect to exceed the existing requirements before the expiration of
their licenses, while maintaining adequate safety, and may choose to
comply with the final rule as an alternative to complying with the
existing requirements.
DATES: Effective Date: February 3, 2010.
ADDRESSES: You can access publicly available documents related to this
document using the following methods:
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 at 301-492-3668; e-mail
[email protected].
NRC's Public Document Room (PDR): The public may examine publicly
available documents at the NRC's PDR, Public File Area O1-F21, One
White Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR
reproduction contractor will copy documents for a fee.
NRC's Agencywide Documents 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. Matthew Mitchell, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone (301) 415-1467; e-mail: [email protected],
or Mr. Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 251-
7631; e-mail: [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. Discussion
III. Responses to Comments on the Proposed Rule and Supplemental
Proposed Rule
IV. Section-by-Section Analysis
V. Availability of Documents
VI. Agreement State Compatibility
VII. Voluntary Consensus Standards
VIII. Finding of No Significant Environmental Impact: Availability
IX. Paperwork Reduction Act Statement
X. Regulatory Analysis
XI. Regulatory Flexibility Act Certification
XII. Backfit Analysis
XIII. Congressional Review Act
I. Background
PTS events are system transients in a PWR in which there is a rapid
operating temperature cooldown that results in cold vessel temperatures
with or without repressurization of the vessel. The rapid cooling of
the inside surface of the reactor vessel causes thermal stresses. The
thermal stresses can combine with 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 current PTS rule, described in Sec. 50.61, ``Fracture
Toughness Requirements for Protection against Pressurized Thermal Shock
Events,'' 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. These screening criteria
effectively define a limiting level of embrittlement beyond which
operation cannot continue without further plant-specific evaluation.
A licensee may not continue to use a reactor vessel with materials
predicted to exceed the screening criteria in Sec. 50.61 without
implementing 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 the 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 50.66, ``Requirements for Thermal
Annealing of the Reactor Pressure Vessel.''
Currently, no operating PWR vessel is projected to exceed the Sec.
50.61 screening criteria before the expiration of its 40 year operating
license. However, several PWR vessels are approaching the screening
criteria, while others are likely to exceed the screening criteria
during the extended period of operation of their first license renewal.
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 developed a proposed new rule, Sec. 50.61a,
``Alternate Fracture Requirements for Protection against Pressurized
Thermal Shock Events,'' providing 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
[[Page 14]]
criteria and updated embrittlement correlations would be appropriate.
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 rule for public comment in the
Federal Register on October 3, 2007 (72 FR 56275). Following the
closure of the comment period on the proposed rule and during the
development of the PTS final rule, the NRC determined that several
changes to the October 3, 2007 proposed rule language were desirable to
adequately address issues raised in stakeholder's comments. Because
these modifications may not have represented a logical outgrowth from
the October 2007 proposed rule's provisions, the NRC requested
stakeholder feedback on the modified provisions in a supplemental
proposed rule published in August 11, 2008 (73 FR 46557). In the
supplemental proposed rule, the NRC proposed modifications to the
provisions related to the applicability of the rule and the evaluation
of reactor vessel surveillance data. In addition, the NRC requested
comments on the adjustments of volumetric examination data to
demonstrate compliance with the rule. After consideration of the
October 2007 proposed rule, the August 2008 supplemental proposed rule
and the stakeholder comments received on both, the NRC has decided to
adopt the PTS final rule as described further in this document.
II. Discussion
The NRC completed a research program that concluded that the risk
of through-wall cracking due to a PTS event is much lower than
previously estimated. This finding indicates that the screening
criteria in Sec. 50.61 are unnecessarily conservative and may impose
an unnecessary burden on some licensees. Therefore, the NRC developed a
final rule, Sec. 50.61a, that can be implemented by PWR licensees.
The Sec. 50.61a alternate screening criteria and corresponding
embrittlement correlations are based on a technical basis as documented
in the following reports: (1) NUREG-1806, ``Technical Basis for
Revision of the Pressurized Thermal Shock (PTS) Screening Limits in the
PTS Rule (10 CFR 50.61): Summary Report,'' (ADAMS Accession No.
ML061580318); (2) NUREG-1874, ``Recommended Screening Limits for
Pressurized Thermal Shock (PTS),'' (ADAMS Accession No. ML070860156);
(3) Memorandum from Elliot to Mitchell, dated April 3, 2007,
``Development of Flaw Size Distribution Tables for Draft Proposed Title
10 of the Code of Federal Regulations (10 CFR) 50.61a,'' (ADAMS
Accession No. ML070950392); (4) ``Statistical Procedures for Assessing
Surveillance Data for 10 CFR Part 50.61a,'' (ADAMS Accession No.
ML081290654); and (5) ``A Physically Based Correlation of Irradiation
Induced Transition Temperature Shifts for RPV Steel,'' (ADAMS Accession
No. ML081000630).
Applicability of the Final Rule
The final rule is based on, in part, analyses of information from
three currently operating PWRs. Because the severity of the risk-
significant transient classes (e.g., primary side pipe breaks, stuck
open valves on the primary side that may later re-close) is controlled
by factors that are common to PWRs in general, the NRC concluded that
the results and screening criteria developed from the analysis of these
three plants can be applied with confidence to the entire fleet of
operating PWRs. This conclusion is based on an understanding of
characteristics of the dominant transients that drive their risk
significance and on an evaluation of a larger population of high
embrittlement PWRs. This evaluation revealed no design, operational,
training, or procedural factors that could credibly increase either the
severity of these transients or the frequency of their occurrence in
the general PWR population above the severity and frequency
characteristic of the three plants that were modeled in detail. The NRC
also concluded that insignificant PTS events are not expected to become
dominant.
The final rule is applicable to licensees whose construction
permits were issued before February 3, 2010 and whose reactor vessels
were designed and fabricated to the American Society of Mechanical
Engineers Boiler and Pressure Vessel Code (ASME Code), 1998 Edition or
earlier. This would include applicants for plants such as Watts Bar
Unit 2 who have not yet received an operating license. However, it
cannot be demonstrated, a priori, that reactor vessels that were not
designed and fabricated to the specified ASME Code editions will have
material properties, operating characteristics, PTS event sequences and
thermal-hydraulic responses consistent with those evaluated as part of
the technical basis for this rule. Therefore, the NRC determined that
it would not be prudent at this time to extend the use of the rule to
future PWR plants and plant designs such as the Advanced Passive (AP)
1000, Evolutionary Power Reactor (EPR) and U.S. Advanced Pressurized
Water Reactor (US-APWR). These designs have different reactor vessels
than those in the currently operating plants, and the fabrication of
the vessels based on these designs may differ from the vessels
evaluated in the analyses that form the bases for the final rule.
Licensees of reactors who commence commercial power operation after the
effective date of this rule or licensees with reactor vessels that were
not designed and fabricated to the 1998 Edition or earlier of the ASME
Code 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 plant operating characteristics, materials of fabrication, and
welding methods is provided.
Updated Embrittlement Correlation
The technical basis for Sec. 50.61a uses many different models and
parameters to estimate the yearly probability that a PWR will develop a
through-wall crack as a consequence of PTS loading. One of these models
is a revised embrittlement correlation that uses information on the
chemical composition and neutron exposure of low alloy steels in the
reactor vessel's beltline region to estimate the resistance to fracture
of these materials. Although the general trends of the embrittlement
models in Sec. Sec. 50.61 and 50.61a are similar, the form of the
revised embrittlement correlation in Sec. 50.61a differs substantially
from the correlation in Sec. 50.61. The correlation in the Sec.
50.61a final rule has been updated to more accurately represent the
substantial amount of reactor vessel surveillance data that has
accumulated since the embrittlement correlation was last revised during
the 1980s.
In-Service Inspection Volumetric Examination and Flaw Assessments
The Sec. 50.61a final rule differs from Sec. 50.61 in that it
contains a requirement for licensees who choose to follow its
requirements to analyze the results from the ASME Code, Section XI,
inservice inspection volumetric examinations. The examinations and
analyses will determine if the flaw density and size distribution in
the licensee's reactor vessel beltline are bounded by the flaw density
and size distribution used in the technical basis. The technical basis
was developed using a flaw density, spatial distribution, and size
distribution determined from experimental data, as well as from
physical models and expert
[[Page 15]]
elicitation. The experimental data were obtained from samples removed
from reactor vessel materials from cancelled plants (i.e., Shoreham and
the Pressure Vessel Research Users Facility (PVRUF) vessel). The NRC
considers that the analysis of the ASME Code inservice inspection
volumetric examination is needed to confirm that the flaw density and
size distributions in the reactor vessel, to which the final rule may
be applied, are consistent with those in the technical basis.
Paragraph (g)(6)(ii)(C) of 10 CFR 50.55a requires licensees to
implement the ASME Code, Section XI, Appendix VIII, Supplements 4 and
6. Supplement 4 contains qualification requirements for the reactor
vessel inservice inspection volume from the clad-to-base metal
interface to the inner 1.0 inch or 10 percent of the vessel thickness,
whichever is larger. Supplement 6 contains qualification requirements
for reactor vessel weld volumes other than those near the clad-to-base
metal interface. Analysis of the performance by qualified inspectors
indicates that there is an 80 percent or greater probability of
detecting a flaw that contributes to crack initiation from PTS events
when they are inspected using the ASME Code, Section XI, Appendix VIII,
Supplement 4 requirements.\1\
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\1\ Becker, L., ``Reactor Pressure Vessel Inspection
Reliability,'' Proceeding of the Joint EC-IAEA Technical Meeting on
the Improvement in In-Service Inspection Effectiveness, Petten, the
Netherlands, November 2002.
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The true flaw density for flaws with a through-wall extent of
between 0.1 and 0.3 inch can be inferred from the ASME Code examination
results and the probability of detection. The technical basis for the
final rule concludes that flaws as small as 0.1 inch in through-wall
extent contribute to the through-wall crack frequency (TWCF), and
nearly all of the contributions come from flaws buried less than 1 inch
below the inner diameter surface of the reactor vessel. For weld flaws
that exceed the sizes prescribed in the final rule, the risk analysis
indicates that a single flaw can be expected to contribute a
significant fraction of the 1 x 10-6 per reactor year limit
on TWCF. Therefore, if a flaw that exceeds the sizes prescribed in the
final rule is found in a reactor vessel, it is important to assess it
individually.
The technical basis for the final rule also indicates that flaws
buried deeper than 1 inch from the clad-to-base interface are not as
susceptible to brittle fracture as similar size flaws located closer to
the inner surface. Therefore, the final rule does not require the
comparison of the density of these flaws, but still requires large
flaws, if discovered, to be evaluated for contributions to TWCF if they
are within the inner three-eighths of the vessel thickness. The
limitation for flaw acceptance, specified in ASME Code, Section XI,
Table IWB-3510-1, approximately corresponds to the threshold for flaw
sizes that can make a significant contribution to TWCF if present in
reactor vessel material at this depth. Therefore, the final rule
requires that flaws exceeding the size limits in ASME Code, Section XI,
Table IWB-3510-1 be evaluated for contribution to TWCF in addition to
the other evaluations for such flaws that are prescribed in the ASME
Code.
The numerical values in Tables 2 and 3 of the final rule represent
the number of flaws in each size range that were derived from the
technical basis. Verifying that a plant that intends to implement this
rule has weld, plate and/or forging flaw distributions which are
consistent with those assumed in the technical basis is necessary to
ensure the applicability of the rule to that plant. If one or more
larger flaws are found in a reactor vessel, they must be evaluated to
ensure that they are not causing the TWCF to exceed the regulatory
limit.
The final rule also clarifies that, to be consistent with ASME
Code, Section XI, Appendix VIII, the smallest flaws that must be sized
are 0.075 inches in through-wall extent. For each flaw detected that
has a through-wall extent equal to or greater than 0.075 inches, the
licensee shall document the dimensions of the flaw, its orientation and
its location within the reactor vessel, and its depth from the clad-to-
base metal interface. Those planar flaws for which the major axis of
the flaw is identified by an ultrasonic transducer oriented in the
circumferential direction must be documented as ``axial.'' All other
planar flaws may be categorized as ``circumferential.'' The NRC may
also use this information to evaluate whether plant-specific
information gathered suggests that the NRC staff should generically re-
examine the technical basis for the rule.
Surface cracks that penetrate through the stainless steel clad and
more than 0.070 inch into the welds or the adjacent base metal were not
included in the technical basis because these types of flaws have not
been observed in the beltline of any operating PWR vessel. However,
flaws of this type were observed in the Quad Cities Unit 2 reactor
vessel head in 1990 (NUREG-1796, ``Safety Evaluation Report Related to
the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3
and Quad Cities Nuclear Power Station, Units 1 and 2,'' dated October
31, 2004). The observed cracks had a maximum depth into the base metal
of approximately 0.24 inch and penetrated through the stainless steel
clad. Quad Cities Units 2 and 3 are boiling water reactors which are
not susceptible to PTS events and hence are not subject to the
requirements of 10 CFR 50.61. The cracking at Quad Cities Unit 2 was
attributed to intergranular stress corrosion cracking of the stainless
steel cladding, which has not been observed in PWR vessels, and hot
cracking of the low alloy steel base metal. If these cracks were in the
beltline region of a PWR, they would be a significant contributor to
TWCF because of their size and location. The final rule requires
licensees to determine if cracks of this type exist in the beltline
weld region at each ASME Code, Section XI, ultrasonic examination.
Nondestructive Examination (NDE)-Related Uncertainties
The flaw sizes in Tables 2 and 3 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.
As a result of stakeholder feedback received on the NRC
solicitation for comments published in the August 2008 supplemental
proposed rule, the final rule will permit licenses to adjust the flaw
sizes estimated by inspectors qualified under the ASME Code, Section
XI, Appendix VIII, Supplement 4 and Supplement 6.
The NRC determined that, in addition to the NDE sizing
uncertainties, licensees should be allowed to consider other NDE
uncertainties, such as probability of detection and flaw density and
location, because these uncertainties may affect the ability of a
licensee to demonstrate compliance with the rule. As a result, the
language in Sec. 50.61a(e) will allow licensees to account for the
effects of NDE-related uncertainties in meeting the flaw size and
density requirements of Tables 2 and 3. The methodology to account for
the effects of NDE-related uncertainties must be based on statistical
data collected from ASME Code inspector qualification tests or any
other tests that measure the difference between the actual flaw size
and the size determined from the ultrasonic examination. Verification
that a licensee's flaw size
[[Page 16]]
and density distribution are upper-bounded by the distribution of
Tables 2 and 3 is required to confirm that the risk associated with PTS
is acceptable. Collecting, evaluating, and using data from ASME Code
inspector qualification tests will require extensive engineering
judgment. Therefore, the methodology used to adjust flaw sizes to
account for the effects of NDE-related uncertainties must be reviewed
and approved by the Director of the Office of Nuclear Reactor
Regulation (NRR).
Surveillance Data
Paragraph (f) of the final rule defines the process for calculating
the values for the reference temperature properties (i.e., defined as
RTMAX-X) for a particular reactor vessel. These values must
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).
The rule includes 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 be the case
most often. 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
trend, and to estimate RTMAX-X, for the conditions being
assessed.
The NRC is modifying the final rule to include three statistical
tests to determine the significance of the differences between heat-
specific surveillance data and the embrittlement trend curve. The NRC
determined that a single test is not sufficient to ensure that the
temperature shift predicted by the embrittlement trend curve represents
well the heat-specific surveillance data. Specifically, this single
statistical test cannot determine if the temperature shift from the
surveillance data show a more rapid increase after significant
radiation exposure than the progression predicted by the generic
embrittlement trend curve. This potential deficiency could be
particularly important during a plant's period of extended operation.
The deviations from the generic embrittlement trend curve are best
assessed by licensees on a case-by-case basis, which would be submitted
for the review of the Director of NRR.
The results of the first statistical test will determine if, on
average, the temperature shifts from the surveillance data are
significantly higher than the temperature shifts from the generic
embrittlement trend curve. The results of the second and third tests
will determine if the temperature shift from the surveillance data show
a more rapid increase after significant radiation exposure than the
progression predicted by the generic embrittlement trend curve.
III. Responses to Comments on the Proposed Rule and Supplemental
Proposed Rule
The NRC received 5 comment letters for a total of 54 comments on
the proposed rule published on October 3, 2007, and 3 comment letters
for a total of 5 comments on the supplemental proposed rule published
on August 11, 2008. All the comments on the proposed rule and
supplemental proposed rule were submitted by industry stakeholders. A
detailed discussion of the public comments and the NRC's responses are
contained in a separate document (see Section V, ``Availability of
Documents,'' of this document). This section only discusses the more
significant comments received on the proposed rule and supplemental
proposed rule provisions and the substantive changes made to develop
the final rule requirements. The NRC also requested stakeholder
feedback on one question in the supplemental proposed rule. This
section discusses the comments received from the NRC inquiry and the
changes made to the final rule language as a result of these comments.
Comments are discussed by subject.
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 that this rule is applicable to
the current PWR fleet and not the new plant designs.
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 plants whose construction permit was issued after February 3,
2010, and whose reactor vessel was designed and fabricated to ASME Code
Editions later than the 1998 Edition will have material properties,
operating characteristics, PTS event sequences and thermal-hydraulic
responses 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, would also be consistent
with the underlying technical basis of 10 CFR 50.61a. As a result of
this comment, the NRC modified Sec. 50.61a(b) and the statement of
considerations of the rule to reflect this position to allow the use of
the rule only to plants whose construction permit was issued before
February 3, 2010 and whose reactor vessel was designed and fabricated
to the 1998 Edition or earlier of the ASME Code.
Comments on Surveillance Data:
Comment: The commenters stated that there is little added value in
the requirement to assess the surveillance 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 then 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.
Response: The NRC does not agree with the proposed change. The NRC
believes that there is added value in the
[[Page 17]]
requirement to assess reactor vessel surveillance data. Although
variability has been accounted for in the derivation of the
embrittlement correlation, it is the NRC's view that the surveillance
data 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 of the Director of NRR. 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 enhanced the procedure
described in Sec. 50.61a(f)(6) to, among other things, detect trends
from plant- and heat-specific surveillance data that may emerge at high
fluences 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 data
assessment procedure is described in the document located at ADAMS
Accession No. ML081290654.
Comment: The second surveillance data check described in the
supplemental proposed rule should be eliminated from the rule because
the slope change evaluation appears to be of limited value.
The second required surveillance data check is to address a slope
change. The intent of this section appears to identify potential
increases in the embrittlement rate at high fluence. The industry
intends to move forward with an initiative to populate the power
reactor vessel surveillance program database with higher neutron
fluence surveillance data (i.e., extending to fluence values equivalent
to 60-80 effective full power year (EFPY)) that will adequately cover
materials variables for the entire PWR fleet. This database should
provide a more effective means of evaluating the potential for enhanced
embrittlement rates at high fluence values rather than using an
individual surveillance data set to modify the trend with fluence. Data
from this initiative will be available in the next few years to assess
the likelihood of enhanced embrittlement rates for the PWR fleet.
Response: The NRC does not agree with the commenters' statement
that the slope test (i.e., Sec. 50.61a(f)(6)(iii)) has limited value
and that it should be eliminated from the rule. The NRC believes that
the slope test provides a method for determining whether high neutron
fluence surveillance data is consistent with the [Delta]T30
model in the rule. Because there are currently only a few surveillance
data points from commercial power reactors at high neutron fluences and
the slope test will provide meaningful information, the NRC determines
that the slope test should not be eliminated from the rule.
The NRC agrees with the industry initiative to obtain additional
power reactor data at higher fluences. The NRC will review this data
and the information available to evaluate the effects of high neutron
fluence exposure when it becomes available. At that point, the NRC will
determine if modifications to the embrittlement model and/or the
surveillance data checks in Sec. 50.61a should be made.
No changes were made to the rule language as a result of this
comment.
Comments Related to the NRC Inquiry Related to the Adjustment of
Volumetric Examination Data:
Comment: Sec. 50.61a(e) should be modified to allow licensees to
account for the effects of flaw sizing uncertainties and other
uncertainties in meeting the requirements of Tables 2 and 3. The rule
language should allow the use of applicable data from ASME
qualification tests, vendor-specific performance demonstration tests,
and other current and future data that may be applicable for assessing
these uncertainties. The rule language should permit flaw sizes to be
adjusted to account for the sizing uncertainties and other
uncertainties before comparing the estimated size and density
distribution to the acceptable size and density distributions in Tables
2 and 3.
The industry will provide guidance to enable licensees to account
for the effects of sizing uncertainties and other uncertainties in
meeting the requirements of Tables 2 and 3 of the rule. Guidance to
ensure that the risk associated with PTS is acceptable will be provided
to the Director of NRR for review and approval when completed.
Response: The NRC agrees that, in addition to the NDE sizing
uncertainties, licensees should be allowed to consider other NDE
uncertainties (e.g., probability of detection, flaw density and
location) in meeting the requirements of the rule as these
uncertainties may affect the ability of a licensee to demonstrate
compliance with the rule. As a result, the language in Sec. 50.61a(e)
was modified to allow licensees to account for the effects of NDE-
related uncertainties in meeting the flaw size and density requirements
of Tables 2 and 3. This requirement would be accomplished by requiring
licensees to base their methodology to account for the NDE
uncertainties on statistical data collected from ASME Code inspector
qualification tests and any other tests that measure the difference
between the actual flaw size and the size determined from the
ultrasonic examination. Collecting, evaluating, and using data from
these tests will require extensive engineering judgment. Therefore, the
methodology would have to be reviewed and approved by the Director of
NRR.
Lastly, the commenters proposed to provide industry guidance to
enable licensees to account for the effects of NDE uncertainties. The
NRC determined that the rule language clearly states the information
that must specifically be provided for NRC review and approval if
licensees choose to account for NDE uncertainties. However, if industry
guidance documents are developed, the NRC will consider them when
submitted for review and approval.
IV. Section-by-Section Analysis
The following section-by-section analysis discusses the sections
that are being modified as a result of this final rulemaking.
Section 50.8(b)--Information collection requirements: OMB approval
This paragraph is modified to include the amended information
collection requirements as a result of this final rule.
[[Page 18]]
Section 50.61--Fracture toughness requirements for protection against
pressurized thermal shock events
Section 50.61 contains the current requirements for PTS screening
limits and embrittlement correlations. Paragraph (b) of this section is
modified to reference Sec. 50.61a as a voluntary alternative to
compliance with the requirements of Sec. 50.61. No changes are made to
the current PTS screening criteria, embrittlement correlations, or any
other related requirements in this section.
Section 50.61a--Alternate fracture toughness requirements for
protection against pressurized thermal shock events
A new Sec. 50.61a is added. Section 50.61a contains PTS screening
limits based on updated probabilistic fracture mechanics analyses. This
section provides requirements on PTS analogous to that of Sec. 50.61,
fracture toughness requirements for protection against PTS events for
PWRs. However, Sec. 50.61a differs extensively in how the licensee
determines the resistance to fractures initiating from different flaws
at different locations in the vessel beltline, as well as in the
fracture toughness screening criteria. The final rule requires
quantifying PTS reference temperatures (RTMAX-X) for flaws
along axial weld fusion lines, plates, forgings, and circumferential
weld fusion lines, and comparing the quantified value against the
RTMAX-X screening criteria. Although comparing quantified
values to the screening criteria is also required by the current Sec.
50.61, the new Sec. 50.61a provides screening criteria that vary
depending on material product form and vessel wall thickness. Further,
the embrittlement correlation and the method of calculation of
RTMAX-X values in Sec. 50.61a differ significantly from
that in Sec. 50.61 as described in the technical basis for this rule.
The new embrittlement correlation was developed using multivariable
surface-fitting techniques based on pattern recognition, understanding
of the underlying physics, and engineering judgment. The embrittlement
database used for this analysis was derived primarily from reactor
vessel material surveillance data from operating reactors that are
contained in the Power Reactor Embrittlement Data Base (PR-EDB)
developed at Oak Ridge National Laboratory. The updated
RTMAX-X estimation procedures provide a better (compared to
the existing regulation) method for estimating the fracture toughness
of reactor vessel materials over the lifetime of the plant. However, if
extensive mixed oxide (MOX) fuels with a high plutonium component are
to be used, the neutron irradiation of the vessel material will contain
more neutrons per unit energy produced and those neutrons will have
higher energies. Extensive use of MOX fuel would result in a change in
the Reactor Core Fuel Assembly (RCFA) design. Thus, in accordance to
Sec. 50.90, licensees are required to submit a license amendment
before changing the RCFA design. The Sec. 50.61a final rule requires
that licensees verify an appropriate RTMAX-X value has been
calculated for each reactor vessel beltline material considering plant-
specific information that could affect the use of the model. A licensee
using MOX fuel would use its surveillance data to meet the requirements
of Sec. 50.61a and must justify the applicability of the model
expressed by Equations 5, 6, and 7 listed in the final rule.
Section 50.61a(a)
This paragraph contains definitions for terms used in Sec. 50.61a.
It explains that terms defined in Sec. 50.61 have the same meaning in
Sec. 50.61a, unless otherwise noted.
Section 50.61a(b)
This paragraph sets forth the applicability of the final rule and
specifies that its provisions apply only to those holders of operating
licenses whose construction permits were issued before February 3,
2010, and whose reactor vessels were designed and fabricated to the
1998 Edition or earlier of the ASME Code. Both elements must be
satisfied in order for a licensee to take advantage of Sec. 50.61a.
The rule does not apply to any combined license issued under Part 52
for two reasons: (1) the combined license would be issued after
February 3, 2010, and (2) none of the reactor vessels for the nuclear
power reactors covered by these combined licenses would have been
designed and fabricated to the 1998 Edition or earlier of the ASME
Code. The same logic also explains why Sec. 50.61a would not apply to
any design certification or manufacturing license issued under Part 52.
Section 50.61a(c)
This paragraph establishes the requirements governing NRC approval
of a licensee's use of Sec. 50.61a. The licensee has to make a formal
request to the NRC via a license amendment, and would only be allowed
to implement Sec. 50.61a upon NRC approval. The license amendment
request must provide information that includes: (1) Calculations of the
values of RTMAX-X values as required by Sec. 50.61a(c)(1);
(2) examination and assessment of flaws discovered by ASME Code
inspections as required by Sec. 50.61a(c)(2); and (3) comparison of
the RTMAX-X values against the applicable screening criteria
as required by Sec. 50.61a(c)(3). In doing so, the licensee also would
be required to use Sec. Sec. 50.61a(e), (f) and (g) to perform the
necessary calculations, comparisons, examinations, assessments, and
analyses.
Section 50.61a(d)
This paragraph defines the requirements for subsequent examinations
and flaw assessments after initial approval to use Sec. 50.61a has
been obtained under the requirements of Sec. 50.61a(c). It also
defines the required compensatory measures or analyses to be taken if a
licensee determines that the screening criteria will be exceeded.
Paragraph (d)(1) defines the requirements for subsequent
RTMAX-X assessments consistent with the requirements of
Sec. Sec. 50.61a(c)(1) and (c)(3). Paragraph (d)(2) defines the
requirements for subsequent examination and flaw assessments using the
requirements of Sec. 50.61a(e). Paragraphs (d)(3) through (d)(7)
define the requirements for implementing compensatory measures or
plant-specific analyses should the value of RTMAX-X be
projected to exceed the PTS screening criteria in Table 1 of this
section.
Section 50.61a(e)
This paragraph defines the requirements for verifying that the PTS
screening criteria in Sec. 50.61a are applicable to a particular
reactor vessel. The final rule requires that the verification be based
on an analysis of test results from ultrasonic examination of the
reactor vessel beltline materials required by ASME Code, Section XI.
Section 50.61a(e)(1)
This paragraph establishes limits on flaw density and size
distributions within the volume described in ASME Code, Section XI,
Figures IWB-2500-1 and IWB-2500-2, and limited to a depth of
approximately 1 inch from the clad-to-base metal interface or 10
percent of the vessel thickness, whichever is greater. Flaws in this
inspection volume contribute approximately 97 to 99 percent to the TWCF
at the screening limit.
The verification shall be performed line-by-line for Tables 2 and
3. For example, for the second line in Table 2, the licensee would
tabulate all of the flaws detected in the relevant inspection volume in
welds and would tally the
[[Page 19]]
number that have through-wall extents between the minimum
(TWEMIN) and maximum (TWEMAX) values for line 2
(0.075 inches and 0.475 inches), would divide that total number by the
number of thousands of inches of weld length examined to get a density,
and would compare the resulting density to the limit in line 2, column
3 (which is 166.70 flaws per 1000 inches of weld metal). The licensee
would then perform a similar analysis for line 3 in Table 2 by tallying
the number of the flaws that have through-wall extents between the
TWEMIN and TWEMAX values for line 3 (0.125 inches
and 0.475 inches), would divide the total number by the number of
thousands of inches of weld length examined to get a density, and would
compare the resulting density to the limit in line 3, column 3 (which
is 90.80 flaws per 1000 inches of weld metal). This process would be
repeated for each line in the tables.
This paragraph allows licensees to adjust test results from the
volumetric examination to account for the effects of NDE-related
uncertainties. If test data is adjusted to account for NDE-related
uncertainties, the methodology and statistical data used to account for
these uncertainties must be submitted for review and approval by the
Director of NRR.
This paragraph also states that if the licensee's flaw density and
size distribution exceeds the values in Tables 2 and 3, a neutron
fluence map would have to be submitted in accordance with Sec.
50.61a(e)(6).
Sections 50.61a(e)(1)(i) and (e)(1)(ii)
These paragraphs describe the flaw density limits for welds and for
plates and forgings, respectively.
Section 50.61a(e)(1)(iii)
This paragraph describes the specific ultrasonic examination
information to be submitted to the NRC. This paragraph establishes the
documenting requirement for axial and circumferential flaws with a
through-wall extent equal to or greater than 0.075 inches. Licensees
must document indications that have been observed through ultrasonic
inspections intended to locate axially-oriented flaws as ``axial''
(i.e., an axial flaw would be one identified by an ultrasonic
transducer oriented in the circumferential direction). All other
indications may be categorized as ``circumferential.'' The NRC will use
this information to evaluate whether plant-specific information
gathered in accordance with this rule suggests that the NRC should
generically re-examine the technical basis for the rule.
Section 50.61a(e)(2)
This paragraph requires that licensees verify that clad-to-base
metal interface flaws do not open to the inside surface of the vessel.
These types of flaws could have a substantial effect on the TWCF.
Section 50.61a(e)(3)
This paragraph establishes limits for flaws that are between the
clad-to-base metal interface and three-eights of the reactor vessel
wall thickness from the interior surface. Flaws exceeding these limits
could affect the TWCF. Flaws greater than three-eights of the reactor
vessel wall thickness from the interior surface of the reactor vessel
thickness do not contribute to the TWCF at the screening limit.
Section 50.61a(e)(4)
This paragraph establishes requirements to be met if flaws exceed
the limits in Sec. Sec. 50.61a(e)(1) and (e)(3), or open to the inside
surface of the reactor vessel. This section requires an analysis to
demonstrate that the reactor vessel would have a TWCF of less than 1 x
10-\6\ per reactor year. The analysis could be a complete,
plant-specific, probabilistic fracture mechanics analysis or could be a
simplified analysis of flaw size, orientation, location and
embrittlement to demonstrate that the actual flaws in the reactor
vessel are not in locations, and/or do not have orientations, that
would cause the TWCF to be greater than 1 x 10-\6\ per
reactor year. With specific regard to circumferentially-oriented flaws
that exceed the limits of Sec. Sec. 50.61a(e)(1) and (e)(3), it may be
noted that even if a reactor pressure vessel has a circumferential weld
at the RTMAX-CW limits of Table 1, this weld only
contributes 1 x 10-\8\ per reactor year to the TWCF
predicted for the vessel. Licensees must comply with this if the
requirements of Sec. Sec. 50.61a(e)(1), (e)(2), and (e)(3) are not
satisfied.
Section 50.61a(e)(5)
This paragraph describes the critical parameters to be addressed if
flaws exceed the limits in Sec. Sec. 50.61a(e)(1) and (e)(3) or if the
flaws would open to the inside surface of the reactor vessel. This
paragraph will be required to be implemented if the requirements of
Sec. Sec. 50.61a(e)(1), (e)(2), and (e)(3) are not satisfied.
Section 50.61a(e)(6)
This paragraph establishes the requirements for submitting a
neutron fluence map if the flaw density and sizes are greater than
those specified in Tables 2 and 3. Regulatory Guide 1.190 provides an
acceptable methodology for determining the reactor vessel neutron
fluence.
Section 50.61a(f)(1) through (f)(5)
These paragraphs define the process for calculating the values for
the material properties (i.e., RTMAX-X) for a particular
reactor vessel. These values are based on the vessel's copper,
manganese, phosphorus, and nickel weight percentages, reactor cold leg
temperature, and neutron flux and fluence values, as well as the
unirradiated RTNDT of the product form in question.
Section 50.61a(f)(6)
This paragraph requires licensees to consider the plant-specific
information that could affect the use of the embrittlement model
established in the final rule.
Section 50.61a(f)(6)(i)
This paragraph establishes 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.
Licensees are required to 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).
Section 50.61a(f)(6)(ii)
This paragraph establishes the requirements to perform an estimate
of the mean deviation of the surveillance data set from the
embrittlement model. The mean deviation for the surveillance data set
must be compared to values given in Table 5 or Equation 10. The
surveillance data analysis must follow the criteria in Sec. Sec.
50.61a(f)(6)(v) and (f)(6)(vi).
Section 50.61a(f)(6)(iii)
This paragraph establishes 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
must estimate the slope using Equation 11 and compare this value to the
maximum permissible value in Table 6. This surveillance data analysis
must follow the criteria in Sec. Sec. 50.61a(f)(6)(v) and (f)(6)(vi).
Section 50.61a(f)(6)(iv)
This paragraph establishes the requirements to estimate an outlier
deviation from the embrittlement model for the specific data set using
Equations 8 and 12. The licensee must compare the normalized residuals
to the allowable values in Table 7. This
[[Page 20]]
surveillance data analysis must follow the criteria in Sec. Sec.
50.61a(f)(6)(v) and (f)(6)(vi).
Section 50.61a(f)(6)(v)
This paragraph establishes the criteria to be satisfied in order to
calculate the [Delta]T30 shift values.
Section 50.61a(f)(6)(vi)
This paragraph establishes the actions to be taken by a licensee if
the criteria in Sec. 50.61a(f)(6)(v) are not met. The licensee must
submit an evaluation of the surveillance data and propose values for
[Delta]T30, considering their plant-specific surveillance
data, for review and approval by the Director of NRR. The licensee must
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 of NRR no later than 2 years after the capsule
is withdrawn from the vessel.
Section 50.61a(g)
This paragraph provides the necessary equations and variables
required by Sec. 50.61a(f). These equations were calibrated to the
surveillance database collected in accordance with the requirements of
10 CFR part 50, Appendix H. This database contained data occupying the
range of variables detailed in the table below.
----------------------------------------------------------------------------------------------------------------
Values characterizing the surveillance database
------------------------------------------------
Variable Symbol Units Standard
Average deviation Minimum Maximum
----------------------------------------------------------------------------------------------------------------
Neutron Fluence (E>1MeV)........ [phi]t n/cm \2\ 1.24E+19 1.19E+19 9.26E+15 1.07E+20
Neutron Flux (E>1MeV)........... [phi] n/cm \2\/sec 8.69E+10 9.96E+10 2.62E+08 1.63E+12
Irradiation Temperature......... T [deg]F 545 11 522 570
Copper content.................. Cu weight % 0.140 0.084 0.010 0.410
Nickel content.................. Ni weight % 0.56 0.23 0.04 1.26
Manganese content............... Mn weight % 1.31 0.26 0.58 1.96
Phosphorus content.............. P weight % 0.012 0.004 0.003 0.031
----------------------------------------------------------------------------------------------------------------
Tables 1 through 7
Table 1 provides the PTS screening criteria for comparison with the
licensee's calculated RTMAX-X values. Tables 2 and 3 provide
values to be used in Sec. 50.61a(e). Tables 4 through 7 provide values
to be used in Sec. 50.61a(f).
V. Availability of Documents
The documents identified below are available to interested persons
through one or more of the following methods, as indicated.
Public Document Room (PDR). The NRC PDR 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 www.nrc.gov/reading-rm.html.
----------------------------------------------------------------------------------------------------------------
Document PDR Web ERR (ADAMS)
----------------------------------------------------------------------------------------------------------------
Federal Register Notice--Proposed Rule: x NRC-2007-0008 ML072750659
Alternate Fracture Toughness
Requirements for Protection Against
Pressurized Thermal Shock Events (RIN
3150-AI01), 72 FR 56275, October 3,
2007.
Regulatory History for RIN 3150-AI01, x ............................. ML072880444
Proposed Rulemaking Alternate Fracture
Toughness Requirements for Protection
Against Pressurized Thermal Shock
Events.
Letter from Thomas P. Harrall, Jr., x NRC-2007-0008 ML073521542
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 x NRC-2007-0008 ML073521545
17, 2007, ``10 CFR 50.55a Proposed
Rulemaking Comments RIN 3150-AI01''
[Identified as EPRI].
Letter from James H. Riley, dated x NRC-2007-0008 ML073521543
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 x NRC-2007-0008 ML073521547
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, x NRC-2007-0008 ML073610558
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].
Federal Register Notice--Supplemental x NRC-2007-0008 ML081440656
Proposed Rule: Alternate Fracture
Toughness Requirements for Protection
Against Pressurized Thermal Shock
Events (RIN 3150-AI01), 73 FR 46557
August 11, 2008.
Supplemental Regulatory Analysis........ x NRC-2007-0008 ML081440673
Supplemental OMB Supporting Statement... x NRC-2007-0008 ML081440736
[[Page 21]]
Regulatory History Related to x NRC-2007-008 ML082740222
Supplemental Proposed Rule: Alternate
Fracture Toughness Requirements for
Protection Against Pressurized Thermal
Shock Events, 10 CFR 50.61a (RIN
3150[dash]AI01).
E-mail from Todd A. Henderson, FENOC, x NRC-2007-0008 ML082600288
dated September 15, 2008, ``RIN
3150[dash]AI01: Comments on Alternate
Fracture Toughness Requirements for
Protection Against Pressurized Thermal
Shock Events'' [Identified as FENOC].
Letter from Dennis E. Buschbaum, dated x NRC-2007-0008 ML082550705
September 9, 2008, ``Transmittal of
PWROG Additional Comments on the NRC
`Proposed Rule on Alternate Fracture
Toughness Requirements for Protection
Against Pressurized Thermal Shock
Events', RIN 3150-AI01, PA-MSC0421''
[Identified as PWROG2].
Letter from Jack Spanner, dated x NRC-2007-0008 ML082550710
September 10, 2008, ``Proposed
Rulemaking Comments RIN 3150-AI01''
[Identified as EPRI2].
``Statistical Procedures for Assessing x ............................. ML081290654
Surveillance Data for 10 CFR Part
50.61a''.
``A Physically Based Correlation of x ............................. ML081000630
Irradiation Induced Transition
Temperature Shifts for RPV Steel''.
NUREG-1806, ``Technical Basis for x ............................. ML061580318
Revision of the Pressurized Thermal
Shock (PTS) Screening Limits in the PTS
Rule (10 CFR 50.61): Summary Report''.
NUREG-1874, ``Recommended Screening x ............................. ML070860156
Limits for Pressurized Thermal Shock
(PTS)''.
Memorandum from Elliot to Mitchell, x ............................. ML070950392
dated April 3, 2007, ``Development of
Flaw Size Distribution Tables for Draft
Proposed Title 10 of the Code of
Federal Regulations (10 CFR) 50.61a''.
Memo from J. Uhle, dated May 15, 2008, x ............................. ML081120253
``Embrittlement Trend Curve Development
for Reactor Pressure Vessel Materials''.
Draft ``Technical Basis for Revision of x ............................. ML081120289
Regulatory Guide 1.99: NRC Guidance on
Methods to Estimate the Effects of
Radiation Embrittlement on the Charpy
V[dash]Notch Impact Toughness of
Reactor Vessel Materials''.
``Comparison of the Predictions of RM-9 x ............................. ML081120365
to the IVAR and RADAMO Databases''.
Memo from M. Erickson Kirk, dated x ............................. ML081120380
December 12, 2007, ``New Data from
Boiling Water Reactor Vessel Integrity
Program (BWRVIP) Integrated
Surveillance Project (ISP)''.
``Further Evaluation of High Fluence x ............................. ML081120600
Data''.
Regulatory Guide (RG) 1.154, ``Format x ............................. ML003740028
and Content of Plant-Specific
Pressurized Thermal Shock Analysis
Reports for Pressurized Water
Reactors''.
Final OMB Supporting Statement Related x NRC-2007-0008 ML092710534
to Final Rule: Alternate Fracture
Toughness Requirements for Protection
Against Pressurized Thermal Shock
Events, 10 CFR 50.61a (RIN 3150-AI01).
Regulatory Analysis Related to Final x NRC-2007-0008 ML092710544
Rule: Alternate Fracture Toughness
Requirements for Protection Against
Pressurized Thermal Shock Events, 10
CFR 50.61a (RIN 3150-AI01).
Summary and Analysis of Public Comments x NRC-2007-0008 ML092710402
Related to the Alternate Fracture
Toughness Requirements for Protection
Against Pressurized Thermal Shock
Events.
----------------------------------------------------------------------------------------------------------------
VI. Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement States Programs,'' approved by the Commission on June 20,
1997, and published in the Federal Register (62 FR 46517) on September
3, 1997, this rule is classified as compatibility category ``NRC.''
Agreement State Compatibility is not required for Category ``NRC''
regulations. The NRC program elements in this category are those that
relate directly to areas of regulation reserved to the NRC by the
Atomic Energy Act or the provisions of Title 10 of the Code of Federal
Regulations. Although an Agreement State may not adopt program elements
reserved to NRC, it may wish to inform its licensees of certain
requirements via a mechanism that is consistent with the particular
State's administrative procedure laws. Category ``NRC'' regulations do
not confer regulatory authority on the State.
VII. 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 used 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 final 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 final rule because it does
not represent the broad range of conditions necessary to justify a
revision to the regulations.
The ASME Code requirements are used as part of the volumetric
examination analysis requirements of the final 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 used to determine 30-
foot-pound transition temperatures. These standards were selected for
use in the final rule based on their use in other regulations within 10
CFR part 50 and their applicability to the subject of the desired
requirements.
VIII. 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 is not 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.
Section 50.61a would maintain the same functional requirements for the
facility
[[Page 22]]
as the existing PTS rule in Sec. 50.61. This final rule establishes
screening criteria, limiting levels of embrittlement beyond which plant
operation cannot continue without further plant-specific evaluation or
modifications. This provides reasonable assurance that licensees
operating below the screening criteria could endure a PTS event without
fracture of vessel materials, thus assuring integrity of the reactor
pressure vessel. In addition, the final rule is risk-informed and
sufficient safety margins are maintained to ensure that any potential
increases in core damage frequency and large early release frequency
resulting from implementation of Sec. 50.61a are negligible. The final
rule will not significantly increase the probability or consequences of
accidents, result in changes being made in the types of any effluents
that may be released off site, or result in a significant increase in
occupational or public radiation exposure. Therefore, there are no
significant radiological environmental impacts associated with this
final rule. Nonradiological plant effluents are not affected as a
result of this final rule.
The NRC requested the views of the States on the environmental
assessment for this rule. No comments were received. Therefore, the
environmental assessment determination published on October 3, 2007 (72
FR 56275) remains unchanged.
IX. Paperwork Reduction Act Statement
This final rule contains new or amended information collection
requirements contained in 10 CFR part 50, that are subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501, et seq.). These
requirements were approved by the Office of Management and Budget
(OMB), approval number 3150-0011.
The burden to PWR licensees using the requirements of 10 CFR 50.61a
in lieu of the requirements of 10 CFR 50.61 for these information
collections is estimated to average 363 hours per response. This
includes the time for reviewing instructions, searching existing data
sources, gathering and maintaining the data needed, and completing and
reviewing the information collection.
Send comments on any aspect of these information collections,
including suggestions for reducing the burden, to the Records and FOIA/
Privacy Services Branch (T-5 F53), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by e-mail to
[email protected]; and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of
Management and Budget, Washington, DC 20503, or by e-mail to
[email protected].
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.
X. Regulatory Analysis
The NRC has prepared a regulatory analysis of this regulation. The
analysis examines the costs and benefits of the alternatives considered
by the NRC. The NRC concluded that implementing the final rule would
provide savings to licensees projected to exceed the PTS screening
criteria established in Sec. 50.61 in their plant lifetimes.
Availability of the regulatory analysis is provided in Section V,
``Availability of Documents'' of this document. No public comments were
received on the proposed or supplemental regulatory analyses.
XI. 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 have a significant
economic impact on a substantial number of small entities. This final
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).
XII. Backfit Analysis
The NRC has determined that the requirements in this final rule
would not constitute backfitting as defined in 10 CFR 50.109(a)(1).
Therefore, a backfit analysis has not been prepared for this 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 final rule. The requirements of the alternate PTS rule would
not be required, but could be used by current PWR licensees at their
option. Current PWR licensees choosing to implement the alternate PTS
rule are required to comply with its requirements as an alternative to
complying with the requirements of the current PTS rule. Because the
alternate PTS rule would not be mandatory for any PWR licensee, but
rather could be voluntarily implemented, the NRC has determined that
this rulemaking would not constitute backfitting.
XIII. Congressional Review Act
Under the Congressional Review Act of 1996, the NRC has determined
that this action is not a major rule and has verified this
determination with the Office of Information and Regulatory Affairs of
the OMB.
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.
0
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. 552 and 553; the NRC is adopting the
following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
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); Energy policy Act
of 2005, Pub. L. No. 109-58, 119 Stat. 194 (2005). 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).
0
2. Section 50.8(b) is revised to read as follows:
[[Page 23]]
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.
* * * * *
0
3. In Sec. 50.61, paragraph (b)(1) is revised to read as follows:
Sec. 50.61 Fracture toughness requirements for protection against
pressurized thermal shock events.
* * * * *
(b) Requirements. (1) For each pressurized water nuclear power
reactor for which an operating license has been issued under this part
or a combined license issued under Part 52 of this chapter, other than
a nuclear power reactor facility for which the certification required
under Sec. 50.82(a)(1) has been submitted, the licensee shall have
projected values of RTPTS or RTMAX-X, accepted by
the NRC, for each reactor vessel beltline material. For pressurized
water nuclear power reactors for which a construction permit was issued
under this part before February 3, 2010 and whose reactor vessel was
designed and fabricated to the 1998 Edition or earlier of the ASME
Code, the projected values must be in accordance with this section or
Sec. 50.61a. For pressurized water nuclear power reactors for which a
construction permit is issued under this part after February 3, 2010
and whose reactor vessel is designed and fabricated to an ASME Code
after the 1998 Edition, or for which a combined license is issued under
Part 52, the projected values must be in accordance with this section.
When determining compliance with this section, the assessment of
RTPTS must use the calculation procedures described in
paragraph (c)(1) and perform the evaluations described in paragraphs
(c)(2) and (c)(3) of this section. The assessment must specify the
bases for the projected value of RTPTS for each vessel
beltline material, including the assumptions regarding core loading
patterns, and must specify the copper and nickel contents and the
fluence value used in the calculation for each beltline material. This
assessment must be updated whenever there is a significant \2\ change
in projected values of RTPTS, or upon request for a change
in the expiration date for operation of the facility.
---------------------------------------------------------------------------
\2\ Changes to RTPTS values are considered
significant if either the previous value or the current value, or
both values, exceed the screening criterion before the expiration of
the operating license or the combined license under Part 52 of this
chapter, including any renewed term, if applicable for the plant.
---------------------------------------------------------------------------
* * * * *
0
4. 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 presented 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,
RTMAX-CW, or sum of RTMAX-AW and
RTMAX-PL, for a particular reactor vessel.
(7) [phi]t means fast neutron fluence for neutrons with energies
greater than 1.0 MeV. [phi]t is utilized under the provisions of
paragraph (g) of this section and has units of n/cm\2\.
(8) [phi] means average neutron flux for neutrons with energies
greater than 1.0 MeV. [phi] is utilized 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 at the 30 ft-lb energy level produced by
irradiation. The [Delta]T30 value is utilized 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, 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.
(12) CRP means the copper rich precipitate term in the
embrittlement model from this section. The CRP term is defined in
paragraph (g) of this section.
(13) MD means the matrix damage term in the embrittlement model for
this section. The MD term is defined in paragraph (g) of this section.
(b) Applicability. The requirements of this section apply to each
holder of an operating license for a pressurized water nuclear power
reactor whose construction permit was issued before February 3, 2010
and whose reactor vessel was designed and fabricated to the ASME Boiler
and Pressure Vessel Code, 1998 Edition or earlier. The requirements of
this section may be implemented as an alternative to the requirements
of 10 CFR 50.61.
(c) Request for Approval. Before the implementation of this
section, each licensee shall submit a request for approval in the form
of an application for a license amendment in accordance with Sec.
50.90 together with the documentation required by paragraphs (c)(1),
(c)(2), and (c)(3) of this section for review and approval by the
Director of the Office of Nuclear Reactor Regulation (Director). The
application 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.
[[Page 24]]
(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.
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); 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. Assessments
performed under paragraphs (f)(6) and (f)(7) of this section, shall be
submitted by the licensee to the Director in its license amendment
application to utilize Sec. 50.61a.
(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), (e)(1), (e)(2), and (e)(3) of this section have been
met. The licensee must submit to the Director, in its application to
use Sec. 50.61a, the adjustments made to the volumetric test data to
account for NDE-related uncertainties as described in paragraph (e)(1)
of this section, all information required by paragraph (e)(1)(iii) of
this section, and, if applicable, analyses performed under paragraphs
(e)(4), (e)(5) and (e)(6) of this section.
(3) Each licensee shall compare the projected RTMAX-X
values for plates, forgings, axial welds, and circumferential welds to
the PTS screening criteria in Table 1 of this section, 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
use 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, so that the previous value, the current value, or
both values, exceed the screening criteria before the expiration of the
plant operating license; or upon the licensee's request for a change in
the expiration date for operation of the facility; a re-assessment of
RTMAX-X values documented consistent with the requirements
of paragraph (c)(1) and (c)(3) of this section must be submitted in the
form of a license amendment for review and approval by the Director. If
the surveillance data used to perform the re-assessment of
RTMAX-X values meet the requirements of paragraph (f)(6)(v)
of this section, the licensee shall submit the data and the results of
the analysis of the data to the Director for review and approval within
one year after the capsule is withdrawn from the vessel. If the
surveillance data meet the requirements of paragraph (f)(6)(vi) of this
section, the licensee shall submit the data, the results of the
analysis of the data, and proposed [Delta]T30 and
RTMAX-X values considering the surveillance data in the form
of a license amendment to the Director for review and approval within
two years after the capsule is withdrawn from the vessel. 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, before operation
beyond the PTS screening criteria in Table 1 of this section.
(2) The licensee shall verify that the requirements of paragraphs
(e), (e)(1), (e)(2), and (e)(3) of this section have been met. The
licensee must submit, within 120 days after completing a volumetric
examination of reactor vessel beltline materials as required by ASME
Code, Section XI, the adjustments made to the volumetric test data to
account for NDE-related uncertainties as described in paragraph (e)(1)
of this section and all information required by paragraph (e)(1)(iii)
of this section in the form of a license amendment for review and
approval by the Director. If a licensee is required to implement
paragraphs (e)(4), (e)(5), and (e)(6) of this section, the information
required in these paragraphs must be submitted in the form of a license
amendment for review and approval by the Director within one year after
completing a volumetric examination of reactor vessel materials as
required by ASME Code, Section XI.
(3) If the value of RTMAX-X is projected to exceed the
PTS screening criteria, then the licensee shall implement those flux
reduction programs that are reasonably practicable to avoid exceeding
the PTS screening criteria. The schedule for implementation of flux
reduction measures may take into account the schedule for review and
anticipated approval by the Director of detailed plant-specific
analyses which demonstrate acceptable risk with RTMAX-X
values above the PTS screening criteria due to plant modifications, new
information, or new analysis techniques.
(4) If the analysis required by paragraph (d)(3) of this section
indicates that no reasonably practicable flux reduction program will
prevent the RTMAX-X value for one or more reactor vessel
beltline materials from exceeding the PTS screening criteria, then the
licensee shall perform a safety analysis to determine what, if any,
modifications to equipment, systems, and operation are necessary to
prevent the potential for an unacceptably high probability of failure
of the reactor vessel as a result of postulated PTS events. 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 and the description of the modifications must
be submitted to the Director in the form of a license amendment 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. The Director
shall impose the modifications to equipment, systems and operations
described to meet paragraph (d)(4) of this section.
(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, before 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. The
licensee
[[Page 25]]
must show that the proposed alternatives provide reasonable assurance
of adequate protection of the public health and safety.
(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
examination 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, as specified in 10 CFR
50.55a(b)(2)(xv).
(1) The licensee shall verify that the flaw density and size
distributions within the volume described in ASME Code, Section XI,\1\
Figures IWB-2500-1 and IWB-2500-2 and limited to a depth from the clad-
to-base metal interface of 1-inch or 10 percent of the vessel
thickness, whichever is greater, do not exceed the limits in Tables 2
and 3 of this section based on the test results from the volumetric
examination. The values in Tables 2 and 3 represent actual flaw sizes.
Test results from the volumetric examination may be adjusted to account
for the effects of NDE-related uncertainties. The methodology to
account for NDE-related uncertainties must be based on statistical data
from the qualification tests and any other tests that measure the
difference between the actual flaw size and the NDE detected flaw size.
Licensees who adjust their test data to account for NDE-related
uncertainties to verify conformance with the values in Tables 2 and 3
shall prepare and submit the methodology used to estimate the NDE
uncertainty, the statistical data used to adjust the test data and an
explanation of how the data was analyzed for review and approval by the
Director in accordance with paragraphs (c)(2) and (d)(2) of this
section. The verification of the flaw density and size distributions
shall be performed line-by-line for Tables 2 and 3. If the flaw density
and size distribution exceeds the limitations specified in Tables 2 and
3 of this section, the licensee shall perform the analyses required by
paragraph (e)(4) of this section. If analyses are required in
accordance with paragraph (e)(4) of this section, the licensee must
address the effects on through-wall crack frequency (TWCF) in
accordance with paragraph (e)(5) of this section and must prepare and
submit a neutron fluence map in accordance with the requirements of
paragraph (e)(6) of this section.
---------------------------------------------------------------------------
\1\ For forgings susceptible to underclad cracking the
determination of the flaw density for that forging from the
licensee's inspection shall exclude those indications identified as
underclad cracks.
---------------------------------------------------------------------------
(i) The licensee shall determine the allowable number of weld flaws
in 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 in their reactor vessel beltline by multiplying the
values in Table 3 of this section by the total surface area of the
reactor vessel beltline plates or forgings that were volumetrically
inspected and dividing by 1000 square inches.
(iii) For each flaw detected in the inspection volume described in
paragraph (e)(1) with a through-wall extent equal to or greater than
0.075 inches, the licensee shall document the dimensions of the flaw,
including through-wall extent and length, whether the flaw is axial or
circumferential in orientation and its location within the reactor
vessel, including its azimuthal and axial positions and its depth
embedded from the clad-to-base metal interface.
(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
flaws within the inspection volume described in paragraph (e)(1) of
this section that are equal to or greater than 0.075 inches in through-
wall depth, axially-oriented, and located at the clad-to-base metal
interface. The licensee shall verify that these flaws do not open to
the vessel inside surface using surface or visual examination technique
capable of detecting and characterizing service induced cracking of the
reactor vessel cladding.
(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, that all flaws
between the clad-to-base metal interface and three-eights of the
reactor vessel thickness from the interior surface are within the
allowable values in ASME Code, Section XI, Table IWB-3510-1.
(4) The licensee shall perform analyses to demonstrate that the
reactor vessel will have a 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 flaw density and size in the inspection volume described in
paragraph (e)(1) exceed the limits in Tables 2 or 3 of this section;
(ii) There are axial flaws that penetrate through the clad into the
low alloy steel reactor vessel shell, at a depth equal to or greater
than 0.075 inches in through-wall extent from the clad-to-base metal
interface; or
(iii) Any flaws between the clad-to-base metal interface and three-
eighths \2\ of the vessel thickness exceed the size allowable in ASME
Code, Section XI, Table IWB-3510-1.
---------------------------------------------------------------------------
\2\ Because flaws greater than three-eights of the vessel wall
thickness from the inside surface do not contribute to TWCF, flaws
greater than three-eights of the vessel wall thickness from the
inside surface need not be analyzed for their contribution to PTS.
---------------------------------------------------------------------------
(5) The analyses required by paragraph (e)(4) of this section must
address the effects on TWCF of the known sizes and locations of all
flaws detected by the ASME Code, Section XI, Appendix VIII, Supplement
4 and Supplement 6 ultrasonic examination out to three-eights of the
vessel thickness from the inner surface, and may also take into account
other reactor vessel-specific information, including fracture toughness
information.
(6) For all flaw assessments performed in accordance with paragraph
(e)(4) of this section, the licensee shall prepare and submit 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 flaws.
(f) Calculation of RTMAX-X values. Each licensee shall
calculate RTMAX-X values for each reactor vessel beltline
material using [phi]t. The neutron flux ([phi][t]), must be calculated
using a methodology that has been benchmarked to experimental
measurements and with quantified uncertainties and possible biases.\3\
---------------------------------------------------------------------------
\3\ Regulatory Guide 1.190 dated March 2001, establishes
acceptable methods for determining neutron flux.
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[[Page 26]]
(1) The values of RTMAX-AW, RTMAX-PL,
RTMAX-FO, and RTMAX-CW must be determined using
Equations 1 through 4 of this section. When calculating
RTMAX-AW using Equation 1, RTMAX-AW is the
maximum value of (RTNDT(U) + [Delta]T30) for the
weld and for the adjoining plates. When calculating RTMAX-CW
using Equation 4, RTMAX-CW is the maximum value of
(RTNDT(U) + [Delta]T30) for the circumferential
weld and for the adjoining plates or forgings.
(2) The values of [Delta]T30 must be determined using
Equations 5, 6 and 7 of this section, unless the conditions specified
in paragraph (f)(6)(v) of this section are not met, for each axial
weld, plate, forging, and circumferential weld. The
[Delta]T30 value for each axial weld calculated as specified
by Equation 1 of this section must be calculated for the maximum
fluence ([phi]tAXIAL-WELD) occurring along a particular
axial weld at the clad-to-base metal interface. The
[Delta]T30 value for each plate calculated as specified by
Equation 1 of this section must also be calculated using the same value
of [phi]tAXIAL-WELD used for the axial weld. The
[Delta]T30 values in Equation 1 shall be calculated for the
weld itself and each adjoining plate. The [Delta]T30 value
for each plate or forging calculated as specified by Equations 2 and 3
of this section must be calculated for the maximum fluence
([phi]tMAX) occurring at the clad-to-base metal interface
over the entire area of each plate or forging. In Equation 4, the
fluence ([phi]tWELD-CIRC) value used for calculating the
plate, forging, and circumferential weld [Delta]T30 value is
the maximum fluence occurring for each material along the
circumferential weld at the clad-to-base metal interface. The
[Delta]T30 values in Equation 4 shall be calculated for the
circumferential weld and for the adjoining plates or forgings. If the
conditions specified in paragraph (f)(6)(v) of this section are not
met, licensees must propose [Delta]T30 and
RTMAX-X values in accordance with paragraph (f)(6)(vi) of
this section.
(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. For a
plate or forging, the best estimate value is normally the mean of the
measured values for that plate or forging. For a weld, the best
estimate value is normally the mean of the measured values for a weld
deposit made using the same weld wire heat number as the critical
vessel weld. If these values are not available, either the upper
limiting values given in the material specifications to which the
vessel material was fabricated, or conservative estimates (i.e., 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 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 time-weighted 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 by considering plant-specific information that could
affect the use of the model (i.e., Equations 5, 6 and 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. For surveillance data sets with greater than 8 data 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. For
surveillance data sets with greater than 15 data 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.
(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.
(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
[[Page 27]]
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. If
any of these criteria is not satisfied, the licensee must propose
[Delta]T30 and RTMAX-X values in accordance with
paragraph (f)(6)(vi) of this section.
(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 5,
6, and 7 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 to the NRC
and shall 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 must be submitted for review and approval by the Director
in the form of a license amendment in accordance with the requirements
of paragraphs (c)(1) and (d)(1) of this section.
(7) The licensee shall report any information that significantly
influences the RTMAX-X value to the Director in accordance
with the requirements of paragraphs (c)(1) and (d)(1) of this section.
(g) Equations and variables used in this section.
Equation 1: RTMAX-AW = MAX {[RTNDT(U)-plate +
[Delta]T30-plate],
[RTNDT(U)-axial weld +
[Delta]T30-axial weld]{time}
Equation 2: RTMAX-PL = RTNDT(U)-plate +
[Delta]T30-plate
Equation 3: RTMAX-FO = RTNDT(U)-forging +
[Delta]T30-forging
Equation 4: RTMAX-CW = MAX {[RTNDT(U)-plate +
[Delta]T30-plate],
[RTNDT(U)-circweld + [Delta]T30-circweld],
[RTNDT(U)-forging +
[Delta]T30-forging]{time}
Equation 5: [Delta]T30 = MD + CRP
Equation 6: MD = A x (1-0.001718 x TC) x (1 + 6.13 x P x
Mn2.471) x [phi]te0.5
Equation 7: CRP = B x (1 + 3.77 x Ni1.191) x
f(Cue,P) x g(Cue,Ni,[phi]te)
Where:
P [wt-%] = phosphorus content
Mn [wt-%] = manganese content
Ni [wt-%] = nickel content
Cu [wt-%] = copper content
A = 1.140 x 10-7 for forgings
= 1.561 x 10-7 for plates
= 1.417 x 10-7 for welds
B = 102.3 for forgings
= 102.5 for plates in non-Combustion Engineering manufactured
vessels
= 135.2 for plates in Combustion Engineering vessels
= 155.0 for welds
[phi]te = [phi]t for [phi] >= 4.39 x 1010 n/
cm2/sec
= [phi]t x (4.39 x 1010/[phi]) 0.2595 for
[phi] < 4.39 x 1010 n/cm2/sec
Where:
[phi] [n/cm2/sec] = average neutron flux
t [sec] = time that the reactor has been in full power operation
[phi]t [n/cm2] = [phi] 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,[phi]te) = 0.5 + (0.5 x tanh
{[log10([phi]te) + (1.1390 x Cue)-
(0.448 x Ni)-18.120]/0.629{time}
Equation 8: Residual (r) = measured [Delta]T30-predicted
[Delta]T30 (by Equations 5, 6 and 7)
Equation 9: Mean deviation for a data set of n data points =
[GRAPHIC] [TIFF OMITTED] TR04JA10.098
Equation 10: Maximum credible heat-average residual = 2.33[sigma]/
n0.5
Where:
n = number of surveillance 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] TR04JA10.099
Where:
m is 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)) is the least squares estimate of the standard-error
associated with the estimated slope value m.
[GRAPHIC] [TIFF OMITTED] TR04JA10.100
Where:
r is defined using Equation 8 and [sigma] is given in Table 5.
---------------------------------------------------------------------------
\6\ Wall thickness is the beltline wall thickness including the
clad thickness.
\7\ Forgings without underclad cracks apply to forgings for
which no underclad cracks have been detected and that were
fabricated in accordance with Regulatory Guide 1.43.
\8\ RTPTS limits contribute 1 x 10-8 per
reactor year to the reactor vessel TWCF.
\9\ Forgings with underclad cracks apply to forgings that have
detected underclad cracking or were not fabricated in accordance
with Regulatory Guide 1.43.
Table 1--PTS Screening Criteria
--------------------------------------------------------------------------------------------------------------------------------------------------------
RTMAX X limits [[deg]F] for different vessel wall thicknesses \6\ (TWALL)
Product form and RTMAX X Values --------------------------------------------------------------------------------------------
TWALL <= 9.5 in. 9.5 in. < TWALL <= 10.5 in. 10.5 in. < TWALL <= 11.5 in.
--------------------------------------------------------------------------------------------------------------------------------------------------------
Axial Weld RTMAX-AW........................................ 269 230 222
Plate RTMAX-PL............................................. 356 305 293
Forging without underclad cracks RTMAX-FO \7\.............. 356 305 293
Axial Weld and Plate RTMAX-AW + RTMAX-PL................... 538 476 445
Circumferential Weld RTMAX-CW \8\.......................... 312 277 269
Forging with underclad cracks RTMAX-FO \9\................. 246 241 239
--------------------------------------------------------------------------------------------------------------------------------------------------------
[[Page 28]]
Table 2--Allowable Number of Flaws in Welds
------------------------------------------------------------------------
Through-wall extent, TWE [in.] Maximum number of
----------------------------------------------------- flaws per 1000-
inches of weld
length in the
inspection volume
TWEMIN TWEMAX that are greater
than or equal to
TWEMIN and less
than TWEMAX
------------------------------------------------------------------------
0............................... 0.075............. No Limit
0.075........................... 0.475............. 166.70
0.125........................... 0.475............. 90.80
0.175........................... 0.475............. 22.82
0.225........................... 0.475............. 8.66
0.275........................... 0.475............. 4.01
0.325........................... 0.475............. 3.01
0.375........................... 0.475............. 1.49
0.425........................... 0.475............. 1.00
0.475........................... Infinite.......... 0.00
------------------------------------------------------------------------
Table 3--Allowable Number of Flaws in Plates and Forgings
------------------------------------------------------------------------
Through[dash]wall extent, TWE [in.] Maximum number of
----------------------------------------------------- flaws per 1000
square-inches of
inside surface
area in the
inspection volume
that are greater
than or equal to
TWEMIN TWEMAX TWEMIN and less
than TWEMAX. This
flaw density does
not include
underclad cracks
in forgings.
------------------------------------------------------------------------
0............................... 0.075............. No Limit
0.075........................... 0.375............. 8.05
0.125........................... 0.375............. 3.15
0.175........................... 0.375............. 0.85
0.225........................... 0.375............. 0.29
0.275........................... 0.375............. 0.08
0.325........................... 0.375............. 0.01
0.375........................... Infinite.......... 0.00
------------------------------------------------------------------------
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
------------------------------------------------------------------------
[[Page 29]]
Table 7--Threshold Values for the Outlier Deviation Test
[Significance Level = 1%]
------------------------------------------------------------------------
Second largest Largest allowable
Number of available allowable normalized normalized residual
data points (n) 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 28th day of December 2009.
For the Nuclear Regulatory Commission.
Andrew L. Bates,
Acting Secretary of the Commission.
[FR Doc. E9-31146 Filed 12-31-09; 8:45 am]
BILLING CODE 7590-01-P