Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 13-29 [E9-31146]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 1 (Monday, January 4, 2010)]
[Rules and Regulations]
[Pages 13-29]
From the Federal Register Online via the Government Printing 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 https://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 
Carol.Gallagher@nrc.gov.
    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 https://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain 
entry into ADAMS, which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the NRC's PDR 
reference staff at 1-800-397-4209, or (301) 415-4737, or by e-mail to 
PDR.Resource@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Ms. Veronica M. Rodriguez, Office of 
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; telephone (301) 415-3703; e-mail: 
Veronica.Rodriguez@nrc.gov, Mr. Matthew Mitchell, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone (301) 415-1467; e-mail: Matthew.Mitchell@nrc.gov, 
or Mr. Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 251-
7631; e-mail: Mark.Kirk@nrc.gov.

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

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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 https://www.regulations.gov, Docket number NRC-2007-0008.
    NRC's Electronic Reading Room (ERR). The NRC's public electronic 
reading room is located at 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