[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.

---------------------------------------------------------------------------

[[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