[Federal Register Volume 73, Number 155 (Monday, August 11, 2008)]
[Proposed Rules]
[Pages 46557-46569]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-18429]


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Proposed Rules
                                                Federal Register
________________________________________________________________________

This section of the FEDERAL REGISTER contains notices to the public of 
the proposed issuance of rules and regulations. The purpose of these 
notices is to give interested persons an opportunity to participate in 
the rule making prior to the adoption of the final rules.

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Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / 
Proposed Rules

[[Page 46557]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AI01
[NRC-2007-0008]


Alternate Fracture Toughness Requirements for Protection Against 
Pressurized Thermal Shock Events

AGENCY: Nuclear Regulatory Commission.

ACTION: Supplemental Proposed Rule.

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

SUMMARY: The Nuclear Regulatory Commission (NRC) is considering the 
adoption of provisions regarding applicability of the rule and new 
provisions regarding procedures to perform surveillance data checks 
related to the updated fracture toughness requirements for protection 
against pressurized thermal shock (PTS) events for pressurized water 
reactor (PWR) pressure vessels. The NRC is considering these provisions 
as an alternative to the provisions previously noticed for public 
comment on October 3, 2007 (72 FR 56275).

DATES: Submit comments on this proposed rule by September 10, 2008. 
Submit comments on the information collection aspects on this proposed 
rule by September 10, 2008.

ADDRESSES: You may submit comments by any one of the following methods. 
Please include the following number RIN 3150-AI01 in the subject line 
of your comments. Comments submitted in writing or in electronic form 
will be made available for public inspection. Because your comments 
will not be edited to remove any identifying or contact information, 
the NRC cautions you against including any information in your 
submission that you do not want to be publicly disclosed.
    Federal e Rulemaking Portal: Go to http://www.regulations.gov and 
search for documents filed under Docket ID NRC-2007-0008. Address 
questions about NRC dockets to Carol Gallagher (301) 415-5905; e-mail 
<a href="/cdn-cgi/l/email-protection#084b697a6764264f696464696f6d7a48667a6b266f677e"><span class="__cf_email__" data-cfemail="387b594a5754167f595454595f5d4a78564a5b165f574e">[email protected]</span></a>.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
    E-mail comments to: <a href="/cdn-cgi/l/email-protection#e3b1968f868e82888a8d84cda08c8e8e868d9790a38d9180cd848c95"><span class="__cf_email__" data-cfemail="c99bbca5aca4a8a2a0a7aee78aa6a4a4aca7bdba89a7bbaae7aea6bf">[email protected]</span></a>. If you do not 
receive a reply e-mail confirming that we have received your comments, 
contact us directly at (301) 415-1966.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 am and 4:15 pm during Federal workdays. (Telephone 
(301) 415-1966).
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
(301) 415-1101.
    You can access publicly available documents related to this 
document using the following methods:
    NRC's Public Document Room (PDR): The public may examine publicly 
available documents at the NRC's PDR, Public File Area O-F21, One White 
Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR 
reproduction contractor will copy documents for a fee.
    NRC's Agencywide Document Access and Management System (ADAMS): 
Publicly available documents created or received at the NRC are 
available electronically at the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain 
entry into ADAMS, which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the NRC's PDR 
reference staff at 1-800-397-4209, or (301) 415-4737, or by e-mail to 
<a href="/cdn-cgi/l/email-protection#2a7a6e7804784f59455f58494f6a445849044d455c"><span class="__cf_email__" data-cfemail="38687c6a166a5d4b574d4a5b5d78564a5b165f574e">[email protected]</span></a>.

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: 
<a href="/cdn-cgi/l/email-protection#633506110c0d0a00024d310c07110a04160619230d11004d040c15"><span class="__cf_email__" data-cfemail="0751627568696e646629556863756e6072627d4769756429606871">[email protected]</span></a>, Mr. Barry Elliot, Office of Nuclear Reactor 
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; telephone (301) 415-2709; e-mail: <a href="/cdn-cgi/l/email-protection#d290b3a0a0abfc97bebebbbda692bca0b1fcb5bda4"><span class="__cf_email__" data-cfemail="57153625252e79123b3b3e38231739253479303821">[email protected]</span></a>, or Mr. 
Mark Kirk, Office of Nuclear Regulatory Research, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001; telephone (301) 415-
6015; e-mail: <a href="/cdn-cgi/l/email-protection#d69bb7a4bdf89dbfa4bd96b8a4b5f8b1b9a0"><span class="__cf_email__" data-cfemail="3e735f4c551075574c557e504c5d10595148">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

I. Introduction
II. Background
III. Discussion
IV. Responses to Comments on the Proposed Rule
V. Section-by-Section Analysis
VI. Specific Request for Comments
VII. Availability of Documents
VIII. Plain Language
IX. Voluntary Consensus Standards
X. Finding of No Significant Environmental Impact: Availability
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Act Certification
XIV. Backfit Analysis

I. Introduction

    The NRC published a proposed rule on alternate fracture toughness 
requirements for protection against Pressurized Thermal Shock (PTS) for 
public comments in the Federal Register on October 3, 2007 (72 FR 
56275). This rule provides new PTS requirements based on updated 
analysis methods. This action is desirable because the existing 
requirements are based on unnecessarily conservative probabilistic 
fracture mechanics analyses. This action would reduce regulatory burden 
for licensees, specifically those licensees that expect to exceed the 
existing requirements before the expiration of their licenses, while 
maintaining adequate safety. These new requirements would be utilized 
by any Pressurized Water Reactor (PWR) licensee as an alternative to 
complying with the existing requirements.
    During the development of the PTS final rule, the NRC determined 
that several changes to the proposed rule language may be needed to 
adequately address issues raised in stakeholder's comments. The NRC 
also determined, in response to a stakeholder comment, that the 
characteristics of advanced PWR designs were not considered in the 
technical analysis made for the proposed rule. The NRC does not have 
assurance that reactors that commence commercial power operation after 
the effective date of this rule will have operating characteristics and 
materials of fabrication similar to those evaluated as part of the 
technical basis for the proposed rule. Therefore, the NRC has concluded 
that it would be prudent to

[[Page 46558]]

limit the applicability and the use of Sec.  50.61a to currently-
operating plants only, and proposes to modify the applicability 
provisions of the proposed rule accordingly.
    Also, several stakeholders questioned the accuracy and validity of 
the generic embrittlement curves in the proposed rule. The NRC wants to 
ensure that the predicted values from the proposed embrittlement trend 
curves provide an adequate basis for implementation of the rule. 
Therefore, the NRC has continued to work on statistical procedures to 
identify deviations from generic embrittlement trends, such as those 
described in Sec.  50.61a(f)(6) of the proposed rule. Based on this 
work, the NRC is considering enhancing the procedure described in 
paragraph Sec.  50.61a(f)(6) to, among other things, detect signs from 
the plant- and heat-specific surveillance data of embrittlement trends 
that are not reflected by Equations 5, 6 and 7 of the rule that may 
emerge at high fluences.
    Because these proposed modifications may not represent a logical 
outgrowth from the October 2007 proposed rule's provisions, the NRC 
concludes that obtaining stakeholder feedback on the proposed 
alternative provisions through the use of a supplemental proposed rule 
is appropriate. As discussed in Section VI of this notice, the NRC will 
consider comments on Sec. Sec.  50.61a(b); (f)(6)(i) through 
(f)(6)(vi); Equations 10, 11, and 12 in Sec.  50.61a(g); and Tables 5, 
6, and 7 of this supplemental proposed rule. The NRC is also requesting 
comments on whether there should be additional language added to Sec.  
50.61a(e) to allow licensees to account for the effects of sizing 
errors. This supplemental proposed rule does not reflect other 
modifications or editorial and conforming changes that the NRC is 
considering to incorporate in the final rule as a result of the public 
comments on the October 2007 proposed rule.

II. Background

    PTS events are system transients in a PWR in which severe 
overcooling occurs coincident with high pressure. The thermal stresses 
are caused by rapid cooling of the reactor vessel inside surface, which 
combine with the stresses caused by high pressure. The aggregate effect 
of these stresses is an increase in the potential for fracture if a 
pre-existing flaw is present in a material susceptible to brittle 
failure. The ferritic, low alloy steel of the reactor vessel beltline 
adjacent to the core, where neutron radiation gradually embrittles the 
material over the lifetime of the plant, can be susceptible to brittle 
fracture.
    The PTS rule, described in Sec.  50.61, adopted on July 23, 1985 
(50 FR 29937), establishes screening criteria below which the potential 
for a reactor vessel to fail due to a PTS event is deemed to be 
acceptably low. The screening criteria effectively define a limiting 
level of embrittlement beyond which operation cannot continue without 
further plant-specific evaluation. Regulatory Guide (RG) 1.154, 
``Format and Content of Plant-Specific Pressurized Thermal Shock 
Analysis Reports for Pressurized Water Reactors,'' indicates that 
reactor vessels that exceed the screening criteria in Sec.  50.61 may 
continue to operate provided they can demonstrate a mean through-wall 
crack frequency (TWCF) from PTS-related events of no greater than 5 x 
10<SUP>-6</SUP> per reactor year.
    Any reactor vessel with materials predicted to exceed the screening 
criteria in Sec.  50.61 may not continue to operate without 
implementation of compensatory actions or additional plant-specific 
analyses unless the licensee receives an exemption from the 
requirements of the rule. Acceptable compensatory actions are neutron 
flux reduction, plant modifications to reduce PTS event probability or 
severity, and reactor vessel annealing, which are addressed in 
Sec. Sec.  50.61(b)(3), (b)(4), and (b)(7); and Sec.  50.66, 
``Requirements for Thermal Annealing of the Reactor Pressure Vessel.''
    Currently, no operating PWR reactor vessel is projected to exceed 
the Sec.  50.61 screening criteria before the expiration of its 40 year 
operating license. However, several PWR reactor vessels are approaching 
the screening criteria, while others are likely to exceed the screening 
criteria during their first license renewal periods.
    The NRC's Office of Nuclear Regulatory Research (RES) developed a 
technical basis that supports updating the PTS regulations. This 
technical basis concluded that the risk of through-wall cracking due to 
a PTS event is much lower than previously estimated. This finding 
indicated that the screening criteria in Sec.  50.61 are unnecessarily 
conservative and may impose an unnecessary burden on some licensees. 
Therefore, the NRC created a new rule, Sec.  50.61a, which provides 
alternate screening criteria and corresponding embrittlement 
correlations based on the updated technical basis. The NRC decided that 
providing a new section containing the updated screening criteria and 
updated embrittlement correlations would be appropriate because the 
Commission directed the NRC staff, in a Staff Requirements Memorandum 
(SRM) dated June 30, 2006, to prepare a rulemaking which would allow 
current PWR licensees to implement the new requirements of Sec.  50.61a 
or continue to comply with the current requirements of Sec.  50.61. 
Alternatively, the NRC could have revised Sec.  50.61 to include the 
new requirements, which could be implemented as an alternative to the 
current requirements. However, providing two sets of requirements 
within the same regulatory section was considered confusing and/or 
ambiguous as to which requirements apply to which licensees.
    The NRC published the proposed rulemaking on the alternate fracture 
toughness requirements for protection against PTS for public comment in 
the Federal Register on October 3, 2007 (72 FR 56275). The proposed 
rule provided an alternative to the current rule, which a licensee may 
choose to adopt. This prompted the NRC to keep the current requirements 
separate from the new alternative requirements. As a result, the 
proposed rule retained the current requirements in Sec.  50.61 for PWR 
licensees choosing not to implement the less restrictive screening 
limits, and presented new requirements in Sec.  50.61a as an 
alternative relaxation for PWR licensees.

III. Discussion

    The NRC published a proposed new rule, Sec.  50.61a (October 3, 
2007, 72 FR 56275), that would provide new PTS requirements based on 
updated analysis methods because the existing requirements are based on 
unnecessarily conservative probabilistic fracture mechanics analyses. 
Stakeholders' comments raised concerns related to the applicability of 
the rule and the accuracy and validity of the generic embrittlement 
curves. The NRC reconsidered the technical and regulatory issues in 
these areas and is considering adopting the modified provisions 
regarding the applicability of the rule and new provisions regarding 
procedures to perform surveillance data checks described in this 
supplemental proposed rule. The NRC will consider comments on 
Sec. Sec.  50.61a(b), (f)(6)(i) through (f)(6)(vi); Equations 10, 11 
and 12 in Sec.  50.61a(g); and Tables 5, 6, and 7 of this supplemental 
proposed rule. As described in Section VI of this notice, the NRC is 
also requesting comments on whether there should be additional language 
added to Sec.  50.61a(e) to allow licensees to account for the effects 
of sizing errors. The NRC will consider the October 2007 proposed rule, 
the supplemental proposed rule, and the comments received in response

[[Page 46559]]

to both, when deciding whether to adopt a final PTS rule.

Applicability of the Proposed Rule, Sec.  50.61a(b)

    The supplemental proposed rule differs from the proposed rule and 
from Sec.  50.61 in that it proposes to limit the use of Sec.  50.61a 
to currently operating plants only. It cannot be demonstrated, a 
priori, that reactors which commence commercial power operation after 
the effective date of this rule will have operating characteristics, in 
particular identified PTS event sequences and thermal-hydraulic 
responses, which are consistent with the reactors which were evaluated 
as part of the technical basis for this rule. Other factors, including 
materials of fabrication and welding methods, could also vary. Hence, 
the use of Sec.  50.61a would be limited to currently operating PWR 
facilities which are known to have characteristics consistent with 
those assumed in the technical basis. The NRC also proposes to allow 
the holder of the operating license for Watts Bar Unit 2 to adopt the 
requirements in Sec.  50.61a as this facility has operating 
characteristics consistent with those assumed in the technical basis. 
The NRC recognizes that licensees for reactors who commence commercial 
power operation after the effective date of this rule may, under the 
provisions of Sec.  50.12, seek an exemption from Sec.  50.61a(b) to 
apply this rule if a plant-specific basis analyzing their operating 
characteristics, materials of fabrications, and welding methods is 
provided.

Surveillance Data, Sec.  50.61a(f)

    Section 50.61a(f) of the proposed rule defines the process for 
calculating the values for the material properties (i.e. , 
RT<INF>MAX-X</INF>) for a particular reactor vessel. These values would 
be based on the vessel material's copper, manganese, phosphorus, and 
nickel weight percentages, reactor cold leg temperature, and fast 
neutron flux and fluence values, as well as the unirradiated nil-
ductility transition reference temperature (i.e., RT<INF>NDT</INF>).
    Section 50.61a(f) of the proposed rule included a procedure by 
which the RT<INF>MAX-X</INF> values, which are predicted for plant-
specific materials using a generic temperature shift (i.e., 
[Delta]T<INF>30</INF>) embrittlement trend curve, are compared with 
heat-specific surveillance data that are collected as part of 10 CFR 
Part 50, Appendix H surveillance programs. The purpose of this 
comparison is to assess how well the surveillance data are represented 
by the generic embrittlement trend curve. If the surveillance data are 
close (closeness is assessed statistically) to the generic 
embrittlement trend curve, then the predictions of this embrittlement 
trend curve are used. This is expected to normally be the case. 
However, if the heat-specific surveillance data deviate significantly, 
and non-conservatively, from the predictions of the generic 
embrittlement trend curve, this indicates that alternative methods 
(i.e., other than, or in addition to, the generic embrittlement trend 
curve) may be needed to reliably predict the temperature shift trends, 
and to estimate RT<INF>MAX-X</INF>, for the conditions being assessed. 
However, alternative methods for temperature shift prediction are not 
prescribed by Sec.  50.61a(f) of the proposed rule.
    Although standard and accepted procedures exist to assess the 
statistical significance of the differences between heat-specific 
surveillance data and the generic embrittlement trend curve, similarly 
standard and acceptable procedures are not available to assess the 
practical importance of such differences. The practical importance of 
statistically significant deviations is best assessed by licensees on a 
case-by-case basis, which would be submitted for the review of the 
Director of NRR, as prescribed by Sec.  50.61a(f).
    The method described in the proposed rulemaking to compare the 
heat-specific surveillance data collected as part of 10 CFR part 50, 
Appendix H surveillance programs to the generic temperature shift 
embrittlement trend curve included a single statistical test. This 
statistical test was set forth by Equations 9 and 10, and Table 5. This 
test determined if, on average, the temperature shift from the 
surveillance data was significantly higher than the temperature shift 
of the generic embrittlement trend curve. The NRC has determined that, 
while necessary, this single test is not sufficient to ensure that the 
temperature shift predicted by the embrittlement trend curve well 
represents the heat-specific surveillance data. Specifically, this 
single statistical test cannot determine if the temperature shift from 
the surveillance data shows a more rapid increase after significant 
radiation exposure than the progression predicted by the generic 
embrittlement trend curve. To address this potential deficiency, which 
could be particularly important during a plant's period of extended 
operation, the NRC added two more statistical tests in this 
supplemental proposed rulemaking, which are expressed by Equations 11 
and 12 and by Tables 6 and 7. Together, these two additional tests 
determine if the surveillance data from a particular heat show a more 
rapid increase after significant radiation exposure than the 
progression predicted by the generic embrittlement trend curve.
    The NRC documented the technical basis for the proposed alternative 
in the following reports: (1) ``Statistical Procedures for Assessing 
Surveillance Data for 10 CFR Part 50.61a,'' (ADAMS Accession No. 
ML081290654), and (2) ``A Physically Based Correlation of Irradiation 
Induced Transition Temperature Shifts for RPV Steel,'' (ADAMS Accession 
No. ML081000630).

IV. Responses to Comments on the Proposed Rule

    The NRC received 5 comment letters on the proposed 10 CFR 50.61a 
rule published on October 3, 2007 (72 FR 56275). The following 
paragraphs discuss those comments which are directly associated with 
the supplemental proposed rule's provisions on the applicability of the 
rule and surveillance data procedures. The remainder of the comments 
and the NRC responses will be provided in the Federal Register notice 
for the final rule.

Comments on the Applicability of the Proposed Rule

    Comment: The commenters stated that the rule, as written, is only 
applicable to the existing fleet of PWRs. The characteristics of 
advanced PWR designs were not considered in the analysis. The 
commenters suggested adding a statement to state that this rule is 
applicable to the current PWR fleet and not the new plant designs. 
[PWROG-5, EPRI-5]
    Response: The NRC agrees with the comment that this rule is only 
applicable to the existing fleet of PWRs. The NRC cannot be assured 
that reactors that commence commercial power operation after the 
effective date of this rule will have operating characteristics, in 
particular identified PTS event sequences and thermal-hydraulic 
responses, which are consistent with the reactors that were evaluated 
as part of the technical basis for Sec.  50.61a. Other factors, 
including materials of fabrication and welding methods, could also 
vary. Therefore, the NRC agrees with the commenters that it would be 
prudent to restrict the use of Sec.  50.61a to current plants. As a 
result of this comment, the NRC proposes to modify Sec.  50.61a(b) and 
the statement of considerations of the rule to reflect this position to 
limit the use of the rule to currently operating plants.

Comments on Surveillance Data

    Comment: The commenters stated that there is little added value in 
the requirement to assess the surveillance

[[Page 46560]]

data as a part of this rule because variability in data has already 
been accounted for in the derivation of the embrittlement correlation.
    The commenters also stated that there is no viable methodology for 
adjusting the projected [Delta]T<INF>30</INF> for the vessel based on 
the surveillance data. Any effort to make this adjustment is likely to 
introduce additional error into the prediction. Note that the 
embrittlement correlation described in the basis for the revised PTS 
rule (i.e., NUREG-1874) was derived using all of the currently 
available industry-wide surveillance data.
    In the event that the surveillance data does not match the 
[Delta]T<INF>30</INF> 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]T<INF>30</INF> 
prediction based on the best estimate chemical composition for the heat 
of the material is more reliable than a prediction based on a single 
set of surveillance measurements.
    The commenters suggested removing the requirement to assess 
surveillance data, including Table 5, of this rule. [PWROG-4, EPRI-4, 
NEI-2]
    Response: The NRC does not agree with the proposed change. The NRC 
believes that there is added value in the requirement to assess 
surveillance data. Although variability has been accounted for in the 
derivation of the embrittlement correlation, it is the NRC's view that 
the surveillance assessment required in Sec.  50.61a(f)(6) is needed to 
determine if the embrittlement for a specific heat of material in a 
reactor vessel is consistent with the embrittlement predicted by the 
embrittlement correlation.
    The commenters also assert that there is no viable methodology for 
adjusting the projected [Delta]T<INF>30</INF> 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]T<INF>30</INF> for the vessel based on the surveillance data, it 
is possible, on a case-specific basis, to justify adjustments to the 
generic [Delta]T<INF>30</INF> prediction. For this reason the rule does 
not specify a method for adjusting the [Delta]T<INF>30</INF> value 
based on surveillance data, but rather requires the licensee to propose 
a case-specific [Delta]T<INF>30</INF> adjustment procedure for review 
and approval from the Director. Although the commenters assert that it 
is possible that error could be introduced, it is the NRC view that 
appropriate plant-specific adjustments based upon available 
surveillance data may be necessary to project reactor pressure vessel 
embrittlement for the purpose of this rule.
    As the result of these public comments, the NRC has continued to 
work on statistical procedures to identify deviations from generic 
embrittlement trends, such as those described in Sec.  50.61a(f)(6) of 
the proposed rule. Based on this work, the NRC is considering further 
enhancing the procedure described in paragraph (f)(6) to, among other 
things, detect signs from the plant- and heat-specific surveillance 
data that may emerge at high fluences of embrittlement trends that are 
not reflected by Equations 5, 6, and 7. The empirical basis for the 
NRC's concern regarding the potential for un-modeled high fluence 
effects is described in documents located at ADAMS Accession Nos. 
ML081120253, ML081120289, ML081120365, ML081120380, and ML081120600. 
The technical basis for the enhanced surveillance assessment procedure 
is described in the document located at ADAMS Accession No. 
ML081290654.

V. Section-by-Section Analysis

    The following section-by-section analysis only discusses the 
modifications in the provisions related to the applicability of the 
rule and surveillance data procedures that the NRC is considering as an 
alternative in this supplemental proposed rule. The NRC is only seeking 
comments on these alternative provisions. This supplemental proposed 
rule does not reflect other modifications or editorial and conforming 
changes that the NRC is considering to incorporate as a result of the 
public comments on the proposed rule that were not discussed in this 
notice as they will be provided in the Federal Register notice for the 
final rule.

Proposed Sec.  50.61a(b)

    The proposed language for Sec.  50.61a(b) would establish the 
applicability of the rule. The NRC proposes to modify this paragraph to 
limit the use of this rule to currently-operating plants only.

Proposed Sec.  50.61a(f)(6)(i)

    The proposed language for Sec.  50.61a(f)(6)(i) would establish the 
requirements to perform data checks to determine if the surveillance 
data show a significantly different trend than what the embrittlement 
model in this rule predicts. The NRC proposes to modify Sec.  
50.61a(f)(6)(i)(B) to state that licensees would evaluate the 
surveillance for consistency with the embrittlement model by following 
the procedures specified by Sec. Sec.  50.61a(f)(6)(ii), (f)(6)(iii), 
and (f)(6)(iv) of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(ii)

    The proposed language for Sec.  50.61a(f)(6)(ii) would establish 
the requirements to perform an estimate of the mean deviation of the 
data set from the embrittlement model. The mean deviation for the data 
set would be compared to values given in Table 5 or Equation 10 of this 
section. The NRC proposes to modify this paragraph to state that the 
surveillance data analysis would follow the criteria in Sec. Sec.  
50.61a(f)(6)(v) and (f)(6)(vi) of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(iii)

    The NRC proposes to modify Sec.  50.61a(f)(6)(iii) to establish the 
requirements to estimate the slope of the embrittlement model residuals 
(i.e., the difference between the measured and predicted value for a 
specific data point). The licensee would estimate the slope using 
Equation 11 and compare this value to the maximum permissible value in 
Table 6, both from the supplemental proposed rule. This surveillance 
data analysis would follow the criteria in Sec. Sec.  50.61a(f)(6)(v) 
and (f)(6)(vi) of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(iv)

    The NRC proposes to modify Sec.  50.61a(f)(6)(iv) to establish the 
requirements to estimate an outlier deviation from the embrittlement 
model for the specific data set using Equations 8 and 12. The licensee 
would compare the normalized residuals to the allowable values in Table 
7 of the supplemental proposed rule. This surveillance data analysis 
would follow the criteria in Sec. Sec.  50.61a(f)(6)(v) and (f)(6)(vi) 
of the supplemental proposed rule.

Proposed Sec.  50.61a(f)(6)(v)

    The NRC proposes to add paragraph (f)(6)(v) to establish the 
criteria to be satisfied in order to calculate the 
[Delta]T<INF>30</INF> shift values.

Proposed Sec.  50.61a(f)(6)(vi)

    The NRC proposes to add paragraph (f)(6)(vi) to establish the 
actions to be taken by a licensee if the criteria in paragraph 
(f)(6)(v) of this section are not met. The licensee would need to 
submit an evaluation of the surveillance data and propose values for 
[Delta]T<INF>30</INF>, considering their plant-specific

[[Page 46561]]

surveillance data, for the review and approval by the Director. The 
licensee would need to submit an evaluation of each surveillance 
capsule removed from the vessel after the submittal of the initial 
application for review and approval by the Director no later than 2 
years after the capsule is withdrawn from the vessel.

Proposed Sec.  50.61a(g)

    The proposed language for Sec.  50.61a(g) would provide the 
necessary equations and variables required by the proposed changes in 
Sec.  50.61a(f)(6). The NRC proposes to modify Equation 10 to account 
for 1 percent of significance level. Equations 11 and 12 would be added 
to provide the means for estimating the slope and the outlier deviation 
from the embrittlement model.

Proposed Tables 5, 6, and 7

    Tables 5, 6, and 7 would provide values to be used in the proposed 
changes in Sec.  50.61a(f)(6). The NRC proposes to modify Table 5 to 
account for the use of a 1 percent of significance level. Tables 6 and 
7 would be added to provide the threshold values for the slope and the 
outlier deviation tests.

VI. Specific Request for Comments

    The NRC seeks comments on Sec. Sec.  50.61a(b), (f)(6)(i) through 
(f)(6)(vi); Equations 10, 11, and 12 in Sec.  50.61a(g), and Tables 5, 
6, and 7 of the supplemental proposed rule. The NRC is not seeking 
comments on any other provisions of the proposed Sec.  50.61a which 
remain unchanged from the October 2007 proposed rule. In addition, the 
NRC also requests comments on the following question:

Adjustments of the Inservice Inspection Volumetric Examination and Flaw 
Assessments

    The flaw sizes in Tables 2 and 3 are selected so that reactor 
vessels with flaw sizes less than or equal to those in the tables will 
have a TWCF less than or equal to 1 x 10<SUP>-6</SUP> per reactor year 
at the maximum permissible embrittlement. The NRC recognizes that the 
flaw sizes in these tables represent actual flaw dimensions while the 
results from the ASME Code examinations are estimated dimensions. The 
available information indicates that, for most flaw sizes in Tables 2 
and 3, qualified inspectors will oversize flaws. Comparing oversized 
flaws to the size and density distributions in Tables 2 and 3 is 
conservative and acceptable, but not necessary. Therefore, NRC is 
considering to permit flaw sizes to be adjusted to account for the 
effects of sizing error before comparing the estimated size and density 
distribution to the acceptable size and density distributions in Tables 
2 and 3. This would be accomplished by requiring licensees to base the 
methodology to account for the effects of sizing error on statistical 
data collected from ASME Code inspector qualification tests. An 
acceptable method would include a demonstration, that accounting for 
the effects of sizing error, is unlikely to result in accepting actual 
flaw size distribution that cause the TWCF to exceed the acceptance 
criteria. Adjusting flaw sizes to account for sizing error can change 
an unacceptable examination result into an acceptable result; further, 
collecting, evaluating, and using data from ASME Code inspector 
qualification tests will require extensive engineering judgment. 
Therefore, the methodology would have to be reviewed and approved by 
the Director of the NRC's Office of Nuclear Reactor Regulation (NRR) to 
ensure that the risk associated with PTS is acceptable. The NRC 
requests specific comments on whether there should be additional 
language added to 10 CFR 50.61a(e) to allow licensees to account for 
the effects of sizing errors.

VII. Availability of Documents

    The NRC is making the documents identified below available to 
interested persons through one or more of the following methods, as 
indicated.
    Public Document Room (PDR). The NRC Public Document Room is located 
at 11555 Rockville Pike, Rockville, Maryland 20852.
    Regulations.gov (Web). These documents may be viewed and downloaded 
electronically through the Federal eRulemaking Portal http://www.regulations.gov, Docket number NRC-2007-0008.
    NRC's Electronic Reading Room (ERR). The NRC's public electronic 
reading room is located at http://www.nrc.gov/reading-rm.html.

------------------------------------------------------------------------
          Document               PDR           Web          ERR (ADAMS)
------------------------------------------------------------------------
Federal Register Notice--           X      NRC-2007-0008     ML072750659
 Proposed Rule: Alternate
 Fracture Toughness
 Requirements for Protection
 Against Pressurized Thermal
 Shock Events (RIN 3150-
 AI01), 72 FR 56275, October
 3, 2007....................
Letter from Thomas P.               X      NRC-2007-0008     ML073521542
 Harrall, Jr., dated
 December 17, 2007,
 ``Comments on Proposed Rule
 10 CFR 50, Alternate
 Fracture Toughness
 Requirements for Protection
 Against Pressurized Thermal
 Shock Events, RIN 3150-
 AI01'' [Identified as Duke]
Letter from Jack Spanner,           X      NRC-2007-0008     ML073521545
 dated December 17, 2007,
 ``10 CFR 50.55a Proposed
 Rulemaking Comments RIN
 3150-AI01'' [Identified as
 EPRI]......................
Letter from James H. Riley,         X      NRC-2007-0008     ML073521543
 dated December 17, 2007,
 ``Proposed Rulemaking--
 Alternate Fracture
 Toughness Requirements for
 Protection Against
 Pressurized Thermal Shock
 Events (RIN 3150-AI01), 72
 FR 56275, October 3, 2007
 [Identified as NEI]........
Letter from Melvin L. Arey,         X      NRC-2007-0008     ML073521547
 dated December 17, 2007,
 ``Transmittal of PWROG
 Comments on the NRC
 Proposed Rule on Alternate
 Fracture Toughness
 Requirements for Protection
 Against Pressurized Thermal
 Shock Events'', RIN 3150-
 AI01, PA-MSC-0232
 [Identified as PWROG]......
Letter from T. Moser, dated         X      NRC-2007-0008     ML073610558
 December 17, 2007,
 ``Strategic Teaming and
 Resource Sharing (STARS)
 Comments on RIN 3150-AI01,
 Alternate Fracture
 Toughness Requirements for
 Protection against
 Pressurized Thermal Shock
 Events 72 FR 56275 (October
 3,2007) [Identified as
 STARS].....................
``Statistical Procedures for        X   ................     ML081290654
 Assessing Surveillance Data
 for 10 CFR Part 50.61a''...
``A Physically Based                X   ................     ML081000630
 Correlation of Irradiation
 Induced Transition
 Temperature Shifts for RPV
 Steel''....................
Supplemental Regulatory             X      NRC-2007-0008     ML081440673
 Analysis...................
Supplemental OMB Supporting         X      NRC-2007-0008     ML081440736
 Statement..................
Memo from J. Uhle, dated May        X   ................     ML081120253
 15, 2008, ``Embrittlement
 Trend Curve Development for
 Reactor Pressure Vessel
 Materials''................
Draft ``Technical Basis for         X   ................     ML081120289
 Revision of Regulatory
 Guide 1.99: NRC Guidance on
 Methods to Estimate the
 Effects of Radiation
 Embrittlement on the Charpy
 V-Notch Impact Toughness of
 Reactor Vessel Materials''.

[[Page 46562]]

 
``Comparison of the                 X   ................     ML081120365
 Predictions of RM-9 to the
 IVAR and RADAMO Databases''
Memo from M. Erickson Kirk,         X   ................     ML081120380
 dated December 12, 2007,
 ``New Data from Boiling
 Water Reactor Vessel
 Integrity Program (BWRVIP)
 Integrated Surveillance
 Project (ISP)''............
``Further Evaluation of High        X   ................     ML081120600
 Fluence Data''.............
------------------------------------------------------------------------

VIII. Plain Language

    The Presidential memorandum ``Plain Language in Government 
Writing'' published in June 10, 1998 (63 FR 31883), directed that the 
Government's documents be in clear and accessible language. The NRC 
requests comments on the proposed rule specifically with respect to the 
clarity and effectiveness of the language used. Comments should be sent 
to the NRC as explained in the ADDRESSES heading of this notice.

IX. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113, requires that Federal agencies use technical 
standards that are developed or adopted by voluntary consensus 
standards bodies unless using such a standard is inconsistent with 
applicable law or is otherwise impractical.
    The NRC determined that there is only one technical standard 
developed that could be utilized for characterizing the embrittlement 
correlations. That standard is the American Society for Testing and 
Materials (ASTM) standard E-900, ``Standard Guide for Predicting 
Radiation-Induced Temperature Transition Shift in Reactor Vessel 
Materials.'' This standard contains a different embrittlement 
correlation than that of this supplemental proposed rule. However, the 
correlation developed by the NRC has been more recently calibrated to 
available data. As a result, ASTM standard E-900 is not a practical 
candidate for application in the technical basis for the supplemental 
proposed rule because it does not represent the broad range of 
conditions necessary to justify a revision to the regulations.
    The ASME Code requirements are utilized as part of the volumetric 
examination analysis requirements of the supplemental proposed rule. 
ASTM Standard Practice E 185, ``Standard Practice for Conducting 
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor 
Vessels,'' is incorporated by reference in 10 CFR Part 50, Appendix H 
and utilized to determine 30-foot-pound transition temperatures. These 
standards were selected for use in the supplemental proposed rule based 
on their use in other regulations within 10 CFR Part 50 and their 
applicability to the subject of the desired requirements.
    The NRC will consider using a voluntary consensus standard in the 
final rule if an appropriate standard is identified in the public 
comment period for this supplemental proposed rule.

X. Finding of No Significant Environmental Impact: Availability

    The Commission has determined under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 10 
CFR Part 51, Subpart A, that this rule, if adopted, would not be a 
major Federal action significantly affecting the quality of the human 
environment and, therefore, an environmental impact statement is not 
required.
    The determination of this environmental assessment is that there 
will be no significant offsite impact to the public from this action. 
This determination was made as part of the proposed rulemaking issued 
on October 3, 2007 (72 FR 56275), and remains applicable to this 
supplemental proposed rulemaking.

XI. Paperwork Reduction Act Statement

    This supplemental proposed rule would contain new or amended 
information collection requirements that are subject to the Paperwork 
Reduction Act of 1995 (44 U.S.C. 3501, et seq). This supplemental 
proposed rule has been submitted to the Office of Management and Budget 
for review and approval of the information collection requirements.
    Type of submission, new or revision: Revision.
    The title of the information collection: 10 CFR Part 50, 
``Alternate Fracture Toughness Requirements for Protection against 
Pressurized Thermal Shock Events (10 CFR 50.61 and 50.61a)'' 
supplemental proposed rule.
    The form number if applicable: Not applicable.
    How often the collection is required: Collections would be 
initially required for PWR licensees utilizing the requirements of 10 
CFR 50.61a as an alternative to the requirements of 10 CFR 50.61. 
Collections would also be required, after implementation of the new 10 
CFR 50.61a, when any change is made to the design or operation of the 
facility that affects the calculated RT<INF>MAX-X</INF> value. 
Collections would also be required during the scheduled periodic 
ultrasonic examination of beltline welds.
    Who will be required or asked to report: Licensees of currently 
operating PWRs utilizing the requirements of 10 CFR 50.61a in lieu of 
the requirements of 10 CFR 50.61 would be subject to all of the 
proposed requirements in this rulemaking.
    An estimate of the number of annual responses: 2.
    The estimated number of annual respondents: 1.
    An estimate of the total number of hours needed annually to 
complete the requirement or request: 363 hours (253 hours annually for 
record keeping plus 110 hours annually for reporting).
    Abstract: The NRC is proposing to amend its regulations to provide 
updated fracture toughness requirements for protection against PTS 
events for PWR pressure vessels. The supplemental proposed rule would 
provide new PTS requirements based on updated analysis methods. This 
action is necessary because the existing requirements are based on 
unnecessarily conservative probabilistic fracture mechanics analyses. 
This action is expected to reduce regulatory burden for licensees, 
specifically those licensees that expect to exceed the existing 
requirements before the expiration of their licenses. These new 
requirements would be utilized by licensees of currently operating PWRs 
as an alternative to complying with the existing requirements.
    The NRC is seeking public comment on the potential impact of the 
information collections contained in this supplemental proposed rule 
and on the following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Estimate of burden?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the information collection be minimized, 
including the use of automated collection techniques?

[[Page 46563]]

    A copy of the OMB clearance package may be viewed free of charge at 
the NRC Public Document Room, One White Flint North, 11555 Rockville 
Pike, Room O-1F21, Rockville, MD 20852. The OMB clearance package and 
rule are available at the NRC worldwide Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html. The document will be 
available on the NRC home page site for 60 days after the signature 
date of this notice.
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
above issues, by September 10, 2008. Comments received after this date 
will be considered if it is practical to do so, but assurance of 
consideration cannot be given to comments received after this date. 
Comments submitted in writing or in electronic form will be made 
available for public inspection. Because your comments will not be 
edited to remove any identifying or contact information, the NRC 
cautions you against including any information in your submission that 
you do not want to be publicly disclosed. Comments submitted should 
reference Docket No. NRC-2007-0008. Comments can be submitted in 
electronic form via the Federal e-Rulemaking Portal at http://www.regulations.gov by search for Docket No. NRC-2007-0008. Comments 
can be mailed to NRC Clearance Officer, Russell Nichols (T-5F52), U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001. Questions 
about the information collection requirements may be directed to the 
NRC Clearance Officer, Russell Nichols (T-5 F52), U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, by telephone at (301) 
415-6874, or by e-mail to <a href="/cdn-cgi/l/email-protection#b0f9fef6fff3fffcfcf5f3e4e39ee2d5c3dfc5c2d3d5f0dec2d39ed7dfc6"><span class="__cf_email__" data-cfemail="f6bfb8b0b9b5b9babab3b5a2a5d8a493859983849593b6988495d8919980">[email protected]</span></a>. Comments can 
be mailed to the Desk Officer, Office of Information and Regulatory 
Affairs, NEOB-10202, (3150-0011), Office of Management and Budget, 
Washington, DC 20503, or by e-mail to <a href="/cdn-cgi/l/email-protection#4f012e3b272e2110056110093d2a360f20222d612a203f61282039"><span class="__cf_email__" data-cfemail="5b153a2f333a35041175041d293e221b343639753e342b753c342d">[email protected]</span></a>, or 
by telephone at (202) 395-7345.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
OMB control number.

XII. Regulatory Analysis

    The NRC has issued a supplemental regulatory analysis for this 
supplemental proposed rulemaking. The analysis examines the costs and 
benefits of the alternatives considered by the NRC. The NRC requests 
public comments on this supplemental draft regulatory analysis. 
Availability of the supplemental regulatory analysis is provided in 
Section VII of this notice. Comments on the supplemental draft 
regulatory analysis may be submitted to the NRC as indicated under the 
ADDRESSES heading of this notice.

XIII. Regulatory Flexibility Act Certification

    In accordance with the Regulatory Flexibility Act (5 U.S.C. 
605(b)), the NRC certifies that this rule would not, if promulgated, 
have a significant economic impact on a substantial number of small 
entities. This supplemental proposed rule would affect only the 
licensing and operation of currently operating nuclear power plants. 
The companies that own these plants do not fall within the scope of the 
definition of ``small entities'' set forth in the Regulatory 
Flexibility Act or the size standards established by the NRC (10 CFR 
2.810).

XIV. Backfit Analysis

    The NRC has determined that the requirements in this supplemental 
proposed rule would not constitute backfitting as defined in 10 CFR 
50.109(a)(1). Therefore, a backfit analysis has not been prepared for 
this proposed rule.
    The requirements of the current PTS rule, 10 CFR 50.61, would 
continue to apply to all PWR licensees and would not change as a result 
of this supplemental proposed rule. The requirements of the proposed 
PTS rule, including those in the supplemental proposed rule, would not 
be required, but could be utilized by PWR licensees with currently 
operating plants. Licensees choosing to implement the proposed PTS rule 
would be required to comply with its requirements as an alternative to 
complying with the requirements of the current PTS rule. Because the 
proposed PTS rule would not be mandatory for any PWR licensee, but 
rather could be voluntarily implemented, the NRC finds that this 
amendment would not constitute backfitting.

List of Subjects for 10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
protection, Intergovernmental relations, Nuclear power plants and 
reactors, Radiation protection, Reactor siting criteria, Reporting and 
recordkeeping requirements.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing to 
adopt the following amendments to 10 CFR part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    1. The authority citation for part 50 continues to read as follows:

    Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); sec. 651(e), Pub. 
L. 109-58, 119 Stat. 806-810 (42 U.S.C. 2014, 2021, 2021b, 2111).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42 
U.S.C. 5841). Section 50.10 also issued under secs. 101, 185, 68 
Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-
190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 
50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued 
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 
50.55a and appendix Q also issued under sec. 102, Pub. L. 91-190, 83 
Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued 
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42 
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued 
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

    2. Section 50.8(b) is revised to read as follows:


Sec.  50.8  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  50.30, 50.33, 50.34, 50.34a, 50.35, 
50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 
50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64, 50.65, 50.66, 
50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 
50.91, 50.120, and appendices A, B, E, G, H, I, J, K, M, N,O, Q, R, and 
S to this part.
* * * * *
    3. Section 50.61a is added to read as follows:


Sec.  50.61a  Alternate fracture toughness requirements for protection 
against pressurized thermal shock events.

    (a) Definitions. Terms in this section have the same meaning as 
those set

[[Page 46564]]

forth in 10 CFR 50.61(a), with the exception of the term ``ASME Code''.
    (1) ASME Code means the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code, Section III, Division I, ``Rules for 
the Construction of Nuclear Power Plant Components,'' and Section XI, 
Division I, ``Rules for Inservice Inspection of Nuclear Power Plant 
Components,'' edition and addenda and any limitations and modifications 
thereof as specified in Sec.  50.55a.
    (2) RTMAX	AW means the material property which characterizes the 
reactor vessel's resistance to fracture initiating from flaws found 
along axial weld fusion lines. RT<INF>MAX-AW</INF> 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. 
RT<INF>MAX-PL</INF> 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. 
RT<INF>MAX-FO</INF> 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. RT<INF>MAX-CW</INF> 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 
RT<INF>MAX-AW,</INF> RT<INF>MAX-PL,</INF> RT<INF>MAX-FO</INF>, or 
RT<INF>MAX-CW</INF> for a particular reactor vessel.
    (7) [phis]t means fast neutron fluence for neutrons with energies 
greater than 1.0 MeV. [phis]t is determined under the provisions of 
paragraph (g) of this section and has units of n/cm\2\.
    (8) [phis] means average neutron flux. [phis] is determined under 
the provisions of paragraph (g) of this section and has units of n/
cm\2\/sec.
    (9) [Delta]T30 means the shift in the Charpy V-notch transition 
temperature produced by irradiation defined at the 30 ft-lb energy 
level. The [Delta]T<INF>30</INF> value is determined under the 
provisions of paragraph (g) of this section and has units of [deg]F.
    (10) Surveillance data means any data that demonstrates the 
embrittlement trends for the beltline materials, including, but not 
limited to, data from test reactors or surveillance programs at other 
plants with or without a surveillance program integrated under 10 CFR 
Part 50, Appendix H.
    (11) Tc means cold leg temperature under normal full power 
operating conditions, as a time-weighted average from the start of full 
power operation through the end of licensed operation. T<INF>c</INF> 
has units of [deg]F.
    (b) Applicability. Each licensee of a pressurized water nuclear 
power reactor, whose original operating license was issued prior to 
[EFFECTIVE DATE OF FINAL RULE], and the holder of any operating license 
issued under this part or part 54 for the Watts Bar Unit 2 facility, 
may utilize the requirements of this section as an alternative to the 
requirements of 10 CFR 50.61.
    (c) Request for Approval. Prior to implementation of this section, 
each licensee shall submit a request for approval in the form of a 
license amendment together with the documentation required by 
paragraphs (c)(1), (c)(2), and (c)(3) of this section for review and 
approval to the Director, Office of Nuclear Reactor Regulation 
(Director). The information required by paragraphs (c)(1), (c)(2), and 
(c)(3) of this section must be submitted for review and approval by the 
Director at least three years before the limiting RT<INF>PTS</INF> 
value calculated under 10 CFR 50.61 is projected to exceed the PTS 
screening criteria in 10 CFR 50.61 for plants licensed under this part.
    (1) Each licensee shall have projected values of RT<INF>MAX-X</INF> 
for each reactor vessel beltline material for the EOL fluence of the 
material. The assessment of RT<INF>MAX-X</INF> values must use the 
calculation procedures given in paragraphs (f) and (g) of this section, 
except as provided in paragraphs (f)(6) and (f)(7) of this section. The 
assessment must specify the bases for the projected value of 
RT<INF>MAX-X</INF> for each reactor vessel beltline material, including 
the assumptions regarding future plant operation (e.g., core loading 
patterns, projected capacity factors, etc.); the copper (Cu), 
phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor 
cold leg temperature (T<INF>C</INF>); and the neutron flux and fluence 
values used in the calculation for each beltline material.
    (2) Each licensee shall perform an examination and an assessment of 
flaws in the reactor vessel beltline as required by paragraph (e) of 
this section. The licensee shall verify that the requirements of 
paragraphs (e)(1) through (e)(3) have been met and submit all 
documented indications and the neutron fluence map required by 
paragraph (e)(1)(iii) to the Director in its application to utilize 10 
CFR 50.61a. If analyses performed under paragraph (e)(4) of this 
section are used to justify continued operation of the facility, 
approval by the Director is required prior to implementation.
    (3) Each licensee shall compare the projected RT<INF>MAX-X</INF> 
values for plates, forgings, axial welds, and circumferential welds to 
the PTS screening criteria for the purpose of evaluating a reactor 
vessel's susceptibility to fracture due to a PTS event. If any of the 
projected RT<INF>MAX-X</INF> values are greater than the PTS screening 
criteria in Table 1 of this section, then the licensee may propose the 
compensatory actions or plant-specific analyses as required in 
paragraphs (d)(3) through (d)(7) of this section, as applicable, to 
justify operation beyond the PTS screening criteria in Table 1 of this 
section.
    (d) Subsequent Requirements. Licensees who have been approved to 
utilize 10 CFR 50.61a under the requirements of paragraph (c) of this 
section shall comply with the requirements of this paragraph.
    (1) Whenever there is a significant change in projected values of 
RT<INF>MAX-X</INF>, such that the previous value, the current value, or 
both values, exceed the screening criteria prior to the expiration of 
the plant operating license; or upon the licensee's request for a 
change in the expiration date for operation of the facility; a 
reassessment of RT<INF>MAX-X</INF> values documented consistent with 
the requirements of paragraph (c)(1) and (c)(3) of this section must be 
submitted for review and approval to the Director. If the Director does 
not approve the assessment of RT<INF>MAX-X</INF> values, then the 
licensee shall perform the actions required in paragraphs (d)(3) 
through (d)(7) of this section, as necessary, prior to operation beyond 
the PTS screening criteria in Table 1 of this section.
    (2) Licensees shall determine the impact of the subsequent flaw 
assessments required by paragraphs (e)(1)(i), (e)(1)(ii), (e)(2), and 
(e)(3) of this section and shall submit the assessment for review and 
approval to the Director within 120 days after completing a volumetric 
examination of reactor vessel beltline materials as required by Section 
XI of the ASME Code. If a licensee is required to implement paragraphs 
(e)(4) and (e)(5) of this section, a reanalysis in accordance with 
paragraphs (e)(4) and (e)(5) of this section is required within one 
year of the subsequent ASME Code inspection.
    (3) If the value of RT<INF>MAX-X</INF> is projected to exceed the 
PTS screening criteria, then the licensee shall implement those flux 
reduction

[[Page 46565]]

programs that are reasonably practicable to avoid exceeding the PTS 
screening criteria. The schedule for implementation of flux reduction 
measures may take into account the schedule for review and anticipated 
approval by the Director of detailed plant-specific analyses which 
demonstrate acceptable risk with RT<INF>MAX-X</INF> 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 RT<INF>MAX-X</INF> value for one or more reactor vessel 
beltline materials from exceeding the PTS screening criteria, then the 
licensee shall perform a safety analysis to determine what, if any, 
modifications to equipment, systems, and operation are necessary to 
prevent the potential for an unacceptably high probability of failure 
of the reactor vessel as a result of postulated PTS events if continued 
operation beyond the PTS screening criteria is to be allowed. In the 
analysis, the licensee may determine the properties of the reactor 
vessel materials based on available information, research results and 
plant surveillance data, and may use probabilistic fracture mechanics 
techniques. This analysis must be submitted to the Director at least 
three years before RT<INF>MAX-X</INF> 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 
RT<INF>MAX-X</INF> values in excess of the PTS screening criteria. The 
Director will consider factors significantly affecting the potential 
for failure of the reactor vessel in reaching a decision.
    (6) If the Director concludes, under paragraph (d)(5) of this 
section, that operation of the facility with RT<INF>MAX-X</INF> values 
in excess of the PTS screening criteria cannot be approved on the basis 
of the licensee's analyses submitted under paragraphs (d)(3) and (d)(4) 
of this section, then the licensee shall request a license amendment, 
and receive approval by the Director, prior to any operation beyond the 
PTS screening criteria. The request must be based on modifications to 
equipment, systems, and operation of the facility in addition to those 
previously proposed in the submitted analyses that would reduce the 
potential for failure of the reactor vessel due to PTS events, or on 
further analyses based on new information or improved methodology.
    (7) If the limiting RT<INF>MAX-X</INF> 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 RT<INF>MAX-X</INF> values 
accounting for the effects of annealing and subsequent irradiation.
    (e) Examination and Flaw Assessment Requirements. The volumetric 
examinations results evaluated under paragraphs (e)(1), (e)(2), and 
(e)(3) of this section must be acquired using procedures, equipment and 
personnel that have been qualified under the ASME Code, Section XI, 
Appendix VIII, Supplement 4 and Supplement 6.
    (1) The licensee shall verify that the indication density and size 
distributions within the ASME Code, Section XI, Appendix VIII, 
Supplement 4 inspection volume \1\ are within the flaw density and size 
distributions in Tables 2 and 3 of this section based on the test 
results from the volumetric examination. The allowable number of flaws 
specified in Tables 2 and 3 of this section represent a cumulative flaw 
size distribution for each ASME flaw size increment. The allowable 
number of flaws for a particular ASME flaw size increment represents 
the maximum total number of flaws in that and all larger ASME flaw size 
increments. The licensee shall also demonstrate that no flaw exceeds 
the size limitations specified in Tables 2 and 3 of this section.
---------------------------------------------------------------------------

    \1\ The ASME Code, Section XI, Appendix VIII, Supplement 4 weld 
volume is the weld volume from the clad-to-base metal interface to 
the inner 1.0 inch or 10 percent of the vessel thickness, whichever 
is greater.
---------------------------------------------------------------------------

    (i) The licensee shall determine the allowable number of weld flaws 
for the reactor vessel beltline by multiplying the values in Table 2 of 
this section by the total length of the reactor vessel beltline welds 
that were volumetrically inspected and dividing by 1000 inches of weld 
length.
    (ii) The licensee shall determine the allowable number of plate or 
forging flaws for their reactor vessel beltline by multiplying the 
values in Table 3 of this section by the total plate or forging surface 
area that was volumetrically inspected in the beltline plates or 
forgings and dividing by 1000 square inches.
    (iii) For each indication detected in the ASME Code, Section XI, 
Appendix VIII, Supplement 4 inspection volume, the licensee shall 
document the dimensions of the indication, including depth and length, 
the orientation of the indication relative to the axial direction, and 
the location within the reactor vessel, including its azimuthal and 
axial positions and its depth embedded from the clad-to-base metal 
interface. The licensee shall also document a neutron fluence map, 
projected to the date of license expiration, for the reactor vessel 
beltline clad-to-base metal interface and indexed in a manner that 
allows the determination of the neutron fluence at the location of the 
detected indications.
    (2) The licensee shall identify, as part of the examination 
required by paragraph (c)(2) of this section and any subsequent ASME 
Code, Section XI ultrasonic examination of the beltline welds, any 
indications within the ASME Code, Section XI, Appendix VIII, Supplement 
4 inspection volume that are located at the clad-to-base metal 
interface. The licensee shall verify that such indications do not open 
to the vessel inside surface using a qualified surface or visual 
examination.
    (3) The licensee shall verify, as part of the examination required 
by paragraph (c)(2) of this section and any subsequent ASME Code, 
Section XI ultrasonic examination of the beltline welds, all 
indications between the clad-to-base metal interface and three-eighths 
of the reactor vessel thickness from the interior surface are within 
the allowable values in ASME Code, Section XI, Table IWB-3510-1.
    (4) The licensee shall perform analyses to demonstrate that the 
reactor vessel will have a through-wall crack frequency (TWCF) of less 
than 1 x 10<SUP>-6</SUP> per reactor year if the ASME Code, Section XI 
volumetric examination required by paragraph (c)(2) or (d)(2) of this 
section indicates any of the following:
    (i) The indication density and size in the ASME Code, Section XI, 
Appendix VIII, Supplement 4 inspection volume is not within the flaw 
density and size limitations specified in Tables 2 and 3 of this 
section;
    (ii) Any indication in the ASME Code, Section XI, Appendix VIII, 
Supplement 4 inspection volume that is larger \2\ than

[[Page 46566]]

the sizes in Tables 2 and 3 of this section;
---------------------------------------------------------------------------

    \2\ Table 2 for the weld flaws is limited to flaw sizes that are 
expected to occur and were modeled from the technical basis 
supporting this rule. Similarly, Table 3 for the plate and forging 
flaws stops at the maximum flaw size modeled for these materials in 
the technical basis supporting this rule.
---------------------------------------------------------------------------

    (iii) There are linear indications that penetrate through the clad 
into the low alloy steel reactor vessel shell; or
    (iv) Any indications between the clad-to-base metal interface and 
three-eighths \3\ of the vessel thickness exceed the size allowable in 
ASME Code, Section XI, Table IWB-3510-1.
---------------------------------------------------------------------------

    \3\ Because flaws greater than three-eighths of the vessel wall 
thickness from the inside surface do not contribute to TWCF, flaws 
greater than three-eighths of the vessel wall thickness from the 
inside surface need not be analyzed for their contribution to PTS.
---------------------------------------------------------------------------

    (5) The analyses required by paragraph (e)(4) of this section must 
address the effects on TWCF of the known sizes and locations of all 
indications detected by the ASME Code, Section XI, Appendix VIII, 
Supplement 4 and Supplement 6 ultrasonic examination out to three-
eighths of the vessel thickness from the inner surface, and may also 
take into account other reactor vessel-specific information, including 
fracture toughness information.
    (f) Calculation of RTMAX	X values. Each licensee shall calculate 
RTMAX	X values for each reactor vessel beltline material using [phis]t. 
[phis]t must be calculated using an NRC-approved methodology.
    (1) The values of RTMAX	AW, RTMAX	PL, RTMAX	FO, and RTMAX	CW must 
be determined using Equations 1 through 4 of this section.
    (2) The values of [Delta]T30 must be determined using Equations 5 
through 7 of this section, unless the conditions specified in paragraph 
(f)(6)(vi) of this section are met, for each axial weld fusion line, 
plate, and circumferential weld fusion line. The [Delta]T30 value for 
each axial weld fusion line calculated as specified by Equation 1 of 
this section must be calculated for the maximum fluence ([phis]tFL) 
occurring along a particular axial weld fusion line. The [Delta]T30 
value for each plate calculated as specified by Equation 1 of this 
section must be calculated for [phis]tFL occurring along a particular 
axial weld fusion line. The [Delta]T30 value for each plate or forging 
calculated as specified by Equations 2 and 3 of this section are 
calculated for the maximum fluence ([phis]tMAX) occurring at the clad-
to-base metal interface of each plate or forging. In Equation 4, the 
[phis]tFL value used for calculating the plate, forging, and 
circumferential weld RTMAX	CW value is the maximum [phis]t occurring 
for each material along the circumferential weld fusion line.
    (3) The values of Cu, Mn, P, and Ni in Equations 6 and 7 of this 
section must represent the best estimate values for the material weight 
percentages. For a plate or forging, the best estimate value is 
normally the mean of the measured values for that plate or forging. For 
a weld, the best estimate value is normally the mean of the measured 
values for a weld deposit made using the same weld wire heat number as 
the critical vessel weld. If these values are not available, either the 
upper limiting values given in the material specification to which the 
vessel material was fabricated, or conservative estimates (mean plus 
one standard deviation) based on generic data \4\ as shown in Table 4 
of this section for P and Mn, must be used.
---------------------------------------------------------------------------

    \4\ Data from reactor vessels fabricated to the same material 
specification in the same shop as the vessel in question and in the 
same time period is an example of ``generic data.''
---------------------------------------------------------------------------

    (4) The values of RTNDT(U) must be evaluated according to the 
procedures in the ASME Code, Section III, paragraph NB-2331. If any 
other method is used for this evaluation, the licensee shall submit the 
proposed method for review and approval by the Director along with the 
calculation of RTMAX	X values required in paragraph (c)(1) of this 
section.
    (i) If a measured value of RTNDT(U) is not available, a generic 
mean value of RTNDT(U) for the class \5\ of material must be used if 
there are sufficient test results to establish a mean.
---------------------------------------------------------------------------

    \5\ The class of material for estimating RT<INF>NDT(U)</INF> 
must be determined by the type of welding flux (Linde 80, or other) 
for welds or by the material specification for base metal.
---------------------------------------------------------------------------

    (ii) The following generic mean values of RTNDT(U) must be used 
unless justification for different values is provided: 0 [deg]F for 
welds made with Linde 80 weld flux; and -56 [deg]F for welds made with 
Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes.
    (5) The value of Tc in Equation 6 of this section must represent 
the weighted time average of the reactor cold leg temperature under 
normal operating full power conditions from the beginning of full power 
operation through the end of licensed operation.
    (6) The licensee shall verify that an appropriate RTMAX	X value has 
been calculated for each reactor vessel beltline material. The licensee 
shall consider plant-specific information that could affect the use of 
Equations 5 though 7 of this section for the determination of a 
material's [Delta]T30 value.
    (i) The licensee shall evaluate the results from a plant-specific 
or integrated surveillance program if the surveillance data satisfy the 
criteria described in paragraphs (f)(6)(i)(A) and (f)(6)(i)(B) of this 
section:
    (A) The surveillance material must be a heat-specific match for one 
or more of the materials for which RTMAX	X is being calculated. The 30-
foot-pound transition temperature must be determined as specified by 
the requirements of 10 CFR Part 50, Appendix H.
    (B) If three or more surveillance data points measured at three or 
more different neutron fluences exist for a specific material, the 
licensee shall determine if the surveillance data show a significantly 
different trend than the embrittlement model predicts. This must be 
achieved by evaluating the surveillance data for consistency with the 
embrittlement model by following the procedures specified by paragraphs 
(f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of this section. If fewer than 
three surveillance data points exist for a specific material, then the 
embrittlement model must be used without performing the consistency 
check.
    (ii) The licensee shall estimate the mean deviation from the 
embrittlement model for the specific data set (i.e. , a group of 
surveillance data points representative of a given material). The mean 
deviation from the embrittlement model for a given data set must be 
calculated using Equations 8 and 9 of this section. The mean deviation 
for the data set must be compared to the maximum heat-average residual 
given in Table 5 or derived using Equation 10 of this section. The 
maximum heat-average residual is based on the material group into which 
the surveillance material falls and the number of surveillance data 
points. The surveillance data analysis must use the criteria in 
paragraphs (f)(6)(v) and (f)(6)(vi) of this section. For surveillance 
data sets with greater than 8 shift points, the maximum credible heat-
average residual must be calculated using Equation 10 of this section. 
The value of [sigma] used in Equation 10 of this section must be 
obtained from Table 5 of this section.
    (iii) The licensee shall estimate the slope of the embrittlement 
model residuals (estimated using Equation 8) plotted as a function of 
the base 10 logarithm of neutron fluence for the specific data set. The 
licensee shall estimate the T-statistic for this slope (TSURV) using 
Equation 11 and compare this value to the maximum permissible T-
statistic (TMAX) in Table 6. The surveillance data analysis must follow 
the criteria in paragraphs (f)(6)(v) and (f)(6)(vi) of this section. 
For surveillance data sets with greater than 15 shift points, the TMAX 
value must be calculated using Student's T distribution with a 
significance level ([alpha]) of 1 percent for a one-tailed test.

[[Page 46567]]

    (iv) The licensee shall estimate the two largest positive 
deviations (i.e. , outliers) from the embrittlement model for the 
specific data set using Equations 8 and 12. The licensee shall compare 
the largest normalized residual (r*) to the appropriate allowable value 
from the third column in Table 7 and the second largest normalized 
residual to the appropriate allowable value from the second column in 
Table 7. The surveillance data analysis must follow the criteria in 
paragraphs (f)(6)(v) and (f)(6)(vi) of this section.
    (v) The [Delta]T30 value must be determined using Equations 5, 6, 
and 7 of this section if all three of the following criteria are 
satisfied:
    (A) The mean deviation from the embrittlement model for the data 
set is equal to or less than the value in Table 5 or the value derived 
using Equation 10 of this section;
    (B) The T-statistic for the slope (TSURV) estimated using Equation 
11 is equal to or less than the maximum permissible T-statistic (TMAX) 
in Table 6; and
    (C) The largest normalized residual value is equal to or less than 
the appropriate allowable value from the third column in Table 7 and 
the second largest normalized residual value is equal to or less than 
the appropriate allowable value from the second column in Table 7.
    (vi) If any of the criteria described in paragraph (f)(6)(v) of 
this section are not satisfied, the licensee shall review the data base 
for that heat in detail, including all parameters used in Equations 4, 
5, and 6 of this section and the data used to determine the baseline 
Charpy V-notch curve for the material in an unirradiated condition. The 
licensee shall submit an evaluation of the surveillance data and shall, 
on the basis of this review, propose [Delta]T30 and RTMAX	X values, 
considering their plant-specific surveillance data, to be used for 
evaluation relative to the acceptance criteria of this rule. These 
evaluations shall be submitted for the review and approval by the 
Director at the time of the initial application. For each surveillance 
capsule removed from the reactor vessel after the submittal of the 
initial application, the licensee shall perform the analyses required 
by paragraph (f)(6) of this section. The analyses must be submitted for 
the review and approval by the Director in the form of a license 
amendment, and must be submitted no later than two years after the 
capsule is withdrawn from the vessel.
    (7) The licensee shall report any information that significantly 
improves the accuracy of the RTMAX	X value to the Director. Any value 
of RTMAX	X that has been modified as specified in paragraph (f)(6)(iv) 
of this section is subject to the approval of the Director when used as 
provided in this section.
    (g) Equations and variables used in this section.
    [GRAPHIC] [TIFF OMITTED] TP11AU08.016
    
    [GRAPHIC] [TIFF OMITTED] TP11AU08.017
    
    [GRAPHIC] [TIFF OMITTED] TP11AU08.018
    
    [GRAPHIC] [TIFF OMITTED] TP11AU08.019
    
    [GRAPHIC] [TIFF OMITTED] TP11AU08.020
    
    [GRAPHIC] [TIFF OMITTED] TP11AU08.021
    
    [GRAPHIC] [TIFF OMITTED] TP11AU08.022
    
Where:

P [wt-%] = phosphorus content
Mn [wt-%] = manganese content
Ni [wt-%] = nickel content
Cu [wt-%] = copper content
A = 1.140 x 10<SUP>-7</SUP> for forgings
    = 1.561 x 10<SUP>-7</SUP> for plates
    = 1.417 x 10<SUP>-7</SUP> for welds
B = 102.3 for forgings
    = 102.5 for plates in non-Combustion Engineering manufactured 
vessels
    = 135.2 for plates in Combustion Engineering vessels
    = 155.0 for welds
[phgr]te = [phgr] for [phgr] >= 4.39 x 10<SUP>10</SUP> n/
cm<SUP>2</SUP>/sec
    = [phgr]t x (4.39 x 10<INF>10</INF> / [phgr])<SUP>0.2595</SUP> 
for [phgr] < 4.39 x 10<SUP>10</SUP> n/cm<SUP>2</SUP>/sec

Where:

[phgr][n/cm<SUP>2</SUP>/sec] = average neutron flux
t[sec] = time that the reactor has been in full power operation
[phgr]t[n/cm<SUP>2</SUP>] = [phgr] x t
f(Cu<INF>e</INF>,P) = 0 for Cu <= 0.072
    = [Cu<INF>e</INF>-0.072]<SUP>0.668</SUP> for Cu > 0.072 and P <= 
0.008
    = [Cu<INF>e</INF>-0.072 + 1.359 x (P-0.008)]<SUP>0.668</SUP> for 
Cu > 0.072 and P > 0.008
and Cu<INF>e</INF> = 0 for Cu <= 0.072
    = MIN (Cu, maximum Cu<INF>e</INF>) for Cu > 0.072
and maximum Cu<INF>e</INF> = 0.243 for Linde 80 welds
    = 0.301 for all other materials
g(Cu<INF>e</INF>,Ni,[phgr]t<INF>e</INF>) = 0.5 + (0.5 x tanh 
{[log<INF>10</INF>([phgr]t<INF>e</INF>) + (1.1390 x Cu<INF>e</INF>)-
(0.448 x Ni)-18.120] / 0.629{time} )

[[Page 46568]]

[GRAPHIC] [TIFF OMITTED] TP11AU08.029

[GRAPHIC] [TIFF OMITTED] TP11AU08.030

[GRAPHIC] [TIFF OMITTED] TP11AU08.031

Where:

n = number of surveillance shift data points (sample size) in the 
specific data set
[sigma] = standard deviation of the residuals about the model for a 
relevant material group given in Table 5.
[GRAPHIC] [TIFF OMITTED] TP11AU08.023

Where:

m = the slope of a plot of all of the r values (estimated using 
Equation 8) versus the base 10 logarithm of the neutron fluence for 
each r value. The slope shall be estimated using the method of least 
squares.
se(m) = the least squares estimate of the standard-error associated 
with the estimated slope value m.
[GRAPHIC] [TIFF OMITTED] TP11AU08.024

Where:

r is defined using Equation 8 and [sigma] is given in Table 5.

                     Table 1--PTS Screening Criteria
------------------------------------------------------------------------
                                   RTMAX-X limits [[deg]F] for different
                                    vessel wall thicknesses \6\ (TWALL)
                                  --------------------------------------
 Product form and RTMAX-X values                  9.5in. <    10.5in. <
                                     TWALL <=     TWALL <=     TWALL <=
                                      9.5in.      10.5in.      11.5in.
------------------------------------------------------------------------
Axial Weld, RTMAX-AW.............          269          230          222
Plate, RTMAX-PL..................          356          305          293
Forging without underclad cracks,          356          305          293
 RTMAX-FO........................
Axial Weld and Plate, RTMAX-AW +           538          476          445
 RTMAX-PL........................
Circumferential Weld, RTMAX-CW\7\          312          277          269
Forging with underclad cracks,             246          241         239
 RTMAX-FO........................
------------------------------------------------------------------------
\6\ Wall thickness is the beltline wall thickness including the clad
  thickness.
\7\ RTPTS limits contributes 1 x 10-8 per reactor year to the ractor
  vessel TWCF.


                                   Table 2--Allowable Number of Flaws in Welds
----------------------------------------------------------------------------------------------------------------
                                                                       Allowable number of cumulative flaws per
  ASME section XI flaw size per IWA-    Range of Through-Wall Extent    1000 inches of weld length in the ASME
                 3200                       (TWE) of flaw [in.]          section XI Appendix VIII supplement 4
                                                                                   inspection volume
----------------------------------------------------------------------------------------------------------------
0.05.................................  0.025 <= TWE < 0.075.........  Unlimited.
0.10.................................  0.075 <= TWE < 0.125.........  166.70.
0.15.................................  0.125 <= TWE < 0.175.........  90.80.
0.20.................................  0.175 <= TWE < 0.225.........  22.82.
0.25.................................  0.225 <= TWE < 0.275.........  8.66.
0.30.................................  0.275 <= TWE < 0.325.........  4.01.
0.35.................................  0.325 <= TWE < 0.375.........  3.01.
0.40.................................  0.375 <= TWE < 0.425.........  1.49.
0.45.................................  0.425 <= TWE < 0.475.........  1.00.
----------------------------------------------------------------------------------------------------------------


                             Table 3--Allowable Number of Flaws in Plates or Forging
----------------------------------------------------------------------------------------------------------------
                                                                       Allowable number of cumulative flaws per
                                                                         1000 square inches of inside diameter
  ASME section XI flaw size per IWA-    Range of Through-Wall Extent   surface area in forgings or plates in the
                 3200                       (TWE) of flaw [in.]       ASME section XI Appendix VIII supplement 4
                                                                                 inspection volume \8\
----------------------------------------------------------------------------------------------------------------
0.05.................................  0.025 <= TWE < 0.075.........  Unlimited
0.10.................................  0.075 <= TWE < 0.125.........  8.049
0.15.................................  0.125 <= TWE < 0.175.........  3.146

[[Page 46569]]

 
0.20.................................  0.175 <= TWE < 0.225.........  0.853
0.25.................................  0.225 <= TWE < 0.275.........  0.293
0.30.................................  0.275 <= TWE < 0.325.........  0.0756
0.35.................................  0.325 <= TWE < 0.375.........  0.0144
----------------------------------------------------------------------------------------------------------------
\8\ Excluding underclad cracks in forgings.


 Table 4--Conservative Estimates for Chemical Element Weight Percentages
------------------------------------------------------------------------
                   Materials                         P            Mn
------------------------------------------------------------------------
Plates........................................        0.014         1.45
Forgings......................................        0.016         1.11
Welds.........................................        0.019         1.63
------------------------------------------------------------------------


 Table 5--Maximum Heat-Average Residual [[deg]F] for Relevant Material Groups by Number of Available Data Points
                                            [Significance level = 1%]
----------------------------------------------------------------------------------------------------------------
                                                                       Number of available data points
                 Material group                    [sigma] -----------------------------------------------------
                                                  [[deg]F]     3        4        5        6        7        8
----------------------------------------------------------------------------------------------------------------
Welds, for Cu > 0.072...........................      26.4     35.5     30.8     27.5     25.1     23.2     21.7
Plates, for Cu > 0.072..........................      21.2     28.5     24.7     22.1     20.2     18.7     17.5
Forgings, for Cu > 0.072........................      19.6     26.4     22.8     20.4     18.6     17.3     16.1
Weld, Plate or Forging, for Cu <= 0.072.........      18.6     25.0     21.7     19.4     17.7     16.4     15.3
----------------------------------------------------------------------------------------------------------------


            Table 6--TMAX Values for the Slope Deviation Test
                        [Significance level = 1%]
------------------------------------------------------------------------
            Number of available data points (n)                  TMAX
------------------------------------------------------------------------
3..........................................................        31.82
4..........................................................         6.96
5..........................................................         4.54
6..........................................................         3.75
7..........................................................         3.36
8..........................................................         3.14
9..........................................................         3.00
10.........................................................         2.90
11.........................................................         2.82
12.........................................................         2.76
14.........................................................         2.68
15.........................................................         2.65
------------------------------------------------------------------------


 Table 7--Threshold Values for the Outlier Deviation Test (Significance
                               Level = 1%)
------------------------------------------------------------------------
                                                   Second
                                                  largest      Largest
                                                 allowable    allowable
      Number of available data points (n)        normalized   normalized
                                                  residual     residual
                                                 value (r*)   value (r*)
------------------------------------------------------------------------
3.............................................         1.55         2.71
4.............................................         1.73         2.81
5.............................................         1.84         2.88
6.............................................         1.93         2.93
7.............................................         2.00         2.98
8.............................................         2.05         3.02
9.............................................         2.11         3.06
10............................................         2.16         3.09
11............................................         2.19         3.12
12............................................         2.23         3.14
13............................................         2.26         3.17
14............................................         2.29         3.19
15............................................         2.32         3.21
------------------------------------------------------------------------


    Dated at Rockville, Maryland, this 24th day of July 2008.

    For the Nuclear Regulatory Commission.
R.W. Borchardt,
Executive Director for Operations.
 [FR Doc. E8-18429 Filed 8-8-08; 8:45 am]
BILLING CODE 7590-01-P


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