[Federal Register Volume 72, Number 191 (Wednesday, October 3, 2007)]
[Proposed Rules]
[Pages 56275-56287]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 07-4887]


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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. 72, No. 191 / Wednesday, October 3, 2007 / 
Proposed Rules

[[Page 56275]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AI01


Alternate Fracture Toughness Requirements for Protection Against 
Pressurized Thermal Shock Events

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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

SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend 
its regulations to provide updated fracture toughness requirements for 
protection against pressurized thermal shock (PTS) events for 
pressurized water reactor (PWR) pressure vessels. The proposed rule 
would provide 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 voluntarily utilized by any PWR 
licensee as an alternative to complying with the existing requirements.

DATES: Submit comments by December 17, 2007. Submit comments specific 
to the information collection aspects of this rule by November 2, 2007. 
Comments received after these dates will be considered if it is 
practical to do so, but assurance of consideration cannot be given to 
comments received after these dates.

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 on rulemakings submitted in writing or 
in electronic form will be made available for public inspection. 
Because your comment 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.
    Submit comments via the Federal e-Rulemaking Portal http://www.regulations.gov. 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#e9baacaab0a9879b8ac78e869f"><span class="__cf_email__" data-cfemail="4b180e08120b253928652c243d">[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. Address questions about our rulemaking Web 
site to Carol Gallagher (301) 415-5905; E-mail <a href="/cdn-cgi/l/email-protection#9ddedcdaddf3effeb3faf2eb"><span class="__cf_email__" data-cfemail="f6b5b7b1b6988495d8919980">[email protected]</span></a>.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 a.m. and 4:15 p.m. Federal workdays (telephone 
(301) 415-1966).
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
(301) 415-1101.
    You may submit comments on the information collections by the 
methods indicated in the Paperwork Reduction Act Statement.
    Publicly available documents related to this rulemaking may be 
viewed electronically on the public computers located at the NRC's 
Public Document Room (PDR), O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, MD 20852-2738. The PDR reproduction 
contractor will copy documents for a fee.
    Publicly available documents created or received at the NRC after 
November 1, 1999, are available electronically at the NRC's Electronic 
Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this 
site, the public can gain entry into the NRC's Agencywide Document 
Access and Management System (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 PDR Reference staff at 1-800-397-4209, 301-415-4737, or 
by e-mail to <a href="/cdn-cgi/l/email-protection#bbebffe9fbd5c9d895dcd4cd"><span class="__cf_email__" data-cfemail="7a2a3e283a140819541d150c">[email protected]</span></a>.

FOR FURTHER INFORMATION CONTACT: Mr. George Tartal, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone (301) 415-0016; e-mail: <a href="/cdn-cgi/l/email-protection#d4939980e594baa6b7fab3bba2"><span class="__cf_email__" data-cfemail="1a5d574e2b5a746879347d756c">[email protected]</span></a>, or 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#24666e61644a56470a434b52"><span class="__cf_email__" data-cfemail="1d5f57585d736f7e337a726b">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

Table of Contents

I. Background
II. Section-by-Section Analysis
III. Agreement State Compatibility
IV. Availability of Documents
V. Plain Language
VI. Voluntary Consensus Standards
VII. Finding of No Significant Environmental Impact: Environmental 
Assessment
VIII.Paperwork Reduction Act Statement
IX. Regulatory Analysis
X. Regulatory Flexibility Certification
XI. Backfit Analysis

I. Background

    Pressurized thermal shock events are system transients in a 
pressurized water reactor (PWR) in which severe overcooling occurs 
coincident with high pressure. The thermal stresses caused by rapid 
cooling of the reactor vessel inside surface combine with the stresses 
caused by high pressure. The aggregate effect of these stresses is an 
increase in the potential for fracture if a preexisting 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 may be such a material.
    The toughness of ferritic reactor vessel materials is characterized 
by a ``reference temperature for nil ductility transition'' 
(RT<INF>NDT</INF>). RT<INF>NDT</INF> is referred to as a ductile-to-
brittle transition temperature. At temperatures below RT<INF>NDT</INF> 
fracture occurs very rapidly, by cleavage, a behavior referred to as 
``brittle.'' As temperatures increase above RT<INF>NDT</INF>, 
progressively larger amounts of deformation occur before rapid cleavage 
fracture occurs. Eventually, at temperatures above approximately 
RT<INF>NDT</INF> + 60 [deg]F, there is no longer adequate stress 
intensification to promote cleavage and fracture occurs by the slower 
mechanism of micro-void initiation, growth, and coalescence into the 
crack, a behavior referred to as ``ductile.''
    At normal operating temperature, ferritic reactor vessel materials 
are usually tough. However, neutron

[[Page 56276]]

radiation embrittles the material over time, causing a shift in 
RT<INF>NDT</INF> to higher temperatures. Correlations based on test 
results for unirradiated and irradiated specimens have been developed 
to calculate the shift in RT<INF>NDT</INF> as a function of neutron 
fluence (the integrated neutron flux over a specified time of plant 
operation) for various material compositions. The value of RT<INF>NDT</INF> 
at a given time in a reactor vessel's life is used in fracture 
mechanics calculations to determine the probability that assumed pre-
existing flaws would propagate when the reactor vessel is stressed.
    The Pressurized Thermal Shock (PTS) rule, 10 CFR 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 the rule 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 10 CFR 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, other plant modifications to reduce PTS event 
probability or severity, and reactor vessel annealing, which are 
addressed in 10 CFR 50.61(b)(3), (b)(4), and (b)(7); and 10 CFR 50.66, 
respectively.
    No currently operating PWR reactor vessel is projected to exceed 
the 10 CFR 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.

Technical Basis for the Proposed Amendment

    The NRC's Office of Nuclear Regulatory Research (RES) has completed 
a research program to update the PTS regulations. The results of this 
research program conclude 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 10 CFR 50.61 are unnecessarily 
conservative and may impose an unnecessary burden on some licensees. 
Therefore, the NRC is proposing a new rule, 10 CFR 50.61a, which would 
provide alternative screening criteria and corresponding embrittlement 
correlations based on the updated technical basis. The updated 
embrittlement correlation is the projected increase in the Charpy V-
notch 30 ft-lb transition temperature for reactor vessel materials 
resulting from neutron radiation and is calculated using equations 5 
through 7 of the proposed rule. The proposed rule would be voluntary 
for all holders of a PWR operating license under 10 CFR part 50 or a 
combined license under 10 CFR part 52, although it is intended for 
licensees with reactor vessels that cannot demonstrate compliance with 
the more restrictive criteria in 10 CFR 50.61. The requirements of 10 
CFR 50.61 would continue to apply to licensees who choose not to 
implement 10 CFR 50.61a.
    The following two reports provide the technical basis for this 
rulemaking: (1) NUREG-1806, ``Technical Basis for Revision of the 
Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 
50.61): Summary Report,'' and (2) NUREG-1874, ``Recommended Screening 
Limits for Pressurized Thermal Shock (PTS).'' These reports summarize 
and reference several additional reports on the same topic. The updated 
technical basis indicates that, after 60 years of operation, the risk 
of reactor vessel failure due to a PTS event is much lower than 
previously estimated. The updated analyses were based on information 
from three currently operating PWRs. Because the severity of the risk-
significant transient classes (i.e., 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 concludes that 
the TWCF results and resultant RT-based screening criteria developed 
from their analysis of 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/
frequency characteristic of the three plants that were modeled in 
detail.
    The current guidance provided by Regulatory Guide 1.174, Revision 
1, ``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis,'' 
for large early release frequency (LERF) was used to relate the PTS 
screening criteria in 10 CFR 50.61a to an acceptable yearly limit of 1 
x 10<SUP>-6</SUP> per reactor year on reactor vessel TWCF. Although 
many post-through-wall cracking accident progressions are expected to 
lead only to core damage (which suggests a 1 x 10<SUP>-5</SUP> events 
per year limit on TWCF per Regulatory Guide 1.174), uncertainties in 
the accident progression analysis led to the recommendation of adopting 
the more conservative TWCF limit of 1 x 10<SUP>-6</SUP> per reactor 
year based on LERF.
    The updated technical basis 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 10 CFR 50.61 and the proposed rule are similar, the form of 
the revised embrittlement correlation differs substantially from the 
correlation in the existing 10 CFR 50.61. The correlation in 10 CFR 
50.61a 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.
    This proposed rule would differ from the current rule in that it 
would contain a requirement for licensees who choose to follow its 
requirements to analyze the results from the American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) 
Section XI in service inspection volumetric examinations. This 
requirement would be provided in paragraph (e) of the proposed rule. 
The examinations and analyses would confirm that the flaw density and 
size in the licensee's reactor vessel beltline are bounded by the flaw 
density and size utilized in the technical basis. The technical basis 
was developed using a flaw density, spatial distribution, and size 
distribution determined from a small amount of experimental data, as

[[Page 56277]]

well as from physical models and expert elicitation. The experimental 
data included 22,210 cubic inches of weld metal, 3845 cubic inches of 
plate, and 1650 cubic inches of clad. The experimental data were 
obtained from samples removed from reactor vessel materials from 
cancelled plants (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 proposed rule may be applied are consistent with 
those in the technical basis because the experimental data was obtained 
from a limited number of reactor vessels.
    Paragraph (g)(6)(ii)(c) of 10 CFR 50.55a requires licensees to 
implement Supplements 4 and 6 in Appendix VIII to ASME BPV Code Section 
XI after November 22, 2000. 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.
    The performance of inspectors who have gone through the Supplement 
4 qualification process has been documented in a paper by Becker 
(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). Analysis of the results reported in this paper indicates that an 
inspector using a Supplement 4 qualification procedure would have an 80 
percent probability of detecting a flaw with a through-wall extent of 
0.1 inch and would have an approximately 99 percent probability of 
detecting a flaw with a through-wall extent of 0.3 inch. Therefore, 
there is an 80 percent or greater probability of detecting a flaw that 
contributes to crack initiation from PTS events in reactor vessels with 
embrittlement conditions characteristic of 1 x 10<SUP>-6</SUP> per 
reactor-year TWCF when they are inspected using ASME BPV Code Section 
XI, Appendix VIII, Supplement 4 requirements.
    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 proposed rule would 
require licensees to determine if:
    (1) The indication density and size within the weld and base metal 
inservice inspection volume from the clad-to-base metal interface to 
the inner 1.0 inch or 10 percent of the vessel thickness are within the 
flaw density and size distributions that were used in the technical 
basis represented in Tables 2 and 3 in the proposed rule;
    (2) Any indications within the weld and base metal inservice 
inspection volume from the clad-to-base metal interface to the inner 
1.0 inch or 10 percent of the vessel thickness are larger than the 
sizes in Tables 2 and 3;
    (3) Any indications between the clad-to-base metal interface and 
three-eights of the vessel thickness exceed the size allowable in ASME 
BPV Code Section XI, Table IWB-3510-1; or
    (4) Any linear indications that penetrate through the clad into the 
welds or the adjacent base metal.
    The technical basis for the proposed rule concludes that flaws as 
small as 0.1 inch deep contribute to TWCF and that nearly all of the 
contributions come from flaws in the range below 1 inch deep for 
reactor vessels with embrittlement characteristics of TWCF equal to 1 x 
10<SUP>-6</SUP> per reactor year. The peak contribution comes from 
flaws between 0.1 and 0.2 inch deep, because that is the range that has 
the maximum combined effect from the number of flaws, which is 
decreasing with flaw size, and their susceptibility to brittle 
fracture, which is increasing with flaw size. For weld flaws that 
exceed the sizes in the table, the risk analysis indicates that a 
single flaw can be expected to contribute a significant fraction of the 
1 x 10<SUP>-6</SUP>/reactor-year limit on TWCF. Therefore, if a flaw of 
that size is found in a reactor vessel, it is important to more 
accurately assess if its size and location with respect to the local 
level of embrittlement challenge the regulatory limit.
    The technical basis for the proposed rule indicates that flaws 
buried deeper than 1 inch from the inner surface of the reactor vessel 
are not as susceptible to brittle fracture as similar size flaws 
located closer to the inner surface. Therefore, the proposed rule would 
not require the comparison of the density of such flaws, but still 
would require large flaws, if discovered, to be evaluated for 
contributions to TWCF if they are within the inner three-eights of the 
vessel thickness. This requirement would be provided in paragraph 
(e)(4)(iv) of the proposed rule. 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, this proposed rule would require these flaws to 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 proposed rule would 
represent the number of flaws in each size range that were derived from 
the technical basis. Table 2 for the weld flaws is limited to flaw 
sizes that are frequent enough to be expected to occur in most plants. 
Similarly, Table 3 for the plate and forging flaws stops at the maximum 
flaw size that was modeled for these materials in the technical basis. 
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 for that reactor 
vessel to exceed the regulatory limit.
    Surface cracks that penetrate through the stainless steel clad 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 an operating PWR reactor 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''). The observed cracks had 
a maximum depth into the base metal of approximately 6 mm (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 10 CFR 50.61. The cracking at Quad 
Cities Unit 2 was attributed to intergranular stress corrosion cracking 
(IGSCC) of the stainless steel cladding, which has not been observed in 
PWR reactor vessels, and hot cracking of the low alloy steel metal 
base. 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 proposed rule would require licensees to determine if 
cracks of this type exist in the beltline weld region at each ASME Code 
Section XI ultrasonic examination. This requirement would be provided 
in paragraph (e)(2) of the proposed rule.

Development of Tables 2 and 3 Flaw Density and Size Screening Criteria

    The ASME Code specifies that the dimension of flaws detected by 
nondestructive examination be

[[Page 56278]]

expressed to the nearest 0.05 inch for indications less than 1 inch. 
Hence, the examination results from the ASME Code volumetric 
examination will be reported in multiples of 0.05 inch with a range of 
<plus-minus>0.025 inch. Therefore, Tables 2 and 3 in the proposed rule 
describe the flaw density in multiples of 0.05 inch with a size range 
of <plus-minus>0.025 inch.
    The ASME Code standard for reporting flaw sizes did not match the 
size increments in the technical basis. Therefore, the NRC staff 
developed a procedure to distribute the flaws used in the technical 
basis into ASME Code-sized ranges. This is explained in greater detail 
in the NRC staff document ``Development of Flaw Size Distribution 
Tables for Draft Proposed Title 10 of the Code of Federal Regulations 
(10 CFR) 50.61a'' (refer to ADAMS accession number ML070950392).
    The values in Tables 2 and 3 of the proposed rule exceed the values 
for those size ranges that were developed from the laboratory analyses 
of the two reactor vessels. It was decided to allow licensees to use 
the Table 2 and 3 values instead of the values that would come from the 
laboratory results because it is still conservative to model all of the 
flaws as if they were the largest size for each of the ASME Code size 
ranges. In effect, some of the conservatism that was in the original 
risk modeling is being made available to licensees for demonstrating 
that the results of an individual plant's ASME Code examinations are 
consistent with the underlying technical basis.

Rulemaking Initiation

    In SECY-06-0124, dated May 26, 2006, the NRC staff presented a 
rulemaking plan to the Commission to amend fracture toughness 
requirements for PWRs. In this SECY paper, the NRC staff proposed four 
options for rulemaking. The NRC staff recommended Option 3, which would 
allow licensees to voluntarily implement the less restrictive screening 
limits based on the updated technical basis and insert the updated 
embrittlement correlation into 10 CFR 50.61 to maintain regulatory 
consistency and implement the best state-of-the-art embrittlement 
correlation in both 10 CFR 50.61 and 10 CFR 50.61a. This recommendation 
was based on providing the necessary relief to licensees that would 
otherwise expend considerable resources to justify continued plant 
operation beyond the screening criteria in 10 CFR 50.61 (via 
compensatory actions, plant-specific analyses, annealing or exemption), 
while also requiring all licensees to recalculate their embrittlement 
metric to ensure that all plants' analyses are consistent.
    In a Staff Requirements Memorandum (SRM) dated June 30, 2006, the 
Commission approved the initiation of the rulemaking as specified in 
Option 2 of the rulemaking plan. This option would require licensees to 
continue to meet the requirements of 10 CFR 50.61, which provides 
adequate protection against PTS events, without implementing the 
updated embrittlement correlation. For licensees whose reactor vessels 
do not meet the requirements of 10 CFR 50.61, Option 2 would allow 
licensees to voluntarily implement 10 CFR 50.61a which utilizes the 
less restrictive screening limits based on the updated technical basis 
as well as the updated embrittlement correlation. Accordingly, the 
proposed rule provides for a voluntary alternative to the current set 
of PTS requirements for any PWR licensee. The NRC considered requiring 
new plants to use the best available embrittlement correlation (i.e., 
the embrittlement correlation developed for the new rule). The NRC 
believes that such a requirement was not necessary to provide adequate 
protection of public health and safety. The NRC believes that imposing 
the existing 10 CFR 50.61, without modification, on new reactors would 
ensure that adequate protection concerns would be met. The NRC believes 
that the proposed rule's requirements should be a voluntary alternative 
available to new plants, if needed.
    In implementing the rulemaking plan, the proposed rule would 
provide a new section, 10 CFR 50.61a, for the new set of fracture 
toughness requirements. 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 to prepare a rulemaking which would allow current PWR 
licensees to implement the new requirements of 10 CFR 50.61a or 
continue to comply with the current requirements of 10 CFR 50.61. 
Alternatively, the NRC could have revised 10 CFR 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 
proposed rule would provide a voluntary alternative to the current 
rule, which further prompted the NRC to keep the current, mandatory 
requirements separate from the new, voluntarily-implemented 
requirements. As a result, the proposed new rule would retain the 
current requirements in 10 CFR 50.61 for PWR licensees choosing not to 
implement the less restrictive screening limits, and would present new 
requirements in 10 CFR 50.61a as a voluntary relaxation for any PWR 
licensee.

II. Section-by-Section Analysis

Section 50.61--Fracture Toughness Requirements for Protection Against 
Pressurized Thermal Shock Events

    Section 50.61 contains the current requirements for pressurized 
thermal shock screening limits and embrittlement correlations. 
Paragraph (b) of this section would be modified to reference the 
proposed new section, Sec.  50.61a, as a voluntary alternative to 
compliance with the requirements of Sec.  50.61. No changes are made to 
the current pressurized thermal shock 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

    Proposed new Sec.  50.61a would contain pressurized thermal shock 
screening limits based on updated probabilistic fracture mechanics 
analyses. This new section would provide similar requirements to that 
of Sec.  50.61, fracture toughness requirements for protection against 
pressurized thermal shock events for pressurized water nuclear power 
reactors. However, Sec.  50.61a would differ 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 proposed rule 
would require quantifying PTS reference temperatures (RT<INF>MAX-
X</INF>) for flaws along axial weld fusion lines, plates, forgings, and 
circumferential weld fusion lines, and comparing the quantified value 
against the RT<INF>MAX-X</INF> screening criteria. Although comparing 
quantified values to the screening criteria is also required by the 
current Sec.  50.61, the proposed Sec.  50.61a would provide screening 
criteria that vary depending on material product form and vessel wall 
thickness. Further, the embrittlement correlation and the method of 
calculation of RT<INF>MAX-X</INF> values in Sec.  50.61a would 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

[[Page 56279]]

surface-fitting techniques based on pattern recognition, understanding 
of mechanisms, and engineering judgement. The embrittlement database 
used for this analysis was derived primarily from the Power Reactor 
Embrittlement Data Base (PR-EDB) developed at Oak Ridge National 
Laboratory. The updated RT<INF>MAX-X</INF> estimation procedures 
provide a more realistic (compared to the existing regulation) method 
for estimating the fracture toughness of reactor vessel materials over 
the lifetime of the plant.
    Paragraph (a) would contain definitions for terms used in Sec.  
50.61a. It would also provide that terms defined in Sec.  50.61 also 
have the same meaning in Sec.  50.61a unless otherwise noted.
    Paragraph (b) would describe the applicability of Sec.  50.61a to 
PWRs as an alternative to the requirements of Sec.  50.61. The 
requirements of this section would provide a voluntarily-implemented 
alternative to the current requirements of Sec.  50.61 for any current 
PWR licensee or future holder of a PWR operating license or combined 
license.
    Paragraph (c) would set forth the requirements governing NRC 
approval of a licensee's use of Sec.  50.61a. The licensee would make 
the formal request to the NRC via a license amendment, and only upon 
approval of the license amendment by the NRC would a licensee be 
permitted to implement Sec.  50.61a. In the licensee's amendment 
request, the required information would include (a) calculating the 
values of RT<INF>MAX-X</INF> values as required by paragraph (c)(1), 
(b) examining and assessing flaws discovered by ASME Code inspections 
as required by paragraph (c)(2), and (c) comparing the RT<INF>MAX-X</INF> 
values against the applicable screening criteria as required by 
paragraph (c)(3). In doing so, the licensee would also be required to 
utilize paragraphs (e)(1) through (e)(3), paragraph (f), and paragraph 
(g) in order to perform the necessary calculations, comparisons, 
examinations, assessments, and analyses.
    Paragraph (d) would define the requirements for subsequent 
examinations and flaw assessments after initial approval to use Sec.  
50.61a has been obtained under the requirements of paragraph (c). It 
would also define the required compensatory measures or analyses to be 
taken if a licensee determines that the screening criteria will be 
exceeded. Paragraph (d)(1) would define the requirements for subsequent 
RT<INF>MAX-X</INF> assessments consistent with the requirements of 
paragraphs (c)(1) and (c)(3). Paragraph (d)(2) would define the 
requirements for subsequent examination and flaw assessments utilizing 
the requirements of paragraphs (e)(1), (e)(1)(i), (e)(1)(ii), (e)(2), 
and (e)(3). Paragraphs (d)(3) through (d)(7) would define the 
requirements for implementing compensatory measures or plant-specific 
analyses should the value of RT<INF>MAX-X</INF> be projected to exceed 
the PTS screening criteria in Table 1 of this section.
    Paragraph (e) would define the requirements for verifying that the 
PTS screening criteria in Sec.  50.61a are applicable to a particular 
reactor vessel. The proposed rule would require that verification be 
based on an analysis of test results from ultrasonic examination of the 
reactor vessel beltline materials required by Section XI of the ASME 
Code.
    Paragraph (e)(1) would establish cumulative limits on flaw density 
and size within the ASME Code, Section XI, Appendix VIII, Supplement 4 
inspection volume, which corresponds to a depth of approximately one 
inch from the clad-to-base metal interface. The allowable number of 
flaws provided in Tables 2 and 3 are cumulative values. If flaws exist 
in larger increments, the allowable number of flaws is the value in 
Table 2 or 3 for that increment minus the total number of flaws in all 
larger increments. Flaws in this inspection volume contribute 
approximately 97-99 percent to the TWCF at the screening limit.
    Paragraph (e)(1)(i) would describe the flaw density limits for 
welds.
    Paragraph (e)(1)(ii) would describe the flaw density limits for 
plates and forgings.
    Paragraph (e)(1)(iii) would describe the specific ultrasonic 
examination and neutron fluence information to be submitted to the NRC. 
The NRC would utilize this information to evaluate whether plant-
specific information gathered in accordance with this rule suggests 
that the NRC staff should generically re-examine the technical basis 
for the rule.
    Paragraph (e)(2) would require that licensees verify that no clad-
base metal interface flaws within the ASME Code, Section XI, Appendix 
VIII, Supplement 4 inspection volume open to the vessel inside surface. 
These types of flaws could have a substantial effect on the TWCF.
    Paragraph (e)(3) would establish limits on flaw density and size 
beyond the ASME Code, Section XI, Appendix VIII, Supplement 4 
inspection volume to three-eights of the reactor vessel thickness from 
the interior surface. Flaws in this inspection volume contribute 
approximately 1-3 percent to the TWCF at the screening criteria. Flaws 
exceeding this limit could affect the TWCF. Flaws greater than three-
eights of the reactor vessel thickness from the interior surface do not 
contribute to the TWCF at the screening limit.
    Paragraph (e)(4) would establish requirements to be met if flaws 
exceed the limits in (e)(1) and (e)(3) or open to the inside surface of 
the reactor vessel. This section requires an analysis to demonstrate 
the reactor vessel would have a TWCF of less than 1 x 10<SUP>-6</SUP> 
per reactor-year. The analysis could be a complete, plant-specific, 
probabilistic fracture mechanics analysis or could be a simplified 
analysis of flaw size, location and embrittlement to demonstrate that 
the actual flaws in the reactor vessel are not in locations that would 
cause the TWCF to be greater than 1 x 10<SUP>-6</SUP> per reactor-year. 
This paragraph would be required to be implemented if the requirements 
of (e)(1) through (e)(3) are not satisfied.
    Paragraph (e)(5) would describe the critical parameters to be 
addressed if flaws exceed the limits in (e)(1) and (e)(3) or if the 
flaws would open to the inside surface of the reactor vessel. This 
paragraph would be required to be implemented if the requirements of 
(e)(1) through (e)(3) are not satisfied.
    Paragraph (f) would define the process for calculating RT<INF>MAX-X</INF> 
values. These values would be 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 RT<INF>NDT</INF> of the product form in question.
    Paragraph (g) would provide the necessary equations and variables 
required by paragraph (f) of this section.
    Table 1 would provide the PTS screening criteria for comparison 
with the licensee's calculated RT<INF>MAX-X</INF> values. Tables 2 and 
3 would provide values to be used in paragraph (e) of this section. 
Tables 4 and 5 would provide values to be used in paragraph (f) of this 
section.

III. 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; 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

[[Page 56280]]

Atomic Energy Act or the provisions of Title 10 of the Code of Federal 
Regulations (10 CFR). 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, but does not confer 
regulatory authority on the State.

IV. Availability of Documents

    The following table lists documents relating to this rulemaking 
which are available to the public and how they may be obtained.
    Public Document Room (PDR). The NRC's Public Document Room is 
located at the NRC's headquarters at 11555 Rockville Pike, Rockville, 
MD 20852.
    NRC's Electronic Reading Room (ERR). The NRC's electronic reading 
room is located at http://www.nrc.gov/reading-rm.html.

----------------------------------------------------------------------------------------------------------------
                   Document                         PDR           Web                   ERR (ADAMS)
----------------------------------------------------------------------------------------------------------------
Regulatory Analysis..........................            X             X   ML070570383
OMB Supporting Statement.....................            X             X   ML070570446
SECY-06-0124, May 26, 2006, Rulemaking Plan              X   ............  ML060530624
 Request for Commission Approval.
SRM-SECY-06-0124, June 30, 2006, Staff                   X   ............  ML061810148
 Requirements--Commission Approval of
 Rulemaking Plan.
NUREG-1796, ``Safety Evaluation Report                   X   ............  ML043060581
 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''.
NUREG-1806, ``Technical Basis for Revision of            X   ............  ML061580318
 the Pressurized Thermal Shock (PTS)
 Screening Limits in the PTS Rule (10 CFR
 50.61): Summary Report''.
NUREG-1874, ``Recommended Screening Limits               X   ............  ML070860156
 for Pressurized Thermal Shock (PTS)''.
Regulatory Guide 1.154, ``Format and Content             X   ............  ML003740028
 of Plant-Specific Pressurized Thermal Shock
 Analysis Reports for Pressurized Water
 Reactors''.
Regulatory Guide 1.174, ``An Approach for                X   ............  ML023240437
 Using Probabilistic Risk Assessment in Risk-
 Informed Decisions on Plant-Specific Changes
 to the Licensing Basis''.
Memorandum from Elliot to Mitchell, dated                X   ............  ML070950392
 April 3, 2007, ``Development of Flaw Size
 Distribution Tables for Draft Proposed Title
 10 of the Code of Federal Regulations (10
 CFR) 50.61a''.
----------------------------------------------------------------------------------------------------------------

V. Plain Language

    The Presidential memorandum dated June 1, 1998, entitled ``Plain 
Language in Government Writing'' directed that the Government's writing 
be in plain language. This memorandum was published on June 10, 1998 
(63 FR 31883). 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 address listed under the 
ADDRESSES caption of the preamble of this document.

VI. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, Pub. 
L. 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 considered using 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 proposed rule. However, the correlation developed by RES 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 proposed rule because it does not represent the broad 
range of conditions necessary to justify a revision to the regulations.
    American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code requirements are utilized as part of the volumetric 
examination analysis requirements of the 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 50 Appendix H and utilized to determine 30-foot-
pound transition temperatures. These standards were selected for use in 
the proposed rule based on their use in other regulations within Part 
50 and their applicability to the subject of the desired requirements.
    The NRC will consider using other voluntary consensus standards if 
appropriate standards are identified.

VII. Finding of No Significant Environmental Impact: Environmental 
Assessment

    The Commission has determined under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 
Subpart A of 10 CFR part 51, 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 basis for this determination is as follows:

Environmental Impacts of the Action

    This environmental assessment focuses on those aspects of Sec.  
50.61a where there is a potential for an environmental impact. The NRC 
has concluded that there will be no significant radiological 
environmental impacts associated with implementation of the rule 
requirements for the following reasons:
    (1) Section 50.61a would maintain the same functional requirements 
for the facility as the existing PTS rule in Sec.  50.61 as a voluntary 
alternative to the existing rule. This proposed rule would establish 
screening criteria, limiting levels of embrittlement beyond which 
operation cannot continue without further plant-specific evaluation or 
modifications, as well as require calculation of the maximum 
embrittlement predicted at the end of the licensed period of operation. 
The screening criteria provide reasonable assurance that licensees 
operating below (predicted embrittlement less than) the screening 
criteria could endure a pressurized thermal shock event without 
fracture of vessel materials, thus assuring integrity of the reactor 
pressure vessel.
    (2) The new rule is risk-informed and in accordance with the NRC's 
1995 PRA policy statement and risk-informed regulation guidance. 
Sufficient safety margins are maintained to ensure that any potential 
increases in core damage frequency (CDF) and large early release 
frequency (LERF) resulting from

[[Page 56281]]

implementation of Sec.  50.61a are negligible.
    The action 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 action.
    With regard to potential nonradiological impacts, implementation of 
the rule requirements has no impact on the facility other than to 
provide a more realistic method of calculating PWR vessel fracture 
toughness with associated limits. Nonradiological plant effluents are 
not affected and there are no other environmental impacts. Therefore, 
the NRC concludes that there are no significant environmental impacts 
associated with the action.

Alternatives to the Action

    As an alternative to the rulemaking described above, the NRC 
considered not taking the action (i.e., the ``no-action'' alternative). 
Not adopting the more realistic and less conservative regulation would 
result in no change in environmental impacts for current PWRs or those 
that would be expected for future PWRs under 10 CFR 50.61.

Agencies and Persons Consulted

    The NRC staff developed the proposed rule and this environmental 
assessment. Under the NRC's stated policy, a copy of this environmental 
assessment will be provided to the state liaison officials as part of 
the publication of the proposed rule for public comment.

Conclusion

    On the basis of this environmental assessment, the NRC concludes 
that the action would not have a significant effect on the quality of 
the human environment. Accordingly, the NRC has determined not to 
prepare an environmental impact statement for the action.
    The determination of this environmental assessment is that no 
significant offsite impact to the public from this action would occur. 
However, the general public should note that the NRC is seeking public 
participation. Comments on any aspect of the environmental assessment 
may be submitted to the NRC as indicated under the ADDRESSES heading.
    The NRC has sent a copy of this proposed rule to every State 
Liaison Officer and requested their comments on the environmental 
assessment.

VIII. Paperwork Reduction Act Statement

    This 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 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 60.61 and 50.61a)'' 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 a voluntary alternative to the requirements of 10 CFR 
50.61. Collections would also be required, after voluntary 
implementation of the new Sec.  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: Any PWR licensee 
voluntarily 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: 264 hours (24 hours annually for 
recordkeeping plus 240 hours annually for reporting).
    Abstract: The NRC is proposing to amend its regulations to provide 
updated fracture toughness requirements for protection against 
pressurized thermal shock (PTS) events for pressurized water reactor 
(PWR) pressure vessels. The 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 
would 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 
voluntarily utilized by any PWR licensee as an alternative to complying 
with the existing requirements.
    The U.S. Nuclear Regulatory Commission is seeking public comment on 
the potential impact of the information collections contained in this 
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?
    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-1 F21, 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 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 November 2, 2007 to the Records and FOIA/Privacy 
Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, or by Internet electronic mail to 
<a href="/cdn-cgi/l/email-protection#7d34333b323e323131383e292e3d332f3e533a322b"><span class="__cf_email__" data-cfemail="054c4b434a464a494940465156454b57462b424a53">[email protected]</span></a> and to the Desk Officer, Office of Information and 
Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and 
Budget, Washington, DC 20503. 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. You may also 
comment by telephone at (202) 395-3087.

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.

[[Page 56282]]

IX. Regulatory Analysis

    The Commission has prepared a draft regulatory analysis on this 
proposed regulation. The analysis examines the costs and benefits of 
the alternatives considered by the Commission. The Commission requests 
public comments on this draft regulatory analysis. Availability of the 
regulatory analysis is provided in Section IV. Comments on the draft 
regulatory analysis may be submitted to the NRC as indicated under the 
ADDRESSES heading of this document.
    In addition, the Commission also requests public comments on the 
cost and benefit of requiring PWR licensees to revise their vessel 
analyses if the updated embrittlement correlation were imposed in 10 
CFR 50.61. This would differ from the proposed rule, which leaves the 
technical content of 10 CFR 50.61 unchanged.

X. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the 
Commission certifies that this rule would not, if promulgated, have a 
significant economic impact on a substantial number of small entities. 
This proposed rule would affect only the licensing and operation of 
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).

XI. Backfit Analysis

    The NRC has determined that the requirements in this proposed rule 
do 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 proposed rule. The requirements of the proposed PTS 
rule, 10 CFR 50.61a, would not be required, but could be voluntarily 
utilized, by any PWR licensee. Licensees choosing to implement the 
proposed PTS rule would be required to comply with its requirements as 
a voluntary 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 by 
any PWR licensee, the NRC finds that this amendment would not 
constitute backfitting.

List of Subjects in 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). Section 50.7 also 
issued under Pub. L. 95-601, sec. 10, 92 Stat. 2951 (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. 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 certifications required 
under Sec.  50.82(a)(1) have been submitted, the licensee shall have 
projected values of RT<INF>PTS</INF> or RT<INF>MAX-X</INF>, accepted by 
the NRC, for each reactor vessel beltline material for the EOL fluence 
of the material in accordance with this section or Sec.  50.61a. For a 
licensee choosing to comply with this section, the assessment of 
RT<INF>PTS</INF> must use the calculation procedures given in paragraph 
(c)(1) of this section, except as provided in paragraphs (c)(2) and 
(c)(3) of this section. The assessment must specify the bases for the 
projected value of RT<INF>PTS</INF> 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 RT<INF>PTS</INF>, or upon request for a change in the expiration 
date for operation of the facility.
* * * * *
    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 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.

[[Page 56283]]

    (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]T30 value is determined under the provisions of 
paragraph (g) of this section and has units of [deg]F.
    (10) Surveillance data means any data that demonstrates the 
embrittlement trends for the beltline materials, including, but not 
limited to, data from test reactors or surveillance programs at other 
plants with or without a surveillance program integrated under 10 CFR 
part 50, Appendix H.
    (11) TC means cold leg temperature under normal full power 
operating conditions, as a time-weighted average from the start of full 
power operation through the end of licensed operation. T<INF>C</INF> 
has units of [deg]F.
    (b) Applicability. Each holder of an operating license under this 
part or holder of a combined license under part 52 of this chapter of a 
pressurized water nuclear power reactor 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 10 CFR 
part 50 or 10 CFR part 52.
    (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 re-
assessment 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 re-analysis 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 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

[[Page 56284]]

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-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 through-wall crack frequency (TWCF) of less 
than 1x10<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 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-eights \3\ of the vessel thickness exceed the size allowable in 
ASME Code, Section XI, Table IWB-3510-1.
---------------------------------------------------------------------------

    \3\ 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 
indications 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.
    (f) Calculation of RT<INF>MAX-X</INF> values. Each licensee shall 
calculate RT<INF>MAX-X</INF> values for each reactor vessel beltline 
material using [phi]t. [phi]t must be calculated using an NRC-approved 
methodology.
    (1) The values of RT<INF>MAX-AW</INF>, RT<INF>MAX-PL</INF>, 
RT<INF>MAX-FO</INF>, and RT<INF>MAX-CW</INF> must be

[[Page 56285]]

determined using Equations 1 through 4 of this section.
    (2) The values of [Delta]T<INF>30</INF> must be determined using 
Equations 5 through 7 of this section, unless the conditions specified 
in paragraph (f)(6)(iv) of this section are met, for each axial weld 
fusion line, plate, and circumferential weld fusion line. The 
[Delta]T<INF>30</INF> value for each axial weld fusion line calculated 
as specified by Equation 1 of this section must be calculated for the 
maximum fluence ([phi]<INF>FL</INF>) occurring along a particular axial 
weld fusion line. The [Delta]T<INF>30</INF> value for each plate 
calculated as specified by Equation 1 of this section must be 
calculated for tFL occurring along a particular axial weld fusion line. 
The [Delta]T<INF>30</INF> value for each plate or forging calculated as 
specified by Equations 2 and 3 of this section are calculated for the 
maximum fluence ([phi]t<INF>MAX</INF>) occurring at the clad-to-base 
metal interface of each plate or forging. In Equation 4, the 
[phi]t<INF>FL</INF> value used for calculating the plate, forging, and 
circumferential weld RT<INF>MAX-CW</INF> value is the maximum [phi] 
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 specifications 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 RT<INF>NDT(u)</INF> 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 RT<INF>MAX-X</INF> values required in paragraph 
(c)(1) of this section.
    (i) If a measured value of RT<INF>NDT(u)</INF> is not available, a 
generic mean value of RT<INF>NDT(u)</INF> 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 RT<INF>NDT(u)</INF> 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 T<INF>c</INF> 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 RT<INF>MAX-X</INF> 
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]T<INF>30</INF> value.
    (i) The licensee shall evaluate the results from a plant-specific 
or integrated surveillance program if the surveillance data has been 
deemed consistent as judged by the following criteria:
    (A) The surveillance material must be a heat-specific match for one 
or more of the materials for which RT<INF>MAX-X</INF> is being 
calculated. The 30-foot-pound transition temperature must be determined 
as specified by the requirements of 10 CFR 50 Appendix H.
    (B) If three or more surveillance data points exist for a specific 
material, the surveillance data must be evaluated for consistency with 
the model in Equations 5, 6, and 7 as specified by paragraph (f)(6)(ii) 
of this section. If fewer than three surveillance data points exist for 
a specific material, then Equations 5, 6, and 7 of this section must be 
used without performing the consistency check.
    (ii) The licensee shall estimate the mean deviation from the model 
(Equations 5, 6 and 7 of this section) for the specific data set (i.e., 
a group of surveillance data points representative of a given 
material). The mean deviation from the 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 Equation 10 of this section and based on 
the material group into which the surveillance material falls and the 
number of available data points. The licensee shall determine, based on 
this comparison, if the surveillance data show a significantly 
different trend than the model predicts. The surveillance data analysis 
must follow the criteria in paragraphs (f)(6)(iii) through (f)(6)(iv) 
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 comply with Table 5 of this section.
    (iii) If the mean deviation from the model for the data set is 
equal to or less than the value in Table 5 or the value using Equation 
10 of this section, then the [Delta]T<INF>30</INF> value must be 
determined using Equations 5, 6, and 7 of this section.
    (iv) If the mean deviation from the model for the data set is 
greater than the value in Table 5 or the value using Equation 10 of 
this section, the [Delta]T<INF>30</INF> value must be determined using 
the surveillance data. If the mean deviation from the model for the 
data set is outside the limits specified in Equation 10 of this section 
or in Table 5 of this section, 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 its 
[Delta]T<INF>30</INF> and RT<INF>MAX-X</INF> values for review and 
approval by the Director no later than one year after the surveillance 
capsule is withdrawn from the reactor vessel.
    (7) The licensee shall report any information that significantly 
improves the accuracy of the RT<INF>MAX-X</INF> value to the Director. 
Any value of RT<INF>MAX-X</INF> 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.

Equation 1: RT<INF>MAX-AW</INF> = MAX {[RT<INF>NDT(u)-plate</INF> + 
[Delta]T<INF>30-plate</INF>([phi]t<INF>FL</INF>)],
    [RT<INF>NDT(u)-axialweld</INF> + 
[Delta]T<INF>30-axialweld</INF>([phi]t<INF>FL</INF>)]{time} 
Equation 2: RT<INF>MAX-PL</INF> = RT<INF>NDT(u)-plate</INF> + 
[Delta]T<INF>30-plate</INF>([phi]t<INF>MAX</INF>)
Equation 3: RT<INF>MAX-FO</INF> = RT<INF>NDT(u)-forging</INF> + 
[Delta]T<INF>30-forging</INF>([phi]t<INF>MAX</INF>)
Equation 4: RT<INF>MAX-CW</INF> = MAX {[RT<INF>NDT(u)-plate</INF> + 
[Delta]T<INF>30-plate</INF>([phi]t<INF>MAX</INF>)],
    [RT<INF>NDT(u)-circweld</INF> + 
[Delta]T<INF>30-circweld</INF>([phi]t<INF>MAX</INF>)],
    [RT<INF>NDT(u)-forging</INF> + 
[Delta]T<INF>30-forging</INF>([phi]t<INF>MAX</INF>)]{time} 
Equation 5: [Delta]T<INF>30</INF> = MD + CRP
Equation 6: MD = A [middot] (1 - 0.001718 [middot] T<INF>C</INF>) 
[middot] (1 + 6.13 [middot] P [middot] Mn<SUP>2.471</SUP>) [middot] 
[phi]t<INF>e</INF><SUP>0.5</SUP>
Equation 7: CRP = B [middot] (1 + 3.77 [middot] Ni<SUP>1.191</SUP>) 
[middot] f(Cu<INF>e</INF>,P) [middot] 
g(Cu<INF>e</INF>,Ni,[phi]t<INF>e</INF>) VVVVVVV


[[Page 56286]]


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

[phi]t<INF>e</INF> = [phi]t for [phi] greater than or equal to 4.39 x 
10<SUP>10</SUP> n/cm<SUP>2</SUP>/sec
    = [phi]t [middot] (4.39 x 10<SUP>10</SUP>/[phi])<SUP>0.2595</SUP> 
for [phi] less than 4.39 x 10<SUP>10</SUP> n/cm<SUP>2</SUP>/sec

Where:

[phi] [n/cm<SUP>2</SUP>/sec] = average neutron flux
t [sec] = time that the reactor has been in full power operation
[phi]t [n/cm<SUP>2</SUP>] = [phi] [middot] 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 [middot] (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,[phi]t<INF>e</INF>) = 0.5 + 0.5 [middot] 
tanh{[log<INF>10</INF>([phi]t<INF>e</INF>) + 1.1390 [middot] 
Cu<INF>e</INF> - 0.448 [middot] Ni - 18.120] / 0.629{time} 

Equation 8: Residual [reg] = measured [Delta]T<INF>30</INF> - predicted 
[Delta]T<INF>30</INF> (by Equations 5, 6, and 7)
Equation 9: Mean deviation for a data set of n data points =
[GRAPHIC] [TIFF OMITTED] TP03OC07.013

Equation 10: Maximum credible heat-average residual = 3[sigma]/
n<SUP>0.5</SUP>

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.

                                        Table 1.--PTS Screening Criteria
----------------------------------------------------------------------------------------------------------------
                                                            RT MAX-X limits [[deg]F] for different vessel wall
                                                                         thicknesses \6\ (TWALL)
             Product form and RT MAX-Values             --------------------------------------------------------
                                                                             9.5 in. < TWALL    10.5 in. < TWALL
                                                          TWALL <= 9.5 in.     <= 10.5 in.        <= 11.5 in.
----------------------------------------------------------------------------------------------------------------
Axial Weld RT<INF>MAX-AW</INF>....................................                269                230                222
Plate RT<INF>MAX-PL</INF>.........................................                356                305                293
Forging without underclad cracks RT<INF>MAX-FO</INF>..............                356                305                293
Axial Weld and Plate RT<INF>MAX-AW</INF> + RT<INF>MAX-PL</INF>...............                538                476                445
Circumferential Weld RT<INF>MAX-CW</INF> \7\......................                312                277                269
Forging with underclad cracks RT<INF>MAX-FO</INF>.................                246                241                239
----------------------------------------------------------------------------------------------------------------


                                  Table 2.--Allowable Number of Flaws in Welds
----------------------------------------------------------------------------------------------------------------
                                                                                Allowable number of cumulative
                                                                                 flaws per 1000 inches of weld
   ASME section XI flaw size per IWA-3200      Range of through-wall extent      length in the ASME section XI
                                                   (TWE) of flaw  (in.)           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
                                               Range of through-wall extent     inside diameter surface area in
   ASME section XI flaw size per IWA-3200           (TWE) of flaw (in.)         forgings or plates in the 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
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
----------------------------------------------------------------------------------------------------------------


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

[[Page 56287]]

 
Welds...........................              0.019                1.63
------------------------------------------------------------------------


Table 5.--Maximum Heat-Average Residual [[deg]F] for Relevant Material Groups by Number of Available Data Points
----------------------------------------------------------------------------------------------------------------
                                                                       Number of available data points
                 Material group                    [sigma] -----------------------------------------------------
                                                  [[deg]F]     3        4        5        6        7        8
----------------------------------------------------------------------------------------------------------------
Welds, for Cu > 0.072...........................      26.4     45.7     39.6     35.4     32.3     29.9     28.0
Plates, for Cu > 0.072..........................      21.2     36.7     31.8     28.4     26.0     24.0     22.5
Forgings, for Cu > 0.072........................      19.6     33.9     29.4     26.3     24.0     22.2     20.8
Weld, Plate or Forging, for Cu <= 0.072.........      18.6     32.2     27.9     25.0     22.8     21.1     19.7
----------------------------------------------------------------------------------------------------------------


    Dated at Rockville, Maryland, this 27th day of September 2007.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
---------------------------------------------------------------------------

    \6\ Wall thickness is the beltline wall thickness including the 
clad thickness.
    \7\ RT<INF>PTS</INF> limits contributes 1 x 10<SUP>-8</SUP> per 
reactor year to the reactor vessel TWCF.
    \8\ Excluding underclad cracks in forgings.
---------------------------------------------------------------------------

[FR Doc. 07-4887 Filed 10-2-07; 8:45 am]
BILLING CODE 7590-01-P


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