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.

Proposed Rules Federal Register

56275 

Vol. 72, No. 191 

Wednesday, October 3, 2007 

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 eRulemaking 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: SECY@nrc.gov. 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 CAG@nrc.gov. 

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 

PDR@nrc.gov. 

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: 

GMT1@nrc.gov, or Mr. Barry Elliot, 

Office of Nuclear Reactor Regulation, 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001; telephone 

(301) 415–2709; e-mail: BJE@nrc.gov. 

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’’ (RTNDT). RTNDT is referred to 

as a ductile-to-brittle transition 

temperature. At temperatures below 

RTNDT fracture occurs very rapidly, by 

cleavage, a behavior referred to as 

‘‘brittle.’’ As temperatures increase 

above RTNDT, progressively larger 

amounts of deformation occur before 

rapid cleavage fracture occurs. 

Eventually, at temperatures above 

approximately RTNDT + 60 °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 


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radiation embrittles the material over 

time, causing a shift in RTNDT to higher 

temperatures. Correlations based on test 

results for unirradiated and irradiated 

specimens have been developed to 

calculate the shift in RTNDT as a 

function of neutron fluence (the 

integrated neutron flux over a specified 

time of plant operation) for various 

material compositions. The value of 

RTNDT 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 plantspecific 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 × 10¥6 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 Vnotch 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 × 10¥6

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 × 10¥5 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 × 10¥6 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 


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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 × 10¥6 per reactoryear 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 cladto-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 × 10¥6 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 × 10¥6/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 


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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 ±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 ±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 Codesized 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, 

§ 50.61a, as a voluntary alternative to 

compliance with the requirements of 

§ 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 § 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 § 50.61, fracture 

toughness requirements for protection 

against pressurized thermal shock 

events for pressurized water nuclear 

power reactors. However, § 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 

(RTMAX–X) for flaws along axial weld 

fusion lines, plates, forgings, and 

circumferential weld fusion lines, and 

comparing the quantified value against 

the RTMAX–X screening criteria. 

Although comparing quantified values 

to the screening criteria is also required 

by the current § 50.61, the proposed 

§ 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 

RTMAX–X values in § 50.61a would differ 

significantly from that in § 50.61 as 

described in the technical basis for this 

rule. The new embrittlement correlation 

was developed using multivariable 


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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 RTMAX–X

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 § 50.61a. It 

would also provide that terms defined 

in § 50.61 also have the same meaning 

in § 50.61a unless otherwise noted. 

Paragraph (b) would describe the 

applicability of § 50.61a to PWRs as an 

alternative to the requirements of 

§ 50.61. The requirements of this section 

would provide a voluntarilyimplemented alternative to the current 

requirements of § 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 § 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 

§ 50.61a. In the licensee’s amendment 

request, the required information would 

include (a) calculating the values of 

RTMAX–X 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 

RTMAX–X 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 § 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 RTMAX–X 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 plantspecific analyses should the value of 

RTMAX–X 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 § 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 threeeights 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 × 10¥6 per 

reactor-year. The analysis could be a 

complete, plant-specific, probabilistic 

fracture mechanics analysis or could be 

a simplified analysis of flaw size, 

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 × 10¥6 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 RTMAX–X 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 RTNDT 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 RTMAX–X

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 


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56280 Federal Register / Vol. 72, No. 191 / Wednesday, October 3, 2007 / Proposed Rules 

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 Request for Commission Approval .................. X .................... ML060530624 

SRM–SECY–06–0124, June 30, 2006, Staff Requirements—Commission Approval of Rulemaking Plan.X .................... ML061810148 

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’’.X .................... ML043060581 

NUREG–1806, ‘‘Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limits in the PTS Rule (10 CFR 50.61): Summary Report’’.X .................... ML061580318 

NUREG–1874, ‘‘Recommended Screening Limits for Pressurized Thermal Shock (PTS)’’ ............ X .................... ML070860156 

Regulatory Guide 1.154, ‘‘Format and Content of Plant-Specific Pressurized Thermal Shock 

Analysis Reports for Pressurized Water Reactors’’.

X .................... ML003740028 

Regulatory Guide 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis’’.X .................... ML023240437 

Memorandum from Elliot to Mitchell, dated April 3, 2007, ‘‘Development of Flaw Size Distribution 

Tables for Draft Proposed Title 10 of the Code of Federal Regulations (10 CFR) 50.61a’’.

X .................... ML070950392 

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 § 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 

§ 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 


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Federal Register / Vol. 72, No. 191 / Wednesday, October 3, 2007 / Proposed Rules 56281 

implementation of § 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 ‘‘noaction’’ 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 

§ 50.61a, when any change is made to 

the design or operation of the facility 

that affects the calculated RTMAX-X

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 

INFOCOLLECTS@NRC.GOV 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. 


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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 § 50.61, paragraph (b)(1) is 

revised to read as follows: 

§ 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 

§ 50.82(a)(1) have been submitted, the 

licensee shall have projected values of 

RTPTS or RTMAX–X, accepted by the 

NRC, for each reactor vessel beltline 

material for the EOL fluence of the 

material in accordance with this section 

or § 50.61a. For a licensee choosing to 

comply with this section, the 

assessment of RTPTS 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 RTPTS for each vessel beltline 

material, including the assumptions 

regarding core loading patterns, and 

must specify the copper and nickel 

contents and the fluence value used in 

the calculation for each beltline 

material. This assessment must be 

updated whenever there is a 

significant 2 change in projected values 

of RTPTS, or upon request for a change 

in the expiration date for operation of 

the facility. 

* * * * * 

3. Section 50.61a is added to read as 

follows: 

§ 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 

§ 50.55a. 

(2) RTMAX–AW means the material 

property which characterizes the reactor 

vessel’s resistance to fracture initiating 

from flaws found along axial weld 

fusion lines. RTMAX–AW is determined 

under the provisions of paragraph (f) of 

this section and has units of °F. 

(3) RTMAX–PL means the material 

property which characterizes the reactor 

vessel’s resistance to fracture initiating 

from flaws found in plates in regions 

that are not associated with welds found 

in plates. RTMAX–PL is determined under 

the provisions of paragraph (f) of this 

section and has units of °F. 

(4) RTMAX–FO means the material 

property which characterizes the reactor 

vessel’s resistance to fracture initiating 

from flaws in forgings that are not 

associated with welds found in forgings. 

RTMAX–FO is determined under the 

provisions of paragraph (f) of this 

section and has units of °F. 


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(5) RTMAX–CW means the material 

property which characterizes the reactor 

vessel’s resistance to fracture initiating 

from flaws found along the 

circumferential weld fusion lines. 

RTMAX–CW is determined under the 

provisions of paragraph (f) of this 

section and has units of °F. 

(6) RTMAX–X means any or all of the 

material properties RTMAX–AW, RTMAX–

PL, RTMAX–FO, or RTMAX–CW for a 

particular reactor vessel. 

(7) jt means fast neutron fluence for 

neutrons with energies greater than 1.0 

MeV. jt is determined under the 

provisions of paragraph (g) of this 

section and has units of n/cm2. 

(8) j means average neutron flux. j is 

determined under the provisions of 

paragraph (g) of this section and has 

units of n/cm2/sec. 

(9) DT30 means the shift in the Charpy 

V-notch transition temperature 

produced by irradiation defined at the 

30 ft-lb energy level. The DT30 value is 

determined under the provisions of 

paragraph (g) of this section and has 

units of °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. 

TC has units of °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 

RTPTS value calculated under 10 CFR 

50.61 is projected to exceed the PTS 

screening criteria in 10 CFR 50.61 for 

plants licensed under 10 CFR part 50 or 

10 CFR part 52. 

(1) Each licensee shall have projected 

values of RTMAX–X for each reactor 

vessel beltline material for the EOL 

fluence of the material. The assessment 

of RTMAX–X values must use the 

calculation procedures given in 

paragraphs (f) and (g) of this section, 

except as provided in paragraphs (f)(6) 

and (f)(7) of this section. The assessment 

must specify the bases for the projected 

value of RTMAX–X for each reactor vessel 

beltline material, including the 

assumptions regarding future plant 

operation (e.g., core loading patterns, 

projected capacity factors, etc.); the 

copper (Cu), phosphorus (P), manganese 

(Mn), and nickel (Ni) contents; the 

reactor cold leg temperature (TC); and 

the neutron flux and fluence values 

used in the calculation for each beltline 

material. 

(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 RTMAX–X 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 RTMAX–X

values are greater than the PTS 

screening criteria in Table 1 of this 

section, then the licensee may propose 

the compensatory actions or plantspecific 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 RTMAX–X, 

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 RTMAX–X

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 RTMAX–X 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 RTMAX–X is 

projected to exceed the PTS screening 

criteria, then the licensee shall 

implement those flux reduction 

programs that are reasonably practicable 

to avoid exceeding the PTS screening 

criteria. The schedule for 

implementation of flux reduction 

measures may take into account the 

schedule for review and anticipated 

approval by the Director of detailed 

plant-specific analyses which 

demonstrate acceptable risk with 

RTMAX–X values above the PTS 

screening criteria due to plant 

modifications, new information, or new 

analysis techniques. 

(4) If the analysis required by 

paragraph (d)(3) of this section indicates 

that no reasonably practicable flux 

reduction program will prevent the 

RTMAX–X value for one or more reactor 

vessel beltline materials from exceeding 

the PTS screening criteria, then the 

licensee shall perform a safety analysis 

to determine what, if any, modifications 

to equipment, systems, and operation 

are necessary to prevent the potential 

for an unacceptably high probability of 

failure of the reactor vessel as a result 

of postulated PTS events 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 RTMAX–X is 


ebenthall on PRODPC61 with PROPOSALS
56284 Federal Register / Vol. 72, No. 191 / Wednesday, October 3, 2007 / Proposed Rules 

1The 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. 

2Table 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. 

3Because 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. 

projected to exceed the PTS screening 

criteria. 

(5) After consideration of the 

licensee’s analyses, including effects of 

proposed corrective actions, if any, 

submitted under paragraphs (d)(3) and 

(d)(4) of this section, the Director may, 

on a case-by-case basis, approve 

operation of the facility with RTMAX–X

values in excess of the PTS screening 

criteria. The Director will consider 

factors significantly affecting the 

potential for failure of the reactor vessel 

in reaching a decision. 

(6) If the Director concludes, under 

paragraph (d)(5) of this section, that 

operation of the facility with RTMAX–X

values in excess of the PTS screening 

criteria cannot be approved on the basis 

of the licensee’s analyses submitted 

under paragraphs (d)(3) and (d)(4) of 

this section, then the licensee shall 

request a license amendment, and 

receive approval by the Director, 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 RTMAX–X value of 

the facility is projected to exceed the 

PTS screening criteria and the 

requirements of paragraphs (d)(3) 

through (d)(6) of this section cannot be 

satisfied, the reactor vessel beltline may 

be given a thermal annealing treatment 

under the requirements of § 50.66 to 

recover the fracture toughness of the 

material. The reactor vessel may be used 

only for that service period within 

which the predicted fracture toughness 

of the reactor vessel beltline materials 

satisfy the requirements of paragraphs 

(d)(1) through (d)(6) of this section, with 

RTMAX–X values accounting for the 

effects of annealing and subsequent 

irradiation. 

(e) Examination and Flaw Assessment 

Requirements. The volumetric 

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. 

(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 1×10-6

per reactor-year if the ASME Code, 

Section XI volumetric examination 

required by paragraph (c)(2) or (d)(2) of 

this section indicates any of the 

following: 

(i) The 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; 

(iii) There are linear indications that 

penetrate through the clad into the low 

alloy steel reactor vessel shell; or 

(iv) Any indications between the cladto-base metal interface and three-eights 3

of the vessel thickness exceed the size 

allowable in ASME Code, Section XI, 

Table IWB–3510–1. 

(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 RTMAX–X values. 

Each licensee shall calculate RTMAX–X

values for each reactor vessel beltline 

material using jt. jt must be calculated 

using an NRC-approved methodology. 

(1) The values of RTMAX–AW, RTMAX–

PL, RTMAX–FO, and RTMAX–CW must be 


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Federal Register / Vol. 72, No. 191 / Wednesday, October 3, 2007 / Proposed Rules 56285 

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.’’

5The 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. 

determined using Equations 1 through 4 

of this section. 

(2) The values of DT30 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 DT30 value for 

each axial weld fusion line calculated as 

specified by Equation 1 of this section 

must be calculated for the maximum 

fluence (jFL) occurring along a 

particular axial weld fusion line. The 

DT30 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 DT30 value for each plate or forging 

calculated as specified by Equations 2 

and 3 of this section are calculated for 

the maximum fluence (jtMAX) occurring 

at the clad-to-base metal interface of 

each plate or forging. In Equation 4, the 

jtFL value used for calculating the plate, 

forging, and circumferential weld 

RTMAX–CW value is the maximum j 

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) The values of RTNDT(u) must be 

evaluated according to the procedures 

in the ASME Code, Section III, 

paragraph NB–2331. If any other 

method is used for this evaluation, the 

licensee shall submit the proposed 

method for review and approval by the 

Director along with the calculation of 

RTMAX–X values required in paragraph 

(c)(1) of this section. 

(i) If a measured value of RTNDT(u) is 

not available, a generic mean value of 

RTNDT(u) for the class 5 of material must 

be used if there are sufficient test results 

to establish a mean. 

(ii) The following generic mean values 

of RTNDT(u) must be used unless 

justification for different values is 

provided: 0 °F for welds made with 

Linde 80 weld flux; and ¥56 °F for 

welds made with Linde 0091, 1092, and 

124 and ARCOS B–5 weld fluxes. 

(5) The value of Tc in Equation 6 of 

this section must represent the weighted 

time average of the reactor cold leg 

temperature under normal operating full 

power conditions from the beginning of 

full power operation through the end of 

licensed operation. 

(6) The licensee shall verify that an 

appropriate RTMAX–X value has been 

calculated for each reactor vessel 

beltline material. The licensee shall 

consider plant-specific information that 

could affect the use of Equations 5 

though 7 of this section for the 

determination of a material’s DT30 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 RTMAX–X 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 s 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 DT30 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 DT30

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 DT30 and 

RTMAX–X 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 RTMAX–X value to 

the Director. Any value of RTMAX–X that 

has been modified as specified in 

paragraph (f)(6)(iv) of this section is 

subject to the approval of the Director 

when used as provided in this section. 

(g) Equations and variables used in 

this section. 

Equation 1: RTMAX–AW = MAX 

{[RTNDT(u)¥plate + DT30¥plate(jtFL)], 

[RTNDT(u)¥axialweld + 

DT30¥axialweld(jtFL)]} 

Equation 2: RTMAX–PL = RTNDT(u)¥plate + 

DT30¥plate(jtMAX) 

Equation 3: RTMAX–FO = RTNDT(u)¥forging

+ DT30¥forging(jtMAX) 

Equation 4: RTMAX–CW = MAX 

{[RTNDT(u)¥plate + DT30¥plate(jtMAX)], 

[RTNDT(u)¥circweld + 

DT30¥circweld(jtMAX)], 

[RTNDT(u)¥forging + 

DT30¥forging(jtMAX)]} 

Equation 5: DT30 = MD + CRP 

Equation 6: MD = A · (1 ¥ 0.001718 ·

TC) · (1 + 6.13 · P · Mn2.471) · jte

0.5

Equation 7: CRP = B · (1 + 3.77 · Ni1.191) 

· f(Cue,P) · g(Cue,Ni,jte) VVVVVVV 


ebenthall on PRODPC61 with PROPOSALS
56286 Federal Register / Vol. 72, No. 191 / Wednesday, October 3, 2007 / Proposed Rules 

Where: 

P [wt¥%] = phosphorus content 

Mn [wt¥%] = manganese content 

Ni [wt¥%] = nickel content 

Cu [wt¥%] = copper content 

A = 1.140 × 10¥7 for forgings 

= 1.561 × 10¥7 for plates 

= 1.417 × 10¥7 for welds 

B = 102.3 for forgings 

= 102.5 for plates in non-Combustion 

Engineering manufactured vessels 

= 135.2 for plates in Combustion 

Engineering vessels 

= 155.0 for welds 

jte = jt for j greater than or equal to 

4.39 × 1010 n/cm2/sec 

= jt · (4.39 × 1010/j)0.2595 for j less 

than 4.39 × 1010 n/cm2/sec 

Where: 

j [n/cm2/sec] = average neutron flux 

t [sec] = time that the reactor has been in full 

power operation 

jt [n/cm2] = j · t 

f(Cue,P) = 0 for Cu ≤ 0.072 

= [Cue ¥ 0.072]0.668 for Cu > 0.072 and P 

≤ 0.008 

= [Cue ¥ 0.072 + 1.359 · (P¥0.008)]0.668 for 

Cu > 0.072 and P > 0.008 

and Cue = 0 for Cu ≤ 0.072 

= MIN (Cu, maximum Cue) for Cu > 0.072 

and maximum Cue = 0.243 for Linde 80 

welds 

= 0.301 for all other materials 

g(Cue,Ni,jte) = 0.5 + 0.5 · tanh{[log10(jte) + 

1.1390 · Cue ¥ 0.448 · Ni ¥ 18.120] / 

0.629} 

Equation 8: Residual  = measured DT30

¥ predicted DT30 (by Equations 5, 

6, and 7) 

Equation 9: Mean deviation for a data 

set of n data points = 

r n i i

n

/ =

∑

1

Equation 10: Maximum credible heataverage residual = 3s/n0.5

Where: 

n = number of surveillance shift data points 

(sample size) in the specific data set 

s = standard deviation of the residuals about 

the model for a relevant material group 

given in Table 5. 

TABLE 1.—PTS SCREENING CRITERIA

Product form and RT MAX–Values 

RT MAX–X limits [°F] for different vessel wall thicknesses 6

(TWALL) 

TWALL ≤ 9.5 in. 9.5 in. < TWALL ≤ 

10.5 in. 

10.5 in. < TWALL ≤ 

11.5 in. 

Axial Weld RTMAX–AW ............................................................................................... 269 230 222 

Plate RTMAX–PL ......................................................................................................... 356 305 293 

Forging without underclad cracks RTMAX–FO ............................................................ 356 305 293 

Axial Weld and Plate RTMAX–AW + RTMAX–PL .......................................................... 538 476 445 

Circumferential Weld RTMAX–CW 7 ............................................................................. 312 277 269 

Forging with underclad cracks RTMAX–FO ................................................................. 246 241 239 

TABLE 2.—ALLOWABLE NUMBER OF FLAWS IN WELDS

ASME section XI flaw size per IWA–3200 Range of through-wall extent (TWE) of flaw 

(in.) 

Allowable number of cumulative 

flaws per 1000 inches of weld length 

in the ASME section XI appendix VIII 

supplement 4 inspection volume 

0.05 ........................................................................... 0.025 ≤ TWE < 0.075 ............................................... Unlimited 

0.10 ........................................................................... 0.075 ≤ TWE < 0.125 ............................................... 166.70 

0.15 ........................................................................... 0.125 ≤ TWE < 0.175 ............................................... 90.80 

0.20 ........................................................................... 0.175 ≤ TWE < 0.225 ............................................... 22.82 

0.25 ........................................................................... 0.225 ≤ TWE < 0.275 ............................................... 8.66 

0.30 ........................................................................... 0.275 ≤ TWE < 0.325 ............................................... 4.01 

0.35 ........................................................................... 0.325 ≤ TWE < 0.375 ............................................... 3.01 

0.40 ........................................................................... 0.375 ≤ TWE < 0.425 ............................................... 1.49 

0.45 ........................................................................... 0.425 ≤ TWE < 0.475 ............................................... 1.00 

TABLE 3.—ALLOWABLE NUMBER OF FLAWS IN PLATES OR FORGING

ASME section XI flaw size per IWA–3200 Range of through-wall extent (TWE) of flaw 

(in.) 

Allowable number of cumulative 

flaws per 1000 square inches of inside diameter surface area 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 


ebenthall on PRODPC61 with PROPOSALSEP03OC07.013</GPH>
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6Wall thickness is the beltline wall thickness 

including the clad thickness. 

7RTPTS limits contributes 1 × 10¥8 per reactor 

year to the reactor vessel TWCF. 

8Excluding underclad cracks in forgings. 

TABLE 4.—CONSERVATIVE ESTIMATES FOR CHEMICAL ELEMENT WEIGHT PERCENTAGES—Continued 

Materials P Mn 

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

TABLE 5.—MAXIMUM HEAT-AVERAGE RESIDUAL [°F] FOR RELEVANT MATERIAL GROUPS BY NUMBER OF AVAILABLE DATA

POINTS

Material group s [°F] 

Number of available data points 

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. 

[FR Doc. 07–4887 Filed 10–2–07; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 52 

RIN 3150–AI19 

Consideration of Aircraft Impacts for 

New Nuclear Power Reactor Designs 

AGENCY: U.S. Nuclear Regulatory 

Commission. 

ACTION: Proposed rule. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC or the Commission) 

is proposing to amend its regulations to 

require applicants for new standard 

design certifications that do not 

reference a standard design approval; 

new standard design approvals; 

combined licenses that do not reference 

a standard design certification, standard 

design approval, or manufactured 

reactor; and new manufacturing licenses 

that do not reference a standard design 

certification or standard design approval 

to assess the effects of the impact of a 

large, commercial aircraft on the nuclear 

power plant. Based on the insights 

gained from this assessment, the 

applicant shall include in its 

application a description and evaluation 

of design features, functional 

capabilities, and strategies to avoid or 

mitigate, to the extent practicable, the 

effects of the aircraft impact with 

reduced reliance on operator actions. 

The impact of a large, commercial 

aircraft is a beyond-design-basis event, 

and the NRC’s requirements applicable 

to the design, construction, testing, 

operation, and maintenance of design 

features, functional capabilities, and 

strategies for design basis events would 

not be applicable to design features, 

functional capabilities, or strategies 

selected by the applicant solely to meet 

the requirements of this rule. The 

objective of this rule is to require 

nuclear power plant designers to 

perform a rigorous assessment of design 

features that could provide additional 

inherent protection to avoid or mitigate, 

to the extent practicable, the effects of 

an aircraft impact, with reduced 

reliance on operator actions. 

DATES: Submit comments on this 

proposed rule by December 17, 2007. 

Submit comments on the information 

collection aspects on this proposed rule 

by November 2, 2007. Comments 

received after the above 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–AI19 in the subject line of 

your comments. Comments on 

rulemakings submitted in writing or in 

electronic form will be made available 

to the public in their entirety on the 

NRC rulemaking Web site. Personal 

information, such as your name, 

address, telephone number, e-mail 

address, etc., will not be removed from 

your submission. 

Submit comments via the Federal 

eRulemaking 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: SECY@nrc.gov. If 

you do not receive a reply e-mail 

confirming that we have received your 

comments, contact us directly at 301–

415–1966. 

Hand deliver comments to: 11555 

Rockville Pike, Rockville, Maryland 

20852, between 7:30 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, 

Maryland. 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 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 PDR 

Reference staff at 1–800–397–4209, 301–

415–4737 or by e-mail to pdr@nrc.gov. 

FOR FURTHER INFORMATION CONTACT: Mr. 

Stewart Schneider, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone 301–415–

1462; e-mail: sxs4@nrc.gov or Ms. 

Nanette Gilles, Office of New Reactors, 

U.S. Nuclear Regulatory Commission, 


ebenthall on PRODPC61 with PROPOSALS