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

46557 

Vol. 73, No. 155 

Monday, August 11, 2008 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 50 

RIN 3150–AI01 

[NRC–2007–0008] 

Alternate Fracture Toughness 

Requirements for Protection Against 

Pressurized Thermal Shock Events 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Supplemental Proposed Rule. 

SUMMARY: The Nuclear Regulatory 

Commission (NRC) is considering the 

adoption of provisions regarding 

applicability of the rule and new 

provisions regarding procedures to 

perform surveillance data checks related 

to the updated fracture toughness 

requirements for protection against 

pressurized thermal shock (PTS) events 

for pressurized water reactor (PWR) 

pressure vessels. The NRC is 

considering these provisions as an 

alternative to the provisions previously 

noticed for public comment on October 

3, 2007 (72 FR 56275). 

DATES: Submit comments on this 

proposed rule by September 10, 2008. 

Submit comments on the information 

collection aspects on this proposed rule 

by September 10, 2008. 

ADDRESSES: You may submit comments 

by any one of the following methods. 

Please include the following number 

RIN 3150–AI01 in the subject line of 

your comments. Comments submitted in 

writing or in electronic form will be 

made available for public inspection. 

Because your comments will not be 

edited to remove any identifying or 

contact information, the NRC cautions 

you against including any information 

in your submission that you do not want 

to be publicly disclosed. 

Federal e Rulemaking Portal: Go to 

http://www.regulations.gov and search 

for documents filed under Docket ID 

NRC–2007–0008. Address questions 

about NRC dockets to Carol Gallagher 

(301) 415–5905; e-mail 

Carol.Gallager@nrc.gov. 

Mail comments to: Secretary, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, ATTN: 

Rulemakings and Adjudications Staff. 

E-mail comments to: 

Rulemaking.Comments@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 am and 4:15 pm 

during Federal workdays. (Telephone 

(301) 415–1966). 

Fax comments to: Secretary, U.S. 

Nuclear Regulatory Commission at (301) 

415–1101. 

You can access publicly available 

documents related to this document 

using the following methods: 

NRC’s Public Document Room (PDR): 

The public may examine publicly 

available documents at the NRC’s PDR, 

Public File Area O–F21, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland. The PDR 

reproduction contractor will copy 

documents for a fee. 

NRC’s Agencywide Document Access 

and Management System (ADAMS): 

Publicly available documents created or 

received at the NRC are available 

electronically at the NRC’s Electronic 

Reading Room at http://www.nrc.gov/ 

reading-rm/adams.html. From this page, 

the public can gain entry into ADAMS, 

which provides text and image files of 

NRC’s public documents. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC’s 

PDR reference staff at 1–800–397–4209, 

or (301) 415–4737, or by e-mail to 

PDR.Resource@nrc.gov. 

FOR FURTHER INFORMATION CONTACT: Ms. 

Veronica M. Rodriguez, Office of 

Nuclear Reactor Regulation, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001; telephone 

(301) 415–3703; e-mail: 

Veronica.Rodriguez@nrc.gov, Mr. Barry 

Elliot, Office of Nuclear Reactor 

Regulation, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001; telephone (301) 415–2709; e-mail: 

Barry.Elliot@nrc.gov, or Mr. Mark Kirk, 

Office of Nuclear Regulatory Research, 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001; telephone 

(301) 415–6015; e-mail: 

Mark.Kirk@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

II. Background 

III. Discussion 

IV. Responses to Comments on the Proposed 

Rule 

V. Section-by-Section Analysis 

VI. Specific Request for Comments 

VII. Availability of Documents 

VIII. Plain Language 

IX. Voluntary Consensus Standards 

X. Finding of No Significant Environmental 

Impact: Availability 

XI. Paperwork Reduction Act Statement 

XII. Regulatory Analysis 

XIII. Regulatory Flexibility Act Certification 

XIV. Backfit Analysis 

I. Introduction 

The NRC published a proposed rule 

on alternate fracture toughness 

requirements for protection against 

Pressurized Thermal Shock (PTS) for 

public comments in the Federal 

Register on October 3, 2007 (72 FR 

56275). This rule provides new PTS 

requirements based on updated analysis 

methods. This action is desirable 

because the existing requirements are 

based on unnecessarily conservative 

probabilistic fracture mechanics 

analyses. This action would reduce 

regulatory burden for licensees, 

specifically those licensees that expect 

to exceed the existing requirements 

before the expiration of their licenses, 

while maintaining adequate safety. 

These new requirements would be 

utilized by any Pressurized Water 

Reactor (PWR) licensee as an alternative 

to complying with the existing 

requirements. 

During the development of the PTS 

final rule, the NRC determined that 

several changes to the proposed rule 

language may be needed to adequately 

address issues raised in stakeholder’s 

comments. The NRC also determined, in 

response to a stakeholder comment, that 

the characteristics of advanced PWR 

designs were not considered in the 

technical analysis made for the 

proposed rule. The NRC does not have 

assurance that reactors that commence 

commercial power operation after the 

effective date of this rule will have 

operating characteristics and materials 

of fabrication similar to those evaluated 

as part of the technical basis for the 

proposed rule. Therefore, the NRC has 

concluded that it would be prudent to 


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limit the applicability and the use of 

§ 50.61a to currently-operating plants 

only, and proposes to modify the 

applicability provisions of the proposed 

rule accordingly. 

Also, several stakeholders questioned 

the accuracy and validity of the generic 

embrittlement curves in the proposed 

rule. The NRC wants to ensure that the 

predicted values from the proposed 

embrittlement trend curves provide an 

adequate basis for implementation of 

the rule. Therefore, the NRC has 

continued to work on statistical 

procedures to identify deviations from 

generic embrittlement trends, such as 

those described in § 50.61a(f)(6) of the 

proposed rule. Based on this work, the 

NRC is considering enhancing the 

procedure described in paragraph 

§ 50.61a(f)(6) to, among other things, 

detect signs from the plant- and heatspecific surveillance data of 

embrittlement trends that are not 

reflected by Equations 5, 6 and 7 of the 

rule that may emerge at high fluences. 

Because these proposed modifications 

may not represent a logical outgrowth 

from the October 2007 proposed rule’s 

provisions, the NRC concludes that 

obtaining stakeholder feedback on the 

proposed alternative provisions through 

the use of a supplemental proposed rule 

is appropriate. As discussed in Section 

VI of this notice, the NRC will consider 

comments on §§ 50.61a(b); (f)(6)(i) 

through (f)(6)(vi); Equations 10, 11, and 

12 in § 50.61a(g); and Tables 5, 6, and 

7 of this supplemental proposed rule. 

The NRC is also requesting comments 

on whether there should be additional 

language added to § 50.61a(e) to allow 

licensees to account for the effects of 

sizing errors. This supplemental 

proposed rule does not reflect other 

modifications or editorial and 

conforming changes that the NRC is 

considering to incorporate in the final 

rule as a result of the public comments 

on the October 2007 proposed rule. 

II. Background 

PTS events are system transients in a 

PWR in which severe overcooling 

occurs coincident with high pressure. 

The thermal stresses are caused by rapid 

cooling of the reactor vessel inside 

surface, which combine with the 

stresses caused by high pressure. The 

aggregate effect of these stresses is an 

increase in the potential for fracture if 

a pre-existing flaw is present in a 

material susceptible to brittle failure. 

The ferritic, low alloy steel of the 

reactor vessel beltline adjacent to the 

core, where neutron radiation gradually 

embrittles the material over the lifetime 

of the plant, can be susceptible to brittle 

fracture. 

The PTS rule, described in § 50.61, 

adopted on July 23, 1985 (50 FR 29937), 

establishes screening criteria below 

which the potential for a reactor vessel 

to fail due to a PTS event is deemed to 

be acceptably low. The screening 

criteria effectively define a limiting 

level of embrittlement beyond which 

operation cannot continue without 

further plant-specific evaluation. 

Regulatory Guide (RG) 1.154, ‘‘Format 

and Content of Plant-Specific 

Pressurized Thermal Shock Analysis 

Reports for Pressurized Water Reactors,’’

indicates that reactor vessels that exceed 

the screening criteria in § 50.61 may 

continue to operate provided they can 

demonstrate a mean through-wall crack 

frequency (TWCF) from PTS-related 

events of no greater than 5 × 10¥6 per 

reactor year. 

Any reactor vessel with materials 

predicted to exceed the screening 

criteria in § 50.61 may not continue to 

operate without implementation of 

compensatory actions or additional 

plant-specific analyses unless the 

licensee receives an exemption from the 

requirements of the rule. Acceptable 

compensatory actions are neutron flux 

reduction, plant modifications to reduce 

PTS event probability or severity, and 

reactor vessel annealing, which are 

addressed in §§ 50.61(b)(3), (b)(4), and 

(b)(7); and § 50.66, ‘‘Requirements for 

Thermal Annealing of the Reactor 

Pressure Vessel.’’

Currently, no operating PWR reactor 

vessel is projected to exceed the § 50.61 

screening criteria before the expiration 

of its 40 year operating license. 

However, several PWR reactor vessels 

are approaching the screening criteria, 

while others are likely to exceed the 

screening criteria during their first 

license renewal periods. 

The NRC’s Office of Nuclear 

Regulatory Research (RES) developed a 

technical basis that supports updating 

the PTS regulations. This technical basis 

concluded that the risk of through-wall 

cracking due to a PTS event is much 

lower than previously estimated. This 

finding indicated that the screening 

criteria in § 50.61 are unnecessarily 

conservative and may impose an 

unnecessary burden on some licensees. 

Therefore, the NRC created a new rule, 

§ 50.61a, which provides alternate 

screening criteria and corresponding 

embrittlement correlations based on the 

updated technical basis. The NRC 

decided that providing a new section 

containing the updated screening 

criteria and updated embrittlement 

correlations would be appropriate 

because the Commission directed the 

NRC staff, in a Staff Requirements 

Memorandum (SRM) dated June 30, 

2006, to prepare a rulemaking which 

would allow current PWR licensees to 

implement the new requirements of 

§ 50.61a or continue to comply with the 

current requirements of § 50.61. 

Alternatively, the NRC could have 

revised § 50.61 to include the new 

requirements, which could be 

implemented as an alternative to the 

current requirements. However, 

providing two sets of requirements 

within the same regulatory section was 

considered confusing and/or ambiguous 

as to which requirements apply to 

which licensees. 

The NRC published the proposed 

rulemaking on the alternate fracture 

toughness requirements for protection 

against PTS for public comment in the 

Federal Register on October 3, 2007 (72 

FR 56275). The proposed rule provided 

an alternative to the current rule, which 

a licensee may choose to adopt. This 

prompted the NRC to keep the current 

requirements separate from the new 

alternative requirements. As a result, the 

proposed rule retained the current 

requirements in § 50.61 for PWR 

licensees choosing not to implement the 

less restrictive screening limits, and 

presented new requirements in § 50.61a 

as an alternative relaxation for PWR 

licensees. 

III. Discussion 

The NRC published a proposed new 

rule, § 50.61a (October 3, 2007, 72 FR 

56275), that would provide new PTS 

requirements based on updated analysis 

methods because the existing 

requirements are based on unnecessarily 

conservative probabilistic fracture 

mechanics analyses. Stakeholders’

comments raised concerns related to the 

applicability of the rule and the 

accuracy and validity of the generic 

embrittlement curves. The NRC 

reconsidered the technical and 

regulatory issues in these areas and is 

considering adopting the modified 

provisions regarding the applicability of 

the rule and new provisions regarding 

procedures to perform surveillance data 

checks described in this supplemental 

proposed rule. The NRC will consider 

comments on §§ 50.61a(b), (f)(6)(i) 

through (f)(6)(vi); Equations 10, 11 and 

12 in § 50.61a(g); and Tables 5, 6, and 

7 of this supplemental proposed rule. 

As described in Section VI of this 

notice, the NRC is also requesting 

comments on whether there should be 

additional language added to § 50.61a(e) 

to allow licensees to account for the 

effects of sizing errors. The NRC will 

consider the October 2007 proposed 

rule, the supplemental proposed rule, 

and the comments received in response 


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to both, when deciding whether to 

adopt a final PTS rule. 

Applicability of the Proposed Rule, 

§ 50.61a(b) 

The supplemental proposed rule 

differs from the proposed rule and from 

§ 50.61 in that it proposes to limit the 

use of § 50.61a to currently operating 

plants only. It cannot be demonstrated, 

a priori, that reactors which commence 

commercial power operation after the 

effective date of this rule will have 

operating characteristics, in particular 

identified PTS event sequences and 

thermal-hydraulic responses, which are 

consistent with the reactors which were 

evaluated as part of the technical basis 

for this rule. Other factors, including 

materials of fabrication and welding 

methods, could also vary. Hence, the 

use of § 50.61a would be limited to 

currently operating PWR facilities 

which are known to have characteristics 

consistent with those assumed in the 

technical basis. The NRC also proposes 

to allow the holder of the operating 

license for Watts Bar Unit 2 to adopt the 

requirements in § 50.61a as this facility 

has operating characteristics consistent 

with those assumed in the technical 

basis. The NRC recognizes that licensees 

for reactors who commence commercial 

power operation after the effective date 

of this rule may, under the provisions of 

§ 50.12, seek an exemption from 

§ 50.61a(b) to apply this rule if a plantspecific basis analyzing their operating 

characteristics, materials of fabrications, 

and welding methods is provided. 

Surveillance Data, § 50.61a(f) 

Section 50.61a(f) of the proposed rule 

defines the process for calculating the 

values for the material properties (i.e. , 

RTMAX–X) for a particular reactor vessel. 

These values would be based on the 

vessel material’s copper, manganese, 

phosphorus, and nickel weight 

percentages, reactor cold leg 

temperature, and fast neutron flux and 

fluence values, as well as the 

unirradiated nil-ductility transition 

reference temperature (i.e., RTNDT). 

Section 50.61a(f) of the proposed rule 

included a procedure by which the 

RTMAX–X values, which are predicted for 

plant-specific materials using a generic 

temperature shift (i.e., DT30) 

embrittlement trend curve, are 

compared with heat-specific 

surveillance data that are collected as 

part of 10 CFR Part 50, Appendix H 

surveillance programs. The purpose of 

this comparison is to assess how well 

the surveillance data are represented by 

the generic embrittlement trend curve. If 

the surveillance data are close 

(closeness is assessed statistically) to the 

generic embrittlement trend curve, then 

the predictions of this embrittlement 

trend curve are used. This is expected 

to normally be the case. However, if the 

heat-specific surveillance data deviate 

significantly, and non-conservatively, 

from the predictions of the generic 

embrittlement trend curve, this 

indicates that alternative methods (i.e., 

other than, or in addition to, the generic 

embrittlement trend curve) may be 

needed to reliably predict the 

temperature shift trends, and to estimate 

RTMAX–X, for the conditions being 

assessed. However, alternative methods 

for temperature shift prediction are not 

prescribed by § 50.61a(f) of the proposed 

rule. 

Although standard and accepted 

procedures exist to assess the statistical 

significance of the differences between 

heat-specific surveillance data and the 

generic embrittlement trend curve, 

similarly standard and acceptable 

procedures are not available to assess 

the practical importance of such 

differences. The practical importance of 

statistically significant deviations is best 

assessed by licensees on a case-by-case 

basis, which would be submitted for the 

review of the Director of NRR, as 

prescribed by § 50.61a(f). 

The method described in the 

proposed rulemaking to compare the 

heat-specific surveillance data collected 

as part of 10 CFR part 50, Appendix H 

surveillance programs to the generic 

temperature shift embrittlement trend 

curve included a single statistical test. 

This statistical test was set forth by 

Equations 9 and 10, and Table 5. This 

test determined if, on average, the 

temperature shift from the surveillance 

data was significantly higher than the 

temperature shift of the generic 

embrittlement trend curve. The NRC has 

determined that, while necessary, this 

single test is not sufficient to ensure that 

the temperature shift predicted by the 

embrittlement trend curve well 

represents the heat-specific surveillance 

data. Specifically, this single statistical 

test cannot determine if the temperature 

shift from the surveillance data shows a 

more rapid increase after significant 

radiation exposure than the progression 

predicted by the generic embrittlement 

trend curve. To address this potential 

deficiency, which could be particularly 

important during a plant’s period of 

extended operation, the NRC added two 

more statistical tests in this 

supplemental proposed rulemaking, 

which are expressed by Equations 11 

and 12 and by Tables 6 and 7. Together, 

these two additional tests determine if 

the surveillance data from a particular 

heat show a more rapid increase after 

significant radiation exposure than the 

progression predicted by the generic 

embrittlement trend curve. 

The NRC documented the technical 

basis for the proposed alternative in the 

following reports: (1) ‘‘Statistical 

Procedures for Assessing Surveillance 

Data for 10 CFR Part 50.61a,’’ (ADAMS 

Accession No. ML081290654), and (2) 

‘‘A Physically Based Correlation of 

Irradiation Induced Transition 

Temperature Shifts for RPV Steel,’’

(ADAMS Accession No. ML081000630). 

IV. Responses to Comments on the 

Proposed Rule 

The NRC received 5 comment letters 

on the proposed 10 CFR 50.61a rule 

published on October 3, 2007 (72 FR 

56275). The following paragraphs 

discuss those comments which are 

directly associated with the 

supplemental proposed rule’s 

provisions on the applicability of the 

rule and surveillance data procedures. 

The remainder of the comments and the 

NRC responses will be provided in the 

Federal Register notice for the final 

rule. 

Comments on the Applicability of the 

Proposed Rule 

Comment: The commenters stated 

that the rule, as written, is only 

applicable to the existing fleet of PWRs. 

The characteristics of advanced PWR 

designs were not considered in the 

analysis. The commenters suggested 

adding a statement to state that this rule 

is applicable to the current PWR fleet 

and not the new plant designs. 

[PWROG–5, EPRI–5] 

Response: The NRC agrees with the 

comment that this rule is only 

applicable to the existing fleet of PWRs. 

The NRC cannot be assured that reactors 

that commence commercial power 

operation after the effective date of this 

rule will have operating characteristics, 

in particular identified PTS event 

sequences and thermal-hydraulic 

responses, which are consistent with the 

reactors that were evaluated as part of 

the technical basis for § 50.61a. Other 

factors, including materials of 

fabrication and welding methods, could 

also vary. Therefore, the NRC agrees 

with the commenters that it would be 

prudent to restrict the use of § 50.61a to 

current plants. As a result of this 

comment, the NRC proposes to modify 

§ 50.61a(b) and the statement of 

considerations of the rule to reflect this 

position to limit the use of the rule to 

currently operating plants. 

Comments on Surveillance Data 

Comment: The commenters stated 

that there is little added value in the 

requirement to assess the surveillance 


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data as a part of this rule because 

variability in data has already been 

accounted for in the derivation of the 

embrittlement correlation. 

The commenters also stated that there 

is no viable methodology for adjusting 

the projected DT30 for the vessel based 

on the surveillance data. Any effort to 

make this adjustment is likely to 

introduce additional error into the 

prediction. Note that the embrittlement 

correlation described in the basis for the 

revised PTS rule (i.e., NUREG–1874) 

was derived using all of the currently 

available industry-wide surveillance 

data. 

In the event that the surveillance data 

does not match the DT30 value predicted 

by the embrittlement correlation, the 

best estimate value for the pressure 

vessel material is derived using the 

embrittlement correlation. The likely 

source of the discrepancy is an error in 

the characterization of the surveillance 

material or of the irradiation 

environment. Therefore, unless the 

discrepancy can be resolved, obtaining 

the DT30 prediction based on the best 

estimate chemical composition for the 

heat of the material is more reliable than 

a prediction based on a single set of 

surveillance measurements. 

The commenters suggested removing 

the requirement to assess surveillance 

data, including Table 5, of this rule. 

[PWROG–4, EPRI–4, NEI–2] 

Response: The NRC does not agree 

with the proposed change. The NRC 

believes that there is added value in the 

requirement to assess surveillance data. 

Although variability has been accounted 

for in the derivation of the 

embrittlement correlation, it is the 

NRC’s view that the surveillance 

assessment required in § 50.61a(f)(6) is 

needed to determine if the 

embrittlement for a specific heat of 

material in a reactor vessel is consistent 

with the embrittlement predicted by the 

embrittlement correlation. 

The commenters also assert that there 

is no viable methodology for adjusting 

the projected DT30 for the vessel based 

on the surveillance data, and that any 

adjustment is likely to introduce 

additional error into the prediction. The 

NRC believes that although there is no 

single methodology for adjusting the 

projected DT30 for the vessel based on 

the surveillance data, it is possible, on 

a case-specific basis, to justify 

adjustments to the generic DT30

prediction. For this reason the rule does 

not specify a method for adjusting the 

DT30 value based on surveillance data, 

but rather requires the licensee to 

propose a case-specific DT30 adjustment 

procedure for review and approval from 

the Director. Although the commenters 

assert that it is possible that error could 

be introduced, it is the NRC view that 

appropriate plant-specific adjustments 

based upon available surveillance data 

may be necessary to project reactor 

pressure vessel embrittlement for the 

purpose of this rule. 

As the result of these public 

comments, the NRC has continued to 

work on statistical procedures to 

identify deviations from generic 

embrittlement trends, such as those 

described in § 50.61a(f)(6) of the 

proposed rule. Based on this work, the 

NRC is considering further enhancing 

the procedure described in paragraph 

(f)(6) to, among other things, detect 

signs from the plant- and heat-specific 

surveillance data that may emerge at 

high fluences of embrittlement trends 

that are not reflected by Equations 5, 6, 

and 7. The empirical basis for the NRC’s 

concern regarding the potential for unmodeled high fluence effects is 

described in documents located at 

ADAMS Accession Nos. ML081120253, 

ML081120289, ML081120365, 

ML081120380, and ML081120600. The 

technical basis for the enhanced 

surveillance assessment procedure is 

described in the document located at 

ADAMS Accession No. ML081290654. 

V. Section-by-Section Analysis 

The following section-by-section 

analysis only discusses the 

modifications in the provisions related 

to the applicability of the rule and 

surveillance data procedures that the 

NRC is considering as an alternative in 

this supplemental proposed rule. The 

NRC is only seeking comments on these 

alternative provisions. This 

supplemental proposed rule does not 

reflect other modifications or editorial 

and conforming changes that the NRC is 

considering to incorporate as a result of 

the public comments on the proposed 

rule that were not discussed in this 

notice as they will be provided in the 

Federal Register notice for the final 

rule. 

Proposed § 50.61a(b) 

The proposed language for § 50.61a(b) 

would establish the applicability of the 

rule. The NRC proposes to modify this 

paragraph to limit the use of this rule to 

currently-operating plants only. 

Proposed § 50.61a(f)(6)(i) 

The proposed language for 

§ 50.61a(f)(6)(i) would establish the 

requirements to perform data checks to 

determine if the surveillance data show 

a significantly different trend than what 

the embrittlement model in this rule 

predicts. The NRC proposes to modify 

§ 50.61a(f)(6)(i)(B) to state that licensees 

would evaluate the surveillance for 

consistency with the embrittlement 

model by following the procedures 

specified by §§ 50.61a(f)(6)(ii), (f)(6)(iii), 

and (f)(6)(iv) of the supplemental 

proposed rule. 

Proposed § 50.61a(f)(6)(ii) 

The proposed language for 

§ 50.61a(f)(6)(ii) would establish the 

requirements to perform an estimate of 

the mean deviation of the data set from 

the embrittlement model. The mean 

deviation for the data set would be 

compared to values given in Table 5 or 

Equation 10 of this section. The NRC 

proposes to modify this paragraph to 

state that the surveillance data analysis 

would follow the criteria in 

§§ 50.61a(f)(6)(v) and (f)(6)(vi) of the 

supplemental proposed rule. 

Proposed § 50.61a(f)(6)(iii) 

The NRC proposes to modify 

§ 50.61a(f)(6)(iii) to establish the 

requirements to estimate the slope of the 

embrittlement model residuals (i.e., the 

difference between the measured and 

predicted value for a specific data 

point). The licensee would estimate the 

slope using Equation 11 and compare 

this value to the maximum permissible 

value in Table 6, both from the 

supplemental proposed rule. This 

surveillance data analysis would follow 

the criteria in §§ 50.61a(f)(6)(v) and 

(f)(6)(vi) of the supplemental proposed 

rule. 

Proposed § 50.61a(f)(6)(iv) 

The NRC proposes to modify 

§ 50.61a(f)(6)(iv) to establish the 

requirements to estimate an outlier 

deviation from the embrittlement model 

for the specific data set using Equations 

8 and 12. The licensee would compare 

the normalized residuals to the 

allowable values in Table 7 of the 

supplemental proposed rule. This 

surveillance data analysis would follow 

the criteria in §§ 50.61a(f)(6)(v) and 

(f)(6)(vi) of the supplemental proposed 

rule. 

Proposed § 50.61a(f)(6)(v) 

The NRC proposes to add paragraph 

(f)(6)(v) to establish the criteria to be 

satisfied in order to calculate the DT30

shift values. 

Proposed § 50.61a(f)(6)(vi) 

The NRC proposes to add paragraph 

(f)(6)(vi) to establish the actions to be 

taken by a licensee if the criteria in 

paragraph (f)(6)(v) of this section are not 

met. The licensee would need to submit 

an evaluation of the surveillance data 

and propose values for DT30, 

considering their plant-specific 


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surveillance data, for the review and 

approval by the Director. The licensee 

would need to submit an evaluation of 

each surveillance capsule removed from 

the vessel after the submittal of the 

initial application for review and 

approval by the Director no later than 2 

years after the capsule is withdrawn 

from the vessel. 

Proposed § 50.61a(g) 

The proposed language for § 50.61a(g) 

would provide the necessary equations 

and variables required by the proposed 

changes in § 50.61a(f)(6). The NRC 

proposes to modify Equation 10 to 

account for 1 percent of significance 

level. Equations 11 and 12 would be 

added to provide the means for 

estimating the slope and the outlier 

deviation from the embrittlement 

model. 

Proposed Tables 5, 6, and 7 

Tables 5, 6, and 7 would provide 

values to be used in the proposed 

changes in § 50.61a(f)(6). The NRC 

proposes to modify Table 5 to account 

for the use of a 1 percent of significance 

level. Tables 6 and 7 would be added to 

provide the threshold values for the 

slope and the outlier deviation tests. 

VI. Specific Request for Comments 

The NRC seeks comments on 

§§ 50.61a(b), (f)(6)(i) through (f)(6)(vi); 

Equations 10, 11, and 12 in § 50.61a(g), 

and Tables 5, 6, and 7 of the 

supplemental proposed rule. The NRC 

is not seeking comments on any other 

provisions of the proposed § 50.61a 

which remain unchanged from the 

October 2007 proposed rule. In 

addition, the NRC also requests 

comments on the following question: 

Adjustments of the Inservice Inspection 

Volumetric Examination and Flaw 

Assessments 

The flaw sizes in Tables 2 and 3 are 

selected so that reactor vessels with flaw 

sizes less than or equal to those in the 

tables will have a TWCF less than or 

equal to 1 × 10¥6 per reactor year at the 

maximum permissible embrittlement. 

The NRC recognizes that the flaw sizes 

in these tables represent actual flaw 

dimensions while the results from the 

ASME Code examinations are estimated 

dimensions. The available information 

indicates that, for most flaw sizes in 

Tables 2 and 3, qualified inspectors will 

oversize flaws. Comparing oversized 

flaws to the size and density 

distributions in Tables 2 and 3 is 

conservative and acceptable, but not 

necessary. Therefore, NRC is 

considering to permit flaw sizes to be 

adjusted to account for the effects of 

sizing error before comparing the 

estimated size and density distribution 

to the acceptable size and density 

distributions in Tables 2 and 3. This 

would be accomplished by requiring 

licensees to base the methodology to 

account for the effects of sizing error on 

statistical data collected from ASME 

Code inspector qualification tests. An 

acceptable method would include a 

demonstration, that accounting for the 

effects of sizing error, is unlikely to 

result in accepting actual flaw size 

distribution that cause the TWCF to 

exceed the acceptance criteria. 

Adjusting flaw sizes to account for 

sizing error can change an unacceptable 

examination result into an acceptable 

result; further, collecting, evaluating, 

and using data from ASME Code 

inspector qualification tests will require 

extensive engineering judgment. 

Therefore, the methodology would have 

to be reviewed and approved by the 

Director of the NRC’s Office of Nuclear 

Reactor Regulation (NRR) to ensure that 

the risk associated with PTS is 

acceptable. The NRC requests specific 

comments on whether there should be 

additional language added to 10 CFR 

50.61a(e) to allow licensees to account 

for the effects of sizing errors. 

VII. Availability of Documents 

The NRC is making the documents 

identified below available to interested 

persons through one or more of the 

following methods, as indicated. 

Public Document Room (PDR). The 

NRC Public Document Room is located 

at 11555 Rockville Pike, Rockville, 

Maryland 20852. 

Regulations.gov (Web). These 

documents may be viewed and 

downloaded electronically through the 

Federal eRulemaking Portal http:// 

www.regulations.gov, Docket number 

NRC–2007–0008. 

NRC’s Electronic Reading Room 

(ERR). The NRC’s public electronic 

reading room is located at http:// 

www.nrc.gov/reading-rm.html. 

Document PDR Web ERR (ADAMS) 

Federal Register Notice—Proposed Rule: Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (RIN 3150–AI01), 72 FR 56275, October 3, 

2007 ................................................................................................................................................. X NRC–2007–0008 ML072750659 

Letter from Thomas P. Harrall, Jr., dated December 17, 2007, ‘‘Comments on Proposed Rule 10 

CFR 50, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, RIN 3150–AI01’’ [Identified as Duke] ............................................................... X NRC–2007–0008 ML073521542 

Letter from Jack Spanner, dated December 17, 2007, ‘‘10 CFR 50.55a Proposed Rulemaking 

Comments RIN 3150–AI01’’ [Identified as EPRI] ............................................................................ X NRC–2007–0008 ML073521545 

Letter from James H. Riley, dated December 17, 2007, ‘‘Proposed Rulemaking—Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (RIN 

3150–AI01), 72 FR 56275, October 3, 2007 [Identified as NEI] ..................................................... X NRC–2007–0008 ML073521543 

Letter from Melvin L. Arey, dated December 17, 2007, ‘‘Transmittal of PWROG Comments on the 

NRC Proposed Rule on Alternate Fracture Toughness Requirements for Protection Against 

Pressurized Thermal Shock Events’’, RIN 3150–AI01, PA-MSC–0232 [Identified as PWROG] .... X NRC–2007–0008 ML073521547 

Letter from T. Moser, dated December 17, 2007, ‘‘Strategic Teaming and Resource Sharing 

(STARS) Comments on RIN 3150–AI01, Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events 72 FR 56275 (October 3,2007) [Identified as 

STARS] ............................................................................................................................................ X NRC–2007–0008 ML073610558 

‘‘Statistical Procedures for Assessing Surveillance Data for 10 CFR Part 50.61a’’ ........................... X ML081290654 

‘‘A Physically Based Correlation of Irradiation Induced Transition Temperature Shifts for RPV 

Steel’’ ................................................................................................................................................ X ML081000630 

Supplemental Regulatory Analysis ...................................................................................................... X NRC–2007–0008 ML081440673 

Supplemental OMB Supporting Statement ......................................................................................... X NRC–2007–0008 ML081440736 

Memo from J. Uhle, dated May 15, 2008, ‘‘Embrittlement Trend Curve Development for Reactor 

Pressure Vessel Materials’’ .............................................................................................................. X ML081120253 

Draft ‘‘Technical Basis for Revision of Regulatory Guide 1.99: NRC Guidance on Methods to Estimate the Effects of Radiation Embrittlement on the Charpy V-Notch Impact Toughness of Reactor Vessel Materials’’ ........................................................................................................................ X ML081120289 


rwilkins on PROD1PC63 with PROPOSALS
46562 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 

Document PDR Web ERR (ADAMS) 

‘‘Comparison of the Predictions of RM–9 to the IVAR and RADAMO Databases’’ ........................... X ML081120365 

Memo from M. Erickson Kirk, dated December 12, 2007, ‘‘New Data from Boiling Water Reactor 

Vessel Integrity Program (BWRVIP) Integrated Surveillance Project (ISP)’’ .................................. X ML081120380 

‘‘Further Evaluation of High Fluence Data’’ ......................................................................................... X ML081120600 

VIII. Plain Language 

The Presidential memorandum ‘‘Plain 

Language in Government Writing’’

published in June 10, 1998 (63 FR 

31883), directed that the Government’s 

documents be in clear and accessible 

language. The NRC requests comments 

on the proposed rule specifically with 

respect to the clarity and effectiveness 

of the language used. Comments should 

be sent to the NRC as explained in the 

ADDRESSES heading of this notice. 

IX. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995, Public 

Law 104–113, requires that Federal 

agencies use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless 

using such a standard is inconsistent 

with applicable law or is otherwise 

impractical. 

The NRC determined that there is 

only one technical standard developed 

that could be utilized for characterizing 

the embrittlement correlations. That 

standard is the American Society for 

Testing and Materials (ASTM) standard 

E–900, ‘‘Standard Guide for Predicting 

Radiation-Induced Temperature 

Transition Shift in Reactor Vessel 

Materials.’’ This standard contains a 

different embrittlement correlation than 

that of this supplemental proposed rule. 

However, the correlation developed by 

the NRC has been more recently 

calibrated to available data. As a result, 

ASTM standard E–900 is not a practical 

candidate for application in the 

technical basis for the supplemental 

proposed rule because it does not 

represent the broad range of conditions 

necessary to justify a revision to the 

regulations. 

The ASME Code requirements are 

utilized as part of the volumetric 

examination analysis requirements of 

the supplemental proposed rule. ASTM 

Standard Practice E 185, ‘‘Standard 

Practice for Conducting Surveillance 

Tests for Light-Water Cooled Nuclear 

Power Reactor Vessels,’’ is incorporated 

by reference in 10 CFR Part 50, 

Appendix H and utilized to determine 

30-foot-pound transition temperatures. 

These standards were selected for use in 

the supplemental proposed rule based 

on their use in other regulations within 

10 CFR Part 50 and their applicability 

to the subject of the desired 

requirements. 

The NRC will consider using a 

voluntary consensus standard in the 

final rule if an appropriate standard is 

identified in the public comment period 

for this supplemental proposed rule. 

X. Finding of No Significant 

Environmental Impact: Availability 

The Commission has determined 

under the National Environmental 

Policy Act of 1969, as amended, and the 

Commission’s regulations in 10 CFR 

Part 51, Subpart A, that this rule, if 

adopted, would not be a major Federal 

action significantly affecting the quality 

of the human environment and, 

therefore, an environmental impact 

statement is not required. 

The determination of this 

environmental assessment is that there 

will be no significant offsite impact to 

the public from this action. This 

determination was made as part of the 

proposed rulemaking issued on October 

3, 2007 (72 FR 56275), and remains 

applicable to this supplemental 

proposed rulemaking. 

XI. Paperwork Reduction Act 

Statement 

This supplemental proposed rule 

would contain new or amended 

information collection requirements that 

are subject to the Paperwork Reduction 

Act of 1995 (44 U.S.C. 3501, et seq). 

This supplemental proposed rule has 

been submitted to the Office of 

Management and Budget for review and 

approval of the information collection 

requirements. 

Type of submission, new or revision: 

Revision. 

The title of the information collection: 

10 CFR Part 50, ‘‘Alternate Fracture 

Toughness Requirements for Protection 

against Pressurized Thermal Shock 

Events (10 CFR 50.61 and 50.61a)’’

supplemental proposed rule. 

The form number if applicable: Not 

applicable. 

How often the collection is required: 

Collections would be initially required 

for PWR licensees utilizing the 

requirements of 10 CFR 50.61a as an 

alternative to the requirements of 10 

CFR 50.61. Collections would also be 

required, after implementation of the 

new 10 CFR 50.61a, when any change 

is made to the design or operation of the 

facility that affects the calculated 

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: Licensees of currently operating 

PWRs utilizing the requirements of 10 

CFR 50.61a in lieu of the requirements 

of 10 CFR 50.61 would be subject to all 

of the proposed requirements in this 

rulemaking. 

An estimate of the number of annual 

responses: 2. 

The estimated number of annual 

respondents: 1. 

An estimate of the total number of 

hours needed annually to complete the 

requirement or request: 363 hours (253 

hours annually for record keeping plus 

110 hours annually for reporting). 

Abstract: The NRC is proposing to 

amend its regulations to provide 

updated fracture toughness 

requirements for protection against PTS 

events for PWR pressure vessels. The 

supplemental proposed rule would 

provide new PTS requirements based on 

updated analysis methods. This action 

is necessary because the existing 

requirements are based on unnecessarily 

conservative probabilistic fracture 

mechanics analyses. This action is 

expected to reduce regulatory burden 

for licensees, specifically those 

licensees that expect to exceed the 

existing requirements before the 

expiration of their licenses. These new 

requirements would be utilized by 

licensees of currently operating PWRs as 

an alternative to complying with the 

existing requirements. 

The NRC is seeking public comment 

on the potential impact of the 

information collections contained in 

this supplemental proposed rule and on 

the following issues: 

1. Is the proposed information 

collection necessary for the proper 

performance of the functions of the 

NRC, including whether the information 

will have practical utility? 

2. Estimate of burden? 

3. Is there a way to enhance the 

quality, utility, and clarity of the 

information to be collected? 

4. How can the burden of the 

information collection be minimized, 

including the use of automated 

collection techniques? 


rwilkins on PROD1PC63 with PROPOSALS
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 46563 

A copy of the OMB clearance package 

may be viewed free of charge at the NRC 

Public Document Room, One White 

Flint North, 11555 Rockville Pike, Room 

O–1F21, Rockville, MD 20852. The 

OMB clearance package and rule are 

available at the NRC worldwide Web 

site: http://www.nrc.gov/public-involve/

doc-comment/omb/index.html. The 

document will be available on the NRC 

home page site for 60 days after the 

signature date of this notice. 

Send comments on any aspect of 

these proposed information collections, 

including suggestions for reducing the 

burden and on the above issues, by 

September 10, 2008. Comments received 

after this date will be considered if it is 

practical to do so, but assurance of 

consideration cannot be given to 

comments received after this date. 

Comments submitted in writing or in 

electronic form will be made available 

for public inspection. Because your 

comments will not be edited to remove 

any identifying or contact information, 

the NRC cautions you against including 

any information in your submission that 

you do not want to be publicly 

disclosed. Comments submitted should 

reference Docket No. NRC–2007–0008. 

Comments can be submitted in 

electronic form via the Federal eRulemaking Portal at http:// 

www.regulations.gov by search for 

Docket No. NRC–2007–0008. Comments 

can be mailed to NRC Clearance Officer, 

Russell Nichols (T–5F52), U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001. Questions about the 

information collection requirements 

may be directed to the NRC Clearance 

Officer, Russell Nichols (T–5 F52), U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, by 

telephone at (301) 415–6874, or by email to INFOCOLLECTS.Resource@

nrc.gov. Comments can be mailed to the 

Desk Officer, Office of Information and 

Regulatory Affairs, NEOB–10202, 

(3150–0011), Office of Management and 

Budget, Washington, DC 20503, or by email to Nathan_J._Frey@omb.eop.gov, or 

by telephone at (202) 395–7345. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a request for information or an 

information collection requirement 

unless the requesting document 

displays a currently valid OMB control 

number. 

XII. Regulatory Analysis 

The NRC has issued a supplemental 

regulatory analysis for this 

supplemental proposed rulemaking. The 

analysis examines the costs and benefits 

of the alternatives considered by the 

NRC. The NRC requests public 

comments on this supplemental draft 

regulatory analysis. Availability of the 

supplemental regulatory analysis is 

provided in Section VII of this notice. 

Comments on the supplemental draft 

regulatory analysis may be submitted to 

the NRC as indicated under the 

ADDRESSES heading of this notice. 

XIII. Regulatory Flexibility Act 

Certification 

In accordance with the Regulatory 

Flexibility Act (5 U.S.C. 605(b)), the 

NRC certifies that this rule would not, 

if promulgated, have a significant 

economic impact on a substantial 

number of small entities. This 

supplemental proposed rule would 

affect only the licensing and operation 

of currently operating nuclear power 

plants. The companies that own these 

plants do not fall within the scope of the 

definition of ‘‘small entities’’ set forth in 

the Regulatory Flexibility Act or the size 

standards established by the NRC (10 

CFR 2.810). 

XIV. Backfit Analysis 

The NRC has determined that the 

requirements in this supplemental 

proposed rule would not constitute 

backfitting as defined in 10 CFR 

50.109(a)(1). Therefore, a backfit 

analysis has not been prepared for this 

proposed rule. 

The requirements of the current PTS 

rule, 10 CFR 50.61, would continue to 

apply to all PWR licensees and would 

not change as a result of this 

supplemental proposed rule. The 

requirements of the proposed PTS rule, 

including those in the supplemental 

proposed rule, would not be required, 

but could be utilized by PWR licensees 

with currently operating plants. 

Licensees choosing to implement the 

proposed PTS rule would be required to 

comply with its requirements as an 

alternative to complying with the 

requirements of the current PTS rule. 

Because the proposed PTS rule would 

not be mandatory for any PWR licensee, 

but rather could be voluntarily 

implemented, the NRC finds that this 

amendment would not constitute 

backfitting. 

List of Subjects for 10 CFR Part 50 

Antitrust, Classified information, 

Criminal penalties, Fire protection, 

Intergovernmental relations, Nuclear 

power plants and reactors, Radiation 

protection, Reactor siting criteria, 

Reporting and recordkeeping 

requirements. 

For the reasons set out in the 

preamble and under the authority of the 

Atomic Energy Act of 1954, as amended; 

the Energy Reorganization Act of 1974, 

as amended; and 5 U.S.C. 553; the NRC 

is proposing to adopt the following 

amendments to 10 CFR part 50. 

PART 50—DOMESTIC LICENSING OF 

PRODUCTION AND UTILIZATION 

FACILITIES 

1. The authority citation for part 50 

continues to read as follows: 

Authority: Secs. 102, 103, 104, 105, 161, 

182, 183, 186, 189, 68 Stat. 936, 937, 938, 

948, 953, 954, 955, 956, as amended, sec. 

234, 83 Stat. 444, as amended (42 U.S.C. 

2132, 2133, 2134, 2135, 2201, 2232, 2233, 

2236, 2239, 2282); secs. 201, as amended, 

202, 206, 88 Stat. 1242, as amended, 1244, 

1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 

112 Stat. 2750 (44 U.S.C. 3504 note); sec. 

651(e), Pub. L. 109–58, 119 Stat. 806–810 (42 

U.S.C. 2014, 2021, 2021b, 2111). 

Section 50.7 also issued under Pub. L. 95–

601, sec. 10, 92 Stat. 2951 as amended by 

Pub. L. 102–486, sec. 2902, 106 Stat. 3123 (42 

U.S.C. 5841). Section 50.10 also issued under 

secs. 101, 185, 68 Stat. 955, as amended (42 

U.S.C. 2131, 2235); sec. 102, Pub. L. 91–190, 

83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 

50.54(dd), and 50.103 also issued under sec. 

108, 68 Stat. 939, as amended (42 U.S.C. 

2138). Sections 50.23, 50.35, 50.55, and 50.56 

also issued under sec. 185, 68 Stat. 955 (42 

U.S.C. 2235). Sections 50.33a, 50.55a and 

appendix Q also issued under sec. 102, Pub. 

L. 91–190, 83 Stat. 853 (42 U.S.C. 4332). 

Sections 50.34 and 50.54 also issued under 

sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). 

Sections 50.58, 50.91, and 50.92 also issued 

under Pub. L. 97–415, 96 Stat. 2073 (42 

U.S.C. 2239). Section 50.78 also issued under 

sec. 122, 68 Stat. 939 (42 U.S.C. 2152). 

Sections 50.80–50.81 also issued under sec. 

184, 68 Stat. 954, as amended (42 U.S.C. 

2234). Appendix F also issued under sec. 

187, 68 Stat. 955 (42 U.S.C. 2237). 

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

follows: 

§ 50.8 Information collection 

requirements: OMB approval. 

* * * * * 

(b) The approved information 

collection requirements contained in 

this part appear in §§ 50.30, 50.33, 

50.34, 50.34a, 50.35, 50.36, 50.36a, 

50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 

50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 

50.61a, 50.62, 50.63, 50.64, 50.65, 50.66, 

50.68, 50.69, 50.70, 50.71, 50.72, 50.74, 

50.75, 50.80, 50.82, 50.90, 50.91, 50.120, 

and appendices A, B, E, G, H, I, J, K, M, 

N,O, Q, R, and S to this part. 

* * * * * 

3. Section 50.61a is added to read as 

follows: 

§ 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 


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46564 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 

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. 

(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) ∆T30 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 licensee of a 

pressurized water nuclear power 

reactor, whose original operating license 

was issued prior to [EFFECTIVE DATE 

OF FINAL RULE], and the holder of any 

operating license issued under this part 

or part 54 for the Watts Bar Unit 2 

facility, may utilize the requirements of 

this section as an alternative to the 

requirements of 10 CFR 50.61. 

(c) Request for Approval. Prior to 

implementation of this section, each 

licensee shall submit a request for 

approval in the form of a license 

amendment together with the 

documentation required by paragraphs 

(c)(1), (c)(2), and (c)(3) of this section for 

review and approval to the Director, 

Office of Nuclear Reactor Regulation 

(Director). The information required by 

paragraphs (c)(1), (c)(2), and (c)(3) of 

this section must be submitted for 

review and approval by the Director at 

least three years before the limiting 

RTPTS value calculated under 10 CFR 

50.61 is projected to exceed the PTS 

screening criteria in 10 CFR 50.61 for 

plants licensed under this part. 

(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 reassessment 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 reanalysis in 

accordance with paragraphs (e)(4) and 

(e)(5) of this section is required within 

one year of the subsequent ASME Code 

inspection. 

(3) If the value of RTMAX–X is 

projected to exceed the PTS screening 

criteria, then the licensee shall 

implement those flux reduction 


rwilkins on PROD1PC63 with PROPOSALS
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 46565 

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 

Continued 

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 

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-eighths of the 

reactor vessel thickness from the 

interior surface are within the allowable 

values in ASME Code, Section XI, Table 

IWB–3510–1. 

(4) The licensee shall perform 

analyses to demonstrate that the reactor 

vessel will have a through-wall crack 

frequency (TWCF) of less than 1 × 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 


rwilkins on PROD1PC63 with PROPOSALS
46566 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 

stops at the maximum flaw size modeled for these 

materials in the technical basis supporting this rule. 

3Because flaws greater than three-eighths of the 

vessel wall thickness from the inside surface do not 

contribute to TWCF, flaws greater than threeeighths of the vessel wall thickness from the inside 

surface need not be analyzed for their contribution 

to PTS. 

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. 

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 threeeighths 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-eighths of the 

vessel thickness from the inner surface, 

and may also take into account other 

reactor vessel-specific information, 

including fracture toughness 

information. 

(f) Calculation of RTMAX–X values. 

Each licensee shall calculate RTMAX–X

values for each reactor vessel beltline 

material using 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 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)(vi) 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 (jtFL) 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 jtFL 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 jt 

occurring for each material along the 

circumferential weld fusion line. 

(3) The values of Cu, Mn, P, and Ni 

in Equations 6 and 7 of this section 

must represent the best estimate values 

for the material weight percentages. For 

a plate or forging, the best estimate 

value is normally the mean of the 

measured values for that plate or 

forging. For a weld, the best estimate 

value is normally the mean of the 

measured values for a weld deposit 

made using the same weld wire heat 

number as the critical vessel weld. If 

these values are not available, either the 

upper limiting values given in the 

material specification to which the 

vessel material was fabricated, or 

conservative estimates (mean plus one 

standard deviation) based on generic 

data 4 as shown in Table 4 of this section 

for P and Mn, must be used. 

(4) 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 satisfy the criteria 

described in paragraphs (f)(6)(i)(A) and 

(f)(6)(i)(B) of this section: 

(A) The surveillance material must be 

a heat-specific match for one or more of 

the materials for which RTMAX–X is 

being calculated. The 30-foot-pound 

transition temperature must be 

determined as specified by the 

requirements of 10 CFR Part 50, 

Appendix H. 

(B) If three or more surveillance data 

points measured at three or more 

different neutron fluences exist for a 

specific material, the licensee shall 

determine if the surveillance data show 

a significantly different trend than the 

embrittlement model predicts. This 

must be achieved by evaluating the 

surveillance data for consistency with 

the embrittlement model by following 

the procedures specified by paragraphs 

(f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of this 

section. If fewer than three surveillance 

data points exist for a specific material, 

then the embrittlement model must be 

used without performing the 

consistency check. 

(ii) The licensee shall estimate the 

mean deviation from the embrittlement 

model for the specific data set (i.e. , a 

group of surveillance data points 

representative of a given material). The 

mean deviation from the embrittlement 

model for a given data set must be 

calculated using Equations 8 and 9 of 

this section. The mean deviation for the 

data set must be compared to the 

maximum heat-average residual given in 

Table 5 or derived using Equation 10 of 

this section. The maximum heat-average 

residual is based on the material group 

into which the surveillance material 

falls and the number of surveillance 

data points. The surveillance data 

analysis must use the criteria in 

paragraphs (f)(6)(v) and (f)(6)(vi) of this 

section. For surveillance data sets with 

greater than 8 shift points, the 

maximum credible heat-average residual 

must be calculated using Equation 10 of 

this section. The value of s used in 

Equation 10 of this section must be 

obtained from Table 5 of this section. 

(iii) The licensee shall estimate the 

slope of the embrittlement model 

residuals (estimated using Equation 8) 

plotted as a function of the base 10 

logarithm of neutron fluence for the 

specific data set. The licensee shall 

estimate the T-statistic for this slope 

(TSURV) using Equation 11 and compare 

this value to the maximum permissible 

T-statistic (TMAX) in Table 6. The 

surveillance data analysis must follow 

the criteria in paragraphs (f)(6)(v) and 

(f)(6)(vi) of this section. For surveillance 

data sets with greater than 15 shift 

points, the TMAX value must be 

calculated using Student’s T 

distribution with a significance level (a) 

of 1 percent for a one-tailed test. 


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Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 46567 

(iv) The licensee shall estimate the 

two largest positive deviations (i.e. , 

outliers) from the embrittlement model 

for the specific data set using Equations 

8 and 12. The licensee shall compare 

the largest normalized residual (r*) to 

the appropriate allowable value from 

the third column in Table 7 and the 

second largest normalized residual to 

the appropriate allowable value from 

the second column in Table 7. The 

surveillance data analysis must follow 

the criteria in paragraphs (f)(6)(v) and 

(f)(6)(vi) of this section. 

(v) The DT30 value must be 

determined using Equations 5, 6, and 7 

of this section if all three of the 

following criteria are satisfied: 

(A) The mean deviation from the 

embrittlement model for the data set is 

equal to or less than the value in Table 

5 or the value derived using Equation 10 

of this section; 

(B) The T-statistic for the slope 

(TSURV) estimated using Equation 11 is 

equal to or less than the maximum 

permissible T-statistic (TMAX) in Table 

6; and 

(C) The largest normalized residual 

value is equal to or less than the 

appropriate allowable value from the 

third column in Table 7 and the second 

largest normalized residual value is 

equal to or less than the appropriate 

allowable value from the second column 

in Table 7. 

(vi) If any of the criteria described in 

paragraph (f)(6)(v) of this section are not 

satisfied, the licensee shall review the 

data base for that heat in detail, 

including all parameters used in 

Equations 4, 5, and 6 of this section and 

the data used to determine the baseline 

Charpy V-notch curve for the material in 

an unirradiated condition. The licensee 

shall submit an evaluation of the 

surveillance data and shall, on the basis 

of this review, propose DT30 and 

RTMAX–X values, considering their 

plant-specific surveillance data, to be 

used for evaluation relative to the 

acceptance criteria of this rule. These 

evaluations shall be submitted for the 

review and approval by the Director at 

the time of the initial application. For 

each surveillance capsule removed from 

the reactor vessel after the submittal of 

the initial application, the licensee shall 

perform the analyses required by 

paragraph (f)(6) of this section. The 

analyses must be submitted for the 

review and approval by the Director in 

the form of a license amendment, and 

must be submitted no later than two 

years after the capsule is withdrawn 

from the vessel. 

(7) The licensee shall report any 

information that significantly improves 

the accuracy of the RTMAX–X value to 

the Director. Any value of RTMAX–X that 

has been modified as specified in 

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

subject to the approval of the Director 

when used as provided in this section. 

(g) Equations and variables used in 

this section. 

Equation 1: RT MAX RT MAX-AW NDT(u) - plate - = +  ∆T t 30 plate FL ( ) φ  

  + ( )   { } , RT T t NDT u FL ( ) - axial weld - axial weld ∆ 30 φ

Equation 2: RTMAX-PL - - = + RT T t NDT u plate plate MAX ( ) ∆ 30 ( ) φ

Equation 3: RTMAX-FO - - = + RT T t NDT u forging forging MAX ( ) ∆ 30 ( ) φ

Equation 4: RTMAX-CW - - plate = +  ( ) MAX RT T t  NDT u plate MAX ( ) ∆ 30 φ 

  + ( )  

 , , RT T t RT NDT u circweld MAX NDT u () () - circweld - - f ∆ 30 φ  orging - + ( )   { ∆T t 30 forging MAX φ }

Equation 5: T MD + CRP ∆ = 30

Equation 6: MD = A 1 ×− × ( ) 0 001718 1 6 13 × + ×× ( )× 2 471 0 5 . . . . T P Mn t C e φ

Equation 7: CRP = B 1 + 3.77 Ni1.191 × × ( )×f Cu P g Cu Ni t ( ) e ee , ,, × ( ) φ

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 = j for j ≥ 4.39 × 1010 n/cm2/sec 

= jt × (4.39 × 1010 / j)0.2595 for j < 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}) 


EP11AU08.016</MATH> EP11AU08.017</MATH> EP11AU08.018</MATH> EP11AU08.019</MATH> EP11AU08.020</MATH> EP11AU08.021</MATH> EP11AU08.022</MATH>rwilkins on PROD1PC63 with PROPOSALS
46568 Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 

Equation 8: Residual (r) = measured T predicted T ( ∆ ∆ 30 30 − by Equations 5, 6 and 7)

Equation 9: Mean deviation for a data set of n data points = 1/n ( )× =

∑ri

i

n

1

Equation 10: Maximum credible heat-average residual = 2.33σ/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. 

Equation 11: T m

se(m) SURV =

Where: 

m = the slope of a plot of all of the r values 

(estimated using Equation 8) versus the 

base 10 logarithm of the neutron fluence 

for each r value. The slope shall be 

estimated using the method of least 

squares. 

se(m) = the least squares estimate of the 

standard-error associated with the 

estimated slope value m. 

Equation 12: r* = r

σ

Where: 

r is defined using Equation 8 and s is given 

in Table 5. 

TABLE 1—PTS SCREENING CRITERIA

Product form and RTMAX-X values 

RTMAX-X limits [°F] for different vessel 

wall thicknesses 6 (TWALL) 

TWALL ≤ 

9.5in. 

9.5in. < 

TWALL ≤ 

10.5in. 

10.5in. < 

TWALL ≤ 

11.5in. 

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-CW7 ........................................................................................................... 312 277 269 

Forging with underclad cracks, RTMAX-FO ............................................................................................... 246 241 239 

6Wall thickness is the beltline wall thickness including the clad thickness. 

7 RTPTS limits contributes 1 × 10¥8 per reactor year to the ractor vessel TWCF. 

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 


rwilkins on PROD1PC63 with PROPOSALSEP11AU08.029</MATH> EP11AU08.030</MATH> EP11AU08.031</MATH> EP11AU08.023</MATH> EP11AU08.024</MATH>
Federal Register / Vol. 73, No. 155 / Monday, August 11, 2008 / Proposed Rules 46569 

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

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

0.20 .................................................... 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 

8Excluding underclad cracks in forgings. 

TABLE 4—CONSERVATIVE ESTIMATES

FOR CHEMICAL ELEMENT WEIGHT

PERCENTAGES

Materials P Mn 

Plates ................ 0.014 1.45 

TABLE 4—CONSERVATIVE ESTIMATES

FOR CHEMICAL ELEMENT WEIGHT

PERCENTAGES—Continued 

Materials P Mn 

Forgings ............ 0.016 1.11 

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

[Significance level = 1%] 

Material group s [°F] 

Number of available data points 

3 4 5 6 7 8 

Welds, for Cu > 0.072 ...................................................................................... 26.4 35.5 30.8 27.5 25.1 23.2 21.7 

Plates, for Cu > 0.072 ...................................................................................... 21.2 28.5 24.7 22.1 20.2 18.7 17.5 

Forgings, for Cu > 0.072 .................................................................................. 19.6 26.4 22.8 20.4 18.6 17.3 16.1 

Weld, Plate or Forging, for Cu ≤ 0.072 ........................................................... 18.6 25.0 21.7 19.4 17.7 16.4 15.3 

TABLE 6—TMAX VALUES FOR THE

SLOPE DEVIATION TEST

[Significance level = 1%] 

Number of available data points 

(n) TMAX

3 ................................................ 31.82 

4 ................................................ 6.96 

5 ................................................ 4.54 

6 ................................................ 3.75 

7 ................................................ 3.36 

8 ................................................ 3.14 

9 ................................................ 3.00 

10 .............................................. 2.90 

11 .............................................. 2.82 

12 .............................................. 2.76 

14 .............................................. 2.68 

15 .............................................. 2.65 

TABLE 7—THRESHOLD VALUES FOR

THE OUTLIER DEVIATION TEST (SIGNIFICANCE LEVEL = 1%) 

Number of available data points 

(n) 

Second 

largest allowable normalized residual value 

(r*) 

Largest allowable normalized residual value 

(r*) 

3 ........................ 1.55 2.71 

4 ........................ 1.73 2.81 

5 ........................ 1.84 2.88 

TABLE 7—THRESHOLD VALUES FOR

THE OUTLIER DEVIATION TEST (SIGNIFICANCE LEVEL = 1%)—Continued 

Number of available data points 

(n) 

Second 

largest allowable normalized residual value 

(r*) 

Largest allowable normalized residual value 

(r*) 

6 ........................ 1.93 2.93 

7 ........................ 2.00 2.98 

8 ........................ 2.05 3.02 

9 ........................ 2.11 3.06 

10 ...................... 2.16 3.09 

11 ...................... 2.19 3.12 

12 ...................... 2.23 3.14 

13 ...................... 2.26 3.17 

14 ...................... 2.29 3.19 

15 ...................... 2.32 3.21 

Dated at Rockville, Maryland, this 24th day 

of July 2008. 

For the Nuclear Regulatory Commission. 

R.W. Borchardt, 

Executive Director for Operations. 

[FR Doc. E8–18429 Filed 8–8–08; 8:45 am] 

BILLING CODE 7590–01–P 

DEPARTMENT OF TRANSPORTATION 

Federal Aviation Administration 

14 CFR Part 39 

[Docket No. FAA–2008–0857; Directorate 

Identifier 2007–NM–317–AD] 

RIN 2120–AA64 

Airworthiness Directives; Dornier 

Model 328–300 Airplanes 

AGENCY: Federal Aviation 

Administration (FAA), Department of 

Transportation (DOT). 

ACTION: Notice of proposed rulemaking 

(NPRM). 

SUMMARY: The FAA proposes to 

supersede an existing airworthiness 

directive (AD) that applies to all AvCraft 

Dornier Model 328–300 airplanes. The 

existing AD currently requires 

modifying the electrical wiring of the 

fuel pumps; installing insulation at the 

flow control and shut-off valves, and 

other components of the environmental 

control system; installing markings at 

fuel wiring harnesses; replacing the 

wiring harness of the auxiliary fuel 

system with a new wiring harness; and 

installing insulated couplings in the fuel 


rwilkins on PROD1PC63 with PROPOSALS