65456 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

NUCLEAR REGULATORY

COMMISSION

10 CFR Part 50

RIN 3150–AD57

Fracture Toughness Requirements for

Light Water Reactor Pressure Vessels

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final rule.

SUMMARY: The Nuclear Regulatory

Commission (NRC) is amending its

regulations for light-water-cooled

nuclear power plants to clarify several

items related to the fracture toughness

requirements for reactor pressure

vessels (RPV). The amendments will

clarify the pressurized thermal shock

(PTS) requirements, make changes to

the Fracture Toughness Requirements

and the Reactor Vessel Material

Surveillance Program Requirements,

and provide new requirements for

thermal annealing of a reactor pressure

vessel.

EFFECTIVE DATE: January 18, 1996.

FOR FURTHER INFORMATION CONTACT:

Alfred Taboada, Division of Engineering

Technology, Office of Nuclear

Regulatory Research, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–00001, telephone: (301) 415–

6014.

SUPPLEMENTARY INFORMATION: On

October 4, 1994 (59 FR 50513), the NRC

published in the Federal Register a

proposed amendment to clarify several

items related to fracture toughness

requirements for reactor pressure

vessels (RPV) and to add a new section

on thermal annealing of a reactor vessel

to 10 CFR Part 50.

Background

Maintaining the structural integrity of

the reactor pressure vessel of lightwater-cooled reactors is a critical

concern related to the safe operation of

nuclear power plants. To assure the

structural integrity of RPVs, NRC

regulations and regulatory guides have

been developed to provide analysis and

measurements methods and procedures

to establish that each RPV has adequate

safety margin for continued operation.

Structural integrity of a RPV is generally

assured through a fracture mechanics

evaluation, including measurement or

estimation of the fracture toughness of

the materials which compose the RPV.

However, the fracture toughness of the

RPV materials varies with time. As the

plant operates, neutrons escaping from

the reactor core impact the vessel

beltline materials (e.g. the materials that

surround the reactor core), causing

embrittlement of those materials. The

NRC’s regulations and regulatory guides

related to RPV integrity provide the

criteria and methods needed to estimate

the extent of the embrittlement, to

evaluate the consequences of the

embrittlement in terms of the structural

integrity of the RPV, and to provide

methods to mitigate the deleterious

effects of the embrittlement.

The NRC has several regulations and

regulatory guides that establish criteria

and procedures for assuring the

structural integrity of RPVs. With the

addition of the thermal annealing

requirements in this rule and several

regulatory guides, the regulatory

documents contribute to a

comprehensive set of regulations and

regulatory guidance pertaining to RPV

integrity.

This final rule adds requirements for

thermal annealing of the RPV as a

method for mitigating the effects of

neutron irradiation (10 CFR 50.66) and

amends the following:

1. The Pressurized Thermal Shock

(PTS) rule (10 CFR 50.61).

2. Appendix G of 10 CFR Part 50,

‘‘Fracture Toughness Requirements.’’

3. Appendix H of 10 CFR Part 50,

‘‘Reactor Vessel Material Surveillance

Program Requirements.’’

Overview of the Final Rule

PTS Rule (10 CFR 50.61)

This amendment to the PTS rule

makes three changes:

1. The rule incorporates in total, and

therefore makes binding by rule, the

method for determining the reference

temperature, RTNDT, including

treatment of the unirradiated RTNDT

value, the margin term, and the explicit

definition of ‘‘credible’’ surveillance

data, which is currently described in

Regulatory Guide 1.99, Revision 2.

2. The section is restructured to

improve clarity, with the requirements

section giving only the requirements for

the value for the reference temperature

for end of life fluence, RTPTS. The

method for calculating RTPTS is moved

to a new paragraph of the rule.

3. Thermal annealing is identified as

a method for mitigating the effects of

neutron irradiation, thereby reducing

RTPTS.

Thermal Annealing Rule (10 CFR 50.66)

The thermal annealing rule, 10 CFR

50.66, provides a consistent set of

requirements for the use of thermal

annealing to mitigate the effects of

neutron irradiation and replaces the

requirements for annealing in the

current Appendix G of 10 CFR Part 50.

The final rule requires, prior to

initiation of thermal annealing,

submittal of a Thermal Annealing

Report containing: (1) A Thermal

Annealing Operating Plan, (2) a

Requalification Inspection and Test

Program, (3) a Fracture Toughness

Recovery and Reembrittlement Trend

Assurance Program, and (4)

Identification of Unreviewed Safety

Questions and Technical Specifications

Changes. The report must be submitted

at least 3 years before the date at which

the limiting fracture toughness criteria

in 50.61 and Appendix G to Part 50

would be exceeded. This 3-year period

is specified to provide the NRC staff

with sufficient time to review the

thermal annealing program. Under

§ 50.66(a), the NRC will, within three

years of submission of a licensee’s

Thermal Annealing Report, document

its views on the plan, including whether

thermal annealing constitutes an

unreviewed safety question.

In order to provide for public

participation in the regulatory process,

Section 50.66(f)(1) requires that the NRC

hold a public meeting a minimum of 30

days before the licensee starts to thermal

anneal the reactor vessel. The

Commission will notify and solicit

comments from cognizant local and

state governments, and will publish a

notice in the Federal Register and in a

forum, such as local newspapers, which

is readily accessible to individuals in

the vicinity of the site, in order to solicit

comments from the public.

The thermal annealing operating plan

must include an evaluation of the effects

of temperature, and of mechanical and

thermal stresses on the reactor and

associated equipment such as

containment, the biological shield, and

attached piping, to demonstrate that the

operability of the reactor will not be

detrimentally affected. The bounding

conditions of the temperatures and

times used in this analysis define the

proposed annealing conditions. If these

conditions are exceeded during the

vessel annealing, then the evaluation

would no longer be valid, and the

acceptability of the actual vessel

annealing would have to be

demonstrated as discussed below in the

next paragraph.

Upon completion of the thermal

annealing, the licensee must confirm in

writing to the Director, Office of Nuclear

Reactor Regulation (NRR), that the

thermal annealing was performed in

accordance with the Thermal Annealing

Operating Plan and the Requalification

Inspection and Test Program. Within 15

days of the licensee’s written

confirmation that the thermal annealing

was completed in accordance with the
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65457

Thermal Annealing Plan, and prior to

restart, the NRC shall: (1) Briefly

document whether the thermal

annealing was performed in compliance

with the licensee’s Thermal Annealing

Operating Plan and the Requalification

Inspection and Test Program, with the

documentation to be placed in the NRC

public document room, and (2) hold a

public meeting to: (1) permit the

licensee to explain the results of the

reactor vessel annealing to the NRC and

the public, (2) allow the NRC to discuss

its inspection of the reactor vessel

annealing, and (3) provide an

opportunity for the public to comment

to the NRC on the thermal annealing.

The licensee may restart its reactor after

the meeting has been completed, unless

the NRC orders otherwise. Within 45

days of the licensee’s written

confirmation that the thermal annealing

was completed in accordance with the

Thermal Annealing Operating plan and

the Requalification Inspection and Test

Program, the NRC staff shall complete

full documentation of the NRC’s

inspection of the licensee’s annealing

process and place the documentation in

the Public Document Room.

If the thermal annealing was

completed but not performed in

accordance with the Thermal Annealing

Operating Plan and the Requalification

Inspection and Test Program, including

the bounding conditions of the

temperature and times as discussed

above, the licensee must submit a

summary of lack of compliance and a

justification for subsequent operations.

The licensee must also identify any

changes to the facility which are

attributable to the noncompliances

which constitute unreviewed safety

questions and any changes to the

technical specifications which are

required for operation as a result of the

noncompliances. This identification

does not relieve the licensee from

complying with applicable requirements

of the Commission regulations and the

operating license, and if, as a result of

the annealing operation, these

requirements cannot be met, the

licensee must obtain the appropriate

exemption per 10 CFR 50.12. If

unreviewed safety questions or changes

to technical specifications are not

identified as necessary for resumed

operation, the licensee may restart after

the NRC staff places a summary of its

inspection of the thermal annealing in

the Public Document Room, and the

NRC holds a public meeting on the

thermal annealing. On the other hand, if

unreviewed safety questions or changes

to technical specifications are identified

as necessary for resumed operation, the

licensee may restart only after the

Director of NRR authorizes restart, the

summary of the NRC staff inspection is

placed in the public document room,

and a public meeting on the thermal

annealing is held.

The final Thermal Annealing Rule

also sets forth the requirements that a

licensee must follow if the thermal

annealing was terminated prior to

completion. In general, the process and

requirements for partial annealing are

analogous to the situations where the

thermal annealing was completed; viz.,

where the partial annealing was

otherwise performed in compliance

with the Thermal Annealing Operating

Plan and relevant portions of the

Requalification Inspection and Test

Program, the licensee submits written

confirmation of such compliance and

may restart following, inter alia, holding

of a public meeting on the annealing. By

contrast, where the partial annealing

was not performed in accordance with

the Thermal Annealing Operating Plan

and relevant portions of the

Requalification Inspection and Test

Program, the licensee is required to

submit a summary of lack of compliance

and a justification for subsequent

operations, and identify any changes to

the facility which are attributable to the

noncompliances which constitute

unreviewed safety questions and

changes to the technical specifications

which are required for operation as a

result of the noncompliances with the

Thermal Annealing Operating Plan and

relevant portions of the Requalification

Inspection and Test Program. If

Unreviewed Safety Questions and/or

changes to technical specifications are

identified as necessary for resumed

operation, the licensee may restart only

after the Director of NRR authorizes

restart and the public meeting on the

thermal annealing is held.

Every licensee that either completes a

thermal annealing or terminates an

annealing but elects to take full or

partial credit for the annealing shall

provide a Thermal Annealing Results

Report detailing: (1) The time and

temperature profile of the actual thermal

anneal, (2) the post-anneal RTNDT and

Charpy upper shelf energy values of the

reactor material to be used in

subsequent operations, (3) the projected

post-anneal reembrittlement trends for

both RTNDT and Charpy upper-shelf

energy, and (4) the projected values of

RTPTS and Charpy upper-shelf energy at

the end of the proposed period of

operation addressed in the application.

The report must be submitted within

three months of completing the thermal

anneal, unless an extension is

authorized by the Director, NRR.

Two items of particular importance to

the overall annealing are the recovery of

fracture toughness and the degree of

reembrittlement of the RPV beltline

materials. This final rule provides

alternative methods for determining

these values, ranging from assessments

using plant-specific materials to an

assessment using a generic computation.

Two methods provided for evaluating

annealing recovery are experimental

methods to determine plant-specific

annealing recovery, and a third method

is a generic computational method.

Experimental methods and the

computational method are also provided

for estimating recovery of RTNDT and

Charpy upper-shelf energy of the

beltline materials. The experimental

methods for estimating recovery of

RTNDT and the Charpy upper-shelf

energy utilize either surveillance

program specimens or material removed

from the vessel beltline. The

experimental methods provide a plantspecific estimate of recovery, rather than

the generic value evaluated from the

computational method. This final rule

requires that surveillance specimens

from ‘‘credible’’ surveillance programs

must be used to develop plant-specific

recovery data, if such specimens are

available. This final rule does not

require the removal of material from the

RPV beltline to permit plant-specific

evaluation of recovery.

As described previously, the

computational method requires

appropriate justification.

Post anneal reembrittlement trends of

both the RTNDT and the Charpy upper

shelf energy must be estimated and

monitored using a surveillance program

described in the Thermal Annealing

Report.

The reactor pressure vessel is perhaps

the most important single component in

the reactor coolant system. As such,

ensuring its integrity is a fundamental

element of plant safety. Thermal

annealing is a positive action that could

be taken to reduce the level of

embrittlement in the pressure vessel

beltline and, thereby, improve the

ability of a pressure vessel to withstand

accident loadings. While thermal

annealing is a positive action, there are

numerous complex technical questions

regarding its application in the U.S. that

are unanswered.

Thermal annealing of a commercial

reactor pressure vessel has never been

accomplished in the United States.

Thermal annealing has been

successfully employed in Eastern

Europe and Russia on Russian-designed

pressure vessels. However, there are

significant differences between the U.S.

and Russian designs in terms of the
65458 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

geometry of the pressure vessels, the

attached piping, and the surrounding

structures. The staff has observed one of

these annealing operations. While

informative, the East European and

Russian experience does not provide

answers to all of the potential questions

related to annealing of U.S. designed

pressure vessels.

Research analyses performed

previously indicated the potential for

plastic deformation of the main coolant

piping for a typical U.S. plant design

and anticipated annealing conditions.

There are also questions regarding how

thermal growth of the pressure vessel is

treated, and the adequacy of the thermal

and stress analyses used to predict

response of the overall system under

thermal annealing conditions.

Additionally, there may be questions in

other areas such as temperature limits

for the concrete structures, and potential

radiological hazards associated with

removing and storing the reactor

internals during the annealing process,

and fire hazards associated with heating

the vessel.

Recognition of the numerous complex

technical questions related to thermal

annealing, and of the potential benefits

for operating nuclear power plants, has

resulted in a cooperative effort, funded

by the U.S. Department of Energy and

the industry, to perform Annealing

Demonstration Projects. Projects are

planned to demonstrate two different

annealing processes, evaluating heater

designs and vessel designs. It is

anticipated that the annealing

demonstration projects will answer

many of the generic questions regarding

thermal annealing of U.S. pressure

vessel and piping designs.

The thermal annealing report,

required by the thermal annealing rule,

is designed to facilitate a detailed

review by the licensee of plant-specific

questions and considerations in

performing a thermal annealing. The

proposed rule specifically discusses the

potential for unreviewed safety

questions and technical specification

changes that may result from or be

related to thermal annealing of the

reactor pressure vessel. With

completion of the demonstration

projects and as the staff and industry

gain experience with thermal annealing,

many of the issues related to annealing

will be better understood and related

questions will be answered. However,

until this experience is realized, the

staff will critically review licensee

determinations regarding unreviewed

safety questions and the need for

technical specification changes

associated with each proposed thermal

annealing.

The thermal annealing rule has been

structured to provide time for the staff

to thoroughly review the licensee’s

annealing plan and determination

regarding unreviewed safety questions

and the need for technical specification

changes. If the staff identifies an

unreviewed safety question or the need

for a technical specification change, the

licensee would be so notified and the

existing NRC regulatory practices would

be invoked to address the issues.

Appendix G of 10 CFR Part 50

Appendix G of 10 CFR Part 50

specifies fracture toughness

requirements for ferritic materials of

pressure-retaining components of the

reactor coolant pressure boundary of

light-water-cooled nuclear power

reactors. These requirements provide

adequate margins of safety during any

condition of normal operation,

including anticipated operational

occurrences and system hydrostatic

tests. The amendments to Appendix G

are principally of a clarifying or a

restructuring nature. Requirements for

‘‘volumetric inspection’’ and

‘‘additional evidence of fracture

toughness’’ have been removed because

they were unnecessary, given the

inspection and performance

demonstration programs currently

required under 10 CFR 50.55a. The

‘‘additional evidence of fracture

toughness’’ requirement in Section

V.C.2 is incorporated in the ‘‘equivalent

margins’’ analysis in Section IV.A.1 as

a provisional method for developing

fracture toughness data needed for that

analysis.

The pressure-temperature and

minimum permissible temperature

requirements in Section IV have been

restructured. The principal feature is the

addition of a table which summarizes

the pressure-temperature limit

requirements and minimum

temperature requirements as a function

of the plant operating condition, the

vessel pressure, whether fuel is in the

vessel, and whether the core is critical.

In addition, Section IV has been

reworded to clarify the minimum

permissible temperature requirement by

indicating the criteria for use in

determining the location in the

component or material which must

satisfy the minimum temperature

requirement. This minimum

temperature is defined in Section IV as

the metal temperature of the controlling

material in the region which has the

least favorable combination of stress and

temperature for the appropriate plant

condition. An explicit statement has

been added to require that pressure and

leak tests of the reactor pressure vessel

required by Section XI of the American

Society of Mechanical Engineers Boiler

& Pressure Vessel (B&PV) Code (ASME

Code) must be completed before the

core is critical.

The requirement that all pressure and

leak tests of the RPV required by Section

XI of the ASME Code must be

completed before the core is critical is

intended to prohibit the use of nuclear

heat, i.e., core criticality, in the conduct

of ASME, Section XI pressure and leak

tests. The use of nuclear heat before the

completion of such tests is not

consistent with basic defense-in-depth

nuclear safety principle for several

reasons, including the hindrance of

finding leaks with the vessel at such a

high temperature and the potential for

exacerbating the consequences of a

vessel rupture (in the extremely

unlikely event that it should occur) by

having the core critical. The explicit

prohibition of nuclear heat in these

cases was discussed in a letter to

Messrs. Reynolds and Stenger of the

Nuclear Utility Backfitting and Reform

Group from James M. Taylor, Executive

Director of Operations, dated February

2, 1990.

The current requirements in 10 CFR

Part 50, Appendix G, Section V. D. with

respect to reactor vessel thermal

annealing are being replaced by a

sentence which references the new

Thermal Annealing rule, 10 CFR 50.66.

Appendix H of 10 CFR Part 50

Appendix H of 10 CFR Part 50,

‘‘Reactor Vessel Material Surveillance

Program Requirements’’ provides the

rules for monitoring the changes in the

fracture toughness properties of the RPV

beltline materials due to irradiation

embrittlement using a surveillance

program. Appendix H references

American Society for Testing and

Materials (ASTM) standard E 185

(‘‘Standard Practice for Conducting

Surveillance Tests for Light-Water

Cooled Nuclear Power Reactor Vessels’’)

for many of the detailed requirements of

surveillance programs, and permits the

use of integrated surveillance programs,

wherein surveillance program capsules

for one reactor are irradiated in another

reactor.

Integrated surveillance programs are

permitted under Section II.C of

Appendix H of 10 CFR Part 50. One

provision of this section is that ‘‘the

amount of testing may be reduced if the

initial results agree with predictions.’’

This provision was deleted, although

previous authorizations granted by the

Director, Office of Nuclear Reactor

Regulation, continue in effect.

A second change to Appendix H

restructures Section II.C to clarify the
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65459

requirements for integrated surveillance

programs.

The other principal change to

Appendix H clarifies the version of

ASTM Standard E 185 that applies to

the various portions of the surveillance

programs. Appendix H recognizes the

need to separate surveillance programs

into two essential parts, specifically the

design of the program and the

subsequent testing and reporting of

results from the surveillance capsules.

Because the design of the surveillance

program cannot be changed once the

program is in place, the requirements

for design of the surveillance program

are static for each plant. However, the

testing and reporting requirements are

updated along with technical

improvements made to ASTM standard

E 185.

Request for Public Comments

At the request of the Commission, the

proposed rule contained a request for

public comments on the following

specific issues related to the proposed

regulation on thermal annealing:

1. The technical adequacy of the

staff’s guidance;

2. The sufficiency of the guidance and

criteria to support a certification that if

satisfied, a plant with an annealed

vessel can safely resume operation;

3. Whether health and safety concerns

are best served by approval of the

thermal annealing plan or of readiness

for restart;

4. The preferred regulatory process

(including opportunities for public

participation) and the commenter’s

basis for recommending a particular

process; and

5. Whether there are health and safety

issues concerning thermal annealing

that cannot be addressed generically

and would warrant plant-specific

consideration.

The supplementary information

section of the proposed rule also

discussed the issue of opportunity for

public participation in regulating

thermal annealing of pressure vessels.

The response to the request for public

comments on these issues, along with

other items, are summarized below.

Summary of Comments

The following includes a summary of

the comments received on the proposed

rule, on the five issues identified by the

Commission, and on the options for

public participation in thermal

annealing.

Comments were received from nine

separate sources. These sources consist

of five utilities, the Nuclear Energy

Institute (NEI), the Nuclear Utility

Backfitting and Reform Group

(NUBARG) represented by the firm

Winston & Strawn, one public citizens

group (Ohio Citizens for Responsible

Energy (OCRE)), and one nuclear steam

system supplier (NSSS).

NEI provided detailed comments on

10 CFR 50.61, 10 CFR 50.66, Appendix

G to 10 CFR Part 50, and Appendix H

to 10 CFR Part 50, responded to the

request for comments on the five issues

related to thermal annealing and

included detailed comments on the

opportunities for public participation.

The five utilities and the NSSS

endorsed the NEI comments. Three of

the five utilities provided additional

comments on 10 CFR 50.61; one of the

five utilities provided additional

comments on 10 CFR Part 50, Appendix

G; two of the utilities provided

additional comments on 10 CFR Part 50,

Appendix H; and one of the five utilities

disagreed with the NEI position on the

opportunity for public participation and

submitted a separate comment. OCRE

provided comments on the opportunity

for public participation. NUBARG

provided comments on the backfitting

aspects of the proposed rule and the

staff’s backfit justification.

NEI and one of the utilities included

comments on the Draft Regulatory

Guide DG–1027, ‘‘Format and Content

of Application for Approval for Thermal

Annealing of Reactor Pressure Vessels,’’

that was discussed in the proposed rule.

These comments on Draft Regulatory

Guide DG–1027 are being reviewed by

the NRC staff and will be addressed

separately in the resolution of

comments on the regulatory guide.

The NRC reviewed the comments

received on the proposed rule, the

comments on the five questions related

to thermal annealing and the issue of

opportunities for public participation.

The resolution of these comments is

presented below.

PTS Rule (10 CFR 50.61)

Sixteen specific comments in the

submittals from NEI and three utilities

addressed 10 CFR 50.61. A general

comment argued that both the existing

10 CFR 50.61 and the proposed

modifications contained an excessive

amount of prescriptive technical detail

that limits licensee compliance

flexibility. The commenters proposed

that these prescriptive technical details

be removed from the rule and placed in

a regulatory guide. These commenters

suggested that the rule not be issued

until it has been written to contain only

those requirements essential to regulate

reactor pressure vessel embrittlement. A

number of comments suggested changes

that were clarifications to the proposed

rule, including proposals to clarify the

procedure for calculating the reference

temperatures in the preservice

condition, RTNDT, and, at end of reactor

life, RTPTS. One comment noted that the

proposed rule omitted part of the

procedure in Regulatory Guide 1.99,

presently being applied by the NRC, that

permits adjustments for differences in

chemistry between surveillance material

and the vessel material when using

credible surveillance data to calculate a

best fit chemistry factor for transition

temperature shifts due to irradiation.

Several comments proposed changes in

the criteria for establishing whether

surveillance material data is credible

that would result in a less restrictive

basis for using surveillance data in

determining the transition temperature

shift. The comments argued that the

proposed rule is ambiguous with respect

to the use of information from other

sources that contain limiting material

for a specific plant and that the NRC

must have the flexibility to approve use

of such information on a case-by-case

basis. Several comments proposed

limiting the basis for making changes of

RTPTS subject to the approval of the

Director, NRR.

The NRC recognizes that 10 CFR

50.61 contains an unusual amount of

prescriptive material and that the

comments proposing simplification

have merit. Some changes to the rule

have been made to provide flexibility,

where appropriate. The NRC staff is

evaluating subsequent changes that

would be more performance based.

However, the NRC staff believes that

this rule, as written, is needed to ensure

that plants apply the appropriate

method for determining RTPTS and that

the appropriate reference to the thermal

annealing rule be applied for the

pressurized thermal shock situation.

A number of clarifications were made

to the rule. The paragraphs dealing with

the determination of RTPTS were

modified to make clear that RTPTS is a

unique, end of life, case of RTNDT and

to clarify the procedure for determining

these values. As suggested, the

adjustment procedure was added to the

rule to permit accounting for differences

in chemistry between surveillance

materials and reactor vessel materials

when calculating chemistry factors.

With respect to the plant specific

material surveillance data that is

permitted to be used in a surveillance

program, the rule was modified to make

clear that such data includes results

from other plant’s surveillance programs

and test reactors. Several clarifications

were made to the criteria for

determining credible material. The NRC

determined that the requirements for

approval by the Director, NRR, for
65460 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

changes in RTPTS are appropriate and

should not be modified.

Thermal Annealing Rule (10 CFR 50.66)

Twelve individual comments were

received on the proposed Thermal

Annealing Rule, 10 CFR 50.66. These

comments included a number of

suggestions for clarification of details of

the proposed rule. Three of the

comments addressed the requirements

that, after the annealing operation, the

reembrittlement rate of the reactor

vessel due to neutron irradiation must

be estimated and must be monitored

using a surveillance program which

conforms to Appendix H of 10 CFR 50,

‘‘Reactor Vessel Materials Surveillance

Program.’’ The comments are

summarized as follows:

(1) The supplementary information

section for the proposed rule is silent on

what is acceptable if limiting material is

not available. The rule should provide

appropriate requirements on the method

for monitoring reembrittlement after

annealing for those plants that do not

have limiting material for their

surveillance program and the

monitoring plans should be consistent

with the preannealing surveillance

program approved by the NRC staff;

(2) Appendix H does not define an

acceptable post-anneal surveillance

program, the reference to Appendix H

should be deleted, and the post-anneal

surveillance program should be defined

in the annealing plan that is approved

by the staff; and

(3) The term reembrittlement rate is

unclear as to the period of time to be

used for its determination, and a

wording change is proposed for the

requirement that would relate change in

toughness to fluence accumulated after

the anneal.

Three of the comments addressed the

requirements in the proposed rule that

the Thermal Annealing Operation Plan

include time-temperature profiles which

represent the annealing conditions that

may not be exceeded during the

annealing operation and are to be used

for determining the amount of recovery

of the fracture toughness of the material

due to annealing. The comments

suggested that, instead of a single timetemperature profile, bounding time and

temperature conditions be established

for the maximum values that would be

used for thermal and stress analysis and

to verify the re-qualification inspection

and test program, and the minimum

values that would be used to establish

the amount of recovery of fracture

toughness and for reembrittlement rate

estimates. The bounding values would

be based upon the estimated

uncertainties in the times and

temperatures and the actual annealing

conditions should fall within these

bounds.

Two comments addressed the section

on Certification of Annealing

Effectiveness. One comment suggested

deleting the requirement in the

proposed rule for certification of the

annealing effectiveness and instead

adding a provision in the Thermal

Annealing Operating Plan that approval

prior to subsequent power operation be

required only if the anneal was not

performed in accordance with the

approved plan. The comment also

suggested that, if the licensee terminates

the annealing before achieving the

specified time but otherwise maintains

the annealing envelop such that no

concern exists for stress or thermal

damage, no additional constraints be

imposed on subsequent operations and

no credit be given for annealing. The

second comment suggested that (1) the

staff’s review of the annealing report

(certification report) need not be

completed prior to reinitiating power

operation if the anneal was performed

in accordance with the approved

Thermal Annealing Operating Plan, (2)

reporting and quantification of the

actual recovery results need not be

reported unless the vessel was at or

above the PTS screening criteria when

annealing was started, and (3) the

Thermal Annealing Operating Plan

should specify the minimum content

and a schedule for reporting the

annealing results. The commenter

provided a proposed list of criteria,

content, and schedule for reporting the

annealing results.

One comment stated that no guidance

was provided in the proposed rule on

what constitutes components ‘‘affected’’

by the annealing operation that are

required to be reported in the Thermal

Annealing Operating Plan. The

comment suggested alternative wording

that components to be reported should

be structures and components that are

expected to experience significant

temperature gradient or stress variations

during the thermal annealing operation.

One comment suggested qualifying the

provision in the proposed rule that the

effects of localized high temperatures

must be evaluated for changes in

thermal and mechanical properties of

the reactor vessel insulation for those

cases where such changes may be

negligible at annealing conditions. One

comment suggested that the use of

applicable material data, such as data

from integrated surveillance programs,

be an optional part of the computational

methods for determining fracture

toughness recovery.

The NRC reviewed the comments

received on the proposed rule in detail.

After consideration, the NRC reached

the conclusion that most of the

comments are not inconsistent with the

intent of the proposed rule and in some

cases reflect a need for clarification of

the rule. In these cases, alternative

wording that clarified the intent of the

rule was substituted in the text. With

respect to the comments on the

requirement that reembrittlement rate

after annealing must be monitored using

a surveillance program, the NRC is

aware that some plants do not have

limiting materials for their existing

preannealing surveillance programs. For

these situations the staff has approved

alternative surveillance plans on a caseby-case basis. Clearly, these plants will

not have limiting material for

surveillance programs for use in

determining reembrittlement rates after

annealing.

The NRC recognizes that Appendix H

of 10 CFR Part 50, which is referenced

in this rule, does not specifically

address the surveillance of an annealed

reactor vessel. However, the

requirements of Appendix H to 10 CFR

Part 50 apply to all reactors including

the specific case of an annealed reactor

vessel. To clarify the surveillance

requirements of an annealed plant, the

final rule has been modified to include,

as suggested, that the post-anneal

reembrittlement is to be monitored

using a surveillance program defined in

the Thermal Annealing Report and that

the surveillance program must conform

to the intent of Appendix H to 10 CFR

Part 50.

The term reembrittlement ‘‘rate’’ in

the proposed rule was intended to mean

the projected amount of reembrittlement

over a specific fluence period. It is

recognized that reembrittlement is not a

straight line function of fluence.

Determination of reembrittlement rate is

discussed in more detail in Draft

Regulatory Guide 1.162, ‘‘Format and

Content of Report for Thermal

Annealing of Reactor Pressure Vessels.’’

In Regulatory Guide 1.162, the approved

method for estimating the

reembrittlement rate, the lateral shift

method, results in the same

embrittlement trend as that used for the

pre-anneal operating period. To avoid

confusion the term ‘‘rate’’ has been

changed to ‘‘trend’’ in the final rule and

the regulatory guide.

The NRC agrees with the comments

that the time and temperature profile

required in the annealing operating plan

should be bounding values. In this

regard, Regulatory Guide DG–1027 calls

for the thermal annealing operating plan

to include identification of the
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65461

limitations and permitted variations in

temperature, time, heatup and

cooldown rate. For clarification, the

final rule has been modified to use the

terms ‘‘bounding conditions for times

and temperatures and heatup and

cooldown schedules’’ to describe

conditions that may not be exceeded

during the annealing operation, and the

lower limit time and temperature of the

actual anneal is used for determining

the projected recovery of fracture

toughness by annealing.

The NRC considers that the intent of

paragraphs (c), Completion or

Termination of Thermal Annealing, and

(d), Thermal Annealing Results Report,

of the final rule to be consistent with the

two comments on that subject. The final

rule does not require that the NRC

approve restart following the annealing

operation if the Thermal Annealing

Operating Plan and the Requalification

Inspection and Test Program was

complied with. The NRC accepts the

suggestion that the rule should be more

specific on the items the licensee should

include in the report and has included

the list in the final rule.

Finally, the NRC agrees with the

suggestion to make clear that a report is

not required if:

(1) The licensee terminates the anneal

prior to completion;

(2) The partial anneal was otherwise

in accordance with the Thermal

Annealing Plan;

(3) The licensee does not elect to take

credit for any recovery. A statement was

added to the Final Rule to cover the

early termination situation.

The NRC has accepted the suggested

clarifications of what constitutes an

‘‘affected’’ component and the

qualification on the requirement to

evaluate changes in properties on

reactor vessel insulation if these are

negligible. The NRC considers it

unnecessary to include a reference in

the rule to data from integrated

surveillance programs as an optional

part of the computational methods to

determine fracture toughness recovery.

Generic computational methods for this

purpose are provided in the Regulatory

Guide 1.162. However, the final rule

does not prohibit use of alternative

methods if adequate justification is

provided.

Appendix G to 10 CFR Part 50

Two comments were received on the

Appendix G to 10 CFR Part 50 of the

proposed rule. The NEI comment,

which was endorsed by five utilities and

one NSSS organization, included a table

with six items on Appendix G. The

other comment on Appendix G was

received from one of the five utilities.

Two of the comments identified

typographical errors and suggested a

change in organization to improve

clarity. One of the comments suggested

revising the rule to change the

definition of reference temperature,

RTNDT, for cases where plants do not

have data to comply with code

procedures for determining RTNDT. One

comment suggested a change in the title

of Table 1, ‘‘Pressure and Temperature

Requirements,’’ by adding to the title

‘‘For the Reactor Pressure Vessel’’ to

make clear that this table does not apply

to other components in the reactor

coolant pressure system and proposed

adding a footnote to the table for the

same purpose. One comment identified

an error in the minimum temperature

requirements for the hydrostatic and

leak testing of the pressure vessel

without fuel when the vessel pressure is

equal or below 20 percent of the vessel

design pressure. One of the comments

suggested that two of the entries in the

table were new requirements when the

table was intended to provide

clarification. The utility’s comment

disagreed with the proposed rule change

to prohibit the use of nuclear heat for

the performance of vessel leak and

hydrostatic testing. The utility

contended that using nuclear heat, by

providing a significant temperature

margin above the pressure and

temperature limit curves, greatly

reduces the probability of brittle fracture

and should be allowed.

The NRC corrected the typographical

errors and corrected the minimum

temperature requirement for the

hydrostatic and leak testing of the

pressure vessel at low vessel pressures

and without fuel. The title to Table 1

was changed, as suggested, for

clarification.

The NRC does not agree with the

proposal to change the definition of

RTNDT. The situation described in the

comment, when data is not available to

comply with code procedures, is

presently handled on a case-by-case

basis in accordance with MEB Branch

position, MEB 5–2. The NRC staff does

not agree with the comment that the two

requirements cited are new

requirements. Item 2.2.c. and Item 2.2.d

of Table 1 are in the existing ASME

code requirement and in Paragraph

IV.A.3. in the rule. The NRC also does

not agree with the utility’s comment

that using nuclear heat greatly reduces

the probability of brittle fracture. The

reasons for this are set forth in the

February 2, 1990, letter to Messrs.

Reynolds and Stenger of NUBARG from

James M. Taylor, Executive Director for

Operations.

Appendix H to 10 CFR Part 50

Three comments were received on

Appendix H to 10 CFR 50. The

comment from NEI was endorsed by the

five utilities and the NSSS. Two of the

five utilities submitted additional

comments. NEI and one utility

commented that the proposed change to

Paragraph III.B.1, which establishes the

applicable edition of ASTM standard E

185 for a reactor surveillance program,

constituted a backfit that would require

a substantial design change in the

surveillance program for those plants

fabricated to a code edition prior to

1973. The other two commenters

suggested new changes to Appendix H

to 10 CFR Part 50. One of the

commenters noted that an existing

provision in Appendix H to 10 CFR Part

50, not part of the proposed rule change,

dealing with requirements for attaching

capsule holders to the vessel wall is a

reiteration of a requirement in the

ASME Code and should be removed.

The other commenter suggested a new

change to Appendix H to 10 CFR Part

50 to add a statement to the criteria for

approval of an integrated surveillance

program that would permit the use of

surveillance specimens for extension of

license purposes. The commenter also

suggested that there is an apparent

conflict between Paragraph III.C.2. and

Paragraph III.C.3. that address

requirements for an integrated

surveillance.

The provision in the proposed rule

was changed and reference to ASTM E

185 73 was deleted to make clear that

the surveillance programs must be

designed to the edition of ASTM 185

that is current on the issue date of the

ASME Code to which the reactor vessel

was purchased or to a later edition

through 1982. The Commission agrees

with the industry comments that

imposing the ASTM E 185 1973 edition

is impractical because vessels

purchased prior to 1973 could not

necessarily comply with all of the

surveillance requirements in the 1973

edition of the ASTM standard. The NRC

staff believes that the provision in the

present rule on requirements for

attaching capsule holders to the reactor

vessel wall is required for clarity and

should not be deleted. The comments

related to the requirements for an

integrated surveillance program were

not persuasive to the NRC staff. The

existing provisions of the rule do not

preclude the application of the

integrated surveillance program for

extension of license purposes. The two

paragraphs purported to be in conflict

address separate items; one addresses

the number of materials to be irradiated,
65462 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

specimen types, and number of

specimens per reactor; the other

addresses amount of testing.

Request for Comments on Issues Related

to Thermal Annealing

Comments were received from NEI on

the five issues on thermal annealing that

were included in the proposed rule at

the Commission’s direction. In addition,

OCRE and one utility, Pacific Gas and

Electric, submitted comments on Issue

4, concerning the preferred regulatory

process (including opportunity for

public participation). Public Comments

on the five issues are summarized

below:

Issue 1: The technical adequacy of the

NRC staff’s guidance.

Comment: The detailed comments

submitted on 10 CFR 50.66 are

summarized in the Summary of

Comments section on the Thermal

Annealing Rule. In addition, NEI

suggested that draft Regulatory Guide,

DG–1027, be revised to include

acceptance criteria where an action is

required, but the acceptance criteria was

not defined. NEI further commented

that the re-embrittlement rate equation

(DG–1027, Equation 1) appeared to be

very conservative and would result in a

post-anneal operating life that is less

than industry believes justified.

Response: The NRC is concurrently

revising the noted draft regulatory guide

and will address this comment in the

resolution of comments for the guide.

Issue 2: The sufficiency of the

guidance and criteria to support a

certification that if satisfied, a plant

with an annealed vessel can safely

resume operation.

Comment: NEI noted that ‘‘The

reactor pressure vessel thermal

annealing rule and guide address

appropriate issues to assure public

health and safety and that the annealed

reactor pressure vessel may be safely

operated. The prior NRC staff approval

of the reactor vessel annealing plan

assures a clear process and criteria to

restart following the vessel anneal. The

licensee needs only to attest to

compliance with the approved plan

prior to resuming operations. The

resumption of operations should not be

needlessly delayed while a report

documenting performance of the vessel

anneal and recovery of the embrittled

material properties is confirmed,

because the vessel anneal will only

improve the material properties. The

final report should be submitted on a

schedule that considers when the vessel

would have exceeded the RTPTS or

uppershelf energy (USE) screening

criteria without an anneal. The material

property recovery will document prior

to the time when the vessel would have

exceeded the screening criteria, thereby

assuring that the vessel is safe to operate

at restart and for the duration justified

by the material embrittlement

recovery.’’

Response: NRC agrees with the NEI

comment, except NRC believes it is

necessary for the licensee to submit the

final report within three months of

completing or terminating the anneal,

unless an extension is authorized by the

Director, Office of Nuclear Reactor

Regulation.

Issue 3: Whether health and safety

concerns are best served by approval of

the thermal annealing plan or of

readiness for restart.

Comment: NEI noted that ‘‘The

performance of a reactor pressure vessel

anneal in accordance with an approved

annealing plan improves the public

health and safety by reducing the

probability of core melt frequency. This

improvement occurs because of the

increase in reactor vessel material

ductility. The amount of recovery

achieved by a thermal anneal will be

documented prior to the original date

when the reactor vessel would have

exceeded the PTS or USE screening

limit. Therefore, a demonstration for

‘‘restart readiness’’ is an extra burden

that will not provide any further

improvement of the public health and

safety.’’

Response: The NRC’s determination

as to the procedures for NRC review of

the Thermal Annealing Operation Plan,

Requalification Inspection and Test

Program and justification for restart

discussed below in further detail in the

Opportunities for Public Participation

section.

Issue 4: The preferred regulatory

process (including opportunities for

public participation) and the

commenter’s basis for recommending a

particular process.

Comment: NEI noted that ‘‘The

industry recommends that a hearing

opportunity be provided, but that it be

a non-adjudicatory, 10 CFR Part 2,

Subpart L type hearing on the docketed

record. The essential features of the

hearing process proposed are as follows.

The NRC would at time of receiving the

licensee proposed annealing plan issue

a Federal Register announcement that

staff is performing the review per 10

CFR 50.66. A Subpart L hearing could

be held, if requested by an intervener,

after the NRC staff has issued a safety

evaluation report on the licensee

annealing plan, but prior to

commencement of the reactor vessel

thermal annealing unless the NRC staff

makes a ‘‘no significant hazards

determination.’’ Enclosure 4 provides

additional details that support this

industry position.’’ Additional detailed

comments by NEI and the comments on

this subject by OCRE are discussed

under the Opportunities for Public

Participation heading.

Response: The rule provides for

public participation in the regulatory

process by incorporating a public

meeting on the Licensee’s Thermal

Annealing Report a minimum of 30 days

before the start of thermal annealing,

and a public meeting after the licensee

completes the anneal but before the

reactor is restarted. The opportunity for

public hearings in thermal annealing

should be limited to those cases where

there is an unreviewed safety question

or a change to the Technical

Specifications or where the licensee did

not comply with the Thermal Annealing

Operating Plan and Requalification

Inspection and Test Program. Expanded

discussion on this issue is provided

below under the Opportunities for

Public Participation heading.

Issue 5: Whether there are health and

safety issues concerning thermal

annealing that cannot be addressed

generically and would warrant plantspecific consideration.

Comment: NEI noted that ‘‘Thermal

annealing to reduce material irradiation

embrittlement is a well understood

metallurgical phenomenon. The

supporting thermal and stress analysis

used to demonstrate that the vessel is

not damaged during the anneal are

standard technologies used at nuclear

plants. Because thermal annealing uses

well understood technology, public

health and safety is reasonably

assured.’’

Response: The NRC agrees with this

comment.

Opportunities for Public Participation

The Supplementary Information

section of the proposed rule discussed

the four options the Commission

considered for structuring the regulatory

process related to public participation

in the NRC’s review and approval of a

licensee’s proposal for thermal

annealing of a reactor vessel. The

proposed rule, at the Commission’s

direction, requested comments on the

preferred regulatory process (including

opportunities for public participation).

The four options included:

(1) No hearings under the rule as

proposed;

(2) Discretionary opportunity for

hearing under rule as proposed in

which situation the Commission would

decide on a case-by-case basis to

determine whether a hearing should be

held;
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65463

(3) Required opportunity for hearing

under rule as proposed, but work could

commence if the NRC were to make a

‘‘no significant hazard determination’’

on the proposed thermal annealing; and

(4) Modify the proposed rule to

require suspension of license prior and

during the thermal annealing at which

time no hearing would be afforded and

the license would only be reinstated if

the licensee demonstrates that it has

addressed the reactor embrittlement

such that it is acceptable to operate the

plant.

Three comments were submitted on

the subject. OCRE and NEI addressed all

of the alternatives in detail and they, as

well as one utility, identified and

discussed individual preferred

alternatives.

NEI commented that each of the four

alternatives has a sufficiently serious

flaw to prevent adoption. With respect

to the no hearing alternative, NEI agrees

that annealing is presently subject to

approval by the Director of NRR in

accordance with Part 50 Appendix G

rather than being the subject of a license

amendment as an unreviewed safety

question under § 50.59. However, NEI

believes that annealing is an important

process from a regulatory standpoint

and that public participation, in the

form of informal hearings, is

appropriate. NEI objected to a

discretionary opportunity for a hearing

because it provides significant

uncertainty in the process for licensees

and members of the public. NEI’s

objection to requiring a hearing, as

discussed in staff Option 3, is that it

would allow those who object to the

resumption of operation, on other than

technical grounds, to use hearings to

delay restart. Option 4 is objectionable

to NEI because it does not provide the

licensee with any stability or

predictability since the licensee would

be required to demonstrate compliance

after the annealing was performed, and

does not provide the public with any

opportunity to express its views.

NEI further commented that a license

amendment is not necessary to approve

a thermal annealing plan because

annealing will not change the reactor

vessel or other components in a manner

inconsistent with the facility technical

specifications nor will it require

changes in the FSAR, and further, that

a licensee is not required to modify its

procedures to address or accommodate

the annealing process. NEI noted that,

while there is an incentive for the

licensee to obtain credit for its improved

P/T curves, and could seek a licensee

amendment to do so, the licensee’s

existing P/T curves could remain in

force.

Despite the conclusion that a license

amendment is not necessary for thermal

annealing, NEI recommended that a

hearing opportunity be provided, but

that it be a non-adjudicatory, Subpart L

type hearing on the record. NEI gave the

following advantages for this approach:

(1) The NRC would be provided with a

clear understanding of the licensee’s

annealing process, and the NRC’s

hearing process; (2) a Subpart L hearing

is held on the written record and

typically does not include the discovery

or live testimony associated with

adjudicatory hearings, but allows the

public to participate in a meaningful

way without consuming the vast NRC,

licensee, and public resources required

for an adjudicatory hearing; and (3) it

would provide predictability and

stability by ensuring that all issues

which could be subject to a hearing are

addressed prior to restart. Any

inspection or test performed in order to

restart would be for the purpose of

confirming compliance with the rule.

OCRE supported the proposed rule

provided that the public hearing rights

were preserved with regard to reactor

pressure vessel annealing. It is OCRE’s

position on the request for public

comment that, based on the Sholly

decision, the NRC must offer the

opportunity for a formal adjudicatory

hearing on the application for annealing

and on the licensee’s justification for

subsequent operation where the licensee

cannot certify that the thermal

annealing was performed in accordance

with the approved application. OCRE

commented that approval by the

Director of NRR of the application for

annealing and restart of the reactor, if

the licensee cannot certify that

annealing was performed in accordance

with the approved application, will give

the licensee the authority to operate in

ways in which they otherwise could

not, and is thus, a de facto license

amendment. OCRE fully supported

Option 3 which requires opportunity for

hearing under the rule as proposed.

OCRE suggested that the adequacy of

the thermal annealing plan, as well as

the vessel’s ability to perform its safety

function after annealing, could be raised

in the hearing on the thermal annealing

plan and that the licensee’s

implementation of the thermal

annealing plan could not commence

until any hearing is concluded or unless

the NRC makes a ‘‘no significant

hazards determination’’ with respect to

thermal annealing.

With respect to Option 1, OCRE

concluded that the informal hearings or

public meetings proposed by the

Commission for the initial thermal

annealing are not a substitute for

adjudicatory hearings required by the

Atomic Energy Act (AEA) and do not

give the interveners the same rights as

they would have in a Section 189a

hearing. OCRE found Option 2

preferable to having no hearing.

However, OCRE contended that this

option is flawed by the assumption that

‘‘Section 189a of the AEA does not

afford an interested member of the

public a right to request a hearing.’’

They contend that approval by the

Director, NRR to anneal the reactor

pressure vessel or to restart after

annealing does constitute a de facto

operating licensing amendment for

which the opportunity for a hearing is

required. OCRE found Options 1 and 4

unacceptable in that they do not provide

the opportunity for a formal

adjudicatory hearing.

The comment from the utility

suggested that Option 1 is the

appropriate approach as long as the

annealing process to be implemented is

approved in advance by the NRC staff

and the utility certifies that they have

complied with the approved annealing

process during the annealing operation,

as provided for in the proposed rule.

The utility further commented that if

Technical Specifications changes or

amendments to the operating license are

required in order to perform the

annealing then the opportunity for

hearings would be required due to the

normal license amendment process and

if the final safety analysis report (FSAR)

were required to be updated to reflect

the thermal annealing process, the

provisions of 10 CFR 50.59 would

apply. The utility suggested that if those

changes did not constitute an

‘‘unreviewed safety question,’’ no

amendment would be needed and the

license amendment process should not

be invoked and that if a member of the

public is concerned about a licensee’s

compliance with the NRC approved

thermal annealing plan, those concerns

could be addressed pursuant to the 10

CFR 2.206 petition process. The utility

commented that, under its proposal,

existing regulatory provisions for public

participation would apply as

appropriate and no new prescriptive

requirements would be necessary.

The Commission has considered the

public comments and has modified the

proposed rule as follows. A licensee that

seeks to utilize thermal annealing to

mitigate the effects of neutron

irradiation of the nuclear reactor vessel

must, at least three years prior to the

date at which the limiting fracture

toughness criteria in § 50.61 or

Appendix G to Part 50 would be

exceeded, submit a Thermal Annealing

Report to the NRC staff for review. The
65464 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

report shall contain four sections: (i)

Thermal Annealing Operating Plan, (ii)

Requalification Inspection and Test

Program, (iii) Program for determining

Fracture Toughness Recovery and

Reembrittlement Trend, and (iv) a

section identifying any changes to the

description of the facility as described

in the updated final safety analysis

report (FSAR) which constitute

unreviewed safety questions (USQs)

under § 50.59, and changes to the

facility’s technical specifications, which

are necessary either to perform the

thermal annealing, or to operate

following completion of the annealing.

Section 50.66(a) provides that the NRC

will, within three years of submission of

a licensee’s annealing report, document

its views on whether the plan for

conducting thermal annealing

constitutes an unreviewed safety

question or otherwise requires a change

to the plant’s technical specifications.

Such a determination is the threshold

determination for whether NRC

approval is required before undertaking

the activity. In the event the NRC were

to conclude, contrary to the licensee,

that an unreviewed safety question is

present or a change to the technical

specifications is necessary, the NRC

would, as a discretionary enforcement

matter, issue an appropriate order to the

licensee prohibiting annealing prior to

issuance of a license amendment. An

opportunity for formal adjudicatory

hearing would be provided in

connection with the license

amendment; however, if the NRC makes

a finding that the proposed change to

the FSAR description or technical

specification constitutes a ‘‘no

significant hazards consideration’’

pursuant to Section 189.(a)(2)(A), the

licensee may conduct the thermal

annealing prior to completion of any

hearing. In any event, at least 30 days

before the licensee starts to thermal

anneal and before the NRC completes its

review, the NRC will hold a public

meeting on the licensee’s proposed

Thermal Annealing Plan and

Requalification Inspection and Test

Program.

Following the completion of the

annealing operation, the licensee must

confirm in writing to the Director, Office

of Nuclear Reactor Regulation, that the

thermal annealing was performed in

accordance with the Thermal Annealing

Operating Plan and the Requalification

and Inspection Test Program. In support

of this confirmation, the licensee must

submit a report, within three months of

completion or termination of the anneal,

that presents the results of the annealing

operation. Within two weeks of the

licensee’s written confirmation that the

thermal annealing was completed in

accordance with the Thermal Annealing

Plan, and prior to restart, the NRC shall:

(1) Place in its public document room a

summary of the NRC staff’s inspection

of the licensee’s thermal annealing

process to confirm that the thermal

annealing was completed in accordance

with the Thermal Annealing Operating

Plan and the Requalification Inspection

and Test Program, and (2) hold a public

meeting with the licensee to permit the

licensee to explain the results of the

reactor vessel annealing to the NRC and

the public, for the NRC to discuss its

inspection of the reactor vessel

annealing process, and to provide an

opportunity for the public to comment

to the NRC on the annealing operation

and the results of the Staff’s inspection.

Within 45 days of the licensee’s

written confirmation that the thermal

annealing was completed, the NRC shall

complete full documentation of the

NRC’s inspection of the licensee’s

annealing process to confirm that the

annealing was completed in accordance

with the Thermal Annealing Operating

Plan and the Requalification Inspection

and Test Program.

The licensee may resume operation if:

(1) The licensee concludes that the

thermal annealing operation was

performed in compliance with the

Thermal Annealing Operating Plan, the

Requalification Inspection and Test

Program, and the provisions of Section

50.66(b), (2) a summary of the NRC’s

inspection of the thermal annealing is

placed in the NRC public document

room as required by Section 50.66(c) (2)

and (3) the NRC holds the public

meeting required by Section 50.66(f)(2),

unless the staff takes action against the

licensee. Since NRC approval to resume

operation is not necessary, an

opportunity for hearing would not be

provided in this situation. If, however,

the licensee cannot conclude that the

thermal annealing was performed in

compliance with the Thermal Annealing

Operating Plan or the Requalification

Inspection and Test Program, the

licensee must submit a justification for

continued operation to the Director. If

the noncompliance presents an

unreviewed safety question, as

determined by the licensee or directed

by the NRC following its review of the

report, then the plant may not restart

until the Director has approved restart.

Those failures to comply with the

Thermal Annealing Operating Plan and

the Requalification Inspection and Test

Program, which either (1) Are

considered to be ‘‘unreviewed safety

questions’’ or (2) require changes to the

technical specifications as a result of the

noncompliances, would also be subject

to an opportunity for a formal

adjudicatory hearing in accordance with

the Commission’s regulations governing

license amendments. However, the

licensee may restart prior to completion

of the hearing if the Director makes a

finding that such restart constitutes a

‘‘no significant hazards consideration,’’

as provided under Section 189.(a)(2)(A)

of the Atomic Energy Act of 1954, as

amended.

The regulatory process for thermal

annealing and the associated hearing

opportunities are consistent with longstanding NRC regulatory practices

defining those matters which present

sufficient potential effect on public

health and safety (e.g., are unreviewed

safety questions) to justify both prior

NRC review of the change, and an

opportunity for hearings (with the

associated time and resource impacts on

both the licensee and the NRC). With

respect to the thermal annealing review

process, the Commission reassessed the

regulatory requirements and processes

for assuring safety. The Commission

determined that the most important

safety matters are normally addressed in

license conditions, technical

specifications, and the FSAR. The

regulatory process for NRC

consideration of licensee-initiated

changes concerning these matters, and

the associated opportunities for hearings

is in 10 CFR 50.59. In view of this wellestablished regulatory process for

important safety information, the

Commission determined that a

regulatory process requiring NRC

approval of a thermal annealing plan is

not necessary, because the licensee is

already required to comply with its

license conditions, technical

specifications, and FSAR. Important

changes to license conditions, technical

specifications, and FSAR from a safety

standpoint are subject to both prior NRC

review and approval and an opportunity

for hearing. With respect to restart

following completion of the annealing,

the 15-day delay period should be

sufficient time for review of the

licensee’s input given the NRC staff’s

understanding of the annealing

operation plan prior to implementation,

ongoing resident inspections and

headquarters inspections of the

implementation of thermal annealing

operating plan. The Commission did not

adopt NEI’s suggestion for informal

hearings where the Director must

approve restart if the Thermal

Annealing Operating Plan and

Requalification Inspection and Test

Program were not complied with,

because the Commission does not see
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65465

any distinction (in terms of safety

implications) between the subject matter

of hearings under this rule, as compared

with other actions under Part 50 which

would require formal hearings.

As discussed earlier in the

supplementary information, previously

performed research analyses indicated

the potential for plastic deformation of

the main coolant piping for a typical

U.S. plant design and anticipated

annealing conditions. There are also

questions regarding how thermal growth

of the pressure vessel is treated, and the

adequacy of the thermal and stress

analyses used to predict response of the

overall system under thermal annealing

conditions. Additionally, there may be

questions in other areas such as

temperature limits for the concrete

structures, and potential radiological

hazards associated with removing and

storing the reactor internals during the

annealing process, and fire hazards

associated with heating the vessel.

Recognition of the numerous complex

technical questions related to 4 thermal

annealing and of the potential benefits

for operating nuclear power plants has

resulted in a cooperative effort, funded

by the U.S. Department of Energy and

the industry, to perform Annealing

Demonstration Projects. Projects are

planned to demonstrate two different

annealing processes, evaluating heater

designs and vessel designs. It is

anticipated that the annealing

demonstration projects will answer

many of the generic questions regarding

thermal annealing of U.S. pressure

vessel and piping designs.

The Thermal Annealing Report,

required by the thermal annealing rule,

is designed to facilitate a detailed

review by the licensee of plant-specific

questions and considerations in

performing a thermal annealing. The

proposed rule specifically discusses the

potential for unreviewed safety

questions and technical specification

changes that may result from or be

related to thermal annealing of the

reactor pressure vessel. With

completion of the demonstration

projects and as the staff and industry

gain experience with thermal annealing,

many of the issues related to annealing

will be better understood and related

questions will be answered. However,

until this experience is realized, the

staff will critically review licensee

determinations regarding unreviewed

safety questions and the need for

technical specification changes

associated with each proposed thermal

annealing. The level of staff effort is

expected to be significantly greater

during its review of the initial proposed

vessel annealings than that which will

be required after experience is gained.

The thermal annealing rule has been

structured to provide time for the staff

to thoroughly review the licensee’s

annealing plan and determination

regarding unreviewed safety questions

and the need for technical specification

changes. If the staff identifies an

unreviewed safety question or the need

for a technical specification change, the

licensee would be so notified and the

existing NRC regulatory practices would

be invoked to address the issues.

Backfitting Issues

Comments were received on

backfitting issues from the Nuclear

Utility Backfitting and Reform Group

(NUBARG). NUBARG commented that

they do not object to the new NRC

position in Appendix G to 10 CFR Part

50 which prohibits core criticality

before completion of hydrostatic

pressure and leak tests as a conservative

measure to enhance safety. However,

they are concerned that amending

Appendix G on the basis of a

compliance exception may set a bad

precedent for avoiding backfitting

analyses. NUBARG stated that ‘‘The

logic of the proposed rule would seem

to allow the NRC to avoid a backfitting

analysis by (1) invoking the intent of

one requirement to override the explicit

provisions of another, (2) using the

compliance exception when the practice

being eliminated seems specifically

contemplated by and specified in the

pertinent regulation, and (3) overlooking

the fact that the NRC has apparently

accepted this position in practice by

some licensees * * *’’ In NUBARG’s

view, this proposed amendment should

be supported by a backfit analysis. The

Commission has reviewed this comment

and has concluded that use of the

compliance exception under § 50.109

for the changes in Appendix G to 10

CFR Part 50 is appropriate. The Backfit

Analysis section contains further

discussion on this subject. The issue of

explicitly prohibiting core criticality

before completing pressure and leak

tests has been addressed previously

(letter from J. M. Taylor, EDO, to N. S.

Reynolds and D. F. Stenger, NUBARG,

dated February 2, 1990) and the

NUBARG comment did not provide new

information. The Commission has

concluded that any backfit requirements

in this amendment are necessary to

bring the facilities into compliance with

licenses, or the rules and orders of the

Commission, or into conformance with

written commitments by the licensees.

Therefore, a backfit analysis is not

required pursuant to 10 CFR

50.109(a)(4)(i).

NUBARG also commented on the

amendment to Appendix H to 10 CFR

Part 50 regarding surveillance that

would preclude reducing the amount of

testing if the initial test results agreed

with predicted results. Although

NUBARG recognizes the change would

be prospective, it believes that NRC

should provide flexibility to allow

continued relief for any licensee who

lacks such an authorization but has

relied on the provision. The

Commission believes that sufficient

flexibility already exists in that

licensees who do not have an

authorization may seek an exemption

under 10 CFR Part 50.12.

Another aspect of the backfitting

concern raised by NUBARG addresses

the proposed amendment to § 50.61

which, based on the adequate protection

exception, would impose a uniform

methodology for calculating the

reference temperature. NUBARG

contends that to rely on the adequate

protection exception is arguably

erroneous because the change in

methodology is not likely an adequate

protection issue (i.e., for most plants,

the screening criteria will not be

approached for many years). As

discussed further under Backfit

Analysis, the Commission believes that

a new backfit analysis is not required for

this conforming change, which corrects

an inadvertent omission from the

previous rulemaking. Therefore, the

Commission concludes that the

adequate protection basis for the backfit

continues to apply from the previous

rulemaking (56 FR 22300; May 15, 1991)

to § 50.61.

Criminal Penalties

For purposes of Section 223 of the

Atomic Energy Act (AEA), the

Commission is issuing the final rule

under one or more of Sections 161b,

161i or 161o of the AEA. Willful

violations of the rule will be subject to

criminal enforcement.

Finding of No Significant

Environmental Impact

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 is not

a major Federal action significantly

affecting the quality of human

environment and, therefore, an

environmental impact statement is not

required.

The individual actions covered in this

final rule would either serve to enhance

safety of the reactor pressure vessel,

thereby decreasing the environmental

impact of plant operation, or have no
65466 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

impact on the environment. Therefore,

in all cases these individual actions will

not have an adverse impact on the

environment.

PTS Rule (10 CFR 50.61)

The inclusion of thermal annealing as

an option for mitigating the effects of

neutron irradiation serves to decrease

the environmental impact of plant

operation by enhancing the safety of the

reactor pressure vessel.

The incorporation of the Regulatory

Guide 1.99, Revision 2, method for

determining RTNDT into the PTS rule

has no impact on the environment

because this change will result in values

of RTPTS which are consistent with

those currently used in plant operation.

The restructuring of the PTS rule is

the type of action described in

categorical exclusion 10 CFR

51.22(c)(2). Therefore, an environmental

assessment is not necessary for this

change.

Thermal Annealing Rule (10 CFR 50.66)

The thermal annealing rule (10 CFR

50.66) permits and provides

requirements for the thermal annealing

of a reactor vessel to restore fracture

properties of the reactor vessel material

which have been degraded by neutron

irradiation. This final rule only applies

when a licensee elects to use it. The

final rule provides an alternative for

assuring compliance with the

requirements in 10 CFR 50.61 and

Appendix G of 10 CFR Part 50.

The application of thermal annealing

to a reactor vessel improves the

condition of the reactor vessel material.

In addition, this rule establishes

requirements to avoid damaging the

reactor system and to protect against

accidents during the annealing

operation.

This rule is one of several regulatory

requirements that will function to

ensure reactor vessel integrity. In that

sense, this rule has a positive impact on

the environment by reducing the

potential for vessel failure. For these

reasons, the Commission has

determined that there is no significant

impact and, therefore, an environmental

statement is not required.

Appendix G to 10 CFR Part 50

The prohibition of core criticality

before completion of the required

pressure and leak tests will serve to

reduce the potential for vessel failure,

and thereby decrease the potential

environmental impact of plant

operation.

The restructuring of Sections IV and

V of Appendix G is clarifying or

corrective in nature, and is the type of

action described in categorical

exclusion 10 CFR 51.22(c)(2). Therefore,

an environmental assessment is not

necessary for this change.

The changing of the reference from

Appendix G of Section III of the ASME

Code to Appendix G of Section XI of the

ASME Code has no impact on the

environment because the requirements

in the Appendices are identical.

Therefore, there is no adverse impact on

the environment from this change.

The referencing of the thermal

annealing rule results in no adverse

impact on the environment because

Appendix G currently permits the use of

thermal annealing to reduce fracture

toughness loss of the RPV materials due

to irradiation embrittlement.

Appendix H to 10 CFR Part 50

Concerning the amendments to

Appendix H to 10 CFR Part 50 in the

final rule, the requirement that all

irradiation surveillance tests be made

(i.e., no reduction in testing is

permitted) will have a positive impact

on the environment in helping to assure

the integrity of the reactor pressure

vessel.

The restructuring of Section II.C is the

type of action described in categorical

exclusion 10 CFR 51.22(c)(2). Therefore,

an environmental assessment is not

necessary for this change.

The clarification of the applicable

version of ASTM Standard E 185 will

result in no adverse impact to the

environment since there will be no

change to current surveillance

programs. Changes to future

surveillance programs will make the

programs more effective in assessing

irradiation embrittlement effects to the

RPV materials, thereby helping to assure

the integrity of the reactor pressure

vessel

Paperwork Reduction Act Statement

This final rule amends information

collection requirements that are subject

to the Paperwork Reduction Act of 1980

(44 U.S.C. 3501 et seq.). These

requirements were approved by the

Office of Management and Budget,

approval number 3150–0011.

The public reporting burden for this

collection of information is estimated to

average 6,000 hours per response,

including the time for reviewing

instructions, searching existing data

sources, gathering and maintaining the

data needed, and completing and

reviewing the collection of information.

Send comments regarding the burden

estimate or any other aspect of this

collection of information, including

suggestions for reducing the burden, to

the Information and Records

Management Branch (T–6 F33), U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, and to the

Desk Officer, Office of Information and

Regulatory Affairs, NEOB–10202,

(3150–0011), Office of Management and

Budget, Washington, DC 20503.

Public Protection Notification

The NRC may not conduct or sponsor,

and a person is not required to respond

to, a collection of information unless it

displays a currently valid OMB control

number.

Regulatory Analysis

The NRC staff has prepared a

regulatory analysis for the amendments

to 10 CFR 50.61, Appendix G of 10 CFR

Part 50, and Appendix H of 10 CFR Part

50 that describes the factors and

alternatives considered by the

Commission in deciding to issue these

amendments. A copy of the regulatory

analysis is available for inspection and

copying for a fee at the NRC Public

Document Room, 2120 L Street NW.

(Lower Level), Washington, DC 20555–

0001. Single copies of the analysis may

be obtained from Alfred Taboada, Office

of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, telephone

(301) 415–6014.

Regulatory Flexibility Act Certification

As required by the Regulatory

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

Commission certifies that this final rule

will not have a significant economic

impact on a substantial number of small

entities. The rules which are affected by

the amendments will: (1) Preclude

brittle fracture of embrittled vessels

during PTS events, (2) provide the

general fracture toughness requirements

for RPVs, including ductile fracture

toughness requirements and pressuretemperature limits, (3) provide the

requirements for surveillance programs

to monitor irradiation embrittlement of

RPV beltline materials, and (4) provide

for a method for restoring the fracture

toughness of RPV beltline materials

used in nuclear facilities licensed under

the provision of 10 CFR 50.21(b) and 10

CFR 50.22. The companies that own

these facilities do not fall within the

scope of the definition of ‘‘small

entities’’ as set forth in the Regulatory

Flexibility Act, the Small Business Size

Standards in regulations issued by the

Small Business Administration at 13

CFR Part 121, or the size standards

established by the NRC at 10 CFR 2.810

(60 FR 18344; April 11, 1995).
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65467

1The Commission is aware that NUBARG has

presented an argument to the NRC that performance

of ASME Code hydrostatic tests are more effective

at the higher temperatures achieved when using

nuclear heat, as compared with the heat sources

normally employed by utilities in performing the

hydrostatic tests. However, for the reasons set forth

in the 1990 letter from James M. Taylor, EDO to N.

S. Reynolds and D.F. Stenger, NUBARG, the

Commission rejects this argument.

Backfit Analysis

PTS Rule (10 CFR 50.61)

The revision to § 50.61 requires

licensees to calculate RTPTS using the

same methodology specified in

Regulatory Guide 1.99, Revision 2, for

determining RTNDT. This change was

logically a requisite part of the previous

rulemaking (56 FR 22300; May 15, 1991)

to § 50.61 that set forth a unified method

for calculating radiation embrittlement

of the reactor beltline materials in Part

50. However, the Commission, at that

time, inadvertently failed to make the

conforming change to § 50.61. The

Commission believes that the backfit

statement for the previous amendment,

which determined that the backfit was

necessary to ensure that the facility

continues to provide adequate

protection to the public health and

safety, is applicable to this conforming

change to § 50.61.

The restructuring of the PTS rule does

not impose any backfits as defined in 10

CFR 50.109(a)(1) because there is no

change in requirements due to this

restructuring.

The inclusion of thermal annealing in

§ 50.61 does not constitute a backfit as

defined in 10 CFR 50.109(a)(1) because

the decision to perform annealing is

voluntary, no annealing has been

conducted in this country, and there are

no staff positions or Commission

requirements relied upon by licensees

that are being changed.

Thermal Annealing Rule (10 CFR 50.66)

The final thermal annealing rule

establishes requirements with respect to

applications for thermal annealing.

However, the Commission has

determined that the rule does not

impose a ‘‘backfit’’ as defined in 10 CFR

50.109(a)(1). The thermal annealing rule

does not require any licensee to perform

thermal annealing. Under existing

requirements, all licensees are required

to evaluate whether they exceed the PTS

screening limits in 10 CFR 50.61 and the

Charpy upper shelf screening limits in

Appendix G of CFR Part 50. However,

these rules provide an alternative means

for meeting these screening limits (e.g.,

performing thermal annealing). No

licensee currently has pending before

the NRC an application for thermal

annealing, nor has any current licensee

been granted permission to conduct

thermal annealing. The rule does not

reflect any new or different NRC staff

position which conflicts with a prior

NRC staff position or Commission rule.

Thus, the final rule will have a purely

prospective effect on future applications

for thermal annealing. The Commission

has stated in other rulemakings

establishing prospective requirements

(10 CFR Part 52 and the License

Renewal Rule, 10 CFR Part 54) that the

Backfit Rule was not intended to protect

the future applicant from current

changes in Commission requirements.

Accordingly, the Commission concludes

that the rule does not impose backfits

and a backfit analysis need not be

prepared for the final thermal annealing

rule.

Appendix G to 10 CFR Part 50

The restructuring of Sections IV and

V of this appendix, referencing of the

thermal annealing rule, changing the

reference from Appendix G of Section

III of the ASME Code to Appendix G of

Section XI of the ASME Code, and

deleting the ‘‘design to permit

annealing’’ requirement do not impose

any backfits as defined in 10 CFR

50.109(a)(1), because they are either

prospective in nature or are of a

clarifying nature.

10 CFR Part 50, Appendix G,

Paragraph IV.2.d. of the final rule

explicitly prohibits core criticality

before completion of ASME Code

hydrostatic pressure and leak tests. This

is intended to make clear that licensees

may not use nuclear heat in order to

perform ASME Code hydrostatic tests.

This amendment can be construed as a

backfit, inasmuch as the prior version of

10 CFR Part 50, Appendix G, Paragraph

IV.A.5 could be read to permit core

criticality during ASME hydrostatic

tests and Section XI of the ASME Code

does not explicitly prohibit core

criticality prior to completion of these

tests. However, the Commission never

intended the disputed language in

Paragraph IV.A.5 of Appendix G to

permit core criticality before successful

completion of the required ASME

hydrostatic tests. The scope of

Appendix G is ‘‘fracture toughness

requirements’’ only; that scope is stated

clearly in the title of Appendix G, and

Appendix G was not intended to specify

system operational requirements. It is

not correct, therefore, to interpret

paragraph IV.A.5. as permitting nuclear

hydrotesting. The final phrase in IV.A.5,

‘‘depending on whether the core is

critical during the test,’’ was included in

the rule for the sake of completeness, to

specify appropriate fracture toughness

requirements in the event that a licensee

for some reason wanted to have the core

critical during hydrotest, and was given

approval to do so (e.g., as in the case of

the Hatch units, where nuclear

hydrotesting was allowed one last time

as an approved exception.) The ASME

Code’s hydrostatic testing provisions for

the reactor coolant pressure boundary

(RCPB) provides the necessary

assurance that GDC–14 is met. GDC–14

inter alia requires RCPB testing in order

to provide an extremely low probability

of RCPB failure, in terms of abnormal

leakage, rapidly propagating failure, and

gross rupture. Using heat produced by a

critical reactor core to perform such

testing essentially undercuts the basic

safety principle embodied in GDC–14

that testing should be completed prior

to nuclear reactor operation. It makes

little sense to allow core criticality—

thereby allowing the reactor to be in an

operational condition where a loss of

coolant could have significant

consequences—prior to successful

completion of tests that are intended to

ensure that the probability of such

coolant losses during such an

operational condition are extremely

low.1 The ASME Code, Section XI,

requires that the System Leakage Test be

performed prior to plant startup

following each refueling outage (Table2500–1, Examination Category B–P,

Note 2). The only way to interpret the

ASME Code as permitting core

criticality prior to completion of the

hydrostatic tests is to read the term,

‘‘plant startup’’ as referring to something

other than reactor criticality. This is

neither the normal industry practice,

nor has it been the NRC staff’s

longstanding interpretation of this

provision of the ASME code. Indeed, it

does not appear that the NRC staff has

construed either Appendix G, Paragraph

IV.A.5 nor Section XI of the ASME Code

as permitting core criticality prior to

successful completion of ASME Code

hydrostatic tests. Moreover, the vast

majority of nuclear utility licensees do

not use nuclear heat to perform ASME

code hydrostatic tests. This suggests that

most licensees hold the same

interpretation of Appendix G and

Section XI of the ASME Code as the

Commission. In sum, the Commission

believes Section XI of the ASME Code,

which is endorsed by 10 CFR 50.55a,

implicitly prohibits core criticality prior

to successful completion of hydrostatic

testing. Therefore, the Commission

concludes that the change in the

language of Appendix G, Paragraph

IV.2.d. is necessary to assure

compliance with 10 CFR 50.55a and the

ASME Code.
65468 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

2Changes to RTPTS values are considered

significant if either the previous value or the

current value, or both values, exceed the screening

criterion prior to the expiration of the operating

license, including any renewed term, if applicable,

for the plant.

The Commission has concluded that

any backfit requirements in this

amendment are necessary to bring the

facilities into compliance with licenses,

or the rules and orders of the

Commission, or into conformance with

written commitments by the licensees.

Therefore, a backfit analysis is not

required pursuant to 10 CFR

50.109(a)(4)(i).

Appendix H to 10 CFR Part 50

The amendments to Appendix H to 10

CFR Part 50 are either prospective in

nature or of a clarifying nature, and

hence do not involve any provisions

which would impose backfits as defined

in 10 CFR 50.109(a)(1).

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 record keeping

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. 552 and 553;

the NRC is adopting the following

amendments to 10 CFR Part 50.

PART 50—DOMESTIC LICENSING OF

PRODUCTION AND UTILIZATION

FACILITIES

1. The general authority citation for

Part 50 is corrected to read as set forth

below, and the section-specific

authority citations continue 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. 1444, 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).

Section 50.7 also issued under Pub. L. 95–

601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851).

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, and

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 § 50.8, paragraph (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,

50.33a, 50.34, 50.34a, 50.35, 50.36,

50.36a, 50.48, 50.49, 50.54, 50.55,

50.55a, 50.59, 50.60, 50.61, 50.63, 50.64,

50.65, 50.66, 70.71, 50.72, 50.73, 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, and R, to this part.

* * * * *

3. Section 50.61 is revised to read as

follows:

§ 50.61 Fracture toughness requirements

for protection against pressurized thermal

shock events.

(a) Definitions. For the purposes of

this section:

(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,’’

edition and addenda and any

limitations and modifications thereof as

specified in § 50.55a.

(2) Pressurized Thermal Shock Event

means an event or transient in

pressurized water reactors (PWRs)

causing severe overcooling (thermal

shock) concurrent with or followed by

significant pressure in the reactor

vessel.

(3) Reactor Vessel Beltline means the

region of the reactor vessel (shell

material including welds, heat affected

zones and plates or forgings) that

directly surrounds the effective height

of the active core and adjacent regions

of the reactor vessel that are predicted

to experience sufficient neutron

radiation damage to be considered in

the selection of the most limiting

material with regard to radiation

damage.

(4) RTNDT means the reference

temperature for a reactor vessel

material, under any conditions. For the

reactor vessel beltline materials, RTNDT

must account for the effects of neutron

radiation.

(5) RTNDT(U) means the reference

temperature for a reactor vessel material

in the pre-service or unirradiated

condition, evaluated according to the

procedures in the ASME Code,

Paragraph NB–2331 or other methods

approved by the Director, Office of

Nuclear Reactor Regulation.

(6) EOL Fluence means the bestestimate neutron fluence projected for a

specific vessel beltline material at the

clad-base-metal interface on the inside

surface of the vessel at the location

where the material receives the highest

fluence on the expiration date of the

operating license.

(7) RTPTS means the reference

temperature, RTNDT, evaluated for the

EOL Fluence for each of the vessel

beltline materials, using the procedures

of paragraph (c) of this section.

(8) PTS Screening Criterion means the

value of RTPTS for the vessel beltline

material above which the plant cannot

continue to operate without

justification.

(b) Requirements.

(1) For each pressurized water nuclear

power reactor for which an operating

license has been issued, the licensee

shall have projected values of RTPTS,

accepted by the NRC, for each reactor

vessel beltline material for the EOL

fluence of the material. 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 a request for a change

in the expiration date for operation of

the facility.

(2) The pressurized thermal shock

(PTS) screening criterion is 270 °F for

plates, forgings, and axial weld

materials, and 300 °F for circumferential

weld materials. For the purpose of

comparison with this criterion, the

value of RTPTS for the reactor vessel

must be evaluated according to the

procedures of paragraph (c) of this

section, for each weld and plate, or

forging, in the reactor vessel beltline.

RTPTS must be determined for each

vessel beltline material using the EOL

fluence for that material.

(3) For each pressurized water nuclear

power reactor for which the value of

RTPTS for any material in the beltline is

projected to exceed the PTS screening

criterion using the EOL fluence, the

licensee shall implement those flux
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65469

3The class of material for estimating RTNDT(U) is

generally determined for welds by the type of

welding flux (Linde 80, or other), and for base metal

by the material specification.

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.’’

reduction programs that are reasonably

practicable to avoid exceeding the PTS

screening criterion set forth in

paragraph (b)(2) of this section. The

schedule for implementation of flux

reduction measures may take into

account the schedule for submittal and

anticipated approval by the Director,

Office of Nuclear Reactor Regulation, of

detailed plant-specific analyses,

submitted to demonstrate acceptable

risk with RTPTS above the screening

limit due to plant modifications, new

information or new analysis techniques.

(4) For each pressurized water nuclear

power reactor for which the analysis

required by paragraph (b)(3) of this

section indicates that no reasonably

practicable flux reduction program will

prevent RTPTS from exceeding the PTS

screening criterion using the EOL

fluence, the licensee shall submit a

safety analysis to determine what, if

any, modifications to equipment,

systems, and operation are necessary to

prevent potential failure of the reactor

vessel as a result of postulated PTS

events if continued operation beyond

the screening criterion is 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 at least three years

before RTPTS is projected to exceed the

PTS screening criterion.

(5) After consideration of the

licensee’s analyses, including effects of

proposed corrective actions, if any,

submitted in accordance with

paragraphs (b)(3) and (b)(4) of this

section, the Director, Office of Nuclear

Reactor Regulation, may, on a case-bycase basis, approve operation of the

facility with RTPTS in excess of the PTS

screening criterion. The Director, Office

of Nuclear Reactor Regulation, will

consider factors significantly affecting

the potential for failure of the reactor

vessel in reaching a decision.

(6) If the Director, Office of Nuclear

Reactor Regulation, concludes, pursuant

to paragraph (b)(5) of this section, that

operation of the facility with RTPTS in

excess of the PTS screening criterion

cannot be approved on the basis of the

licensee’s analyses submitted in

accordance with paragraphs (b)(3) and

(b)(4) of this section, the licensee shall

request and receive approval by the

Director, Office of Nuclear Reactor

Regulation, prior to any operation

beyond the criterion. The request must

be based upon 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 upon further analyses

based upon new information or

improved methodology.

(7) If the limiting RTPTS value of the

plant is projected to exceed the

screening criteria in paragraph (b)(2), or

the criteria in paragraphs (b)(3) through

(b)(6) of this section cannot be satisfied,

the reactor vessel beltline may be given

a thermal annealing treatment to recover

the fracture toughness of the material,

subject to the requirements of § 50.66.

The reactor vessel may continue to be

operated only for that service period

within which the predicted fracture

toughness of the vessel beltline

materials satisfy the requirements of

paragraphs (b)(2) through (b)(6) of this

section, with RTPTS accounting for the

effects of annealing and subsequent

irradiation.

(c) Calculation of RTPTS. RTPTS must

be calculated for each vessel beltline

material using a fluence value, f, which

is the EOL fluence for the material.

RTPTS must be evaluated using the same

procedures used to calculate RTNDT, as

indicated in paragraph (c)(1) of this

section, and as provided in paragraphs

(c)(2) and (c)(3) of this section.

(1) Equation 1 must be used to

calculate values of RTNDT for each weld

and plate, or forging, in the reactor

vessel beltline.

Equation 1:

RTNDT=RTNDT(U)+M+∆RTNDT

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

not available, a generic mean value for

the class 3 of material may be used if

there are sufficient test results to

establish a mean and a standard

deviation for the class.

(ii) For generic values of weld metal,

the following generic mean values must

be used unless justification for different

values is provided: 0°F for welds made

with Linde 80 flux, and ¥56°F for

welds made with Linde 0091, 1092 and

124 and ARCOS B–5 weld fluxes.

(iii) M means the margin to be added

to account for uncertainties in the

values of RTNDT(U), copper and nickel

contents, fluence and the calculational

procedures. M is evaluated from

Equation 2.

Equation M U 2 2 2 2 : = + σ σ ∆

(A) In Equation 2, σU is the standard

deviation for RTNDT(U). If a measured

value of RTNDT(U) is used, then σU is

determined from the precision of the

test method. If a measured value of

RTNDT(U) is not available and a generic

mean value for that class of materials is

used, then σU is the standard deviation

obtained from the set of data used to

establish the mean. If a generic mean

value given in paragraph (c)(1)(i)(B) of

this section for welds is used, then σU

is 17°F.

(B) In Equation 2, σ∆ is the standard

deviation for ∆RTNDT. The value of σ∆ to

be used is 28°F for welds and 17°F for

base metal; the value of σ∆ need not

exceed one-half of ∆RTNDT.

(iv) ∆RTNDT is the mean value of the

transition temperature shift, or change

in RTNDT, due to irradiation, and must

be calculated using Equation 3.

Equation 3: ∆RTNDT=(CF)f(0.28–0.10 log f)

(A) CF (°F) is the chemistry factor,

which is a function of copper and nickel

content. CF is given in Table 1 for welds

and in Table 2 for base metal (plates and

forgings). Linear interpolation is

permitted. In Tables 1 and 2, ‘‘Wt-%

copper’’ and ‘‘Wt-% nickel’’ are the

best-estimate values for the material,

which will normally be the mean of the

measured values for a plate or forging.

For a weld, the best estimate values will

normally be 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, the upper limiting values

given in the material specifications to

which the vessel material was fabricated

may be used. If not available,

conservative estimates (mean plus one

standard deviation) based on generic

data 4 may be used if justification is

provided. If none of these alternatives

are available, 0.35% copper and 1.0%

nickel must be assumed.

(B) f is the best estimate neutron

fluence, in units of 1019 n/cm2 (E greater

than 1 MeV), at the clad-base-metal

interface on the inside surface of the

vessel at the location where the material

in question receives the highest fluence

for the period of service in question. As

specified in this paragraph, the EOL

fluence for the vessel beltline material is

used in calculating KRTPTS.

(v) Equation 4 must be used for

determining RTPTS using equation 3

with EOL fluence values for

determining ∆RTPTS.

Equation 4: RTPTS=RTNDT(U)+M+∆RTPTS

(2) To verify that RTNDT for each

vessel beltline material is a bounding

value for the specific reactor vessel,

licensees shall consider plant-specific

information that could affect the level of
65470 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

5Surveillance program results means any data

that demonstrates the embrittlement trends for the

limiting beltline material, including but not limited

to data from test reactors or from surveillance

programs at other plants with or without

surveillance program integrated per 10 CFR Part 50,

Appendix H.

embrittlement. This information

includes but is not limited to the reactor

vessel operating temperature and any

related surveillance program 5 results.

(i) Results from the plant-specific

surveillance program must be integrated

into the RTNDT estimate if the plantspecific surveillance data has been

deemed credible as judged by the

following criteria:

(A) The materials in the surveillance

capsules must be those which are the

controlling materials with regard to

radiation embrittlement.

(B) Scatter in the plots of Charpy

energy versus temperature for the

irradiated and unirradiated conditions

must be small enough to permit the

determination of the 30-foot-pound

temperature unambiguously.

(C) Where there are two or more sets

of surveillance data from one reactor,

the scatter of ∆RTNDT values must be

less than 28°F for welds and 17°F for

base metal. Even if the range in the

capsule fluences is large (two or more

orders of magnitude), the scatter may

not exceed twice those values.

(D) The irradiation temperature of the

Charpy specimens in the capsule must

equal the vessel wall temperature at the

cladding/base metal interface within

±25°F.

(E) The surveillance data for the

correlation monitor material in the

capsule, if present, must fall within the

scatter band of the data base for the

material.

(ii)(A) Surveillance data deemed

credible according to the criteria of

paragraph (c)(2)(i) of this section must

be used to determine a material-specific

value of CF for use in Equation 3. A

material-specific value of CF is

determined from Equation 5.

Equation CF

A f

f

i i

f

i

n

i

f

i

n

i

i

5

0 28 0 10

1

0 56 0 20

1

:

. . log

. . log

=

× [ ]

[ ]

( ) −

=

( ) −

=

∑

∑

(B) In Equation 5, ‘‘n’’ is the number

of surveillance data points, ‘‘Ai’’ is the

measured value of ∆RTNDT and ‘‘fi’’ is

the fluence for each surveillance data

point. If there is clear evidence that the

copper and nickel content of the

surveillance weld differs from the vessel

weld, i.e. differs from the average for the

weld wire heat number associated with

the vessel weld and the surveillance

weld, the measured values of ∆RTNDT

must be adjusted for differences in

copper and nickel content by

multiplying them by the ratio of the

chemistry factor for the vessel material

to that for the surveillance weld.

(iii) For cases in which the results

from a credible plant-specific

surveillance program are used, the value

of σ∆ to be used is 14°F for welds and

8.5°F for base metal; the value of σ∆

need not exceed one-half of DRTNDT.

(iv) The use of results from the plantspecific surveillance program may result

in an RTNDT that is higher or lower than

those determined in paragraph (c)(1).

(3) Any information that is believed to

improve the accuracy of the RTPTS value

significantly must be reported to the

Director, Office of Nuclear Reactor

Regulation. Any value of RTPTS that has

been modified using the procedures of

paragraph (c)(2) of this section is subject

to the approval of the Director, Office of

Nuclear Reactor Regulation, when used

as provided in this section.

TABLE 1.—CHEMISTRY FACTOR FOR WELD METALS, °F

Copper, wt-%

Nickel, wt-%

0 0.20 0.40 0.60 0.80 1.00 1.20

0 .......................................................................................................... 20 20 20 20 20 20 20

0.01 ..................................................................................................... 20 20 20 20 20 20 20

0.02 ..................................................................................................... 21 26 27 27 27 27 27

0.03 ..................................................................................................... 22 35 41 41 41 41 41

0.04 ..................................................................................................... 24 43 54 54 54 54 54

0.05 ..................................................................................................... 26 49 67 68 68 68 68

0.06 ..................................................................................................... 29 52 77 82 82 82 82

0.07 ..................................................................................................... 32 55 85 95 95 95 95

0.08 ..................................................................................................... 36 58 90 106 108 108 108

0.09 ..................................................................................................... 40 61 94 115 122 122 122

0.10 ..................................................................................................... 44 65 97 122 133 135 135

0.11 ..................................................................................................... 49 68 101 130 144 148 148

0.12 ..................................................................................................... 52 72 103 135 153 161 161

0.13 ..................................................................................................... 58 76 106 139 162 172 176

0.14 ..................................................................................................... 61 79 109 142 168 182 188

0.15 ..................................................................................................... 66 84 112 146 175 191 200

0.16 ..................................................................................................... 70 88 115 149 178 199 211

0.17 ..................................................................................................... 75 92 119 151 184 207 221

0.18 ..................................................................................................... 79 95 122 154 187 214 230

0.19 ..................................................................................................... 83 100 126 157 191 220 238

0.20 ..................................................................................................... 88 104 129 160 194 223 245

0.21 ..................................................................................................... 92 108 133 164 197 229 252
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65471

TABLE 1.—CHEMISTRY FACTOR FOR WELD METALS, °F—Continued

Copper, wt-%

Nickel, wt-%

0 0.20 0.40 0.60 0.80 1.00 1.20

0.22 ..................................................................................................... 97 112 137 167 200 232 257

0.23 ..................................................................................................... 101 117 140 169 203 236 263

0.24 ..................................................................................................... 105 121 144 173 206 239 268

0.25 ..................................................................................................... 110 126 148 176 209 243 272

0.26 ..................................................................................................... 113 130 151 180 212 246 276

0.27 ..................................................................................................... 119 134 155 184 216 249 280

0.28 ..................................................................................................... 122 138 160 187 218 251 284

0.29 ..................................................................................................... 128 142 164 191 222 254 287

0.30 ..................................................................................................... 131 146 167 194 225 257 290

0.31 ..................................................................................................... 136 151 172 198 228 260 293

0.32 ..................................................................................................... 140 155 175 202 231 263 296

0.33 ..................................................................................................... 144 160 180 205 234 266 299

0.34 ..................................................................................................... 149 164 184 209 238 269 302

0.35 ..................................................................................................... 153 168 187 212 241 272 305

0.36 ..................................................................................................... 158 172 191 216 245 275 308

0.37 ..................................................................................................... 162 177 196 220 248 278 311

0.38 ..................................................................................................... 166 182 200 223 250 281 314

0.39 ..................................................................................................... 171 185 203 227 254 285 317

0.40 ..................................................................................................... 175 189 207 231 257 288 320

TABLE 2.—CHEMISTRY FACTOR FOR BASE METALS, °F

Copper, wt-%

Nickel, wt-%

0 0.20 0.40 0.60 0.80 1.00 1.20

0 .......................................................................................................... 20 20 20 20 20 20 20

0.01 ..................................................................................................... 20 20 20 20 20 20 20

0.02 ..................................................................................................... 20 20 20 20 20 20 20

0.03 ..................................................................................................... 20 20 20 20 20 20 20

0.04 ..................................................................................................... 22 26 26 26 26 26 26

0.05 ..................................................................................................... 25 31 31 31 31 31 31

0.06 ..................................................................................................... 28 37 37 37 37 37 37

0.07 ..................................................................................................... 31 43 44 44 44 44 44

0.08 ..................................................................................................... 34 48 51 51 51 51 51

0.09 ..................................................................................................... 37 53 58 58 58 58 58

0.10 ..................................................................................................... 41 58 65 65 67 67 67

0.11 ..................................................................................................... 45 62 72 74 77 77 77

0.12 ..................................................................................................... 49 67 79 83 86 86 86

0.13 ..................................................................................................... 53 71 85 91 96 96 96

0.14 ..................................................................................................... 57 75 91 100 105 106 106

0.15 ..................................................................................................... 61 80 99 110 115 117 117

0.16 ..................................................................................................... 65 84 104 118 123 125 125

0.17 ..................................................................................................... 69 88 110 127 132 135 135

0.18 ..................................................................................................... 73 92 115 134 141 144 144

0.19 ..................................................................................................... 78 97 120 142 150 154 154

0.20 ..................................................................................................... 82 102 125 149 159 164 165

0.21 ..................................................................................................... 86 107 129 155 167 172 174

0.22 ..................................................................................................... 91 112 134 161 176 181 184

0.23 ..................................................................................................... 95 117 138 167 184 190 194

0.24 ..................................................................................................... 100 121 143 172 191 199 204

0.25 ..................................................................................................... 104 126 148 176 199 208 214

0.26 ..................................................................................................... 109 130 151 180 205 216 221

0.27 ..................................................................................................... 114 134 155 184 211 225 230

0.28 ..................................................................................................... 119 138 160 187 216 233 239

0.29 ..................................................................................................... 124 142 164 191 221 241 248

0.30 ..................................................................................................... 129 146 167 194 225 249 257

0.31 ..................................................................................................... 134 151 172 198 228 255 266

0.32 ..................................................................................................... 139 155 175 202 231 260 274

0.33 ..................................................................................................... 144 160 180 205 234 264 282

0.34 ..................................................................................................... 149 164 184 209 238 268 290

0.35 ..................................................................................................... 153 168 187 212 241 272 298

0.36 ..................................................................................................... 158 173 191 216 245 275 303

0.37 ..................................................................................................... 162 177 196 220 248 278 308

0.38 ..................................................................................................... 166 182 200 223 250 281 313

0.39 ..................................................................................................... 171 185 203 227 254 285 317

0.40 ..................................................................................................... 175 189 207 231 257 288 320
65472 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

4. A new § 50.66 is added under the

center heading ‘‘Issuance, Limitations,

and Conditions of Licenses and

Construction Permits’’ to read as

follows:

§ 50.66 Requirements for thermal

annealing of the reactor pressure vessel.

(a) For those light water nuclear

power reactors where neutron radiation

has reduced the fracture toughness of

the reactor vessel materials, a thermal

annealing may be applied to the reactor

vessel to recover the fracture toughness

of the material. The use of a thermal

annealing treatment is subject to the

requirements in this section. A report

describing the licensee’s plan for

conducting the thermal annealing must

be submitted in accordance with § 50.4

at least three years prior to the date at

which the limiting fracture toughness

criteria in § 50.61 or Appendix G to Part

50 would be exceeded. Within three

years of the submittal of the Thermal

Annealing Report and at least thirty

days prior to the start of the thermal

annealing, the NRC will review the

Thermal Annealing Report and place

the results of its evaluation in its Public

Document Room. The licensee may

begin the thermal anneal after:

(1) Submitting the Thermal Annealing

Report required by paragraph (b) of this

section;

(2) the NRC places the results of its

evaluation of the Thermal Annealing

Report in the Public Document Room;

and

(3) the requirements of paragraph

(f)(1) of this section have been satisfied.

(b) Thermal Annealing Report. The

Thermal Annealing Report must

include: a Thermal Annealing Operating

Plan; a Requalification Inspection and

Test Program; a Fracture Toughness

Recovery and Reembrittlement Trend

Assurance Program; and Identification

of Unreviewed Safety Questions and

Technical Specification Changes.

(1) Thermal Annealing Operating

Plan.

The thermal annealing operating plan

must include:

(i) A detailed description of the

pressure vessel and all structures and

components that are expected to

experience significant thermal or stress

effects during the thermal annealing

operation;

(ii) An evaluation of the effects of

mechanical and thermal stresses and

temperatures on the vessel,

containment, biological shield, attached

piping and appurtenances, and adjacent

equipment and components to

demonstrate that operability of the

reactor will not be detrimentally

affected. This evaluation must include:

(A) Detailed thermal and structural

analyses to establish the time and

temperature profile of the annealing

operation. These analyses must include

heatup and cooldown rates, and must

demonstrate that localized

temperatures, thermal stress gradients,

and subsequent residual stresses will

not result in unacceptable dimensional

changes or distortions in the vessel,

attached piping and appurtenances, and

that the thermal annealing cycle will not

result in unacceptable degradation of

the fatigue life of these components.

(B) The effects of localized high

temperatures on degradation of the

concrete adjacent to the vessel and

changes in thermal and mechanical

properties, if any, of the reactor vessel

insulation, and on detrimental effects, if

any, on containment and the biological

shield. If the design temperature

limitations for the adjacent concrete

structure are to be exceeded during the

thermal annealing operation, an

acceptable maximum temperature for

the concrete must be established for the

annealing operation using appropriate

test data.

(iii) The methods, including heat

source, instrumentation and procedures

proposed for performing the thermal

annealing. This shall include any

special precautions necessary to

minimize occupational exposure, in

accordance with the As Low As

Reasonably Achievable (ALARA)

principle and the provisions of

§ 20.1206.

(iv) The proposed thermal annealing

operating parameters, including

bounding conditions for temperatures

and times, and heatup and cooldown

schedules.

(A) The thermal annealing time and

temperature parameters selected must

be based on projecting sufficient

recovery of fracture toughness, using the

procedures of paragraph (e) of this

section, to satisfy the requirements of

§ 50.60 and § 50.61 for the proposed

period of operation addressed in the

application.

(B) The time and temperature

parameters evaluated as part of the

thermal annealing operating plan, and

supported by the evaluation results of

paragraph (b)(1)(ii) of this section,

represent the bounding times and

temperatures for the thermal annealing

operation. If these bounding conditions

for times and temperatures are violated

during the thermal annealing operation,

then the annealing operation is

considered not in accordance with the

Thermal Annealing Operating Plan, as

required by paragraph (c)(1) of this

section, and the licensee must comply

with paragraph (c)(2) of this section.

(2) Requalification Inspection and

Test Program. The inspection and test

program to requalify the annealed

reactor vessel must include the detailed

monitoring, inspections, and tests

proposed to demonstrate that the

limitations on temperatures, times and

temperature profiles, and stresses

evaluated for the proposed thermal

annealing conditions of paragraph

(b)(1)(iv) of this section have not been

exceeded, and to determine the thermal

annealing time and temperature to be

used in quantifying the fracture

toughness recovery. The requalification

inspection and test program must

demonstrate that the thermal annealing

operation has not degraded the reactor

vessel, attached piping or

appurtenances, or the adjacent concrete

structures to a degree that could affect

the safe operation of the reactor.

(3) Fracture Toughness Recovery and

Reembrittlement Trend Assurance

Program. The percent recovery of RTNDT

and Charpy upper-shelf energy due to

the thermal annealing treatment must be

determined based on the time and

temperature of the actual vessel thermal

anneal. The recovery of RTNDT and

Charpy upper-shelf energy provide the

basis for establishing the post-anneal

RTNDT and Charpy upper-shelf energy

for each vessel material. Changes in the

RTNDT and Charpy upper-shelf energy

with subsequent plant operation must

be determined using the post-anneal

values of these parameters in

conjunction with the projected

reembrittlement trend determined in

accordance with paragraph (b)(3)(ii) of

this section. Recovery and

reembrittlement evaluations shall

include:

(i) Recovery Evaluations.

(A) The percent recovery of both

RTNDT and Charpy upper-shelf energy

must be determined by one of the

procedures described in paragraph (e) of

this section, using the proposed lower

bound thermal annealing time and

temperature conditions described in the

operating plan.

(B) If the percent recovery is

determined from testing surveillance

specimens or from testing materials

removed from the reactor vessel, then it

shall be demonstrated that the proposed

thermal annealing parameters used in

the test program are equal to or bounded

by those used in the vessel annealing

operation.

(C) If generic computational methods

are used, appropriate justification must

be submitted as a part of the

application.

(ii) Reembrittlement Evaluations.

(A) The projected post-anneal

reembrittlement of RTNDT must be
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65473

calculated using the procedures in

§ 50.61(c), or must be determined using

the same basis as that used for the preanneal operating period. The projected

change due to post-anneal

reembrittlement for Charpy upper-shelf

energy must be determined using the

same basis as that used for the preanneal operating period.

(B) The post-anneal reembrittlement

trend of both RTNDT and Charpy uppershelf energy must be estimated, and

must be monitored using a surveillance

program defined in the Thermal

Annealing Report and which conforms

to the intent of Appendix H of this part,

‘‘Reactor Vessel Material Surveillance

Program Requirements.’’

(4) Identification of Unreviewed

Safety Questions and Technical

Specification Changes. Any changes to

the facility as described in the updated

final safety analysis report constituting

unreviewed safety questions, and any

changes to the technical specifications,

which are necessary to either conduct

the thermal annealing or operate the

nuclear power reactor following the

annealing, must be identified. The

section shall demonstrate that the

Commission’s requirements continue to

be complied with, and that there is

reasonable assurance of adequate

protection to the public health and

safety following the changes.

(c) Completion or Termination of

Thermal Annealing.

(1) If the thermal annealing was

completed in accordance with the

Thermal Annealing Operating Plan and

the Requalification Inspection and Test

Program, the licensee shall so confirm

in writing to the Director, Office of

Nuclear Reactor Regulation. The

licensee may restart its reactor after the

requirements of paragraph (f)(2) of this

section have been met.

(2) If the thermal annealing was

completed but the annealing was not

performed in accordance with the

Thermal Annealing Operating Plan and

the Requalification Inspection and Test

Program, the licensee shall submit a

summary of lack of compliance with the

Thermal Annealing Operating Plan and

the Requalification Inspection and Test

Program and a justification for

subsequent operation to the Director,

Office of Nuclear Reactor Regulation.

Any changes to the facility as described

in the updated final safety analysis

report which are attributable to the

noncompliances and constitute

unreviewed safety questions, and any

changes to the technical specifications

which are required as a result of the

noncompliances, shall also be

identified.

(i) If no unreviewed safety questions

or changes to technical specifications

are identified, the licensee may restart

its reactor after the requirements of

paragraph (f)(2) of this section have

been met.

(ii) If any unreviewed safety questions

or changes to technical specifications

are identified, the licensee may not

restart its reactor until approval is

obtained from the Director, Office of

Nuclear Reactor Regulation and the

requirements of paragraph (f)(2) of this

section have been met.

(3) If the thermal annealing was

terminated prior to completion, the

licensee shall immediately notify the

NRC of the premature termination of the

thermal anneal.

(i) If the partial annealing was

otherwise performed in accordance with

the Thermal Annealing Operating Plan

and relevant portions of the

Requalification Inspection and Test

Program, and the licensee does not elect

to take credit for any recovery, the

licensee need not submit the Thermal

Annealing Results Report required by

paragraph (d) of this section but instead

shall confirm in writing to the Director,

Office of Nuclear Reactor Regulation

that the partial annealing was otherwise

performed in accordance with the

Thermal Annealing Operating Plan and

relevant portions of the Requalification

Inspection and Test Program. The

licensee may restart its reactor after the

requirements of paragraph (f)(2) of this

section have been met.

(ii) If the partial annealing was

otherwise performed in accordance with

the Thermal Annealing Operating Plan

and relevant portions of the

Requalification Inspection and Test

Program, and the licensee elects to take

full or partial credit for the partial

annealing, the licensee shall confirm in

writing to the Director, Office of Nuclear

Reactor Regulation that the partial

annealing was otherwise performed in

compliance with the Thermal Annealing

Operating Plan and relevant portions of

the Requalification Inspection and Test

Program. The licensee may restart its

reactor after the requirements of

paragraph (f)(2) of this section have

been met.

(iii) If the partial annealing was not

performed in accordance with the

Thermal Annealing Operating Plan and

relevant portions of the Requalification

Inspection and Test Program, the

licensee shall submit a summary of lack

of compliance with the Thermal

Annealing Operating Plan and the

Requalification Inspection and Test

Program and a justification for

subsequent operation to the Director,

Office of Nuclear Reactor Regulation.

Any changes to the facility as described

in the updated final safety analysis

report which are attributable to the

noncompliances and constitute

unreviewed safety questions, and any

changes to the technical specifications

which are required as a result of the

noncompliances, shall also be

identified.

(A) If no unreviewed safety questions

or changes to technical specifications

are identified, the licensee may restart

its reactor after the requirements of

paragraph (f)(2) of this section have

been met.

(B) If any unreviewed safety questions

or changes to technical specifications

are identified, the licensee may not

restart its reactor until approval is

obtained from the Director, Office of

Nuclear Reactor Regulation and the

requirements of paragraph (f)(2) of this

section have been met.

(d) Thermal Annealing Results

Report. Every licensee that either

completes a thermal annealing, or that

terminates an annealing but elects to

take full or partial credit for the

annealing, shall provide the following

information within three months of

completing the thermal anneal, unless

an extension is authorized by the

Director, Office of Nuclear Reactor

Regulation:

(1) The time and temperature profiles

of the actual thermal annealing;

(2) The post-anneal RTNDT and

Charpy upper-shelf energy values of the

reactor vessel materials for use in

subsequent reactor operation;

(3) The projected post-anneal

reembrittlement trends for both RTNDT

and Charpy upper-shelf energy; and

(4) The projected values of RTPTS and

Charpy upper-shelf energy at the end of

the proposed period of operation

addressed in the Thermal Annealing

Report.

(e) Procedures for Determining the

Recovery of Fracture Toughness. The

procedures of this paragraph must be

used to determine the percent recovery

of ∆RTNDT, Rt, and percent recovery of

Charpy upper-shelf energy, Ru. In all

cases, Rt and Ru may not exceed 100.

(1) For those reactors with

surveillance programs which have

developed credible surveillance data as

defined in § 50.61, percent recovery due

to thermal annealing (Rt and Ru) must be

evaluated by testing surveillance

specimens that have been withdrawn

from the surveillance program and that

have been annealed under the same

time and temperature conditions as

those given the beltline material.

(2) Alternatively, the percent recovery

due to thermal annealing (Rt and Ru)

may be determined from the results of
65474 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

6For those cases where materials are removed

from the beltline of the pressure vessel, the stress

limits of the applicable portions of the ASME Code

Section III must be satisfied, including

consideration of fatigue and corrosion, regardless of

the Code of record for the vessel design.

a verification test program employing

materials removed from the beltline

region of the reactor vessel 6 and that

have been annealed under the same

time and temperature conditions as

those given the beltline material.

(3) Generic computational methods

may be used to determine recovery if

adequate justification is provided.

(f) Public information and

participation.

(1) Upon receipt of a Thermal

Annealing Report, and a minimum of 30

days before the licensee starts thermal

annealing, the Commission shall:

(i) Notify and solicit comments from

local and State governments in the

vicinity of the site where the thermal

annealing will take place and any

Indian Nation or other indigenous

people that have treaty or statutory

rights that could be affected by the

thermal annealing,

(ii) Publish a notice of a public

meeting in the Federal Register and in

a forum, such as local newspapers,

which is readily accessible to

individuals in the vicinity of the site, to

solicit comments from the public, and

(iii) Hold a public meeting on the

licensee’s Thermal Annealing Report.

(2) Within 15 days after the NRC’s

receipt of the licensee submissions

required by paragraphs (c)(1), (c)(2) and

(c)(3)(i)–(iii) of this section, the NRC

staff shall place in the NRC Public

Document Room a summary of its

inspection of the licensee’s thermal

annealing, and the Commission shall

hold a public meeting:

(i) For the licensee to explain to NRC

and the public the results of the reactor

pressure vessel annealing,

(ii) for the NRC to discuss its

inspection of the reactor vessel

annealing, and

(iii) for the NRC to receive public

comments on the annealing.

(3) Within 45 days of NRC’s receipt of

the licensee submissions required by

paragraphs (c)(1), (c)(2) and (c)(3)(i)–(iii)

of this section, the NRC staff shall

complete full documentation of its

inspection of the licensee’s annealing

process and place this documentation in

the NRC Public Document Room.

5. In 10 CFR Part 50, Appendix G is

revised to read as follows:

Appendix G to Part 50—Fracture

Toughness Requirements

I. Introduction and scope.

II. Definitions.

III. Fracture toughness tests.

IV. Fracture toughness requirements.

I. Introduction and Scope

This appendix specifies fracture toughness

requirements for ferritic materials of

pressure-retaining components of the reactor

coolant pressure boundary of light water

nuclear power reactors to provide adequate

margins of safety during any condition of

normal operation, including anticipated

operational occurrences and system

hydrostatic tests, to which the pressure

boundary may be subjected over its service

lifetime.

The ASME Code forms the basis for the

requirements of this appendix. ‘‘ASME

Code’’ means the American Society of

Mechanical Engineers Boiler and Pressure

Vessel Code. If no section is specified, the

reference is to Section III, Division 1, ‘‘Rules

for Construction of Nuclear Power Plant

Components.’’ ‘‘Section XI’’ means Section

XI, Division 1, ‘‘Rules for Inservice

Inspection of Nuclear Power Plant

Components.’’ If no edition or addenda are

specified, the ASME Code edition and

addenda and any limitations and

modifications thereof, which are specified in

§ 50.55a, are applicable.

The sections, editions and addenda of the

ASME Boiler and Pressure Vessel Code

specified in § 50.55a have been approved for

incorporation by reference by the Director of

the Federal Register. A notice of any changes

made to the material incorporated by

reference will be published in the Federal

Register. Copies of the ASME Boiler and

Pressure Vessel Code may be purchased from

the American Society of Mechanical

Engineers, United Engineering Center, 345

East 47th Street, New York, NY 10017, and

are available for inspection at the NRC

Library, 11545 Rockville Pike, Two White

Flint North, Rockville, MD 20852–2738.

The requirements of this appendix apply to

the following materials:

A. Carbon and low-alloy ferritic steel plate,

forgings, castings, and pipe with specified

minimum yield strengths not over 50,000 psi

(345 MPa), and to those with specified

minimum yield strengths greater than 50,000

psi (345 MPa) but not over 90,000 psi (621

MPa) if qualified by using methods

equivalent to those described in paragraph

G–2110 of Appendix G of Section XI of the

latest edition and addenda of the ASME Code

incorporated by reference into § 50.55a(b)(2).

B. Welds and weld heat-affected zones in

the materials specified in paragraph I.A. of

this appendix.

C. Materials for bolting and other types of

fasteners with specified minimum yield

strengths not over 130,000 psi (896 MPa).

Note: The adequacy of the fracture

toughness of other ferritic materials not

covered in this section must be demonstrated

to the Director, Office of Nuclear Reactor

Regulation, on an individual case basis.

II. Definitions

A. Ferritic material means carbon and lowalloy steels, higher alloy steels including all

stainless alloys of the 4xx series, and

maraging and precipitation hardening steels

with a predominantly body-centered cubic

crystal structure.

B. System hydrostatic tests means all

preoperational system leakage and

hydrostatic pressure tests and all system

leakage and hydrostatic pressure tests

performed during the service life of the

pressure boundary in compliance with the

ASME Code, Section XI.

C. Specified minimum yield strength

means the minimum yield strength (in the

unirradiated condition) of a material

specified in the construction code under

which the component is built under § 50.55a.

D. RTNDT means the reference temperature

of the material, for all conditions.

(i) For the pre-service or unirradiated

condition, RTNDT is evaluated according to

the procedures in the ASME Code, Paragraph

NB–2331.

(ii) For the reactor vessel beltline materials,

RTNDT must account for the effects of neutron

radiation.

E. RTNDT means the transition temperature

shift, or change in RTNDT, due to neutron

radiation effects, which is evaluated as the

difference in the 30 ft-lb (41 J) index

temperatures from the average Charpy curves

measured before and after irradiation.

F. Beltline or Beltline region of reactor

vessel means the region of the reactor vessel

(shell material including welds, heat affected

zones, and plates or forgings) that directly

surrounds the effective height of the active

core and adjacent regions of the reactor

vessel that are predicted to experience

sufficient neutron radiation damage to be

considered in the selection of the most

limiting material with regard to radiation

damage.

III. Fracture Toughness Tests

A. To demonstrate compliance with the

fracture toughness requirements of Section IV

of this appendix, ferritic materials must be

tested in accordance with the ASME Code

and, for the beltline materials, the test

requirements of Appendix H of this part. For

a reactor vessel that was constructed to an

ASME Code earlier than the Summer 1972

Addenda of the 1971 Edition (under

§ 50.55a), the fracture toughness data and

data analyses must be supplemented in a

manner approved by the Director, Office of

Nuclear Reactor Regulation, to demonstrate

equivalence with the fracture toughness

requirements of this appendix.

B. Test methods for supplemental fracture

toughness tests described in paragraph

IV.A.1.b of this appendix must be submitted

to and approved by the Director, Office of

Nuclear Reactor Regulation, prior to testing.

C. All fracture toughness test programs

conducted in accordance with paragraphs

III.A and III.B must comply with ASME Code

requirements for calibration of test

equipment, qualification of test personnel,

and retention of records of these functions

and of the test data.

IV. Fracture Toughness Requirements

A. The pressure-retaining components of

the reactor coolant pressure boundary that

are made of ferritic materials must meet the

requirements of the ASME Code,

supplemented by the additional requirements

set forth below, for fracture toughness during

system hydrostatic tests and any condition of
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations 65475

1 Defined in ASTM E 185–79 and –82 which are

incorporated by reference in Appendix H to Part 50.

normal operation, including anticipated

operational occurrences. Reactor vessels may

continue to be operated only for that service

period within which the requirements of this

section are satisfied. For the reactor vessel

beltline materials, including welds, plates

and forgings, the values of RTNDT and Charpy

upper-shelf energy must account for the

effects of neutron radiation, including the

results of the surveillance program of

Appendix H of this part. The effects of

neutron radiation must consider the radiation

conditions (i.e., the fluence) at the deepest

point on the crack front of the flaw assumed

in the analysis.

1. Reactor Vessel Charpy Upper-Shelf Energy

Requirements

a. Reactor vessel beltline materials must

have Charpy upper-shelf energy,1 in the

transverse direction for base material and

along the weld for weld material according

to the ASME Code, of no less than 75 ft-lb

(102 J) initially and must maintain Charpy

upper-shelf energy throughout the life of the

vessel of no less than 50 ft-lb (68 J), unless

it is demonstrated in a manner approved by

the Director, Office of Nuclear Reactor

Regulation, that lower values of Charpy

upper-shelf energy will provide margins of

safety against fracture equivalent to those

required by Appendix G of Section XI of the

ASME Code. This analysis must use the latest

edition and addenda of the ASME Code

incorporated by reference into § 50.55a(b)(2)

at the time the analysis is submitted.

b. Additional evidence of the fracture

toughness of the beltline materials after

exposure to neutron irradiation may be

obtained from results of supplemental

fracture toughness tests for use in the

analysis specified in section IV.A.1.a.

c. The analysis for satisfying the

requirements of section IV.A.1 of this

appendix must be submitted, as specified in

§ 50.4, for review and approval on an

individual case basis at least three years prior

to the date when the predicted Charpy uppershelf energy will no longer satisfy the

requirements of section IV.A.1 of this

appendix, or on a schedule approved by the

Director, Office of Nuclear Reactor

Regulation.

2. Pressure-Temperature Limits and

Minimum Temperature Requirements

a. Pressure-temperature limits and

minimum temperature requirements for the

reactor vessel are given in Table 3, and are

defined by the operating condition (i.e.,

hydrostatic pressure and leak tests, or normal

operation including anticipated operational

occurrences), the vessel pressure, whether or

not fuel is in the vessel, and whether the core

is critical. In Table 3, the vessel pressure is

defined as a percentage of the preservice

system hydrostatic test pressure. The

appropriate requirements on both the

pressure-temperature limits and the

minimum permissible temperature must be

met for all conditions.

b. The pressure-temperature limits

identified as ‘‘ASME Appendix G limits’’ in

Table 3 require that the limits must be at

least as conservative as limits obtained by

following the methods of analysis and the

margins of safety of Appendix G of Section

XI of the ASME Code.

c. The minimum temperature requirements

given in Table 3 pertain to the controlling

material, which is either the material in the

closure flange or the material in the beltline

region with the highest reference

temperature. As specified in Table 3, the

minimum temperature requirements and the

controlling material depend on the operating

condition (i.e., hydrostatic pressure and leak

tests, or normal operation including

anticipated operational occurrences), the

vessel pressure, whether fuel is in the vessel,

and whether the core is critical. The metal

temperature of the controlling material, in

the region of the controlling material which

has the least favorable combination of stress

and temperature, must exceed the

appropriate minimum temperature

requirement for the condition and pressure of

the vessel specified in Table 1.

d. Pressure tests and leak tests of the

reactor vessel that are required by Section XI

of the ASME Code must be completed before

the core is critical.

B. If the procedures of Section IV.A. of this

appendix do not indicate the existence of an

equivalent safety margin, the reactor vessel

beltline may be given a thermal annealing

treatment to recover the fracture toughness of

the material, subject to the requirements of

§ 50.66. The reactor vessel may continue to

be operated only for that service period

within which the predicted fracture

toughness of the beltline region materials

satisfies the requirements of Section IV.A. of

this appendix using the values of RTNDT and

Charpy upper-shelf energy that include the

effects of annealing and subsequent

irradiation.

TABLE 1.—PRESSURE AND TEMPERATURE REQUIREMENTS FOR THE REACTOR PRESSURE VESSEL

Operating condition Vessel

pressure 1

Requirements for pressure-temperature

limits

Minimum temperature

requirements

1. Hydrostatic pressure and leak tests (core is not critical):

1.a Fuel in the vessel ...................................................... ≤20% ASME Appendix G Limits ........................ (2)

1.b Fuel in the vessel ...................................................... >20% ASME Appendix G Limits ........................ (2) +90°F (6)

1.c No fuel in the vessel (Preservice Hydrotest Only) .... ALL (Not Applicable) ....................................... (3) +60°F

2. Normal operation (incl. heat-up and cool-down), including

anticipated operational occurrences:

2.a Core not critical ......................................................... ≤20% ASME Appendix G Limits ........................ (2)

2.b Core not critical ......................................................... >20% ASME Appendix G Limits ........................ (2) +120°F (6)

2.c Core critical ................................................................ ≤20% ASME Appendix G Limits + 40°F ............ Larger of [(4)] or [(2) +

40°F]

2.d Core critical ................................................................ >20% ASME Appendix G Limits + 40°F ............ Larger of [(4)] or [(2) +

160°F]

2.e Core critical for BWR (5) ............................................ ≤20% ASME Appendix G Limits + 40°F ............ (2) + 60°F

1Percent of the preservice system hydrostatic test pressure.

2The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload.

3The highest reference temperature of the vessel.

4The minimum permissible temperature for the inservice system hydrostatic pressure test.

5For boiling water reactors (BWR) with water level within the normal range for power operation.

6 Lower temperatures are permissible if they can be justified by showing that the margins of safety of the controlling region are equivalent to

those required for the beltline when it is controlling.
65476 Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 / Rules and Regulations

6. In 10 CFR Part 50, Appendix H is

revised to read as follows:

Appendix H to Part 50—Reactor Vessel

Material Surveillance Program

Requirements

I. Introduction

II. Definitions

III. Surveillance Program Criteria

IV. Report of Test Results

I. Introduction

The purpose of the material surveillance

program required by this appendix is to

monitor changes in the fracture toughness

properties of ferritic materials in the reactor

vessel beltline region of light water nuclear

power reactors which result from exposure of

these materials to neutron irradiation and the

thermal environment. Under the program,

fracture toughness test data are obtained from

material specimens exposed in surveillance

capsules, which are withdrawn periodically

from the reactor vessel. These data will be

used as described in Section IV of Appendix

G to Part 50.

ASTM E 185–73, –79, and –82, ‘‘Standard

Practice for Conducting Surveillance Tests

for Light-Water Cooled Nuclear Power

Reactor Vessels,’’ which are referenced in the

following paragraphs, have been approved

for incorporation by reference by the Director

of the Federal Register. Copies of ASTM E

185–73, –79, and –82, may be purchased

from the American Society for Testing and

Materials, 1916 Race Street, Philadelphia, PA

19103 and are available for inspection at the

NRC Library, 11545 Rockville Pike, Two

White Flint North, Rockville, MD 20852–

2738.

II. Definitions

All terms used in this Appendix have the

same meaning as in Appendix G.

III. Surveillance Program Criteria

A. No material surveillance program is

required for reactor vessels for which it can

be conservatively demonstrated by analytical

methods applied to experimental data and

tests performed on comparable vessels,

making appropriate allowances for all

uncertainties in the measurements, that the

peak neutron fluence at the end of the design

life of the vessel will not exceed 1017 n/cm2

(E > 1 MeV).

B. Reactor vessels that do not meet the

conditions of paragraph III.A of this

appendix must have their beltline materials

monitored by a surveillance program

complying with ASTM E 185, as modified by

this appendix.

1. The design of the surveillance program

and the withdrawal schedule must meet the

requirements of the edition of ASTM E 185

that is current on the issue date of the ASME

Code to which the reactor vessel was

purchased. Later editions of ASTM E 185

may be used, but including only those

editions through 1982. For each capsule

withdrawal, the test procedures and

reporting requirements must meet the

requirements of ASTM E 185–82 to the extent

practicable for the configuration of the

specimens in the capsule.

2. Surveillance specimen capsules must be

located near the inside vessel wall in the

beltline region so that the specimen

irradiation history duplicates, to the extent

practicable within the physical constraints of

the system, the neutron spectrum,

temperature history, and maximum neutron

fluence experienced by the reactor vessel

inner surface. If the capsule holders are

attached to the vessel wall or to the vessel

cladding, construction and inservice

inspection of the attachments and attachment

welds must be done according to the

requirements for permanent structural

attachments to reactor vessels given in

Sections III and XI of the American Society

of Mechanical Engineers Boiler and Pressure

Vessel Code (ASME Code). The design and

location of the capsule holders must permit

insertion of replacement capsules.

Accelerated irradiation capsules may be used

in addition to the required number of

surveillance capsules.

3. A proposed withdrawal schedule must

be submitted with a technical justification as

specified in § 50.4. The proposed schedule

must be approved prior to implementation.

C. Requirements for an Integrated

Surveillance Program.

1. In an integrated surveillance program,

the representative materials chosen for

surveillance for a reactor are irradiated in one

or more other reactors that have similar

design and operating features. Integrated

surveillance programs must be approved by

the Director, Office of Nuclear Reactor

Regulation, on a case-by-case basis. Criteria

for approval include the following:

a. The reactor in which the materials will

be irradiated and the reactor for which the

materials are being irradiated must have

sufficiently similar design and operating

features to permit accurate comparisons of

the predicted amount of radiation damage.

b. Each reactor must have an adequate

dosimetry program.

c. There must be adequate arrangement for

data sharing between plants.

d. There must be a contingency plan to

assure that the surveillance program for each

reactor will not be jeopardized by operation

at reduced power level or by an extended

outage of another reactor from which data are

expected.

e. There must be substantial advantages to

be gained, such as reduced power outages or

reduced personnel exposure to radiation, as

a direct result of not requiring surveillance

capsules in all reactors in the set.

2. No reduction in the requirements for

number of materials to be irradiated,

specimen types, or number of specimens per

reactor is permitted.

3. After (the effective date of this section),

no reduction in the amount of testing is

permitted unless previously authorized by

the Director, Office of Nuclear Reactor

Regulation.

IV. Report of Test Results

A. Each capsule withdrawal and the test

results must be the subject of a summary

technical report to be submitted, as specified

in § 50.4, within one year of the date of

capsule withdrawal, unless an extension is

granted by the Director, Office of Nuclear

Reactor Regulation.

B. The report must include the data

required by ASTM E 185, as specified in

paragraph III.B.1 of this appendix, and the

results of all fracture toughness tests

conducted on the beltline materials in the

irradiated and unirradiated conditions.

C. If a change in the Technical

Specifications is required, either in the

pressure-temperature limits or in the

operating procedures required to meet the

limits, the expected date for submittal of the

revised Technical Specifications must be

provided with the report.

Dated at Rockville MD, this 12th day of

December, 1995.

For the Nuclear Regulatory Commission.

John C. Hoyle,

Secretary of the Commission.

[FR Doc. 95–30665 Filed 12–18–95; 8:45 am]

BILLING CODE 7590–01–P