Fracture Toughness Requirements for Light Water Reactor Pressure Vessels

SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend

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 proposed amendments would 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.

DATES: The comment period expires January 3, 1994. Comments received

after this date will be considered if it is practical to do so, but the

Commission is able to assure consideration only for comments received

on or before this date.

ADDRESSES: Mail comments to: The Secretary of the Commission, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, Attention:

Docketing and Service Branch.

Comments may be delivered to: One White Flint North, 11555

Rockville Pike, Rockville, MD 20852, between 7:30 am and 4:15 pm on

Federal workdays. Comments received on the proposed rules may be

examined at the NRC Public Document Room, 2120 L Street NW. (Lower

Level), Washington, DC.

A free single copy of draft regulatory guides DG-1023, DG-1025, and

DG-1027 may be requested by those considering public comment by writing

to the U.S. Nuclear Regulatory Commission, ATTN: Distribution and Mail

Services Section, Room P-130A, Washington, DC 20555. A copy is also

available for inspection and/or copying in the NRC Public Document

Room, 2120 L Street NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT: Alfred Taboada, Division of

Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear

Regulatory Commission, Washington, DC 20555, telephone: (301) 415-6014.

SUPPLEMENTARY INFORMATION:

Background

Maintaining the structural integrity of the reactor pressure vessel

(RPV) of light-water-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 measurement methods and procedures to

assure that each RPV has adequate safety margins for continued

operation. Structural integrity of a reactor pressure vessel 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 a proposed rule on thermal annealing and

several draft regulatory guides, the existing and proposed regulatory

documents contribute to a comprehensive set of regulations and

regulatory guidance pertaining to RPV integrity, including:

1. The fracture toughness criteria that each RPV must satisfy (10 CFR

50.60, 10 CFR 50.61, and 10 CFR Part 50, Appendix G).

2. Irradiation embrittlement surveillance requirements (10 CFR Part 50,

Appendix H).

3. Guidance for estimating the fracture toughness of the RPV

(Regulatory Guide 1.99 and a draft regulatory guide on dosimetry).

4. Guidance for cases in which the RPV is estimated to exceed specified

screening criteria (Regulatory Guide 1.154 and a draft regulatory guide

on evaluating RPVs with Charpy upper-shelf energy less than 50 ft-lb).

5. Requirements and guidance for using thermal annealing of the RPV as

a method for mitigating the effects of neutron irradiation (proposed 10

CFR 50.66 and a draft regulatory guide).

This notice proposes changes to the following requirements:

a. The Pressurized Thermal Shock (PTS) rule (10 CFR 50.61).

b. Appendix G of 10 CFR Part 50, ``Fracture Toughness Requirements.''

c. Appendix H of 10 CFR Part 50, ``Reactor Vessel Material Surveil-

lance Program Requirements.''

This notice also proposes a thermal annealing requirement, 10 CFR

50.66. In addition to the proposed amendments addressed in this

document, the NRC will publish for public comment a draft regulatory

guide on thermal annealing, DG-1027 (the availability of this draft

regulatory guide for will be announced in the Federal Register at a

later date comment.)

Two related draft regulatory guides have been published for public

comment (58 FR 51392; October 1, 1993). These draft regulatory guides

are:

1. A draft regulatory guide that addresses evaluation of RPVs with

Charpy upper-shelf energy levels less than 50 ft-lb (DG-1023).

2. A draft regulatory guide on dosimetry (DG-1025).

Other regulatory guides pertaining to RPV integrity, Regulatory

Guides 1.99 and 1.154, are under evaluation. Revisions to these guides,

if any, will be addressed in the future.

Other regulatory issues related to reactor pressure vessel

integrity, such as low temperature over-pressure protection system

setpoints, are being addressed as part of a broader scope evaluation of

the pressure vessel regulations and are not part of this proposed

amendment.

Reasons for the Proposed Changes

The needs for these proposed amendments to the fracture toughness

regulations have been identified from three sources:

1. The 1989 Nuclear Utility Backfitting and Reform Group (NUBARG)

appeal concerning use of nuclear heat to warm the RPV for system

leakage and hydrostatic pressure tests;

2. The 1990 review of the RPV integrity of the Yankee Nuclear Power

Station; and

3. A comprehensive review of the regulations by NRC staff, resulting in

the identification of the need for clarifications, corrections, and

improved guidance in certain areas.

The recognition in 1986 by NRC staff that certain boiling water

reactor (BWR) units were using nuclear heat to warm the system prior to

performing leakage and pressure tests led to an NRC staff initiative to

amend Appendix G to 10 CFR Part 50. During the NRC staff review to

determine if this use of nuclear heating was permissible under either

the ASME Code or the NRC regulations, the Nuclear Utility Backfitting

and Reform Group (NUBARG) filed a backfitting claim, and later an

appeal of the determination that a backfit was not involved. Stemming

from this claim and appeal process, the Committee to Review Generic

Requirements recommended to the Executive Director for Operations that

the affected portions of Appendix G be revised to clearly indicate that

all required leakage and pressure tests of the reactor pressure vessel

must be performed when the core is not critical.

In 1990, the NRC began a review of the integrity of the Yankee

Nuclear Power Station (YNPS) RPV. That review, along with stated plans

by the licensee to consider thermal annealing of the RPV, highlighted

the need for the NRC to amend its regulations and guidance pertaining

to RPV integrity. The NRC staff proposed a plan to revise and clarify

the pertinent regulations in SECY-91-333 (October 22, 1991) and SECY-

92-283 (August 14, 1992), including schedules and general descriptions

of the changes contemplated. The proposed changes included

clarifications and corrections planned prior to the YNPS review.

However, the YNPS review identified the need to clarify the

requirements in Sections IV and V of Appendix G to 10 CFR Part 50, and

the need to provide more complete requirements and guidance for thermal

annealing.

The PTS rule, 10 CFR 50.61, was amended on May 15, 1991 (56 FR

22300) to make the method for evaluating irradiation embrittlement

consistent with the recommended procedures of Regulatory Guide 1.99,

Revision 2, ``Radiation Embrittlement of Reactor Vessel Materials.''

Subsequent inquiries to the Commission concerning the appropriate

margin terms and use of surveillance data indicated that the PTS rule

required clarification. A recent review of the rule by NRC staff

concluded that the PTS rule should also be modified to bring the

procedures for evaluating RTPTS into complete agreement with the

recommended procedures in Regulatory Guide 1.99, Revision 2.

Overview of the Proposed Changes

PTS Rule (10 CFR 50.61)

The pressurized thermal shock rule, 10 CFR 50.61, was initially

published in final form on July 23, 1985 (50 FR 29937) and amended on

May 15, 1991 (56 FR 22300). This rule provides a screening criterion

for irradiation embrittlement of RPV beltline materials, above which

the plant cannot continue to be operated without justification.

Historically, a value of reference temperature has been determined for

each vessel beltline material for comparison to the PTS screening

criteria. These values of reference temperature are termed RTPTS

values. However, the method for evaluating RTPTS values has not

been consistent with the embrittlement estimates used for other

purposes, such as pressure-temperature limit calculations. The May 15,

1991, amendment was a step towards unifying the embrittlement estimate

methodology. The amendment included the procedures given in Regulatory

Guide 1.99, Revision 2, for the evaluation of irradiation embrittlement

of the RPV beltline materials. The 1991 amendment left two differences

between the rule and Regulatory Guide 1.99, Revision 2. These two

differences are:

1. Values of unirradiated RTNDT are specified for general

classes of material in the PTS rule, while greater flexibility in

determining unirradiated values is permitted in Regulatory Guide 1.99,

Revision 2; and

2. The margin terms used in the PTS rule are based on assumptions

which are not consistent with the method used in Regulatory Guide 1.99,

Revision 2 for calculating the margin term.

This proposed amendment is intended to make the evaluation of

RTPTS consistent with the recommended methods of Regulatory Guide

1.99, Revision 2, which are used to evaluate RTNDT. In this case,

the RTPTS value for each vessel beltline material is simply the

RTNDT value estimated for the projected end of license fluence.

This proposed amendment to the PTS rule would make three changes:

1. The Regulatory Guide 1.99, Revision 2, method for determining

RTNDT, of which RTPTS is a unique value determined for the

end of license fluence, would be incorporated in total, including

treatment of the unirradiated RTNDT value, the margin term, and

the explicit definition of ``credible'' surveillance data.

2. The section would be restructured to improve clarity, with the

requirements section giving only the requirements for the RTPTS

value. The method for calculating RTPTS would be moved to a new

paragraph of the rule.

3. Thermal annealing would be introduced as a method for mitigating

the effects of neutron irradiation, thereby reducing RTPTS.

Additionally, it should be noted that evaluations of current

surveillance data have indicated that the standard deviation of the

differences between predicted and measured shifts in RTNDT, termed

the residual, are higher than the margin values used in the PTS rule

and in Regulatory Guide 1.99, Revision 2, particularly for plate

materials. However, the mean embrittlement estimation equations in the

rule and in Regulatory Guide 1.99, Revision 2, overestimate the actual

surveillance data by less than 10 deg.F on average. The NRC staff

considered amending the PTS rule to incorporate the revised margin

terms and the overestimation bias, but decided against such an

amendment due to the small number of plants that would be affected by

such a change, and a related NRC research program that addresses the

overall issue of irradiation embrittlement correlations. The number of

plants which would have their RTPTS values change

``significantly'' by such a change to the margin terms is not large;

the impact of the revised margin terms on those plants is being

addressed through other regulatory mechanisms. The effect of the

revised margin on pressure-temperature limits is being handled in a

similar manner.

As noted, this proposed amendment to 10 CFR 50.61 introduces

thermal annealing of the reactor pressure vessel beltline as a method

for mitigating the effects of neutron irradiation and reducing

RTPTS to levels below the screening criteria. As specified in

Sec. 50.61(b)(7) of this proposed rule, the use of thermal annealing

would be subject to the requirements of the proposed new section on

thermal annealing (10 CFR 50.66).

Thermal Annealing Rule (10 CFR 50.66)

The proposed thermal annealing rule, 10 CFR 50.66, would provide a

consistent set of requirements for the use of thermal annealing to

mitigate the effects of neutron irradiation. The proposed rule would

replace the requirements for annealing in the current Appendix G of 10

CFR Part 50 with the proposed consistent set of requirements in this

proposed rule. Also, the PTS rule would be amended to add a new

paragraph (b)(7) which would reference the proposed thermal annealing

rule as a method for mitigating the effects of neutron irradiation,

thereby reducing RTPTS. Therefore, the intent of the thermal

annealing rule, and related changes in 10 CFR 50.61 and Appendix G of

10 CFR Part 50, is to provide requirements for use of thermal annealing

to mitigate the deleterious effects of neutron irradiation on reactor

vessel material properties.

Consistent with guidance in Section V.D of the current Appendix G,

the proposed thermal annealing rule would specify that thermal

annealing would be subject to the approval of the Director, Office of

Nuclear Reactor Regulation (NRR). Section 50.66(a) of the proposed

thermal annealing rule would require submittal of an application

containing three sections: a thermal annealing operating plan, a

requalification inspection and test program, and a fracture toughness

recovery and reembrittlement rate assurance program. This application

would be required to be submitted at least three years before the

proposed date of the annealing operation. This three-year period is

specified only to provide NRC staff with sufficient time to review the

thermal annealing application. The licensee may initiate the thermal

annealing program as soon as NRC approval is given, even if this

approval is given before three years from the date of the application.

The thermal annealing operating plan also must include an

evaluation of the effects of temperature, and of mechanical and thermal

stresses on the reactor and associated equipment to demonstrate that

the operability of the reactor will not be detrimentally affected. 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.

Upon completion of the thermal annealing and before subsequent

operation of the plant, the licensee would be required to certify that

the thermal annealing was performed in accordance with the approved

application, that the approved criteria were satisfied, and that the

proposed annealing conditions were not exceeded. However, in the event

that the licensee cannot make this certification, a justification for

subsequent operation would have to be submitted for approval by the

Director, NRR. However, this provision does not relieve the licensee

from obtaining 10 CFR 50.12 exemptions from any other requirements of

this part that cannot be satisfied.

Two items of particular importance to the overall annealing are the

recovery of fracture toughness and the rate of reembrittlement of the

RPV beltline materials. This proposed rule provides three alternative

methods for determining these values, ranging from assessments using

plant-specific materials to an assessment using a generic computation.

Two methods for evaluating annealing recovery are experimental

methods to determine plant-specific annealing recovery, and the third

method is a generic computational method. The first method would be

required for those plants with ``credible'' surveillance programs and

where surveillance specimens are available. The second method would be

an optional method, in which the licensee may remove material from the

beltline of the RPV to evaluate annealing recovery. This method should

provide the most accurate evaluation of annealing recovery. Presumably,

it would be selected for those plants without credible surveillance

programs or when surveillance specimens are not available. However, for

this method to be acceptable, the vessel must be sufficiently thick so

that the stress limits in Section III of the ASME Code can be

satisfied, considering the effects of fatigue and corrosion.

The third method would use generic computational methods, for which

appropriate justification would be required.

Paragraph (d) of Sec. 50.66 describes the experimental methods and

the computational method 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

plant-specific estimate of recovery, rather than the generic value

evaluated from the computational method. It is the intent of this

proposed rule to require that surveillance specimens from ``credible''

surveillance programs must be used to develop plant-specific recovery

data, if such specimens are available. It is not the intent of this

rule to require the removal of material from the RPV beltline to permit

plant-specific evaluation of recovery.

As described previously, the computational method would require

appropriate justification.

Reembrittlement trends are estimated, and monitored by continued

surveillance in accordance with Appendix H of 10 CFR Part 50.

Paragraph (b)(3)(ii) provides that the reembrittlement rate must be

monitored using a surveillance program which conforms to Appendix H of

this part. Some older plants conform to Appendix H by applying issues

of ASTM Standard E 185 that do not require the use of the vessel

``limiting materials'' in the surveillance program. Within this

context, the term ``limiting materials'' refers to the materials

predicted to have the highest RTNDT or the lowest Charpy upper

shelf energy during the operational lifetime of the plant. It is the

intent of this rule that, as required by later issues of ASTM Standard

E 185, the vessel ``limiting materials'' should be used to monitor

reembrittlement if the materials are available.

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 proposed

amendments to Appendix G are principally of a clarifying or a

restructuring nature. These amendments include:

1. Sections IV and V of Appendix G which would be combined and

rewritten to clarify the Charpy upper-shelf energy requirements, and

the pressure-temperature and minimum permissible temperature

requirements.

2. An explicit statement that would be added to Section IV

requiring 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.

3. The proposed thermal annealing rule, 10 CFR 50.66, that would be

referenced in lieu of the details on thermal annealing previously given

in Section V.D.

4. The reference to the ASME Code that would be changed from

Appendix G of Section III to Appendix G of Section XI of the ASME Code.

5. The ``design to permit annealing'' requirement (Section IV.B),

which would be deleted.

The restructuring of Sections IV and V is intended to promote

clarity of the requirements in these sections. The procedures required

for cases in which the Charpy upper-shelf energy of a RPV beltline

material falls below 50 ft-lb also would be clarified by consolidating

the requirements previously addressed in parts of Sections IV and V.

The provisions in Section V.C concerning requirements for

``volumetric inspection'' and ``additional evidence of fracture

toughness'' would be removed. The volumetric examination requirement

would be removed because it was 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 would be incorporated in the ``equivalent margins''

analysis in Section IV.A.1, as a provisional method for developing

fracture toughness data needed for that analysis. At the present time

there is an adequate generic fracture toughness data base available to

perform these analyses, with appropriate bounding considerations. The

modification would permit a licensee to develop plant-specific data.

Generally, plant-specific data would result in a reduction in the

margin applied to the fracture toughness data, to reflect the reduction

in uncertainties due to the availability of plant-specific data.

However, this must be evaluated on a case-by-case basis.

The pressure-temperature and minimum permissible temperature

requirements in Section IV would be restructured, with the principal

feature being 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 would be 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.

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, before the completion of these tests. The use of

nuclear heat before the completion of such tests is considered unsafe

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 recommended to the

Executive Director for Operations by the Committee to Review Generic

Requirements in a memorandum dated June 7, 1990.

The requirements on thermal annealing contained in the current

Appendix G (Section V.D) would be replaced by a reference to the

proposed Thermal Annealing rule, 10 CFR 50.66.

Changing the reference to Appendix G of the ASME Code from Section

III to Section XI means that the requirements for operating plants will

no longer come from the construction code (Section III of the ASME

Code) but instead will come from Section XI, the in-service inspection

code. Appendix G to Section XI and Appendix G to Section III are

identical, so this amendment would not result in a change in technical

requirements.

Section IV.B of Appendix G requires that:

``Reactor vessels for which the predicted value of upper-shelf

energy at end of life is below 50 ft-lb or for which the predicted

value of adjusted reference temperature at end of life exceeds 200

deg.F (93 deg.C) must be designed to permit a thermal annealing

treatment * * *''

This proposed rule would delete that requirement. This deletion

conforms with Commission direction from 1985 and public comments to

delete this section. An additional consideration to delete this

requirement is that there should be no requirement to ensure the

feasibility of a (future) voluntary activity.

During the Commission review of the revision of Appendix G

published final on May 27, 1983 (48 FR 24009), the requirement to

``design to permit annealing'' was criticized because licensee response

to the requirement was perfunctory and staff review of the responses

was cursory, as detailed in SECY-83-254 (June 27, 1983). Further, there

were no criteria to assess whether a design would permit annealing. An

additional problem cited with the requirement was that it was

misinterpreted to mean that plant operation with an RTNDT greater

than 200 deg.F or a Charpy upper-shelf energy below 50 ft-lb was

unsafe. The Commission indicated that it would seek public comments on

the proposed deletion of the requirement, and this was done

concurrently with the publication of the proposed PTS rule on February

4, 1984 (49 FR 4498). All sixteen of the commenters on this item

recommended deletion of the paragraph, although eight of them urged

that the deletion should not in any way imply that annealing is no

longer an option to increase safety margins. In the notice of final

rulemaking for the PTS rule published on July 23, 1985 (50 FR 29944),

the ``Supplementary Information'' noted that the Commission planned a

separate rulemaking action to delete Section IV.B. That planned

deletion was delayed so that it could be combined with other amendments

to Appendix G.

Appendix H of 10 CFR Part 50

Changes in the fracture toughness properties of the RPV beltline

materials due to irradiation embrittlement are monitored using a

surveillance program, as required in Appendix H of 10 CFR Part 50,

``Reactor Vessel Material Surveillance Program Requirements''. 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. This proposed

amendment would make the following changes:

1. End the provision for ``reducing the amount of testing'' for

integrated surveillance programs,

2. Restructure the section on requirements for integrated surveillance

programs (Section II.C), and

3. Clarify the version of ASTM Standard E 185 that applies to the

surveillance program.

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.'' It is proposed to discontinue this provision as of

the effective date of the Appendix, although previous authorizations

granted by the Director, Office of Nuclear Reactor Regulation, would

continue in effect.

A second proposed change to Appendix H restructures Section II.C to

clarify the 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 subsequent testing and reporting of results from

the surveillance capsules. Since 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. The clarification proposed in

this revision indicates that the design of the program and the

withdrawal schedule must meet the requirements of ASTM E 185-73, or the

edition of ASTM E 185 that is current on the issue date of the ASME

Code to which the reactor vessel was purchased, whichever is latest.

Licensees could choose to comply with later editions of ASTM E 185, up

through the 1982 edition. Further, specimen 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.''

The NRC staff intended that this proposed amendment to Appendix H

would incorporate by reference a version of ASTM standard E 185 updated

from the currently available 1982 version. However, that

standardization process has not been completed, and it was decided to

proceed with this proposed amendment. A subsequent amendment to

Appendix H will be considered after the NRC staff has reviewed the

updated ASTM standard.

Request for Public Comments

On June 13, 1994 (SECY-94-163) the staff requested Commission

approval to publish for public comment these proposed revisions and

provided a discussion of options for public participation related to

thermal annealing. The Commission approved issuance of the proposed

revisions but directed that the staff to (1) include with the proposed

rule package a discussion of options the staff has considered for

structuring of the regulatory process for the proposed thermal

annealing rule (10 CFR 50.66), which is included in the following

section, and (2) request comments on the following 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.

Options the Staff Has Considered for Structuring of the Regulatory

Process Related to Public Participation in Thermal Annealing

A significant issue with respect to thermal annealing, identified

in SECY-92-283 (August 14, 1992), is the nature and timing of public

participation related to the NRC's review and approval of a licensee's

proposal for thermal annealing. The proposed rule does not address

public participation per se, but instead provides the performance

requirements that a licensee would have to meet to gain NRC approval of

a thermal annealing application and to permit subsequent operation.

Under the proposed rule, there are three circumstances that arguably

require an opportunity for hearing pursuant to Section 189 of the

Atomic Energy Act of 1954 (AEA) in connection with NRC review and

approval of thermal annealing. First, a licensee seeking to anneal its

reactor vessel must obtain NRC approval of the content of the thermal

annealing plan prior to implementing the plan (see Sec. 50.66(a) of the

proposed rule). Second, and apart from the NRC approval required under

Sec. 50.66(a), the thermal annealing process as described in the

licensee's plan may necessitate license amendments (including technical

specification changes). License amendments may be required if the

licensee's final safety analysis report (FSAR) needs to be revised to

reflect the thermal annealing process, and the licensee is unable to

conclude that such FSAR changes do not constitute ``unreviewed safety

questions'' under 10 CFR 50.59. Implementation of the thermal annealing

plan may also violate existing technical specifications, necessitating

requests for changes to technical specifications. Any license amendment

and technical specification change must be approved by the NRC before

the licensee may implement the thermal annealing plan. Finally, after

the licensee implements the annealing plan, if he determines that he

cannot meet the criteria specified in Sec. 50.66(c)(1) of the proposed

rule, then NRC approval is needed in order for the licensee to resume

operation (see Sec. 50.66(c)(2) of the proposed rule).

It is clear that any license amendments and technical specification

changes necessitated by the thermal annealing plan would require an

opportunity for hearing, in accordance with Section 189 of the AEA.

However, the scope of such a hearing would normally be limited to

consideration of whether the proposed license amendment and technical

specification changes are in accordance with the Commission's rules,

and therefore provide reasonable assurance of adequate protection to

the public health and safety. Issues related to the more general matter

of the acceptability of the thermal annealing plan proposed by the

licensee would not fall within the scope of any hearing for license

amendment or technical specification change, except as they fall

directly in the scope of the requested amendment or technical

specification change.

However, there is some question whether the AEA requires an

opportunity for hearing in connection with the NRC approval of the

thermal annealing plan or the NRC decision approving resumption of

operation under the proposed rule. There are four primary alternatives

with respect to providing an opportunity for hearing in connection with

thermal annealing. These alternatives are discussed in greater detail

below:

Alternative 1. No Opportunity for Hearing Under Rule as Proposed

Under this alternative, the contention is that Section 189 of the

AEA does not afford an interested member of the public a right to

request a hearing in connection with NRC approvals of thermal annealing

plans and resumption of operation under Sec. 50.66(c)(2). This

alternative is consistent with other provisions in 10 CFR Part 50 where

approval by the Director of NRR is required and hearings are not

routinely offered.

Notwithstanding the lack of a requirement for a public hearing, the

staff anticipates that, with respect to the initial or the first

several applications for thermal annealing, several informal hearings

or public meetings would be held by the staff to permit discussion of

both the thermal annealing plan proposed by the licensee and the

technical issues related to annealing. These hearings or meetings would

ensure that all of the pertinent technical issues have been addressed

by the licensee in its thermal annealing plan and by the staff in its

review of the plan. These hearings or meetings would be noticed in the

Federal Register.

Alternative 2. Discretionary Opportunity for Hearing Under Rule as

Proposed

Under this alternative, the contention is that Section 189 of the

AEA does not afford an interested member of the public a right to

request a hearing. However, as a matter of discretion, the Commission

would determine on a case-by-case basis whether an opportunity for

hearing will be provided in connection with the Director of NRR's

determination on a thermal annealing application under Sec. 50.66(b) of

the proposed rule. In the hearing, the Commission would consider issues

related to the adequacy of the thermal annealing plan, as well as the

vessel's ability to perform its safety function after being annealed.

A case-by-case determination would also be made by the NRC with

respect to providing an opportunity for hearing on the Director of

NRR's determination on the licensee's justification for subsequent

operation under the proposed Sec. 50.66(c)(2).

In both cases, the Commission would publish a notice in the Federal

Register announcing the NRC's approval of the licensee's thermal

annealing plan or approval of resumed operation under Sec. 50.66(c)(2).

Neither implementation of the thermal annealing plan nor resumption of

operation, once approved by the NRC, would be contingent upon

completion of any hearing; i.e., the Commission does not believe that

it is required to make a Section 189 ``no significant hazards

determination'' (``Sholly finding'') when it provides a discretionary

hearing.

Alternative 3. Required Opportunity for Hearing Under Rule as Proposed

Under this alternative, the contention is that a hearing is

required by Section 189 of the AEA for both NRC's approval of the

thermal annealing plan and any NRC approval of resumed operation

following annealing. The adequacy of the thermal annealing plan, as

well as the vessel's ability to perform its safety function after being

annealed, could be raised in the hearing associated with approval of

the thermal annealing plan. Licensee implementation of the thermal

annealing plan could not commence until any hearing is concluded unless

the NRC makes a ``no significant hazards determination'' with respect

to the thermal annealing.

Alternative 4. Modify Proposed Rule to Require Suspension of License

Prior to Thermal Annealing

Under this alternative, the proposed rule's regulatory approach for

thermal annealing would be modified to include a suspension of the

operating license during thermal annealing. The suspension would be

automatic under the rule, without the need for a suspension order,

although a letter confirming the licensee's status under the annealing

rule would be prepared. The rule itself, as is currently drafted, would

specify the conditions for lifting of the suspension (Section

50.66(b)). The licensee would anneal its reactor vessel without prior

NRC approval of its program for conducting the annealing. Upon

completion, the suspension would be lifted only if the licensee

demonstrated that the thermal annealing has addressed the reactor

enbrittlement such that it is acceptable to operate the plant. There

would be no opportunity for hearing associated with the lifting of the

suspension, and since there would be no prior NRC approval of the

annealing program, a hearing opportunity under Section 189 would not be

implicated by any such approval.

Submission of Comments in Electronic Format

Commenters are encouraged to submit, in addition to the original

paper copy, a copy of the letter in electronic format on a DOS-

formatted (IBM compatible) 5.25 or 3.5 inch computer diskette. Text

files should be provided in WordPerfect format or unformatted ASCII

code. The format and version should be identified on the diskette's

external label.

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, if adopted, would not be a

major Federal action significantly affecting the quality of human

environment and, therefore, an environmental impact statement is not

required.

As discussed below, the individual actions covered in this proposed

rulemaking would either serve to enhance safety of the reactor pressure

vessel, thereby decreasing the environmental impact of plant operation,

or have no 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 would serve 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 would have 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 proposed thermal annealing rule (10 CFR 50.66) would permit and

provide 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 rule does not affect all

plants but provides an alternative for assuring compliance with the

requirements in 10 CFR 50.61 and Appendix G of 10 CFR Part 50, and

would only apply when a licensee elects to use it.

The application of thermal annealing to a reactor vessel would

improve the condition of the reactor vessel material. In addition, this

rule would establish requirements to avoid damaging the reactor system

and to protect against accidents during the annealing operation, with

attendant environmental consequences.

This rule is one of several regulatory requirements that will

function to ensure reactor vessel integrity. In that sense, this rule

would have a positive impact on the environment by reducing the

potential for vessel failure. For these reasons, the Commission has

determined that there would be no significant impact and, therefore, an

environmental statement is not required.

Appendix G of 10 CFR Part 50

Concerning the amendments proposed to Appendix G of 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 environmental impact of plant

operation.

The restructuring of Sections IV and V of Appendix G is clarifying

or corrective in nature, and hence 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 since 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 since Appendix G currently permits the use of

thermal annealing to reduce fracture toughness loss of the RPV

materials due to irradiation embrittlement.

The deletion of the ``design to permit annealing'' requirement has

no adverse impact on the environment. This assessment is based on the

fact that annealing is a voluntary activity.

Appendix H of 10 CFR Part 50

Concerning the amendments proposed to Appendix H of 10 CFR Part 50,

the requirement that all irradiation surveillance tests be made (i.e.,

no reduction in testing is permitted) would 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 proposed rule amends information collection requirements that

are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et

seq.). This rule has been submitted to the Office of Management and

Budget for review and approval of the paperwork requirements.

The public reporting burden for this collection of information is

estimated to average 6,000 hours per respondent, 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, D.C. 20555; and to the Desk Officer, Office of Information

and Regulatory Affairs, NEOB-3019, (3150-0011), Office of Management

and Budget, Washington, D.C. 20503.

Regulatory Analysis

The NRC staff has prepared a regulatory analysis for the proposed

amendments to 10 CFR 50.61, 10 CFR Part 50, Appendix G and 10 CFR Part

50 Appendix H, which describes the factors and alternatives considered

by the Commission in deciding to propose 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. Single copies of the analysis may be obtained

from Alfred Taboada, Office of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, telephone, (301)

415-6014.

Single copies of the regulatory analysis prepared for 10 CFR 50.66

may be obtained from Alfred Taboada, Office of Nuclear Regulatory

Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555,

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, if adopted, the proposed amendments would

not have a significant economic impact on a substantial number of small

entities. The rules which would be affected by the proposed amendments:

(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 pressure-

temperature 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 or the Small Business Size Standards in regulations

issued by the Small Business Administration at 13 CFR Part 121.

Backfit Analysis

PTS Rule (10 CFR 50.61)

The proposed revision to Section 50.61 would require licensees to

calculate RTPTS using the same methodology specified in Regulatory

Guide 1.99, Revision 2 for determining RTNDT. This proposal is

logically a requisite part of the 1991 revisions to Sec. 50.61, which

addressed a unified method for calculating radiation enbrittlement of

the reactor beltline materials. However, the Commission inadvertently

failed to make the conforming change to Sec. 50.61. Therefore, the

Commission believes that the backfit statement for the 1991 amendments,

which determined that the backfits were necessary to ensure that the

facility provides adequate protection to the public health and safety,

are applicable to this conforming change to Sec. 50.61.

The restructuring of the PTS rule does not impose any backfits as

defined in 10 CFR 50.109(a)(1), since there is no change in

requirements due to this restructuring.

The inclusion of thermal annealing does not impose any backfits as

defined in 10 CFR 50.109(a)(1), for the reasons set forth below in

``Thermal Annealing Rule (10 CFR 50.66).''

Thermal Annealing Rule (10 CFR 50.66)

The proposed thermal annealing rule would establish new

requirements with respect to applications for thermal annealing.

However, the Commission has determined that the proposed rule would not

impose a ``backfit'' as defined in 10 CFR 50.109(a)(1). The proposed

thermal annealing rule would 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 10 CFR Part

50, Appendix G. However, these rules provide an alternative means to

meet these screening limits, viz., 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. In addition, the proposed rule does not

reflect any new or different Staff position which conflicts with a

prior Staff position or Commission rule. Thus, the proposed rule would

have a purely prospective effect on future applications for thermal

annealing. The Commission has stated in other rulemakings establishing

prospective requirements, e.g., 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

when there are no prior NRC positions upon which the ``substantial

increase in overall protection'' can be measured. Accordingly, the

Commission concludes that the proposed rule does not impose backfits

and a backfit analysis need not be prepared for the proposed thermal

annealing rule.

10 CFR Part 50 Appendix G

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 of a

clarifying nature.

The explicit prohibition on core criticality before the completion

of pressure and leak tests can be construed as a backfit, although NRC

staff intent was never to permit such a procedure (letter from J. M.

Taylor, NRC, to N. S. Reynolds and D. F. Stenger, NUBARG, dated

February 2, 1990). 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). This amendment underscores the prior intent of the

Commission to prohibit the use of nuclear heat before the completion of

leak and pressure tests that is implicit in 10 CFR 50.55a and Section

XI of the ASME Code. The Commission's intent in this regard is

demonstrated by the fact that only a very small minority of licensees

actually used nuclear heat to conduct pressure and leak tests required

by the ASME Code.

10 CFR Part 50 Appendix H

The amendments to Appendix H of 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).

Criminal Penalties

For purposes of Section 223 of the Atomic Energy Act (AEA), the

Commission proposes to issue the proposed rule under one or more of

Sections 161b, 161i or 161o of the AEA. Willful violations of the rule

would be subject to criminal enforcement.

List of Subjects

10 CFR Part 50

Antitrust, Classified information, Criminal penalties, Fire

protection, Incorporation by reference, Intergovernmental relations,

Nuclear power plants and reactors, Radiation protection, Reactor siting

criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of

the Atomic Energy Act of 1954, as amended; the Energy Reorganization

Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing to

adopt the following amendments to 10 CFR Part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION

FACILITIES

1. The authority citation for Part 50 is revised 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. 1244, 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. Section 50.61 is revised to read as follows:

Sec. 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

Sec. 50.55a of this part.

(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.

(6) EOL Fluence means the best-estimate 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, the proposed expiration date if a change in the term

of the operating license has been requested, or the end of a renewal

term if an application for a renewed license under 10 CFR Part 54 has

been submitted.

(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 calculative 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\1\ change in projected values

of RTPTS, or upon a request for a change in the expiration date

for operation of the facility.

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

\1\Changes 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.

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

(2) The pressurized thermal shock (PTS) screening criterion is

270 deg.F for plates, forgings, and axial weld materials, and 300 deg.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 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-by-case 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 Sec. 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 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). RTPTS must be calculated for each vessel beltline

material using a fluence value, f, which is the EOL fluence for the

material.

(1) Equation 1 must be used to calculate values of RTNDT for

each weld and plate, or forging, in the reactor vessel beltline.

TP04OC94.000

(i) RTNDT(U) is the reference temperature, RTNDT, of the

material in the pre-service or unirradiated condition, evaluated

according to the procedures in the ASME Code, Paragraph NB-2331.

(A) If a measured value of RTNDT(U) is not available, a

generic mean value for the class\2\ of material may be used if there

are sufficient test results to establish a mean and a standard

deviation for the class.

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

\2\The 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.

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

(B) For weld metals, the following generic mean values must be

used, unless justification for different values is provided: 0 deg.F

for welds made with Linde 80 flux, and -56 deg.F for welds made with

Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.

(ii) 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.

TP04OC94.001

(A) In Equation 2, U is the standard deviation for

RTNDT(U). If a measured value of RTNDT(U) is used, then

U 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 U 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) for welds

is used, then U is 17 deg.F.

(B) In Equation 2, is the standard deviation

for RTNDT. The value of to be used

is 28 deg.F for welds and 17 deg.F for base metal; the value of

shall not exceed one-half of RTNDT.

(iii) RTNDT is the mean value of the transition

temperature shift, or change in RTNDT, due to irradiation, and

must be calculated using Equation 3.

TP04OC94.002

(A) CF ( deg.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\3\ may be used if justification is provided. If none of

these alternatives are available, 0.35% copper and 1.0% nickel must be

assumed.

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

\3\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.''

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

(B) f is the best estimate neutron fluence, in units of 10\19\ n/

cm\2\ (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 paragraph (c), the EOL fluence for the vessel

beltline material is used in calculating 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

embrittlement. This information includes but is not limited to the

reactor vessel operating temperature and surveillance program results.

(i) Results from the plant-specific surveillance program must be

integrated into the RTNDT estimate if the plant-specific

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 deg.F for welds and 17 deg.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 deg.F.

(E) The surveillance data for the correlation monitor material in

the capsule must fall within the scatter band of the data base for the

material.

(ii) Surveillance data deemed credible according to the criteria of

paragraph (c)(2)(i) 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 4.

TP04OC94.003

In Equation 4, ``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.

(iii) For cases in which the results from a credible plant-specific

surveillance program are used, the value of to be

used is 14 deg.F for welds and 8.5 deg.F for base metal; the value of

may not exceed one-half of RTNDT.

(iv) The use of results from the plant-specific 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) 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, deg.F

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

Nickel, Wt-%

Copper, 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

.02.......... 21 26 27 27 27 27 27

.03.......... 22 35 41 41 41 41 41

.04.......... 24 43 54 54 54 54 54

.05.......... 26 49 67 68 68 68 68

.06.......... 29 52 77 82 82 82 82

.07.......... 32 55 85 95 95 95 95

.08.......... 36 58 90 106 108 108 108

.09.......... 40 61 94 115 122 122 122

.10.......... 44 65 97 122 133 135 135

.11.......... 49 68 101 130 144 148 148

.12.......... 52 72 103 135 153 161 161

.13.......... 58 76 106 139 162 172 176

.14.......... 61 79 109 142 168 182 188

.15.......... 66 84 112 146 175 191 200

.16.......... 70 88 115 149 178 199 211

.17.......... 75 92 119 151 184 207 221

.18.......... 79 95 122 154 187 214 230

.19.......... 83 100 126 157 191 220 238

.20.......... 88 104 129 160 194 223 245

.21.......... 92 108 133 164 197 229 252

.22.......... 97 112 137 167 200 232 257

.23.......... 101 117 140 169 203 236 263

.24.......... 105 121 144 173 206 239 268

.25.......... 110 126 148 176 209 243 272

.26.......... 113 130 151 180 212 246 276

.27.......... 119 134 155 184 216 249 280

.28.......... 122 138 160 187 218 251 284

.29.......... 128 142 164 191 222 254 287

.30.......... 131 146 167 194 225 257 290

.31.......... 136 151 172 198 228 260 293

.32.......... 140 155 175 202 231 263 296

.33.......... 144 160 180 205 234 266 299

.34.......... 149 164 184 209 238 269 302

.35.......... 153 168 187 212 241 272 305

.36.......... 158 172 191 216 245 275 308

.37.......... 162 177 196 220 248 278 311

.38.......... 166 182 200 223 250 281 314

.39.......... 171 185 203 227 254 285 317

.40.......... 175 189 207 231 257 288 320

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

Table 2.--Chemistry Factor for Base Metals, deg.F

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

Nickel, Wt-%

Copper, 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

.02.......... 20 20 20 20 20 20 20

.03.......... 20 20 20 20 20 20 20

.04.......... 22 26 26 26 26 26 26

.05.......... 25 31 31 31 31 31 31

.06.......... 28 37 37 37 37 37 37

.07.......... 31 43 44 44 44 44 44

.08.......... 34 48 51 51 51 51 51

.09.......... 37 53 58 58 58 58 58

.10.......... 41 58 65 65 67 67 67

.11.......... 45 62 72 74 77 77 77

.12.......... 49 67 79 83 86 86 86

.13.......... 53 71 85 91 96 96 96

.14.......... 57 75 91 100 105 106 106

.15.......... 61 80 99 110 115 117 117

.16.......... 65 84 104 118 123 125 125

.17.......... 69 88 110 127 132 135 135

.18.......... 73 92 115 134 141 144 144

.19.......... 78 97 120 142 150 154 154

.20.......... 82 102 125 149 159 164 165

.21.......... 86 107 129 155 167 172 174

.22.......... 91 112 134 161 176 181 184

.23.......... 95 117 138 167 184 190 194

.24.......... 100 121 143 172 191 199 204

.25.......... 104 126 148 176 199 208 214

.26.......... 109 130 151 180 205 216 221

.27.......... 114 134 155 184 211 225 230

.28.......... 119 138 160 187 216 233 239

.29.......... 124 142 164 191 221 241 248

.30.......... 129 146 167 194 225 249 257

.31.......... 134 151 172 198 228 255 266

.32.......... 139 155 175 202 231 260 274

.33.......... 144 160 180 205 234 264 282

.34.......... 149 164 184 209 238 268 290

.35.......... 153 168 187 212 241 272 298

.36.......... 158 173 191 216 245 275 303

.37.......... 162 177 196 220 248 278 308

.38.......... 166 182 200 223 250 281 313

.39.......... 171 185 203 227 254 285 317

.40.......... 175 189 207 231 257 288 320

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

3. A new Sec. 50.66 is added to read as follows:

Sec. 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 treatment may be applied to the reactor

vessel to restore the fracture toughness to acceptable levels. The use

of a thermal annealing treatment is subject to the approval of the

Director, Office of Nuclear Reactor Regulation, and to the requirements

in this section. The application for the Director's approval must be

submitted in accordance with Sec. 50.4, and at least three years prior

to the proposed date of the annealing operation.

(b) Thermal Annealing Application. The content of the application

for approval by the Director, Office of Nuclear Reactor Regulation, for

thermal annealing of the reactor vessel must include: a thermal

annealing operating plan that includes an evaluation of the effects of

mechanical and thermal stresses and temperatures, an inspection and

test program to requalify the annealed reactor vessel, and a program

for demonstrating that the recovery of fracture toughness and the re-

embrittlement rate are adequate to permit subsequent safe operation of

the reactor vessel for the period specified in the application.

(1) Thermal Annealing Operating Plan.

(i) The thermal annealing operating plan must include:

(A) A detailed description of the pressure vessel and all

structures and components that will be affected by the thermal

annealing operation;

(B) The methods, instrumentation and procedures proposed for

performing the thermal annealing;

(C) A description of the heat source to be used; and

(D) The proposed thermal annealing operating parameters, including

temperatures, times, and heatup and cooldown schedules.

(ii) The annealing time and temperature parameters selected must be

based on projecting sufficient recovery of fracture toughness, using

the procedures of paragraph (d) of this section, to satisfy the

requirements of Sec. 50.60 and Sec. 50.61 for the proposed period of

operation addressed in the application. In addition, the operating plan

must describe any special precautions necessary to minimize

occupational exposure, in accordance with the As Low As Reasonably

Achievable (ALARA) principle and the provisions of Sec. 20.1206.

(iii) An evaluation of the effects of mechanical and thermal

stresses and temperatures on the vessel, attached piping and

appurtenances, and adjacent equipment and components must demonstrate

that operability of the reactor will not be detrimentally affected. A

detailed thermal and structural analysis must be performed 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. The effects of

localized high temperatures must be evaluated for degradation of the

concrete adjacent to the vessel and changes in thermal and mechanical

properties of the reactor vessel insulation. If the design temperature

limitations for the adjacent concrete structure are to be exceeded

during the annealing operation, an acceptable maximum temperature for

the concrete must be established for the annealing operation using

appropriate test data.

(iv) The time and temperature profile evaluated as part of the

annealing operating plan, and supported by the evaluation results of

paragraph (b)(1)(iii) of this section, represents the proposed

annealing conditions that may not be exceeded during the annealing

operation. If these conditions are exceeded, then the licensee cannot

certify that the annealing operation was performed in accordance with

the approved application, as required by paragraph (c)(1) of this

section, and must comply with paragraph (c)(2) of this section.

(v) The projected percent recovery of both RTNDT and Charpy

upper-shelf energy must be determined by the procedures described in

paragraph (d) of this section, using the proposed annealing time and

temperature described in the operating plan. The projected post-anneal

RTNDT and Charpy upper-shelf energy must be determined from the

projected percent recovery.

(vi) The projected rate of reembrittlement of RTNDT must be

calculated using the procedures in Sec. 50.61(c), or must be the same

rate as that used for the pre-anneal operating period. The projected

rate of reembrittlement for Charpy upper-shelf energy must be the same

rate as that used for the pre-anneal operating period.

(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 annealing conditions of

paragraph (b)(1)(iv) of this section have not been exceeded, and to

determine the annealing time and temperature to be used in quantifying

the fracture toughness recovery. In addition, the program must

demonstrate that the 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 Rate Assurance

Program. The percent recovery of RTNDT and Charpy upper-shelf

energy obtained by the thermal annealing treatment must be determined

from the time and temperature of the actual vessel annealing. 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 rate determined in accordance with

paragraph (b)(3)(ii) of this section.

(i) The recovery of RTNDT and Charpy upper-shelf energy must

be established using the procedures in paragraph (d) of this section,

using the time and temperature of the actual vessel annealing.

(A) 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 annealing parameters

used in the test program are equal to or bounded by those used in the

vessel annealing operation.

(B) If generic computational methods are used, appropriate

justification must be submitted as a part of the application.

(ii) The reembrittlement rate of both RTNDT and Charpy upper-

shelf energy must be estimated, and must be monitored using a

surveillance program which conforms to Appendix H of this part,

``Reactor Vessel Material Surveil- lance Program Requirements.''

(c) Certification of the Annealing Effectiveness.

(1) Upon completion of the anneal and prior to re-start of the

nuclear power plant, the licensee shall certify to the Director, Office

of Nuclear Reactor Regulation, that the thermal annealing was performed

in accordance with the approved application required by paragraph (a)

of this section, and meets the provisions of paragraph (b) of this

section. In this certification, the licensee shall establish the period

for which the reactor vessel will satisfy the requirements of

Sec. 50.60 and Sec. 50.61, and shall provide:

(i) The post-anneal RTNDT and Charpy upper-shelf energy values

of the reactor vessel materials for use in subsequent reactor

operation;

(ii) The projected reembrittlement trends for both RTNDT and

Charpy upper-shelf energy; and

(iii) The projected values of RTPTS and Charpy upper-shelf

energy at the end of the proposed period of operation addressed in the

application.

(2) If the licensee cannot certify that the thermal annealing was

performed in accordance with the approved application and the

provisions of paragraph (b) of this section, the licensee shall submit

a justification for subsequent operation for approval by the Director,

Office of Nuclear Reactor Regulation.

(d) Procedures for Determining the Recovery of Fracture Toughness.

The procedures of this paragraph must be used to determine the percent

recovery of NDT, 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 Sec. 50.61, percent

recovery due to 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 annealing (Rt

and Ru) may be determined from the results of a verification test

program employing materials removed from the beltline region of the

reactor vessel\1\ and that have been annealed under the same time and

temperature conditions as those given the beltline material.

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

\1\For 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.

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

(3) Generic computational methods may be used to determine recovery

if adequate justification is provided.

4. In 10 CFR Part 50, Appendix G is revised to read as follows:

Appendix G to Part 50--Fracture Toughness Requirements

Table of Contents

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 Sec. 50.55a, are applicable.

The sections, editions and addenda of the ASME Boiler and Pressure

Vessel Code specified in Sec. 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 St., New

York, NY 10017 and are available for inspection at the NRC Library,

11545 Rockville Pike, Two White Flint North, Rockville, Maryland 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 Sec. 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 low-alloy 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 Sec. 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 Sec. 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.

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

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 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 Sec. 50.55a(b)(2) at the time the

analysis is submitted.

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

\1\Defined in ASTM E 185-79 and -82 which are incorporated by

reference in Appendix H to Part 50.

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

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 Sec. 50.4, for

review and approval on an individual case basis at least three years

prior to the date when the predicted Charpy upper-shelf 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 1, 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 or

not the core is critical. In Table 1, 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 1 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 1 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 1, 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

Sec. 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

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

Vessel

Operating condition pressure(\1\) Requirements for pressure- Minimum temperature

temperature limits requirements

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

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

1.a Fuel in the vessel........ 20 ASME Appendix G Limits........... (2) + 60 deg.F.

%

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

1.c No fuel in the vessel ALL (Not Applicable)................. (3)+60 deg.F.

(Preservice Hydrotest Only).

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 deg.F (6).

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

% deg.F].

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

[(2)+160 deg.F].

2.e Core critical for BWR 20 ASME Appendix G Limits+40 deg.F. (2)+60 deg.F.

(\5\). %

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

(\1\)Percent of the preservice system hydrostatic test pressure.

(\2\)The highest reference temperature of the material in the closure flange region that is highly stressed by

the bolt preload.

(\3\)The highest reference temperature of the vessel.

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

(\5\)For 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.

5. In 10 CFR Part 50, Appendix H is revised to read as follows:

Appendix H to Part 50--Reactor Vessel Material Surveillance Program

Requirements

Table of Contents

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 this part.

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 St.,

Philadelphia, PA 19103 and are available for inspection at the NRC

Library, 11545 Rockville Pike, Two White Flint North, Rockville,

Maryland 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 10\17\ n/cm\2\ (E >1 MeV).

B. Reactor vessels that do not meet the conditions of paragraph

II.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 ASTM E 185-73 or the edition of

ASTM E 185 that is current on the issue date of the ASME Code to which

the reactor vessel was purchased, whichever is later. 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 Sec. 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

Sec. 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 26th day of September 1994.

For the Nuclear Regulatory Commission.

John C. Hoyle,

Acting Secretary of the Commission.

[FR Doc. 94-24209 Filed 10-3-94; 8:45 am]

BILLING CODE 7590-01-P