Use of Alternative Source Terms at Operating Reactors

Proposed Rules

Federal Register

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Proposed Rules

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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 21, 50, and 54

RIN 3150-AG12

Use of Alternative Source Terms at Operating Reactors

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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

its regulations to allow holders of operating licenses for nuclear

power plants to voluntarily replace the traditional source term used in

design basis accident analyses with alternative source terms. This

action would allow interested licensees to pursue cost beneficial

licensing actions to reduce unnecessary regulatory burden without

compromising the margin of safety of the facility. The NRC is also

proposing to amend its regulations to revise certain sections to

conform with the final rule published on December 11, 1996, concerning

reactor site criteria.

DATES: The comment period expires on May 25, 1999. Comments received

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

the NRC is able to assure consideration only for comments received on

or before this date.

ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, Mail Stop O16C1.

Deliver comments to: One White Flint North, 11555 Rockville Pike,

Rockville, Maryland, 20852, between 7:30 a.m. and 4:15 p.m. on Federal

workdays.

You may also submit comments via the NRC's interactive rulemaking

web site, ``Rulemaking Forum,'' through the NRC home page (http://

www.nrc.gov)

mmission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, Mail Stop O16C1.

Deliver comments to: One White Flint North, 11555 Rockville Pike,

Rockville, Maryland, 20852, between 7:30 a.m. and 4:15 p.m. on Federal

workdays.

You may also submit comments via the NRC's interactive rulemaking

web site, ``Rulemaking Forum,'' through the NRC home page (http://

www.nrc.gov). This site enables people to transmit comments as files

(in any format, but WordPerfect version 6.1 is preferred), if your web

browser supports that function. Information on the use of the

Rulemaking Forum is available on the website. For additional assistance

on the use of the interactive rulemaking site, contact Ms. Carol

Gallagher, telephone: 301-415-5905; or by Internet electronic mail to

[email protected].

Certain documents related to this rulemaking, including comments

received and the environmental assessment and finding of no significant

impact may be examined at the NRC Public Document Room, 2120 L Street,

NW. (Lower Level), Washington, DC. These same documents also may be

viewed and downloaded electronically via the interactive rulemaking

website established by NRC for this rulemaking.

FOR FURTHER INFORMATION CONTACT: Mr. Stephen F. LaVie, Office of

Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001; telephone: (301) 415-1081; or by Internet

electronic mail to [email protected].

SUPPLEMENTARY INFORMATION:

I. Background

II. Objectives

III. Alternatives

IV. Section-by-Section Analysis

V. Future Regulatory Action

VI. Referenced Documents

VII. Draft Finding of No Significant Environmental Impact;

Availability

VIII. Paperwork Reduction Act Statement

IX. Regulatory Analysis

X. Regulatory Flexibility Certification

XI. Backfit Analysis

I. Background

mail to [email protected].

SUPPLEMENTARY INFORMATION:

I. Background

II. Objectives

III. Alternatives

IV. Section-by-Section Analysis

V. Future Regulatory Action

VI. Referenced Documents

VII. Draft Finding of No Significant Environmental Impact;

Availability

VIII. Paperwork Reduction Act Statement

IX. Regulatory Analysis

X. Regulatory Flexibility Certification

XI. Backfit Analysis

I. Background

A holder of an operating license (i.e., the licensee) for a light-

water power reactor is required by regulations issued by the NRC (or

its predecessor, the U.S. Atomic Energy Commission, (AEC)) to submit a

safety analysis report that contains assessments of the radiological

consequences of potential accidents and an evaluation of the proposed

facility site. The NRC uses this information in its evaluation of the

suitability of the reactor design and the proposed site as required by

its regulations contained in 10 CFR Parts 50 and 100. Section 100.11,

which was adopted by the AEC in 1962 (27 FR 3509; April 12, 1962),

requires an applicant to assume (1) a fission product release from the

reactor core, (2) the expected containment leak rate, and (3) the site

meteorological conditions to establish an exclusion area and a low

population zone. This fission product release is based on a major

accident that would result in substantial release of appreciable

quantities of fission products from the core to the containment

atmosphere. A note to Sec. 100.11 states that Technical Information

Document (TID) 14844, ``Calculation of Distance Factors for Power and

Test Reactors,'' may be used as a source of guidance in developing the

exclusion area, the low population zone, and the population center

distance.

The fission product release from the reactor core into containment

is referred to as the ``source term'' and it is characterized by the

composition and magnitude of the radioactive material, the chemical and

physical properties of the material, and the timing of the release from

the reactor core

e in developing the

exclusion area, the low population zone, and the population center

distance.

The fission product release from the reactor core into containment

is referred to as the ``source term'' and it is characterized by the

composition and magnitude of the radioactive material, the chemical and

physical properties of the material, and the timing of the release from

the reactor core. The accident source term is used to evaluate the

radiological consequences of design basis accidents (DBAs) in showing

compliance with various requirements of the NRC's regulations. Although

originally used for site suitability analyses, the accident source term

is a design parameter for accident mitigation features, equipment

qualification, control room operator radiation doses, and post-accident

vital area access doses. The measurement range and alarm setpoints of

some installed plant instrumentation and the actuation of some plant

safety features are based in part on the accident source term. The TID-

14844 source term was explicitly stated as a required design parameter

for several Three Mile Island (TMI)-related requirements.

The NRC's methods for calculating accident doses, as described in

Regulatory Guide 1.3, ``Assumptions Used for Evaluating the Potential

Radiological Consequences of a Loss of Coolant Accident for Boiling

Water Reactors''; Regulatory Guide 1.4, ``Assumptions Used for

Evaluating the Potential Radiological Consequences of a Loss of Coolant

Accident for Pressurized Water Reactors''; and NUREG-0800, ``Standard

Review Plan for the Review of Safety Analysis Reports for Nuclear Power

Plants,'' were developed to be consistent with the TID-14844 source

term and the whole body and thyroid dose guidelines stated in

Sec. 100.11. In this regulatory framework, the source term is assumed

to be released immediately to the containment at the start of the

postulated accident. The chemical form

-0800, ``Standard

Review Plan for the Review of Safety Analysis Reports for Nuclear Power

Plants,'' were developed to be consistent with the TID-14844 source

term and the whole body and thyroid dose guidelines stated in

Sec. 100.11. In this regulatory framework, the source term is assumed

to be released immediately to the containment at the start of the

postulated accident. The chemical form

of the radioiodine released to the containment atmosphere is assumed to

be predominantly elemental, with the remainder being small fractions of

particulate and organic iodine forms. Radiation doses are calculated at

the exclusion area boundary (EAB) for the first 2-hours and at the low

population zone (LPZ) for the assumed 30-day duration of the accident.

The whole body dose comes primarily from the noble gases in the source

term. The thyroid dose is based on inhalation of radioiodines. In

analyses performed to date, the thyroid dose has generally been

limiting. The design of some engineered safety features, such as

containment spray systems and the charcoal filters in the containment,

the building exhaust, and the control room ventilation systems, are

predicated on these postulated thyroid doses. Subsequently, the NRC

adopted the whole body and thyroid dose criteria in Criterion 19 of 10

CFR Part 50, Appendix A (36 FR 3255; February 20, 1971).

The source term in TID-14844 is representative of a major accident

involving significant core damage and is typically postulated to occur

in conjunction with a large loss-of-coolant accident (LOCA). Although

the LOCA is typically the maximum credible accident, NRC experience in

reviewing license applications has indicated the need to consider other

accident sequences of lesser consequence but higher probability of

occurrence. Some of these additional accident analyses may involve

source terms that are a fraction of those specified in TID-14844

ith a large loss-of-coolant accident (LOCA). Although

the LOCA is typically the maximum credible accident, NRC experience in

reviewing license applications has indicated the need to consider other

accident sequences of lesser consequence but higher probability of

occurrence. Some of these additional accident analyses may involve

source terms that are a fraction of those specified in TID-14844. The

DBAs were not intended to be actual event sequences, but rather, were

intended to be surrogates to enable deterministic evaluation of the

response of the plant engineered safety features. These accident

analyses are intentionally conservative in order to address known

uncertainties in accident progression, fission product transport, and

atmospheric dispersion. Although probabilistic risk assessments (PRAs)

can provide useful insights into system performance and suggest changes

in how the desired defense in depth is achieved, defense in depth

continues to be an effective way to account for uncertainties in

equipment and human performance. The NRC's policy statement on the use

of PRA methods (60 FR 42622; August 16, 1995) calls for the use of PRA

technology in all regulatory matters in a manner that complements the

NRC's deterministic approach and supports the traditional defense-in-

depth philosophy.

Since the publication of TID-14844, significant advances have been

made in understanding the timing, magnitude, and chemical form of

fission product releases from severe nuclear power plant accidents.

Many of these insights developed out of the major research efforts

started by the NRC and the nuclear industry after the accident at Three

Mile Island (TMI). In 1995, the NRC published NUREG-1465, ``Accident

Source Terms for Light-Water Nuclear Power Plants,'' which utilized

this research to provide more physically based estimates of the

accident source term that could be applied to the design of future

light-water power reactors

the major research efforts

started by the NRC and the nuclear industry after the accident at Three

Mile Island (TMI). In 1995, the NRC published NUREG-1465, ``Accident

Source Terms for Light-Water Nuclear Power Plants,'' which utilized

this research to provide more physically based estimates of the

accident source term that could be applied to the design of future

light-water power reactors. The NRC sponsored significant review

efforts by peer reviewers, foreign research partners, industry groups,

and the general public (request for public comment was published in 57

FR 33374).

The information in NUREG-1465 presents a representative accident

source term (``revised source term'') for a boiling-water reactor (BWR)

and for a pressurized-water reactor (PWR). These revised source terms

are described in terms of radionuclide composition and magnitude,

physical and chemical form, and timing of release. Where TID-14844

addressed three categories of radionuclides, the revised source terms

categorize the accident release into eight groups on the basis of

similarity in chemical behavior. Where TID-14844 assumed an immediate

release of the activity, the revised source terms have five release

phases that are postulated to occur over several hours, with the onset

of major core damage occurring after 30 minutes. Where TID-14844

assumed radioiodine to be predominantly elemental, the revised source

terms assume radioiodine to be predominantly cesium iodide (CsI), an

aerosol that is more amenable to mitigation mechanisms.

For DBAs, the NUREG-1465 source terms are comparable to the TID-

14844 source term with regard to the magnitude of the noble gas and

radioiodine release fractions. However, the revised source terms offer

a more representative description of the radionuclide composition and

release timing. The NRC has determined (SECY-94-302, dated December

1994) that design basis analyses will address the first three release

phases--coolant, gap, and in-vessel

to the TID-

14844 source term with regard to the magnitude of the noble gas and

radioiodine release fractions. However, the revised source terms offer

a more representative description of the radionuclide composition and

release timing. The NRC has determined (SECY-94-302, dated December

1994) that design basis analyses will address the first three release

phases--coolant, gap, and in-vessel. The ex-vessel and late in-vessel

phases are considered to be unduly conservative for design basis

analysis purposes. These latter releases could only result from core

damage accidents with vessel failure and core-concrete interactions.

The estimated frequencies of such scenarios are low enough that they

need not be considered for the purpose of meeting the requirements of

Sec. 100.11 or, as proposed herein, Sec. 50.67.

The objective of NUREG-1465 was to define revised accident source

terms for regulatory application for future light water reactors. The

NRC's intent was to capture the major relevant insights available from

severe accident research to provide, for regulatory purposes, a more

realistic portrayal of the amount of the postulated accident source

term. These source terms were derived from examining a set of severe

accident sequences for light water reactors (LWRs) of current design.

Because of general similarities in plant and core design parameters,

these results are considered to be applicable to evolutionary and

passive LWR designs. The revised source term has been used in

evaluating the Westinghouse AP-600 standard design certification

application

terms were derived from examining a set of severe

accident sequences for light water reactors (LWRs) of current design.

Because of general similarities in plant and core design parameters,

these results are considered to be applicable to evolutionary and

passive LWR designs. The revised source term has been used in

evaluating the Westinghouse AP-600 standard design certification

application. (A draft version of NUREG-1465 was used in evaluating

Combustion Engineering's (CE's) System 80+ design.)

The NRC considered the applicability of the revised source terms to

operating reactors and determined that the current analytical approach

based on the TID-14844 source term would continue to be adequate to

protect public health and safety, and that operating reactors licensed

under this approach would not be required to reanalyze accidents using

the revised source terms. The NRC also concluded that some licensees

may wish to use an alternative source term in analyses to support

operational flexibility and cost-beneficial licensing actions. The NRC

initiated several actions to provide a regulatory basis for operating

reactors to voluntarily amend their facility design bases to enable use

of the revised source term in design basis analyses. First, the NRC

solicited ideas on how an alternative source term might be implemented.

In November 1995, the Nuclear Energy Institute (NEI) submitted its

generic framework, Electric Power Research Institute Technical Report

TR-105909, ``Generic Framework for Application of Revised Accident

Source Term to Operating Plants.'' This report and the NRC response

were discussed in SECY-96-242 (November 1996). Second, the NRC

initiated a comprehensive assessment of the overall impact of

substituting the NUREG-1465 source terms for the traditionally used

TID-14844 source term at three typical facilities. This was done to

evaluate the issues involved with applying the revised source terms at

operating plants

ng Plants.'' This report and the NRC response

were discussed in SECY-96-242 (November 1996). Second, the NRC

initiated a comprehensive assessment of the overall impact of

substituting the NUREG-1465 source terms for the traditionally used

TID-14844 source term at three typical facilities. This was done to

evaluate the issues involved with applying the revised source terms at

operating plants. SECY 98-154 (June 1998) described the conclusions of

this assessment. Third, the NRC accepted license amendment requests

related to implementation of the revised source

terms at a small number of pilot plants. Experience has demonstrated

that evaluation of a limited number of plant-specific submittals

improves regulation and regulatory guidance development. The review of

these pilot projects is currently in progress. Insights from these

pilot plant reviews will be incorporated into the regulatory guidance

that will be developed in conjunction with this rulemaking. Fourth, the

NRC initiated an assessment on whether rulemaking would be necessary to

allow operating reactors to use an alternative source term. The

proposed rule and the supporting regulatory guidance that will be

developed as part of this rulemaking have resulted from this

assessment. The NRC plans to issue the supporting regulatory guidance

for public comment on the same day as it publishes the final rule.

This proposed rulemaking for use of alternative source terms is

applicable only to those facilities for which a construction permit was

issued before January 10, 1997, under 10 CFR Part 50, ``Domestic

Licensing of Production and Utilization Facilities.'' The regulations

of this part are supplemented by those in other parts of Chapter I of

Title 10, including Part 100, ``Reactor Site Criteria.'' Part 100

contains language that qualitatively defines a required accident source

term and contains a note that discusses the availability of TID-14844.

With the exception of Sec

Part 50, ``Domestic

Licensing of Production and Utilization Facilities.'' The regulations

of this part are supplemented by those in other parts of Chapter I of

Title 10, including Part 100, ``Reactor Site Criteria.'' Part 100

contains language that qualitatively defines a required accident source

term and contains a note that discusses the availability of TID-14844.

With the exception of Sec. 50.34(f), there are no explicit requirements

in Chapter I of Title 10 to use the TID-14844 accident source term.

Section 50.34(f), which addresses additional TMI-related requirements,

is only applicable to a limited number of construction permit

applications pending on February 16, 1982, and to applications under

Part 52.

An applicant for an operating license is required by Sec. 50.34(b)

to submit a final safety analysis report (FSAR) that describes the

facility and its design bases and limits, and presents a safety

analysis of the structures, systems, and components of the facility as

a whole. Guidance in performing these analyses is given in regulatory

guides. In its review of the more recent applications for operating

licenses, the NRC has used the review procedures in NUREG-0800,

``Standard Review Plan for the Review of Safety Analysis Reports for

Nuclear Power Plants'' (SRP). These review procedures reference or

provide acceptable assumptions and analysis methods. The facility FSAR

documents the assumptions and methods actually used by the applicant in

the required safety analyses. The NRC's finding that a license may be

issued is based on the review of the FSAR, as documented in the

Commission's safety evaluation report (SER). By their inclusion in the

FSAR, the assumptions (including the source term) become part of the

design basis \1\ of the facility. From a regulatory standpoint, the

requirement to use the TID-14844 source term is expressed as a licensee

commitment (typically to Regulatory Guide 1.3 or 1.4) documented in the

facility FSAR, and is subject to the requirements of Sec

safety evaluation report (SER). By their inclusion in the

FSAR, the assumptions (including the source term) become part of the

design basis \1\ of the facility. From a regulatory standpoint, the

requirement to use the TID-14844 source term is expressed as a licensee

commitment (typically to Regulatory Guide 1.3 or 1.4) documented in the

facility FSAR, and is subject to the requirements of Sec. 50.59.

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\1\ As defined in 10 CFR Part 50.2, design bases means that

information which identifies the specific functions to be performed

by a structure, system, or component of a facility, and the specific

values or ranges of values chosen for controlling parameters as

reference bounds for design. These values may be (1) restraints

derived from generally accepted ``state of the art'' practices for

achieving functional goals, or (2) requirements derived from

analysis (based on calculation and/or experiments) of the effects of

a postulated accident for which a structure, system, or component

must meet its functional goals. The NRC considers the accident

source term to be an integral part of the design basis because it

sets forth specific values (or range of values) for controlling

parameters that constitute reference bounds for design.

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In January 1997 (61 FR 65157), the NRC amended its regulations in

10 CFR Parts 21, 50, 52, 54, and 100. That regulatory action produced

site criteria for future sites; presented a stable regulatory basis for

seismic and geologic siting and the engineering design of future

nuclear power plants to withstand seismic events; and relocated source

term and dose requirements for future plants into part 50. Because

these dose requirements tend to affect reactor design rather than

siting, they are more appropriately located in Part 50

site criteria for future sites; presented a stable regulatory basis for

seismic and geologic siting and the engineering design of future

nuclear power plants to withstand seismic events; and relocated source

term and dose requirements for future plants into part 50. Because

these dose requirements tend to affect reactor design rather than

siting, they are more appropriately located in Part 50. This decoupling

of siting from design is consistent with the future licensing of

facilities using standardized plan designs, the design features of

which will be certified in a separate design certification rulemaking.

This decoupling of siting from design was directed by Congress in the

1980 Authorization Act for the NRC. Because the revised criteria would

not apply to operating reactors, the non-seismic and seismic reactor

site criteria for operating reactors were retained as Subpart A and

Appendix A to Part 100, respectively. The revised reactor site criteria

were added as Subpart B in Part 100, and revised source term and dose

requirements were moved to Sec. 50.34. The existing source term and

dose requirements of Subpart A of Part 100 will remain in place as the

licensing bases for those operating reactors that do not elect to use

an alternative source term.

In relocating the source term and dose requirements for future

reactors to Sec. 50.34, the NRC retained the requirements for the

exclusion area and the low population zone, but revised the associated

numerical dose criteria to replace the two different doses for the

whole body and the thyroid gland with a single, total effective dose

equivalent (TEDE) value. The dose criteria for the whole body and the

thyroid, and the immediate 2-hour exposure period were largely

predicated by the assumed source term being predominantly noble gases

and radioiodines instantaneously released to the containment and the

assumed ``single critical organ'' method of modeling the internal dose

used at the time that Part 100 was originally published

ivalent (TEDE) value. The dose criteria for the whole body and the

thyroid, and the immediate 2-hour exposure period were largely

predicated by the assumed source term being predominantly noble gases

and radioiodines instantaneously released to the containment and the

assumed ``single critical organ'' method of modeling the internal dose

used at the time that Part 100 was originally published. However, the

current dose criteria, by focusing on doses to the thyroid and the

whole body, assume that the major contributor to doses will be

radioiodine. Although this may be appropriate with the TID-14844 source

term, as implemented by Regulatory Guides 1.3 and 1.4, it may not be

true for a source term based on a more complete understanding of

accident sequences and phenomenology.

The postulated chemical and physical form of radioiodine in the

revised source terms is more amenable to mitigation and, as such,

radioiodine may not always be the predominant radionuclide in an

accident release. The revised source terms include a larger number of

radionuclides than did the TID-14844 source term as implemented in

regulatory guidance. The whole body and thyroid dose criteria ignore

these contributors to dose. The NRC amended its radiation protection

standards in Part 20 in 1991 (56 FR 23391; May 21, 1991) replacing the

single, critical organ concept for assessing internal exposure with the

TEDE concept that assesses the impact of all relevant nuclides upon all

body organs. TEDE is defined to be the deep dose equivalent (for

external exposure) plus the committed effective dose equivalent (for

internal exposure)

d its radiation protection

standards in Part 20 in 1991 (56 FR 23391; May 21, 1991) replacing the

single, critical organ concept for assessing internal exposure with the

TEDE concept that assesses the impact of all relevant nuclides upon all

body organs. TEDE is defined to be the deep dose equivalent (for

external exposure) plus the committed effective dose equivalent (for

internal exposure). The deep dose equivalent (DDE) is comparable to the

present whole body dose; the committed effective dose equivalent (CEDE)

is the sum of the products of doses (integrated over a 50-year period)

to selected body organs resulting from the intake of radioactive

material multiplied by weighting factors for each organ that are

representative of the radiation risk associated with the particular

organ.

The TEDE, using a risk-consistent methodology, assesses the impact

of all relevant nuclides upon all body organs. Although it is expected

that in many cases the thyroid could still be the limiting organ and

radioiodine the limiting radionuclide, this conclusion cannot be

assured in all potential cases. The revised source terms postulate that

the core inventory is released in a

sequence of phases over 10 hours, with the more significant release

commencing at about 30 minutes from the start of the event. The

assumption that the 2-hour exposure period starts immediately at the

onset of the release is inconsistent with the phased release postulated

in the revised source terms. The proposed rule would extend the future

LWR dose criteria to operating reactors that elect to use an

alternative source term.

An accidental release of radioactivity can result in radiation

exposure to control room operators. Normal ventilation systems may draw

this activity into the control room where it can result in external and

internal exposures. Control room designs differ but, in general, design

features are provided to detect the accident or the activity and

isolate the normal ventilation intake

source term.

An accidental release of radioactivity can result in radiation

exposure to control room operators. Normal ventilation systems may draw

this activity into the control room where it can result in external and

internal exposures. Control room designs differ but, in general, design

features are provided to detect the accident or the activity and

isolate the normal ventilation intake. Emergency ventilation systems

are activated to minimize infiltration of contaminated air and to

remove activity that has entered the control room. Personnel exposures

can also result from radioactivity outside of the control room.

However, because of concrete shielding of the control room, these

latter exposures are generally not limiting. The objective of the

control room design is to provide a location from which actions can be

taken to operate the plant under normal conditions and to maintain it

in a safe condition under accident conditions. General Design Criterion

19 (GDC-19), ``Control Room,'' of Appendix A to 10 CFR part 50 (36 FR

3255; February 20, 1971), establishes minimum requirements for the

design of the control room, including a requirement for radiation

protection features adequate to permit access to and occupancy of the

control room under accident conditions. The GDC-19 criteria were

established for judging the acceptability of the control room design

for protecting control room operators under postulated design basis

accidents, a significant concern being the potential increases in

offsite doses that might result from the inability of control room

personnel to adequately respond to the event.

The GDC-19 criteria are expressed in terms of whole body dose, or

its equivalent to any organ. The NRC did not revise the criteria when

Part 20 was amended (56 FR 23391) instead deferring such action to

individual facility licensing actions (NUREG/CR-6204). This position

was taken in the interest of maintaining the licensing basis for those

facilities already licensed

y respond to the event.

The GDC-19 criteria are expressed in terms of whole body dose, or

its equivalent to any organ. The NRC did not revise the criteria when

Part 20 was amended (56 FR 23391) instead deferring such action to

individual facility licensing actions (NUREG/CR-6204). This position

was taken in the interest of maintaining the licensing basis for those

facilities already licensed. The NRC is proposing to replace the

current GDC-19 dose criteria for future reactors and for operating

reactors that elect to use an alternative source term with a criterion

expressed in terms of TEDE. The rationale for this revision is similar

to the rationale, discussed earlier in this preamble, for revising the

dose criteria for offsite exposures.

On January 10, 1997 (61 FR 65157), the NRC amended 10 CFR Parts 21,

50, 52, 54, and 100 of its regulations to update the criteria used in

decisions regarding power reactor siting for future nuclear power

plants. The NRC intended that future licensing applications in

accordance with Part 52 utilize a source term consistent with the

source term information in NUREG-1465 and the accident TEDE criteria in

Parts 50 and 100. However, during the final design approval (FDA) and

design certification proceeding for the Westinghouse AP-600 advanced

light-water reactor design, the NRC staff and Westinghouse determined

that exemptions were necessary from Secs. 50.34(f)(2)(vii), (viii),

52 utilize a source term consistent with the

source term information in NUREG-1465 and the accident TEDE criteria in

Parts 50 and 100. However, during the final design approval (FDA) and

design certification proceeding for the Westinghouse AP-600 advanced

light-water reactor design, the NRC staff and Westinghouse determined

that exemptions were necessary from Secs. 50.34(f)(2)(vii), (viii),

(xxvi), and (xxviii) and 10 CFR Part 50, Appendix A, GDC-19. This rule

would eliminate the need for these exemptions for future applicants

under Part 52 by making conforming changes to Part 50, Appendix A, GDC-

19 and Sec. 50.34.

II. Objectives

The objectives of this proposed regulatory action are to--

1. Provide a regulatory framework for the voluntary implementation

of alternative source terms as a change to the design basis at

currently licensed power reactors, thereby enabling potential cost-

beneficial licensing actions while continuing to maintain existing

safety margins and defense in depth.

2. Retain the existing regulatory framework for currently licensed

power reactor licensees who choose not to implement an alternative

source term, but continue to comply with their existing source term.

3. Relocate source term and dose requirements that apply primarily

to plant design into 10 CFR Part 50 for operating reactors that choose

to implement an alternative source term, and

4. Implement conforming changes to Sec. 50.34(f) and Part 50,

Appendix A, GDC-19 to eliminate the need for exemptions for future

applicants under Part 52.

III. Alternatives

mply with their existing source term.

3. Relocate source term and dose requirements that apply primarily

to plant design into 10 CFR Part 50 for operating reactors that choose

to implement an alternative source term, and

4. Implement conforming changes to Sec. 50.34(f) and Part 50,

Appendix A, GDC-19 to eliminate the need for exemptions for future

applicants under Part 52.

III. Alternatives

The first alternative considered by the NRC was to continue using

current regulations for accident dose criteria and control room dose

criteria. This is not considered to be an acceptable alternative. As

discussed in the statements of consideration for the final siting rule

(61 FR 65157, 65159; December 11, 1996), the NRC determined that dose

criteria expressed in terms of whole body and thyroid doses were

inconsistent with the use of new source terms not based upon TID-14844.

With regard to the exclusion area dose guideline, the NRC had

previously determined (id. at 65160) that the dose criterion applies to

the 2-hour period resulting in the maximum dose.

The second alternative considered by the NRC was the replacement of

the existing guidelines in Sec. 100.11 and the existing criteria in 10

CFR Part 50 Appendix A, GDC-19 with revised dose criteria. This is not

considered to be a desirable alternative because the provisions of the

existing regulations form part of the licensing bases for many of the

operating reactors. Therefore, these provisions must remain in effect

for operating reactors that do not implement an alternative source

term. In addition, this alternative would also be inconsistent with the

NRC's philosophy of separating plant siting criteria and dose

requirements.

The approach of establishing the requirements for use of

alternative source terms in a new section to Part 50 while retaining

the existing regulations in Part 100 Subpart A and Part 50 Appendix A

GDC-19 was chosen as the best alternative

source

term. In addition, this alternative would also be inconsistent with the

NRC's philosophy of separating plant siting criteria and dose

requirements.

The approach of establishing the requirements for use of

alternative source terms in a new section to Part 50 while retaining

the existing regulations in Part 100 Subpart A and Part 50 Appendix A

GDC-19 was chosen as the best alternative.

The NRC considered alternatives with regard to providing regulatory

guidance to support the new section to Part 50. The first option was to

issue no additional regulatory guidance. This option was not considered

to be acceptable because in the absence of clear regulatory guidance,

licensee efforts in preparing applications and the NRC staff review of

submitted applications, could be hindered by differences in

interpretations and technical positions. This could result in the

inefficient use of licensee and NRC staff resources, could cause

licensing delays, and lead to less uniform and less consistent

regulatory implementation.

The second option was to replace the existing regulatory guides

that address the radiological consequences of accidents with new

revisions. This is not considered to be an acceptable choice because

the provisions of the existing regulatory guides form part of the

licensing bases for many of the operating reactors. Therefore, these

provisions must remain in effect for those operating reactors that do

not implement an alternative source term. The third option was to issue

a new regulatory guide on the implementation

of alternative source terms that would include revised assumptions and

acceptable analysis methods for each design basis accident in a series

of appendices. The approach of issuing a new regulatory guide was

determined to be the best option. To provide review guidance for the

NRC staff, a new section on design basis radiological analyses using

alternative source terms would be added to the Standard Review Plan.

IV. Section-by-Section Analysis

A. Section 50.2

d

acceptable analysis methods for each design basis accident in a series

of appendices. The approach of issuing a new regulatory guide was

determined to be the best option. To provide review guidance for the

NRC staff, a new section on design basis radiological analyses using

alternative source terms would be added to the Standard Review Plan.

IV. Section-by-Section Analysis

A. Section 50.2

The general ``definitions'' section for Part 50 would be

supplemented by adding a definition of source term for the purpose of

Sec. 50.67. In NUREG-1465, the source term is defined by five projected

characteristics: (1) Magnitude of radioactivity release, (2)

radionuclides released, (3) physical form of the radionuclides

released, (4) chemical form of the radionuclides released, and (5)

timing of the radioactivity release. Although all five characteristics

should be addressed in applications proposing the use of an alternative

source term, there may be technically justifiable applications in which

all five characteristics need not be addressed. The NRC intends to

allow licensees flexibility in implementing alternative source terms

consistent with maintaining a conservative, clear, logical, and

consistent plant design basis. The regulatory guide that supports this

proposed rule will contain guidance on an acceptable basis for defining

the characteristics of an alternative source term.

B. Section 50.67(a)

stics need not be addressed. The NRC intends to

allow licensees flexibility in implementing alternative source terms

consistent with maintaining a conservative, clear, logical, and

consistent plant design basis. The regulatory guide that supports this

proposed rule will contain guidance on an acceptable basis for defining

the characteristics of an alternative source term.

B. Section 50.67(a)

This paragraph would define the licensees that may seek to revise

their current radiological source term with an alternative source term.

The proposed rule is applicable only to holders of nuclear power plant

operating licenses that were issued under 10 CFR Part 50 before January

10, 1997. The proposed rule would not require licensees to revise their

current source term. The NRC considered the acceptability of the TID-

14844 source term at current operating reactors and determined that the

analytical approach based on the TID-14844 source term would continue

to be adequate to protect public health and safety, and that operating

reactors licensed under this approach should not be required to

reanalyze design basis accidents using a new source term. The proposed

rule does not explicitly define an alternative source term. In lieu of

an explicit reference to NUREG-1465, Footnote 1 to the proposed rule

identifies the significant characteristics of an accident source term.

The regulatory guide that will be issued to support this proposed rule

will identify the NUREG-1465 source terms as acceptable alternatives to

the source term in TID-14844, and will provide implementation guidance.

This approach would provide for future revised source terms if they are

developed and would allow licensees to propose additional alternatives

for NRC consideration.

C. Section 50.67(b)(1)

atory guide that will be issued to support this proposed rule

will identify the NUREG-1465 source terms as acceptable alternatives to

the source term in TID-14844, and will provide implementation guidance.

This approach would provide for future revised source terms if they are

developed and would allow licensees to propose additional alternatives

for NRC consideration.

C. Section 50.67(b)(1)

This paragraph of Sec. 50.67 would state the information that a

licensee must submit as part of a license amendment application to use

an alternative source term. Because of the extensive use of the

accident source term in the design and operation of a power reactor and

the potential impact on postulated accident consequences and margins of

safety of a change of such a fundamental design assumption, the NRC has

determined that any change to the design basis to use an alternative

source term should be reviewed and approved by the NRC in the form of a

license amendment. Changes to the source term, by itself, would

ordinarily constitute a no significant hazards consideration. In

addition, generic analyses performed by the NRC staff in support of

this proposed rule have indicated that there are potential changes to

the facility as documented in the FSAR which would constitute a no

significant hazards consideration. However, such determinations would

have to be made for each proposed change based upon facility-specific

evaluations. The procedural requirements for processing a license

amendment are given in Secs. 50.90 through 50.92.

The NRC's regulations provide a regulatory mechanism for a licensee

to effect a change in its design basis in Sec. 50.59. That section

allows a licensee to make changes to the facility as described in the

final safety evaluation report (FSAR) without prior NRC approval,

unless the proposed change is deemed to involve an unreviewed safety

question (USQ), or involves a change to the technical specifications

incorporated into the facility license

for a licensee

to effect a change in its design basis in Sec. 50.59. That section

allows a licensee to make changes to the facility as described in the

final safety evaluation report (FSAR) without prior NRC approval,

unless the proposed change is deemed to involve an unreviewed safety

question (USQ), or involves a change to the technical specifications

incorporated into the facility license. If a USQ is determined to exist

or if a change to the technical specifications is involved, the

licensee must request NRC approval of the change using the license

amendment process detailed in Sec. 50.90. The criteria for determining

that a USQ is involved appear in Sec. 50.59. Significant to this

proposed rule is the criterion that a USQ would exist if the proposed

change resulted in an increase in consequences of an accident or

malfunction. In many applications, alternative source terms may reduce

the postulated consequences of the accident or malfunction. For this

reason, the NRC determined that the regulatory framework of Sec. 50.59

does not provide assurance that this change in the design basis would

be recognized by the licensee as needing review by the NRC staff. After

a licensee has been authorized to substitute an alternative source term

in its design basis, subsequent changes to the facility that involve an

alternative source term may be processed under Sec. 50.59 or

Sec. 50.90, as appropriate. However, a subsequent change to the source

term itself could not be implemented under Sec. 50.59; in all cases a

change to the source term must be made through a license amendment.

The proposed rule would require the applicant to perform analyses

of the consequences of applicable design basis accidents previously

analyzed in the safety analysis report and to submit a description of

the analysis inputs, assumptions, methodology, and results of these

analyses for NRC review. Applicable evaluations may include, but are

not limited to, those previously performed to show compliance with

Sec

would require the applicant to perform analyses

of the consequences of applicable design basis accidents previously

analyzed in the safety analysis report and to submit a description of

the analysis inputs, assumptions, methodology, and results of these

analyses for NRC review. Applicable evaluations may include, but are

not limited to, those previously performed to show compliance with

Sec. 100.11, Sec. 50.49, Part 50 Appendix A GDC-19, Sec. 50.34(f), and

NUREG-0737 requirements II.B.2, II.B.3, III.D.3.4. The regulatory guide

that supports this proposed rule will provide guidance on the scope and

extent of analyses used to show compliance with this rule and on the

assumptions and methods used therein. It is not the NRC's intent that

all of the design basis radiological analyses for a facility be

performed again as a prerequisite for approval of the use of an

alternative source term. The NRC does expect that the applicant will

perform sufficient evaluations, supported by calculations as warranted,

to demonstrate the acceptability of the proposed amendment.

D. Sections 50.67(b)(2)(i), (ii), (iii)

These subparagraphs would contain the three criteria for NRC

approval of the license amendment to use an alternative source term. A

detailed rationale for the use of 0.25 Sv (25 rem) TEDE as an accident

dose criterion and the use of the 2-hour exposure period resulting in

the maximum dose for future LWRs is provided at 61 FR 65157; December

11, 1996. The same considerations that formed the basis for that

rationale are similarly applicable to operating reactors that elect to

use an alternative source term. The NRC believes that it is technically

appropriate and logical to extend the philosophy of decoupling of

design and siting, and the dose criteria established

he maximum dose for future LWRs is provided at 61 FR 65157; December

11, 1996. The same considerations that formed the basis for that

rationale are similarly applicable to operating reactors that elect to

use an alternative source term. The NRC believes that it is technically

appropriate and logical to extend the philosophy of decoupling of

design and siting, and the dose criteria established

for future LWRs to operating reactors that elect to use an alternative

source term.

The NRC is proposing to replace the current GDC-19 dose criteria

for operating reactors that elect to use an alternative source term

with a criterion of 0.05 Sv (5 rem) TEDE for the duration of the

accident. This criterion would be included in Sec. 50.67 rather than

GDC-19 in order to co-locate all of the dose requirements associated

with alternative source terms. The bases for the NRC's decision are:

first, that the criteria in GDC-19 and that in the proposed rule are

based on a primary occupational exposure limit. Second, the language in

GDC-19: ``5 rem whole body, or its equivalent to any part of the body''

is subsumed by the definition of TEDE in Sec. 20.1003 and by the 0.05

Sv (5 rem) TEDE annual limit in Sec. 20.1201(a). Although the weighting

factors stated in Sec. 20.1003 for use in determining TEDE differ in

magnitude from the weighting factors implied in the 0.3 Sv (30 rem)

thyroid criteria used for showing compliance with GDC-19, these

differences are the result of improvement in the science of assessing

internal exposures and do not represent a reduction in the level of

protection. Third, as discussed earlier, the use of TEDE in conjunction

with alternative source terms has been deemed appropriate and

necessary. Fourth, the use of TEDE for the control room dose criterion

is consistent with the use of TEDE in the accident dose criteria for

offsite exposure

ovement in the science of assessing

internal exposures and do not represent a reduction in the level of

protection. Third, as discussed earlier, the use of TEDE in conjunction

with alternative source terms has been deemed appropriate and

necessary. Fourth, the use of TEDE for the control room dose criterion

is consistent with the use of TEDE in the accident dose criteria for

offsite exposure.

The NRC is not including a ``capping'' limitation, an additional

requirement that the dose to any individual organ not be in excess of

some fraction of the total as provided for routine occupational

exposures. The bases for the NRC's decision are: first, that this non-

inclusion of a ``capping'' limitation is consistent with the final rule

published in December 11, 1996 (61 FR 65157), with regard to doses to

persons offsite. Second, the use of 0.05 Sv (5 rem) TEDE as the control

room criterion does not imply that this would be an acceptable exposure

during emergency conditions, or that other radiation protection

standards of Part 20, including individual organ dose limits, might not

apply. This criterion is provided only to assess the acceptability of

design provisions for protecting control room operators under

postulated DBA conditions. The DBA conditions assumed in these

analyses, although credible, generally do not represent actual accident

sequences but are specified as conservative surrogates to create

bounding conditions for assessing the acceptability of engineered

safety features. Third, Sec. 20.1206 permits a once-in-a-lifetime

planned special dose of five times the annual dose limits. Also,

Environmental Protection Agency (EPA) guidance sets a limit of five

times the annual dose limits for workers performing emergency services

such as lifesaving or protection of large populations. Considering the

individual organ weighting factors of Sec. 20.1003 and assuming that

only the exposure from a single organ contributed to TEDE, the organ

dose, although exceeding the dose specified in Sec

ironmental Protection Agency (EPA) guidance sets a limit of five

times the annual dose limits for workers performing emergency services

such as lifesaving or protection of large populations. Considering the

individual organ weighting factors of Sec. 20.1003 and assuming that

only the exposure from a single organ contributed to TEDE, the organ

dose, although exceeding the dose specified in Sec. 20.1201(a), would

be less than that considered acceptable as a planned special dose or as

an emergency worker dose. The NRC is not suggesting that control room

dose during an accident can be treated as a planned special exposure or

that the EPA emergency worker dose limits are an alternative to GDC-19

or the proposed rule. However, the NRC does believe that these

provisions offer a useful perspective that supports the conclusion that

the organ doses implied by the proposed 0.05 Sv (5 rem) criterion can

be considered to be acceptable due to the relatively low probability of

the events that could result in doses of this magnitude.

Although the dose criteria in the proposed rule would supersede the

dose criteria in GDC-19, the other provisions of GDC-19 remain

applicable.

E. 10 CFR Part 50, Appendix A, GDC-19

GDC-19 would be changed to include the TEDE dose criterion for

control room design for applicants for construction permits, design

certifications, and combined operating licenses that submitted

applications after January 10, 1997 (the effective date of the 1996

rulemaking adopting the TEDE criterion), and for those licenses using

an alternative source term under Sec. 50.67. The proposed change to

GDC-19 addresses the use of alternative source terms at operating

reactors and a deficiency identified in the regulatory framework for

early site permits, standard design certifications, and combined

licenses under part 52

7 (the effective date of the 1996

rulemaking adopting the TEDE criterion), and for those licenses using

an alternative source term under Sec. 50.67. The proposed change to

GDC-19 addresses the use of alternative source terms at operating

reactors and a deficiency identified in the regulatory framework for

early site permits, standard design certifications, and combined

licenses under part 52. Sections 52.18, 52.48, and 52.81 establish that

applications filed under part 52, Subparts A, B, and C, respectively,

will be reviewed according to the standards given in 10 CFR parts 20,

50, 51, 55, 73, and 100 to the extent that those standards are

technically relevant to the proposed design. Therefore, GDC-19 is

pertinent to applications under part 52. The final rule that became

effective on January 10, 1997 (61 FR 65157; December 11, 1996),

established accident TEDE criteria (in Sec. 50.34) for applicants under

part 52 but did not change the existing control room whole body (or

equivalent) dose criterion in GDC-19. Thus, exemptions from the dose

criteria in the current GDC-19 were necessary in the design

certification process for the Westinghouse AP-600 advanced LWR in order

to use the 0.05 Sv (5 rem) TEDE criterion deemed necessary for use with

alternative source terms. Exemptions would arguably be necessary for

future applicants for construction permits, design certifications, and

combined operating licenses. This proposed change would eliminate the

need for these exemptions.

F. Sections 21.3, 50.2, 50.49(b)(1)(i)(C), 50.65(b)(1), and

54.4(a)(1)(iii)

These sections would be revised to conform with the relocation of

accident dose criteria from Sec. 100.11 to Sec. 50.67 for operating

reactors that have amended their design bases to use an alternative

source term.

G. Section 50.34

ating licenses. This proposed change would eliminate the

need for these exemptions.

F. Sections 21.3, 50.2, 50.49(b)(1)(i)(C), 50.65(b)(1), and

54.4(a)(1)(iii)

These sections would be revised to conform with the relocation of

accident dose criteria from Sec. 100.11 to Sec. 50.67 for operating

reactors that have amended their design bases to use an alternative

source term.

G. Section 50.34

A new footnote to Sec. 50.34 would be added to define what

constitutes an accident source term. This new footnote is identical to

the existing footnote 1 to Sec. 100.11, and is being added to provide

for consistency between Parts 50 and 100.

H. Sections 50.34(f)(2)(vii), (viii), (xxvi) and (xxviii)

These paragraphs would be revised to replace an explicit reference

to the ``TID-14844 source term'' with a more general reference to

``accident source term.'' These changes potentially affect two classes

of applicants. The first affected class is facilities that obtain

combined licenses under part 52. Section 52.47(a)(ii) states that

applications for combined licenses must contain, inter alia,

``demonstration of compliance with any technically-relevant portions of

the Three Mile Island requirements set forth in Sec. 50.34(f).''

Section 50.34(f) contains several references to the TID-14844 source

term. These references would be modified to delete the reference to

TID-14844. This would make it clear that applicants for combined

licenses would not use the TID-14844 source term but would use the

source term in the referenced design certification, or a source term

that is justified in the combined license application.

The second affected class is the small subset of plants that had

construction permits pending on February 16, 1982. With the proposed

change, these plants could use either the TID-14844 source term or an

alternative source term in their operating license applications.

V. Future Regulatory Action

ced design certification, or a source term

that is justified in the combined license application.

The second affected class is the small subset of plants that had

construction permits pending on February 16, 1982. With the proposed

change, these plants could use either the TID-14844 source term or an

alternative source term in their operating license applications.

V. Future Regulatory Action

The NRC is developing the following regulatory guides and Standard

Review Plan sections to provide prospective applicants with the

necessary guidance for implementing the proposed regulation. The draft

guide and draft Standard Review Plan section will be issued to coincide

with the publication of the final regulations that would implement this

proposed rulemaking. A notice of availability for these materials will

be published in the Federal Register at a future date.

1. Draft Guide DG-1081, ``Alternative Radiological Source Terms for

Evaluating the Radiological Consequences of Design Basis Accidents at

Boiling and Pressurized Water Reactors''

This guide is expected to present regulatory guidance on the

implementation of an alternative source term at an operating reactor.

The guide is expected to address issues involving limited or selective

implementation of an alternative source term and probabilistic risk

assessment (PRA) issues related to plant modifications based on an

alternative source term, and to provide guidance on the scope and

extent of affected DBA radiological analyses and associated acceptance

criteria. The guide is expected to include revised assumptions and

methods for each affected DBA in a series of appendices. These

appendices will supersede the guidance in Regulatory Guides 1.3, 1.4,

1.25, and 1.77, and will supplement guidance in Regulatory Guide 1.89

for those facilities using an alternative source term.

2. Standard Review Plan Section, 15.0.1, ``Radiological Consequence

Analyses Using Alternative Source Terms''

de revised assumptions and

methods for each affected DBA in a series of appendices. These

appendices will supersede the guidance in Regulatory Guides 1.3, 1.4,

1.25, and 1.77, and will supplement guidance in Regulatory Guide 1.89

for those facilities using an alternative source term.

2. Standard Review Plan Section, 15.0.1, ``Radiological Consequence

Analyses Using Alternative Source Terms''

This SRP section presents guidance to NRC staff in the review of

the adequacy of licensee submittals requesting approval for use of an

alternative source term.

VI. Referenced Documents

Copies of NUREG-0737, NUREG-0800, NUREG-1465, and NUREG/CR-6204 may

be purchased from the Superintendent of Documents, U.S. Government

Printing Office, Mail Stop SSOP, Washington, DC 20402-9328. Copies also

are available from the National Technical Information Service, 5285

Port Royal Road, Springfield, VA 22161. A copy also is available for

inspection and copying for a fee in the NRC Public Document Room, 2120

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

Copies of issued regulatory guides may be purchased from the

Government Printing Office (GPO) at the current GPO price. Information

on current GPO prices may be obtained by contacting the Superintendent

of Documents, U.S. Government Printing Office, P.O. Box 37082,

Washington, DC 20402-9328. Issued guides also may be purchased from the

National Technical Information Service (NTIS) on a standing order

basis. Details on this service may be obtained by writing NTIS, 5826

Port Royal Road, Springfield, VA 22161.

Copies of SECY-94-302, SECY-96-242, SECY-98-154, TID14844, and TR-

105909 are available for inspection and copying for a fee at the NRC

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

VII. Draft Finding of No Significant Environmental Impact:

Availability

order

basis. Details on this service may be obtained by writing NTIS, 5826

Port Royal Road, Springfield, VA 22161.

Copies of SECY-94-302, SECY-96-242, SECY-98-154, TID14844, and TR-

105909 are available for inspection and copying for a fee at the NRC

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

VII. Draft Finding of No Significant Environmental Impact:

Availability

The NRC has determined under the National Environmental Policy Act

of 1969, as amended, and the NRC's regulations in Subpart A of 10 CFR

Part 51, that this regulation is not a major Federal action

significantly affecting the quality of the human environment and,

therefore, an environmental impact statement is not required. This

proposed rule would allow operating reactors to replace the traditional

TID-14844 source term with a more realistic source term based on the

insights gained from extensive accident research activities. The actual

accident sequence and progression would not be changed; it is the

regulatory assumptions regarding the accident that would be affected by

the change. The use of an alternative source term alone cannot increase

the core damage frequency (CDF) or the large early release frequency

(LERF) or actual offsite or onsite radiation doses. An alternative

source term could be used to justify changes in the plant design that

might have an impact on CDF or LERF or that might increase offsite or

onsite doses. These potential changes are subject to existing

requirements in the NRC's regulations. Thus, the level of protection of

public health and safety provided in NRC regulations would not be

decreased by this proposed rule. The proposed rule would not affect

non-radiological plant effluents and would have no significant

environmental impact.

As discussed above, the determination of the environmental

assessment is that there would be no significant offsite impact on the

public from this action. However, the general public should note that

the NRC welcomes public participation

be

decreased by this proposed rule. The proposed rule would not affect

non-radiological plant effluents and would have no significant

environmental impact.

As discussed above, the determination of the environmental

assessment is that there would be no significant offsite impact on the

public from this action. However, the general public should note that

the NRC welcomes public participation. Also, the NRC has committed

itself to complying in all its actions with Executive Order (E.O.)

12898, ``Federal Actions to Address Environmental Justice in Minority

Populations and Low-Income Populations,'' dated February 11, 1994. In

accordance with that Executive Order, the NRC has determined that there

are no disproportionately high and adverse impacts on minority and low

income parties. In the letter and spirit of E.O. 12898, the NRC is

requesting public comments on any environmental justice considerations

or questions that the public thinks may be related to this proposed

rule, but that somehow were not addressed. The NRC uses the following

working definition of environmental justice: Environmental justice

means the fair treatment and meaningful involvement of all people,

regardless of race, ethnicity, culture, income, or educational level

with respect to the development, implementation and enforcement of

environmental laws, regulations, and policies. Comments on any aspect

of the environmental assessment, including environmental justice, may

be submitted to the NRC as indicated under the ADDRESSES heading.

The draft environmental assessment and the draft finding of no

significant impact on which this determination is based are available

for inspection at the NRC Public Document Room, 2120 L Street NW (Lower

Level), Washington, DC. Single copies of the environmental assessment

and finding of no significant impact are available from Mr. Stephen F.

LaVie, Office of Nuclear Reactor Regulation, U.S

The draft environmental assessment and the draft finding of no

significant impact on which this determination is based are available

for inspection at the NRC Public Document Room, 2120 L Street NW (Lower

Level), Washington, DC. Single copies of the environmental assessment

and finding of no significant impact are available from Mr. Stephen F.

LaVie, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory

NRC, Washington, DC 20555-0001, telephone: 301-415-1081, or by Internet

electronic mail to [email protected].

VIII. Paperwork Reduction Act Statement

This proposed rule increases the burden on licensees by requiring

that when seeking to revise their current accident source term in

design basis radiological consequence analyses, they apply for an

amendment under Sec. 50.90. The public burden for this information

collection is estimated to average 609 hours per request. Because the

burden for this information collection is insignificant, Office of

Management and Budget (OMB) clearance is not required. Existing

requirements were approved by the Office of Management and Budget,

approval number 3150-0011.

Public Protection Notification

If an information collection does not display a currently valid OMB

control number, the NRC may not conduct or sponsor, and a person is not

required to respond to, the information collection.

IX. Regulatory Analysis

The Commission has prepared a regulatory analysis on this

regulation. Interested persons may examine a copy of the regulatory

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

Level), Washington, DC. Single copies of the analysis are available

from Mr. Stephen F. LaVie, Office of Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone:

301-415-1081, or by Internet electronic mail to [email protected].

X. Regulatory Flexibility Certification

he regulatory

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

Level), Washington, DC. Single copies of the analysis are available

from Mr. Stephen F. LaVie, Office of Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone:

301-415-1081, or by Internet electronic mail to [email protected].

X. Regulatory Flexibility Certification

As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.

605(b), the Commission certifies that this regulation will not have a

significant economic impact on a substantial number of small entities.

This proposed regulation will affect only the licensing and operation

of nuclear power plants. The companies that own these plants do not

fall within the definition of ``small entities'' found in the

Regulatory Flexibility Act or within the size standards established by

the NRC (April 11, 1995; 60 FR 18344).

XI. Backfit Analysis

The NRC has determined that the backfit rule in 10 CFR 50.109, does

not apply to this proposed regulation and that a backfit analysis is

not required for this proposed regulation because these amendments do

not involve any provisions that would impose backfits as defined in 10

CFR 50.109(a)(1). This proposed regulation amends the NRC's regulations

by establishing alternate requirements that may be voluntarily adopted

by licensees.

List of Subjects

10 CFR Part 21

Nuclear power plants and reactors, Penalties, Radiation protection,

Reporting and recordkeeping requirements.

10 CFR Part 50

Antitrust, Classified information, Criminal penalties, Fire

protection, Intergovernmental relations, Nuclear power plants and

reactors, Radiation protection, Reactor siting criteria, Reporting and

recordkeeping requirements.

10 CFR Part 54

Administrative practice and procedure, Age-related degradation,

Backfitting, Classified information, Criminal penalties, Environmental

protection, Nuclear power plants and reactors, Reporting and

recordkeeping requirements.

Intergovernmental relations, Nuclear power plants and

reactors, Radiation protection, Reactor siting criteria, Reporting and

recordkeeping requirements.

10 CFR Part 54

Administrative practice and procedure, Age-related degradation,

Backfitting, Classified information, Criminal penalties, Environmental

protection, Nuclear power plants and reactors, Reporting and

recordkeeping requirements.

For the reasons noted 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 the

following amendments to 10 CFR Parts 21, 50, and 54:

PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE

1. The authority citation for part 21 continues to read as follows:

Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83

Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C.

2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as

amended, 1246 (42 U.S.C. 5841, 5846).

Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425,

96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).

2. Section 21.3 is amended by republishing the introductory text

and revising paragraph (1)(i)(C) of the definition of Basic component

to read as follows:

Sec. 21.3 Definitions.

As used in this part:

Basic component. (1)(i) * * *

(C) The capability to prevent or mitigate the consequences of

accidents which could result in potential offsite exposures comparable

to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or

Sec. 100.11 of this chapter, as applicable.

* * * * *

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION

FACILITIES

3. The authority citation for part 50 continues to read as follows:

* * *

(C) The capability to prevent or mitigate the consequences of

accidents which could result in potential offsite exposures comparable

to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or

Sec. 100.11 of this chapter, as applicable.

* * * * *

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION

FACILITIES

3. The authority citation for part 50 continues to read as follows:

Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68

Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,

83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,

2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88

Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L. 95-9601, 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. 91-9190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13,

50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as

amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56

also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections

50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.

91-9190, 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. 97-9415, 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).

4. Section 50.2 is amended by republishing the introductory text,

by revising paragraph (1)(iii) of the definition of Basic component and

by adding in alphabetical order the definition for Source term to read

as follows:

Sec. 50.2 Definitions.

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).

4. Section 50.2 is amended by republishing the introductory text,

by revising paragraph (1)(iii) of the definition of Basic component and

by adding in alphabetical order the definition for Source term to read

as follows:

Sec. 50.2 Definitions.

As used in this part,

* * * * *

Basic component * * *

(1) * * *

(iii) The capability to prevent or mitigate the consequences of

accidents which could result in potential offsite exposures comparable

to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or

Sec. 100.11 of this chapter, as applicable.

* * * * *

Source term refers to the magnitude and mix of radionuclides

released from the reactor core to the reactor containment, their

physical and chemical form, and the timing of their release.

* * * * *

5. Section 50.34 is amended by revising paragraphs (f)(2)(vii),

(viii), (xxvi), and (xxviii) to read as follows:

Sec. 50.34 Contents of applications; technical information.

* * * * *

(f) * * *

(2) * * *

(vii) Perform radiation and shielding design reviews of spaces

around systems that may, as a result of an accident, contain accident

source term 11 radioactive materials, and design as

necessary to permit adequate access to important areas and to protect

safety equipment from the radiation environment. (II.B.2)

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* * *

(vii) Perform radiation and shielding design reviews of spaces

around systems that may, as a result of an accident, contain accident

source term 11 radioactive materials, and design as

necessary to permit adequate access to important areas and to protect

safety equipment from the radiation environment. (II.B.2)

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\11\ The fission product release assumed for these calculations

should be based upon a major accident, hypothesized for purposes of

site analysis or postulated from considerations of possible

accidental events, that would result in potential hazards not

exceeded by those from any accident considered credible. Such

accidents have generally been assumed to result in substantial

meltdown of the core with subsequent release of appreciable

quantities of fission products.

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

(viii) Provide a capability to promptly obtain and analyze samples

from the reactor coolant system and containment that may contain

accident source term 12 radioactive materials without

radiation exposures to any individual exceeding 5 rems to the whole

body or 50 rems to

the extremities. Materials to be analyzed and quantified include

certain radionuclides that are indicators of the degree of core damage

(e.g., noble gases, radioiodines and cesiums, and nonvolatile

isotopes), hydrogen in the containment atmosphere, dissolved gases,

chloride, and boron concentrations. (II.B.3)

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

\12\ See footnote 11 to paragraph (f)(2)(vii) of this section.

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

* * * * *

., noble gases, radioiodines and cesiums, and nonvolatile

isotopes), hydrogen in the containment atmosphere, dissolved gases,

chloride, and boron concentrations. (II.B.3)

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

\12\ See footnote 11 to paragraph (f)(2)(vii) of this section.

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

* * * * *

(xxvi) Provide for leakage control and detection in the design of

systems outside containment that contain (or might contain) accident

source term 13 radioactive materials following an accident.

Applicants shall submit a leakage control program, including an initial

test program, a schedule for re-testing these systems, and the actions

to be taken for minimizing leakage from such systems. The goal is to

minimize potential exposures to workers and public, and to provide

reasonable assurance that excessive leakage will not prevent the use of

systems needed in an emergency. (III.D.1.1)

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\13\ See footnote 11 to paragraph (f)(2)(vii) of this section.

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

* * * * *

(xxviii) Evaluate potential pathways for radioactivity and

radiation that may lead to control room habitability problems under

accident conditions resulting in an accident source term 14

release, and make necessary design provisions to preclude such

problems. (III.D.3.4)

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

\14\ See footnote 11 to paragraph (f)(2)(vii) of this section.

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

6. Section 50.49 is amended by revising paragraph (b)(1)(i)(C) to

read as follows:

Sec. 50.49 Environmental qualification of electric equipment important

to safety for nuclear power plants.

* * * * *

(b) * * *

(1) * * *

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

\14\ See footnote 11 to paragraph (f)(2)(vii) of this section.

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

6. Section 50.49 is amended by revising paragraph (b)(1)(i)(C) to

read as follows:

Sec. 50.49 Environmental qualification of electric equipment important

to safety for nuclear power plants.

* * * * *

(b) * * *

(1) * * *

(i) * * *

(C) The capability to prevent or mitigate the consequences of

accidents that could result in potential offsite exposures comparable

to the guidelines in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or Sec. 100.11

of this chapter, as applicable.

* * * * *

7. Section 50.65 is amended by revising paragraph (b)(1) to read as

follows:

Sec. 50.65 Requirements for monitoring the effectiveness of

maintenance at nuclear power plants.

* * * * *

(b) * * *

(1) Safety-related structures, systems and components that are

relied upon to remain functional during and following design basis

events to ensure the integrity of the reactor coolant pressure

boundary, the capability to shut down the reactor and maintain it in a

safe shutdown condition, or the capability to prevent or mitigate the

consequences of accidents that could result in potential offsite

exposure comparable to the guidelines in Sec. 50.34(a)(1),

Sec. 50.67(b)(2), or Sec. 100.11 of this chapter, as applicable.

* * * * *

8. Part 50 is amended by adding Sec. 50.67 to read as follows:

Sec. 50.67 Accident source term.

(a) Applicability. The requirements of this section apply to all

holders of operating licenses issued prior to January 10, 1997, who

seek to revise the current accident source term used in their design

basis radiological analyses.

those from any accident considered credible. Such

accidents have generally been assumed to result in substantial

meltdown of the core with subsequent release of appreciable

quantities of fission products.

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(2) The NRC may issue the amendment only if the applicant's

analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the

exclusion area for any 2-hour period following the onset of the

postulated fission product release, would not receive a radiation dose

in excess of 0.25 Sv (25 rem) 2 total effective dose

equivalent (TEDE).

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

\2\ The use of 0.25 Sv (25 rem) TEDE is not intended to imply

that this value constitutes an acceptable limit for emergency doses

to the public under accident conditions. Rather, this 0.25 Sv (25

rem) TEDE value has been stated in this section as a reference

value, which can be used in the evaluation of proposed design basis

changes with respect to potential reactor accidents of exceedingly

low probability of occurrence and low risk of public exposure to

radiation.

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

(ii) An individual located at any point on the outer boundary of

the low population zone, who is exposed to the radioactive cloud

resulting from the postulated fission product release (during the

entire period of its passage), would not receive a radiation dose in

excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

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

(ii) An individual located at any point on the outer boundary of

the low population zone, who is exposed to the radioactive cloud

resulting from the postulated fission product release (during the

entire period of its passage), would not receive a radiation dose in

excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to

and occupancy of the control room under accident conditions without

personnel receiving radiation exposures in excess of 0.05 Sv (5 rem)

total effective dose equivalent (TEDE) for the duration of the

accident.

9. Part 50, Appendix A, II., General Design Criterion 19, is

revised to read as follows:

Appendix A to Part 50--General Design

Criteria for Nuclear Power Plants

* * * * *

II. * * *

Criterion 19--Control room. A control room shall be provided

from which actions can be taken to operate the nuclear power unit

safely under normal conditions and to maintain it in a safe

condition under accident conditions, including loss-of-coolant

accidents. Adequate radiation protection shall be provided to permit

access and occupancy of the control room under accident conditions

without personnel receiving radiation exposures in excess of 5 rem

whole body, or its equivalent to any part of the body, for the

duration of the accident.

Equipment at appropriate locations outside the control room

shall be provided (1) with a design capability for prompt hot

shutdown of the reactor, including necessary instrumentation and

controls to maintain the unit in a safe condition during hot

shutdown, and (2) with a potential capability for subsequent cold

shutdown of the reactor through the use of suitable procedures

of the accident.

Equipment at appropriate locations outside the control room

shall be provided (1) with a design capability for prompt hot

shutdown of the reactor, including necessary instrumentation and

controls to maintain the unit in a safe condition during hot

shutdown, and (2) with a potential capability for subsequent cold

shutdown of the reactor through the use of suitable procedures.

Applicants for construction permits under this part or a design

certification or combined license under part 52 of this chapter who

apply on or after January 10, 1997, or holders of operating licenses

using an alternative source term under Sec. 50.67, shall meet the

requirements of this criterion, except that with regard to control

room access and occupancy, adequate radiation protection shall be

provided to ensure that radiation exposures shall not exceed 0.05 Sv

(5 rem) total effective dose equivalent (TEDE) as defined in

Sec. 50.2 for the duration of the accident.

* * * * *

PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR

POWER PLANTS

10. The authority citation for part 54 continues to read as

follows:

Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68

Stat. 936, 937, 938, 948, 953, 954, 955, 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, 202, 206, 88 Stat. 1242,

1244, as amended (42 U.S.C. 5841, 5842), E.O. 12829, 3 CFR, 1993

Comp., p. 570; E.O. 12958, as amended, 3 CFR, 1995 Comp., p. 333;

E.O. 12968, 3 CFR, 1995 Comp., p. 391.

11. Section 54.4 is amended by revising paragraph (a)(1)(iii) to

read as follows:

Sec. 54.4 Scope.

1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,

2232, 2233, 2236, 2239, 2282); secs 201, 202, 206, 88 Stat. 1242,

1244, as amended (42 U.S.C. 5841, 5842), E.O. 12829, 3 CFR, 1993

Comp., p. 570; E.O. 12958, as amended, 3 CFR, 1995 Comp., p. 333;

E.O. 12968, 3 CFR, 1995 Comp., p. 391.

11. Section 54.4 is amended by revising paragraph (a)(1)(iii) to

read as follows:

Sec. 54.4 Scope.

(a) * * *

(1) * * *

(iii) The capability to prevent or mitigate the consequences of

accidents which could result in potential offsite exposures comparable

to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or

Sec. 100.11 of this chapter, as applicable.

* * * * *

Dated at Rockville, Maryland, this 5th day of March 1999.

For the Nuclear Regulatory Commission.

Annette Vietti-Cook,

Secretary of the Commission.

[FR Doc. 99-6058 Filed 3-10-99; 8:45 am]

BILLING CODE 7590-01-U