Use of Alternative Source Terms at Operating Reactors

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: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending 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 will 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

announcing the availability of a draft regulatory guide and a draft

Standard Review Plan section on this subject for public comment. The

NRC is also amending its regulations to revise certain sections to

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

reactor site criteria.

EFFECTIVE DATE: January 24, 2000.

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. Analysis of Public Comments

III. Section-by-Section Analysis

IV. Draft Regulatory Guide; Issuance, Availability

V. Draft Standard Review Plan Section; Issuance, Availability

VI. Referenced Documents

VII. Finding of No Significant Environmental Impact; Availability

VIII. Paperwork Reduction Act Statement

IX. Regulatory Analysis

X. Regulatory Flexibility Act Certification

XI. Backfit Analysis

XII. Small Business Regulatory Enforcement Fairness Act

XIII. National Technology Transfer and Advancement Act

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 (or, for early reactors, a hazard summary

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

[[Page 71991]]

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. Changes to the design of the facility and

the procedures for operating the facility are evaluated in part by

determining whether there are changes to the calculated fission product

release.

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 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. 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

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 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; July 28, 1992).

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 (up to and including the

early in-vessel phase) 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, December 19, 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 inappropriate for design basis analysis purposes.

These latter releases could only result from core damage accidents with

vessel failure and core-concrete interactions.

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

terms for regulatory application for future light water reactors

(LWRs). 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 LWRs of current

[[Page 71992]]

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 AP600 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 concluded that some licensees may

wish to use an alternative source term in analyses to support

operational flexibility and cost-beneficial licensing actions and that

some of these applications could provide concomitant improvements in

overall safety and in reduced occupational exposure. 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 25, 1996). Second, the NRC initiated

an 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 30, 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 have been

incorporated into the regulatory guidance that was 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. This final rule and the

supporting regulatory guidance have resulted from this assessment.

This final rulemaking for use of alternative source terms is

applicable to holders of operating licenses issued prior to January 10,

1997, under 10 CFR Part 50, ``Domestic Licensing of Production and

Utilization Facilities,'' and to holders of renewed licenses under 10

CFR Part 54, ``Requirements for Renewal of Operating Licenses for

Nuclear Power Plants,'' whose initial operating license was issued

prior to January 10, 1997. The regulations of Part 50 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). Fundamental assumptions

that are design inputs, including the source term, were required to be

included in the FSAR and became 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 Sec. 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 1996 (61 FR 65175; December 11, 1996), 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.

This decoupling of siting from design is consistent with the future

licensing of facilities using standardized plant designs, the design

features of which have been or will be certified in a separate design

certification rulemakings. 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

[[Page 71993]]

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). 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 final rule

adopts the future LWR dose criteria for 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. 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; May 21, 1991) instead deferring such

action to individual facility licensing actions (NUREG/CR-6204,

``Questions and Answers Based on the Revised 10 CFR Part 20''). This

position was taken in the interest of maintaining the licensing basis

for those facilities already licensed. The NRC is replacing the current

dose criteria of GDC-19 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 AP600 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 final

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. Analysis of Public Comments

The NRC published a proposed rule in the Federal Register (64 FR

12117, March 31, 1999); that would provide a regulatory framework for

the voluntary implementation of alternative source terms as a change to

the design basis at currently licensed power reactors, while retaining

the existing regulatory framework for currently licensed power reactor

licensees who choose not to implement an alternative source term. The

rule proposed relocating 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. The rule also

proposed 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.

The NRC received seven letters commenting on the proposed rule. All

comments including those received by the NRC after the expiration of

the public comment period but before June 25, 1999, were considered.

The commenters included two State regulatory agencies, two nuclear

industry groups and three utilities. The State of Florida Department of

Community Affairs indicated that they had no comments on the proposed

rule. The State of New Jersey Department of Environmental Protection

concurred with the NRC's position on the use of an AST in emergency

preparedness applications and stated a desire to review the draft

regulatory guidance when issued. Winston & Strawn

[[Page 71994]]

submitted comments on behalf of the Nuclear Utility Backfitting and

Reform Group (NUBARG). The Nuclear Energy Institute (NEI) submitted

comments on behalf of the nuclear industry. Two of the utilities

provided comments, while the third endorsed the comments submitted by

NEI. Copies of these letters are available for public inspection and

copying for a fee at the NRC Public Document Room, 2120 L Street NW.

(Lower Level), Washington, DC.

1. NUBARG Comments

NUBARG supports the rule, noting that the rule as proposed defines

an acceptable regulatory process for implementing more realistic

accident source terms. NUBARG requested clarification in the final rule

of situations in which an alternative source term (AST) may be applied

in future backfitting 2 decisions. First, NUBARG suggests

that the NRC clarify the extent it intends to use the revised source

term in assessing whether new generic requirements provide a cost-

justified, substantial increase in safety in accordance with NRC's

backfitting rule, Sec. 50.109. NUBARG believes that continued use of

the source term in TID-14844 for this purpose in spite of its known

limitations would be inappropriate and could lead to overly

conservative estimates of the safety impact of proposed new

requirements. Second, NUBARG suggests a similar clarification for

plant-specific backfit decisions for plants that have not opted to

implement the revised source term. NUBARG believes that the NRC has

discretion to take all relevant factors into account in making its

safety benefit assessment of the proposed backfit, including the

current state of knowledge concerning the accident source term. NUBARG

suggested that the statements of considerations accompanying the final

rule address these issues. NUBARG also suggests that relevant NRC

guidance should also be revised to reflect NRC policy in these areas.

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\2\ As provided in Sec. 50.109, Backfitting is defined as the

modification of or addition to systems, structures, components, or

the design of a facility; or the design approval or manufacturing

license for a facility; or the procedures or organization required

to design, construct or operate a facility; any of which may result

from a new or amended provision in the Commission rules or the

imposition of a regulatory staff position interpreting the

Commission rules that is either new or different from a previously

applicable staff position.

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NRC Response. When radiological consequence analyses are involved,

the NRC expects to use a technically appropriate AST in evaluating

generic and plant-specific backfitting analyses, including those

proposed for facilities that have not implemented an AST. The NRC

agrees with the NUBARG position that the NRC has discretion to take all

new information on accident source terms into account. The NRC's

guidance for evaluating proposed NRC regulatory actions (including

backfitting) are contained in NUREG/BR-0058, ``Regulatory Analysis

Guidelines of the U.S. Nuclear Regulatory Commission,'' and NUREG/BR-

0184, ``Regulatory Analysis Technical Evaluation Handbook.'' These

documents state that value and impact (including adverse effects on

health and safety) parameters are to be best estimates, preferably mean

or expected values. These documents also provide that analyses are to

be based largely on risk considerations.

2. NEI Comment 1

NEI stated that the Section-by-Section Analysis in the proposed

rule notice is consistent with the NRC's intent to permit limited

application of the new research results. NEI noted that these limited

applications are of two types: (1) application of alternative source

term radiological composition and magnitude in a quantitative analysis

relative to the effect on the performance of a given engineered safety

feature; or (2) application of only the timing aspects in conjunction

with the original TID-14844 source term. NEI stated that proposed

Sec. 50.67 appears to apply to applications where a licensee would use

a completely new source term such as NUREG-1465 in all aspects of the

plant design. The NEI comment acknowledged that further guidance in a

subsequent regulatory guide and standard review plan is helpful and

necessary. Nonetheless, NEI is concerned that licensee pursuit of

either of these limited applications might ultimately require seeking

an exemption to Sec. 50.67, or require extensive analysis. NEI

recommended that the NRC should: (1) revise the proposed rule language

to accommodate limited application of an alternative source term as

done in the Section-By-Section Analysis; (2) provide clarification in

the Statement of Consideration (SOC) for the rule; and (3) for

applications that continue to use the TID source term but incorporate

attributes of newer technical insights such as timing of releases,

specify that the provisions of the proposed rule do not apply.

NRC Response. The language of Sec. 50.67(b) requires an evaluation

of the consequences of applicable design basis accidents. The NRC

believes that the use of the modifier applicable provides the basis for

processing selective implementations. Design basis accidents not

applicable to a particular selective implementation would not be

required to be evaluated. The NRC expects that the licensee will

evaluate all applicable impacts of the proposed AST implementation.

While a selective implementation may result in a reduced scope of

evaluation, the licensee must still demonstrate that the AST

implementation and any associated proposed modifications will not

result in accident conditions exceeding the criteria specified in

Sec. 50.67. Therefore, these criteria are applicable to full and

selective implementations alike. The scope of the required re-analyses

will depend on the specific application proposed by the licensee.

Guidance with regard to this scope is properly provided in the draft

regulatory guide prepared for this rule. Therefore, the NRC has decided

against revising the rule language as suggested by NEI. Consistent with

the second NEI recommendation, the NRC has modified paragraph D of the

section-by-section analysis to clarify this issue.

3. NEI Comment 2

In its second comment, NEI noted that the SOC provides that

licensees may need to perform additional evaluations of equipment

qualifications (Sec. 50.49). The SOC should discuss the circumstances

when such an evaluation may be necessary. NEI recommended that the SOC

should be amended to state that regardless of source term used, the

licensee would be required to re-evaluate the equipment qualification

only when a plant modification alters the plant configuration so that

the underlying assumptions, with respect to dose distribution and

effects, are materially altered. NEI summarized conclusions of several

references in support of its position. NEI stated that there is no

basis to require or expect additional analyses of equipment

qualification if a licensee applied the alternative source term in

limited scope applications, absent a plant configuration change that

materially alters the dose distribution and effects assumed in existing

analyses.

NRC Response. The re-baselining study prepared by the NRC staff

(SECY-98-154, June 30, 1998) considered the impact of an AST on

analyses of the postulated integrated radiation doses for plant

components exposed to containment atmosphere radiation sources and

those exposed to containment sump radiation sources. The staff's

conclusions regarding the atmosphere sources are consistent with those

identified by NEI in its comment. However, the re-baselining study also

concluded that the increased concentration of cesium in the containment

sump water could result in

[[Page 71995]]

an increase in the postulated integrated radiation doses for certain

plant components subject to equipment qualification. It is because of

this conclusion that the NRC included the discussion in the SOC

regarding re-evaluation of equipment environmental qualification. The

NEI comment provides no additional information that would cause the NRC

to change its position on this matter. Further, the NRC has determined

that it is necessary to consider the potential impact of the postulated

cesium concentration in the containment sump water as it applies to all

operating power reactors, not just to those licensees amending their

design basis to use an AST. Since the postulated increase in the

integrated dose occurs only following an accident, there is no adverse

effect on equipment relied upon to perform safety functions immediately

following an accident. Rather, this issue affects equipment that is

required to be operable longer than about 30 days to 4 months after an

accident. As such, the NRC determined that continued plant operation

does not pose an immediate threat to public health and safety. Also,

should such long-term equipment fail there will not be an undue threat

to public health and safety as protective actions for the public would

have already been implemented by the time the postulated failure could

occur. In addition, the time period between the onset of the event and

the projected failure allows compensatory measures to be taken to

prevent the equipment failure or to restore the degraded safety

function. The NRC will evaluate this issue as a generic safety issue to

determine whether further regulatory actions are justified. The final

regulatory guide, or subsequent revisions thereto, is expected to

reflect the resolution of this generic safety issue.

4. NEI Comment 3

NEI recommends that the definition of Source Term in Sec. 50.2 be

revised to ``Source term refers to the magnitude and mix of

radionuclides released from the fuel, their physical and chemical form,

and the timing of their release.'' NEI stated that the language in the

proposed rule would prohibit the use of Sec. 50.67 for accidents such

as the fuel handling accident.

NRC Response. The NRC agrees with the proposed revision. The

proposed definition was consistent with the definition of source term

as used in NUREG-1465, which was written primarily to address loss of

coolant accidents (LOCA). The regulatory guidance for this rule extends

the NUREG-1465 source terms to other accidents which involve core

damage. The definition suggested by NEI is consistent with the proposed

use of the AST. The Sec. 50.2 definition has been revised in the final

rule to reflect the change suggested by NEI and that suggested by

Arizona Public Service Comment 1 below.

5. NEI Comment 4

NEI stated that the proposed rule does not permit new test reactors

to use an alternative source term. New test reactors would have to use

the Part 100 Subpart A, ``Evaluation Factors for Stationary Power

Reactor Site Applications Before January 10, 1997, and for Testing

Reactors,'' even though their application for an operating license

would be filed after January 10, 1997. The use of Section 50.67,

``Accident Source Term,'' is limited to holders of operating licenses

issued before January 10, 1997. This wording prohibits new test

reactors from using the alternative source term. NEI recommended that

Sec. 50.67 be amended to allow new test reactors to use an alternative

source term.

NRC Response. Section 50.67 applies only to holders of licenses for

operating reactors, including test reactors, whose licenses were issued

before January 10, 1997. There is no regulatory requirement for a

specific source term for reactors to be licensed in the future,

including test reactors. Accordingly, no regulatory action is necessary

to accommodate the NEI recommendation.

6. Duke Energy Corporation Comment

Duke Energy Corporation (Duke) endorsed the comments submitted on

behalf of the industry by NEI. Duke stated that the proposed

Sec. 50.67(b)(1) was not clear regarding whether licensees will be

allowed to use a revised source term on a limited basis (e.g., for

analyses of a specific accident or function), or whether they will be

required to review the entire radiological consequence analyses to

apply for the new source term. Duke suggested that necessary guidance

be provided in the draft regulatory guidance to allow for limited use

of the new source terms where such use can be justified.

NRC Response. This comment is similar to NEI Comment 1 addressed

previously. As stated in the SOC, the NRC will consider justifiable

limited (i.e., selective) applications of an AST. Although a selective

implementation may result in a reduced scope of evaluation, the

licensee must still demonstrate that the AST implementation and any

associated proposed modifications will not exceed the criteria

specified in Sec. 50.67. The scope of the required re-analyses will

depend on the specific application proposed by the licensee. Regulatory

guidance on selective implements and the scope of required re-analyses

has been included in the draft guide and are available as announced in

this Federal Register notice.

7. Arizona Public Service Company Comment 1

Arizona Public Service Company (APS) noted that the SOC statement,

``a subsequent change to the source term must be made through a license

amendment'' could be interpreted as requiring prior NRC approval for

any change in the magnitude and mix of radionuclides released from the

reactor core. APS stated that this interpretation could place

additional restrictions on licensee efforts at economical fuel

management, including reload design.

NRC Response. The NRC agrees with the APS comment. The NRC had

intended the phrase ``magnitude and mix'' to refer to the fractions of

the fission product inventory of the radionuclides released from the

reactor fuel. The NRC intent for the provision in question was to

require approval for changes in the radioactivity release fractions,

the radionuclides released, their physical and chemical form, and the

timing of their release. Since ``magnitude and mix'' could be a source

of confusion, the NRC has modified the Sec. 50.2 definition of Source

Term in the final rule to read: ``Source term refers to the magnitude

and mix of the radionuclides released from the fuel, expressed as

fractions of the fission product inventory in the fuel, as well as

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

This is consistent with NUREG-1465 when it refers to ``magnitude and

mix,'' since the NUREG-1465 presents these data in the form of tables

of release fractions and radionuclides. This revised language also

addresses NEI Comment 3 above.

8. Arizona Public Service Company Comment 2

In its second comment, APS noted that NUREG-1465 contains a

disclaimer that the accident source terms provided therein may not be

applicable to fuel irradiated in excess of 40 GWD/MTU. The NRC has

licensed core designs with fuel irradiations of up to 62 GWD/MTU. APS

questioned whether the NRC staff was going to address the affect of

high burnups on a generic basis, or on a facility-by-facility basis.

NRC Response. The AST tabulated in the draft regulatory guidance,

which

[[Page 71996]]

differs in some aspects from that provided in NUREG-1465, is applicable

to peak rod average irradiations up to 62 GWD/MTU. Attachment 1 to the

regulatory analysis for this rulemaking describes the bases of this

extension in fuel irradiation as it applies to the AST. There are some

facility-by-facility considerations. For example, the increase in core

inventory for some long-lived radionuclides and the change in isotopic

mix due to the increase in plutonium fission as the fuel ages is

addressed by the Draft Guide-1081 provision that licensees re-analyze

the core inventory based on current operating parameters, including

fuel burnup.

III. Section-by-Section Analysis

A. Section 50.2

The general ``definitions'' section for Part 50 is 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. The definition of source term in

Sec. 50.2 embodies the NUREG-1465 definition; however, the Sec. 50.2

definition includes the clarifying phrase, ``expressed as fractions of

the fission product inventory in the fuel,'' (see prior response to

Arizona Public Service Comment 1). 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 guidance that supports

this final rule describes an acceptable basis for defining the

characteristics of an alternative source term.

B. Section 50.67(a)

This paragraph defines the licensees that may seek to revise their

current radiological source term with an alternative source term. The

final rule is applicable to holders of operating licenses that were

issued under 10 CFR Part 50 before January 10, 1997, and to holders of

renewed licenses issued under 10 CFR Part 54 whose initial operating

license was issued prior to January 10, 1997. The final rule does 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 final rule does not explicitly

define an alternative source term. In lieu of an explicit reference to

NUREG-1465, Footnote 1 to the final rule identifies the significant

attributes of an accident source term. The regulatory guidance that is

being issued to support this final rule will identify ASTs (based on

the NUREG-1465 source terms) that are acceptable alternatives to the

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

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

developed and will allow licensees to propose additional alternatives

for NRC consideration.

C. Section 50.67(b)(1)

This paragraph of Sec. 50.67 identifies 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 final rule have indicated that there are potential changes to the

facility as documented in the FSAR that will constitute a no

significant hazards consideration. However, these determinations will

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

evaluations. The procedural requirements for processing a license

amendment are presented 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 3 that

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

final safety analysis report (FSAR) without prior NRC approval, if the

proposed change meets certain criteria specified in Sec. 50.59. If the

criteria are not met, the licensee must request NRC approval of the

change using the license amendment process detailed in Sec. 50.90.

Significant to this final rule is the criterion that NRC review is

required if the proposed change would result in a greater than minimal

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 might not

provide assurance that this change in the design basis would be

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

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

\3\ Section 10 CFR 50.59 is being amended in a parallel, but

separate, rulemaking action. That rulemaking, when implemented is

expected to replace the unreviewed safety question (USQ) concept.

Further, the criteria for consequences are being revised from ``may

be increased'' to ``result in more than a minimal increase.'' Those

changes are not expected to invalidate the conclusions drawn in this

analysis.

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

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 fractions of the fission product inventory of the radionuclides

released from the reactor fuel, their chemical and physical form, or

the timing of their release as tabulated in the regulatory guidance

(with deviations proposed by the licensee and approved by the NRC)

could not be implemented under Sec. 50.59. This provision applies only

to these tabulated parameters.

The final rule will 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, ``Clarification of TMI Action Plan Requirements,''

requirements II.B.2, II.B.3, III.D.3.4. The regulatory guidance that

supports this final 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

[[Page 71997]]

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

alternative source term. Nor is it the NRC's intent that EAB, LPZ, and

control room dose calculations be performed for all applications under

Sec. 50.67. 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 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 for future LWRs to

operating reactors that elect to use an alternative source term.

The NRC is replacing 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 is included in Sec. 50.67 as well as in 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 final 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.

The NRC has not included 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. 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 final rule. However, the NRC does

believe that these provisions offer a useful perspective that supports

the conclusion that the organ doses implied by the 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 final rule supersede the dose

criteria in GDC-19, the other provisions of GDC-19 remain applicable.

There may be technically justifiable implementations of an AST that

would not require calculation of the EAB, LPZ, or control room doses.

For example, a proposed modification to change the closure time of a

containment isolation valve from 2 seconds to 5 seconds may be based on

the timing insights of the AST. Although a specific calculation might

not be necessary in this case, the licensee is still required to affirm

with reasonable assurance that the doses would comply with these stated

criteria.

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

GDC-19 is changed to include the TEDE dose criterion for control

room design for applicants for construction permits, design

certifications, and combined 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 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

AP600 advanced LWR in order to use the 0.05 Sv (5 rem) TEDE criterion

deemed necessary for use with alternative source terms. Exemptions will

arguably be necessary for future applicants for construction permits,

design certifications, and combined licenses. This amendment will

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 are 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.

[[Page 71998]]

G. Section 50.34

A new footnote to Sec. 50.34 has been added to define what

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

the existing footnote 1 to Sec. 100.11, and was added to provide for

consistency between Parts 50 and 100.

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

These paragraphs are 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 three

classes of applicants. The first affected class is comprised of

applicants for design certification under Part 52, Subpart B. Section

52.47(a)(1)(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 were modified to delete

the reference to TID-14844. This change makes it clear that applicants

for combined licenses should not use the TID-14844 source term but

should 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 comprised of applicants for combined licenses

under Part 52, Subpart C. Section 52.79(b) makes the requirements of

52.47(a)(1)(i) applicable if a certified design is not referenced.

Thus, the combined license applicant is also subject to the

requirements of Section 50.34(f).

The third 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.

IV. Draft Regulatory Guide; Issuance, Availability

The Nuclear Regulatory Commission is issuing for public comment a

draft of a guide planned for its Regulatory Guide Series. This series

has been developed to describe and make available to the public

information such as methods acceptable to the NRC staff for

implementing specific parts of the Commission's regulations, techniques

used by the staff in evaluating specific problems or postulated

accidents, and data needed by the NRC staff in its review of

applications for permits and licenses. Copies of the draft guide may be

obtained as described in Section VI, ``Referenced Documents,'' of these

statements of consideration. You may also download copies from the

NRC's interactive rulemaking forum website through the NRC home page

(http://ruleforum.llnl.gov/cgi-bin/rulemake).

The draft guide, temporarily identified by its task number DG-1081

(which should be mentioned in all correspondence concerning this draft

guide) is titled ``Alternative Radiological Source Terms for Evaluating

Design Basis Accidents at Nuclear Power Reactors.'' This guide is

intended for Division 1, ``Power Reactors.'' This draft guide is being

developed to provide regulatory guidance on the implementation of an

alternative source term at an operating reactor. The guide addresses

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 provides

guidance on the scope and extent of affected design basis accident

(DBA) radiological analyses and associated acceptance criteria. The

guide includes revised assumptions and methods for each affected DBA in

a series of appendices. These appendices supersede the guidance in

Regulatory Guides 1.3, 1.4, 1.5, 1.25, and 1.77, and supplement

guidance in Regulatory Guide 1.89 for those facilities using an

alternative source term.

The draft guide has not received complete NRC staff review and does

not represent an official NRC staff position.

Previous draft versions of DG-1081 have been made publicly

available to support technical interactions with the public. This

Federal Register announcement provides an opportunity for the public to

provide comments on the DG-1081 guidance. The NRC staff will consider

the public comments in its efforts to finalize the regulatory guidance.

The Commission invites advice and recommendations on the content of

the draft regulatory guide. Comments and suggestion are particularly

requested on the following questions.

A. Scope of Implementation

1. The guidance provided in the draft regulatory guide is intended

to allow licensees the maximum flexibility in pursuing technically

justifiable AST implementations provided that a clear, consistent, and

logical design basis is maintained. Comments are specifically requested

on the following questions.

A. Does the proposed guidance provide the desired flexibility while

providing reasonable assurance that a clear, consistent, and logical

design basis will be maintained?

B. Is there a less complex alternative approach that would provide

the desired flexibility while maintaining a clear, consistent, and

logical design basis?

C. Should the Commission allow licensees that have received

approval for a selective implementation to extend the AST and the TEDE

criteria to other design basis applications (that do not involve

reanalysis of the DBA LOCA) under Sec. 50.59 rather than under

Sec. 50.67 as currently proposed?

2. The guidance would allow selective implementation of the

characteristics (i.e., the fractions of fission product inventory of

the radionuclides released from the reactor fuel, their chemical and

physical form, and the timing of their release) of an AST. The

Commission believes that implementations based only on the timing

insights of an AST may be technically justifiable. The Commission

believes that the other combinations may be internally inconsistent.

Comments are specifically requested on the following questions.

A. What other combinations of AST characteristics are technically

consistent?

B. What plant modifications might be based on these combinations?

B. Scope of Re-Analyses

1. The draft regulatory guide provides guidance on the scope of the

re-analyses that should be performed to support an AST implementation.

Comments are requested on the following questions.

A. Is the proposed guidance on the scope of re-analyses technically

appropriate and clear? How could it be improved?

B. The guidance allows licensees to disposition certain impacts of

an AST on the basis of the NRC staff's re-baselining study. Does this

study or other documents provide a sufficient basis for the Commission

to generically disposition these impacts?

2. It may be possible for licensees to demonstrate that the doses

from certain affected analyses assessed using the prior source term and

dose methodology would be greater than the doses obtained using a

proposed AST and the TEDE methodology. The proposed guidance would

allow the licensee to disposition these affected analyses without re-

calculation. Nonetheless, the design basis would now include the

approved AST and TEDE criteria. The guidance in the draft regulatory

guide would require the licensee to update the calculation to be

consistent with the approved AST and dose methodology described in the

facility design basis in

[[Page 71999]]

the event of a subsequent re-calculation. Comments are requested on the

following questions.

A. Should the Commission allow licensees to continue to use the

prior source term and dose criteria for these analyses and not require

that they be updated on subsequent revisions?

B. If the analyses are not updated, how will licensees assure that

the earlier conclusion that the analyses are limiting remains valid

following subsequent revisions?

3. Analyses of the integrated radiation doses for environmental

qualification of certain equipment important to safety will be affected

by the increased concentration of radioactive cesium in the containment

sump water. The Commission has been considering the position that

licensees proposing to implement an AST must address all impacts of the

proposed implementation, including the impact of the increased cesium

concentration. However, the Commission now believes it may be necessary

for all operating power reactors to address the postulated increase in

the cesium concentration. The Commission will consider this issue as a

generic safety issue. Comments are requested on the following

questions.

A. Is there information that should be considered by the Commission

in resolving this generic issue?

B. If the Commission should conclude that there is safety

significance but that the costs of implementing corrective actions are

not justified on a generic basis, should licensees who are voluntarily

proposing to amend their design basis to use an AST be required to

address the impact of the increased cesium concentration?

C. If a licensee proposes a change in the plant configuration that

would result in an increase in the integrated dose for one or more

components and this licensee is also proposing, or has already

implemented an AST, should the re-analysis of the integrated dose be

based on that AST or on the prior TID14844 source term?

Comments may be accompanied by relevant information or supporting

data. Written comments may be mailed to: Secretary, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, Attention:

Rulemakings and Adjudications Staff. Mail Stop O16C1. Copies of

comments received may be examined at the NRC Public Document Room, 2120

L Street NW., Washington, DC. Comments will be most helpful if received

by March 7, 2000.

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

website through the NRC home page (http://ruleforum.llnl.gov/cgi-bin/

rulemake). This site provides the availability to upload comments as

files (any format), if your web browser supports that function. For

information about the interactive rulemaking website, contact Ms. Carol

Gallagher, (301) 415-5905; or by internet electronic mail to

[email protected]. For information about the draft guide, contact Mr. Stephen

F. LaVie, (301) 415-1081; Internet electronic mail [email protected].

Although a time limit is given for comments on this draft guide,

comments and suggestions in connection with items for inclusion in

guides currently being developed or improvements in all published

guides are encouraged at any time.

V. Draft Standard Review Plan Section; Issuance, Availability

The Nuclear Regulatory Commission is issuing for public comment a

draft of a new section to NUREG-0800, ``Standard Review Plan.''

Standard review plan (SRP) sections are prepared for the guidance of

the Office of Nuclear Reactor Regulation staff responsible for the

review of applications to construct and operate nuclear power plants.

These documents are made available to the public as part of the

Commission's policy to inform the nuclear industry and the general

public of regulatory procedures and policies. The draft SRP Section

15.0.1, is titled ``Radiological Consequence Analyses Using Alternative

Source Terms.'' The SRP section complements draft regulatory guide DG-

1081. The draft SRP section has not received complete NRC staff review

and does not represent an official NRC staff position.

Copies of the draft SRP section may be obtained as described in

Section VI, ``Referenced Documents,'' of these statements of

consideration. You may also download copies from the NRC's interactive

rulemaking forum website through the NRC home page (http://

ruleforum.llnl.gov/cgi-bin/rulemake).

Comments on the content of the draft SRP section are invited.

Comments may be accompanied by relevant information or supporting data.

Comments should be submitted as described above for the draft

regulatory guide. Although a time limit is given for comments on this

draft SRP section, comments and suggestions in connection with items

for inclusion in SRP sections currently being developed or improvements

in all published SRP sections are encouraged at any time.

VI. Referenced Documents

Copies of NUREG-0737, NUREG-0800, NUREG-1465, NUREG/BR-0058, NUREG/

BR-184, 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.

Single copies of regulatory guides, both active and draft may be

obtained free of charge by writing the Reproduction and Distribution

Services Section, OCIO, USNRC, Washington DC 20555-0001, or by fax to

(301) 415-2289, or by email to [email protected]. Active guides may

also be purchased from the National Technical Information Service on a

standing order basis. Details of this service may be obtained by

writing NTIS, 5285 Port Royal Road, Springfield, VA 22161. Copies of

active and draft guides are available for inspection or copying for a

fee from the NRC Public Document Room at 2120 L Street NW., Washington

DC.

Copies of SECY-94-302, SECY-96-242, SECY-98-154, SECY-98-289, TID-

14844, 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. 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

final rule allows 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 are not 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.

Those plant changes that do not

[[Page 72000]]

require prior NRC review and approval pursuant to Sec. 50.59 are not

likely to involve any significant increase in environmental impacts.

The Sec. 50.59 criteria are sufficiently stringent that any potential

change in plant design that could have an adverse environmental impact

in all likelihood could not be made by the licensee without prior NRC

review and approval. Every plant change that requires NRC review and

approval under Sec. 50.59 requires a license amendment and, therefore,

the preparation of an environmental assessment to determine whether the

proposed change involves any significant environmental impact. Thus,

this final rule, by itself, will not result in plant changes that

involve any significant increase in environmental impacts. The final

rule does not affect non-radiological plant effluents.

The NRC requested public comments on any environmental justice

considerations that may be related to this rule. No public comments

relevant to the draft environmental assessment or environmental justice

considerations were received. The NRC requested the views of the States

on the environmental assessment for this rule. No comments relevant to

the draft environmental assessment or environmental justice

considerations were received.

The environmental assessment and 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 final 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 relative to the total burden

estimated, 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 Act 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 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 final rule, and that a backfit analysis is not

required for this rulemaking because these amendments do not involve

any provisions that would impose backfits as defined in 10 CFR

50.109(a)(1). This final rule amends the NRC's regulations by

establishing alternate requirements that may be voluntarily adopted by

licensees, and makes changes to the regulations to conform them to a

1996 rulemaking.

XII. Small Business Regulatory Enforcement Fairness Act

In accordance with the Small Business Regulatory Fairness Act of

1996, the NRC has determined that this action is not a major rule and

has verified this determination with the Office of Information and

Regulatory Affairs, Office of Management and Budget.

XIII. National Technology Transfer and Advancement Act

The National Technology Transfer Act of 1995, Pub. L. 104-113,

requires that Federal agencies use technical standards that are

developed or adopted by voluntary consensus standards bodies unless the

use of such a standard is inconsistent with applicable law or otherwise

impractical. In this final rule the NRC is establishing a government-

unique standard in Section 50.67(b)(2) by specifying accident radiation

dose criteria. These criteria were issued for use by future license

applicants by an earlier rulemaking (61 FR 65157, December 11, 1996)

and, by this final rule, are being applied to operating reactors that

voluntarily use an alternative source term. No voluntary consensus

standard has been identified that could be used instead of the

government-unique standard.

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.

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

[[Page 72001]]

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:

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

and 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 the radionuclides

released from the fuel, expressed as fractions of the fission product

inventory in the fuel, as well as 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)

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

from the reactor coolant system and containment that may contain

accident source term \11\ 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)

* * * * *

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

systems outside containment that contain (or might contain) accident

source term \11\ 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)

* * * * *

(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 \11\ release,

and make necessary design provisions to preclude such problems.

(III.D.3.4)

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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.

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* * * * *

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, and

holders of renewed licenses under part 54 of this chapter whose initial

operating license was issued prior to January 10, 1997, who seek to

revise the current accident source term used in their design basis

radiological analyses.

(b) Requirements. (1) A licensee who seeks to revise its current

accident source term in design basis radiological

[[Page 72002]]

consequence analyses shall apply for a license amendment under

Sec. 50.90. The application shall contain an evaluation of the

consequences of applicable design basis accidents \1\ previously

analyzed in the safety analysis report.

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\1\ The fission product release assumed for these calculations

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

design analyses 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.

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

(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).

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\2\ 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).

(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, section II, ``Protection by Multiple

Fission Product Barriers,'' ``Criterion 19--Control room'' is revised

to read as follows:

Appendix A to Part 50--General Design Criteria for Nuclear Power

Plants

* * * * *

II. Protection by Multiple Fission Product Barriers

* * * * *

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.

Applicants for and holders of construction permits and operating

licenses under this part who apply on or after January 10, 1997,

applicants for design certifications under part 52 of this chapter

who apply on or after January 10, 1997, applicants for and holders

of combined licenses under part 52 of this chapter who do not

reference a standard design certification, 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.

(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 17th day of December 1999.

For the Nuclear Regulatory Commission.

Annette Vietti-Cook,

Secretary of the Commission.

[FR Doc. 99-33283 Filed 12-22-99; 8:45 am]

BILLING CODE 7590-01-P