65018 Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices

(GALL) Report, Standard Review

Plan (SRP), and Regulatory Guide

(Public Meeting) (Contact: Chris

Grimes, 301–415–1183)

This meeting will be webcast live at

the Web address—www.nrc.gov/

live.html.

Note: The schedule for Commission

meetings is subject to change on short notice.

To verify the status of meetings call

(Recording)—(301) 415–1292. Contact Person

for more information: Bill Hill (301) 415–

1661.

SUPPLEMENTARY INFORMATION: By a vote

of 5–0 on October 23, the Commission

determined pursuant to U.S.C. 552b(e)

and § 9.107(a) of the Commission’s rules

that ‘‘Affirmation of Final Rules—10

CFR Part 35, ‘Medical Use of Byproduct

Material’ and 10 CFR Part 20, ‘Standards

for Protection Against Radiation’ ’’ be

held on October 23, and on less than

one week’s notice to the public.

The NRC Commission Meeting

Schedule can be found on the Internet

at: http://www.nrc.gov/SECY/smj/

schedule.htm

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661). In addition, distribution of

this meeting notice over the Internet

system is available. If you are interested

in receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

Dated: October 27, 2000.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 00–28035 Filed 10–27–00; 2:15 pm]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Notice of Availability for Referencing in

License Amendment Applications—

Model Safety Evaluation on Technical

Specification Improvement To

Eliminate Requirements on Post

Accident Sampling Systems Using the

Consolidated Line Item Improvement

Process

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of availability.

SUMMARY: Notice is hereby given that

the staff of the Nuclear Regulatory

Commission (NRC) has prepared a

model safety evaluation (SE) relating to

the elimination of requirements on post

accident sampling imposed on licensees

through orders, license conditions, or

technical specifications. The NRC staff

has also prepared a model no significant

hazards consideration (NSHC)

determination relating to this matter.

The purpose of these models is to

permit the NRC to efficiently process

amendments that propose to remove

requirements for the Post Accident

Sampling System (PASS). Licensees of

nuclear power reactors to which the

models apply may request amendments,

in accordance with Section 50.90 of

Title 10 to the Code of Federal

Regulations, confirming the

applicability of the SE and NSHC

determination to their reactors and

providing the requested plant-specific

verifications and commitments.

DATES: The period during which

licensees may reference the model SE

and NSHC determination expires

October 31, 2001. Applications for

amendments after this date must

include plant-specific justifications for

the proposed changes and an analysis

about the issue of no significant hazards

consideration.

FOR FURTHER INFORMATION CONTACT:

William Reckley, Mail Stop: O–7D1,

Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, telephone 301–415–1323.

SUPPLEMENTARY INFORMATION:

Background

Regulatory Issue Summary 2000–06,

‘‘Consolidated Line Item Improvement

Process for Adopting Standard

Technical Specification Changes for

Power Reactors,’’ was issued on March

20, 2000. The consolidated line item

improvement process (CLIIP) is

intended to improve the efficiency of

NRC licensing processes. This is

accomplished by processing proposed

changes to the Standard Technical

Specifications (STS) in a manner that

supports subsequent license amendment

applications. The CLIIP includes an

opportunity for the public to comment

on proposed changes to the STS

following a preliminary assessment by

the NRC staff and finding that the

change will likely be offered for

adoption by licensees. The CLIIP directs

the NRC staff to evaluate any comments

received for a proposed change to the

STS and to either reconsider the change

or to proceed with announcing the

availability of the change for proposed

adoption by licensees. Those licensees

opting to apply for the subject change to

their technical specifications are

responsible for reviewing the staff’s

evaluation, referencing the applicable

technical justifications, and providing

any necessary plant-specific

information. Each amendment

application made in response to the

notice of availability would be

processed and noticed in accordance

with applicable rules and NRC

procedures.

This proposed change was proposed

for incorporation into the Standard

Technical Specifications by the

Westinghouse Owners Group (WOG)

and the Combustion Engineering

Owners Group (CEOG) participants in

the Technical Specification Task Force

(TSTF) and is designated TSTF–366. A

notice of opportunity to comment on the

use of CLIIP for the elimination of

requirements for PASS and related

administrative controls in technical

specifications for plants with

Westinghouse and Combustion

Engineering designs was published in

the Federal Register on August 11, 2000

(65 FR 49271). The nine comments

submitted to the NRC staff in response

to the solicitation are addressed later in

this notice.

Applicability

This application of the CLIIP to

remove requirements for PASS from

technical specifications (and other

elements of the licensing bases) is

applicable to plants with Westinghouse

and Combustion Engineering designs.

To efficiently process the incoming

license amendment applications, the

staff requests each licensee applying for

the changes addressed by TSTF–366

using the CLIIP to address the plantspecific verifications and regulatory

commitments that are identified in the

model SE. The CLIIP does not prevent

licensees from requesting an alternative

approach or proposing the changes

without the requested verifications and

regulatory commitments. Licensees

choosing to request an approach

different than that described in this

notice should submit applications with

appropriate plant-specific justifications

for the proposed changes and an

analysis about the issue of no significant

hazards consideration. Variations from

the approach recommended in this

notice may require additional review by

the NRC staff and may increase the time

and resources needed for the review.

In making the requested regulatory

commitments, each licensee should

address: (1) That the subject capability

exists (or will be developed) and will be

maintained; (2) where the capability or

procedure will be described (e.g., severe

accident management guidelines,

emergency operating procedures,

emergency plan implementing

Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices 65019

procedures); and (3) a schedule for

implementation. The amendment

request need not provide details about

designs or procedures. Each licensee

should verify that it has, and make a

regulatory commitment to maintain (or

make a regulatory commitment to

develop and maintain):

a. Contingency plans for obtaining

and analyzing highly radioactive

samples from the reactor coolant

system, containment sump, and

containment atmosphere;

b. A capability for classifying fuel

damage events at the Alert level

threshold (typically this is 300 μCi/ml

dose equivalent iodine). This capability

may use the normal sampling system

and/or correlations of sampling or

letdown line dose rates to coolant

concentrations; and

c. The capability to monitor

radioactive iodines that have been

released to offsite environs.

Public Notices

The staff issued a Federal Register

Notice (64 FR 66213, November 24,

1999) that requested public comment on

the NRC’s pending action to approve

topical reports submitted by the WOG

and the CEOG in which they proposed

to eliminate regulatory requirements for

PASS. In particular, the staff sought

comment from offsite emergency

response organizations so that any

impact of the elimination of PASS on

their response could be factored into the

staff’s evaluation. Appendices to the

staff’s safety evaluations for topical

reports submitted by the CEOG and the

WOG contain a synopsis of the public

comments received and the staff’s

evaluation of the comments. The topical

reports as well as the NRC staff’s safety

evaluations for the topical reports may

be examined, and/or copied for a fee, at

the NRC’s Public Document Room,

located at One White Flint North, 11555

Rockville Pike (first floor), Rockville,

Maryland. Publicly available records

will be accessible electronically from

the ADAMS Public Library component

on the NRC Web site, (the Electronic

Reading Room). The staff’s safety

evaluations that address the public

comments about the topical reports are

available on ADAMS (Accession

Numbers ML003715250 dated May 16,

2000, for the CEOG topical report and

ML003723268 dated June 14, 2000, for

the WOG topical report).

A notice soliciting comments from

interested members of the public about

the use of the CLIIP for elimination of

requirements for PASS was published in

the Federal Register on August 11, 2000

(65 FR 49271). The staff received nine

comments (six from individual

licensees, one from the Nuclear Energy

Institute, one from a law firm that

represents licensees, and one from a

member of the public) as a result of the

notice of opportunity to comment about

the subject technical specification

changes. Five of the letters received

included general comments in favor of

the CLIIP and its use in eliminating

requirements for PASS. Specific

comments on the model SE were offered

in four of the comment letters. The

specific comments are discussed below:

1. A licensee suggested that the model

SE include a discussion indicating that

the contingency plans do not have to be

carried out in emergency plans and

exercises. A similar statement was

included in the staff’s SE for the topical

report prepared by the WOG. The staff

agrees with the comment and added a

sentence to the model SE.

2. A licensee stated that some plants

have safety-related hydrogen monitors

with ranges significantly above

hydrogen concentrations of 10% that

could be used for severe accident

conditions. The staff believes that the

model SE provides the necessary

flexibility for plant-specific differences

in the ranges of the monitors by stating

that the appropriate decision-makers

may determine if a grab sample is

necessary and practical during the

management of a severe accident. A

contingency plan for sampling the

containment atmosphere also serves to

confirm the indications from the

monitors and provide information on

parameters other than hydrogen

concentrations (e.g., the mix of

radionuclides) and should, for

consideration of the amendment as part

of the CLIIP, be part of the plant-specific

regulatory commitment discussed in the

model SE. The staff did not revise the

model SE in response to this comment.

3. A licensee suggested that the Alert

level threshold (typically 300 μCi/ml

dose equivalent iodine) recognize an

alternative of 2% to 5% fuel clad

damage and that instrumentation such

as core exit thermocouples or radiation

monitors might also be indicative of fuel

clad damage. The staff did not intend to

preclude the use of other parameters as

an indication of the loss of or challenge

to the fuel clad fission product barrier.

The staff included the regulatory

commitment (item 4.2) in the model SE

to address classifying certain types of

events (such as reactivity excursions or

mechanical damage) which could cause

fuel damage without having an

indication of overheating on core exit

thermocouples. The mention of normal

sampling or letdown line dose rates in

the model SE is intended to be

alternatives for those licensees that

currently use PASS for assessing the 300

μCi/ml does equivalent iodine criterion

for declaration of an Alert. The staff did

not revise the model SE in response to

this comment.

4. A commenter suggested that the use

of the CLIIP to eliminate PASS

requirements be expanded to all

licensed facilities. The staff may choose

to use the CLIIP to address the removal

of PASS from plants with other than

Westinghouse and Combustion

Engineering designs. Such a use of the

CLIIP would follow a specific proposal

and justification from the applicable

owners groups similar to the TSTF

submitted by the WOG and CEOG. The

staff did not revise the model SE in

response to this comment.

This notice is announcing the

availability of the model safety

evaluation and model NSHC

determination for referencing in

applications for amendments to

technical specifications for applicable

plants. Licensees wishing to adopt the

change must submit an application in

accordance with applicable regulatory

requirements. The staff will in turn

issue for each application a notice of

consideration of issuance of amendment

to facility operating license(s), a

proposed NSHC determination, and an

opportunity for a hearing. A notice of

issuance of an amendment to operating

license(s) will also be issued to

announce the elimination of the PASS

requirements for each plant that applies

for and receives the requested change.

Model Safety Evaluation

U.S. Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

Consolidated Line Item Improvement

Technical Specification Task Force

(TSTF) Change TSTF–366

Elimination of Requirements for Post

Accident Sampling System (PASS)

1.0 Introduction

In the aftermath of the accident at

Three Mile Island (TMI), Unit 2, the

Nuclear Regulatory Commission (NRC)

imposed requirements on licensees for

commercial nuclear power plants to

install and maintain the capability to

obtain and analyze post-accident

samples of the reactor coolant and

containment atmosphere. The desired

capabilities of the Post Accident

Sampling System (PASS) were

described in NUREG–0737,

‘‘Clarification of TMI Action Plan

Requirements.’’ The NRC issued orders

to licensees with plants operating at the

time of the TMI accident to confirm the

installation of PASS capabilities

65020 Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices

(generally as they had been described in

NUREG–0737). A requirement for PASS

and related administrative controls was

added to the technical specifications

(TS) of the operating plants and was

included in the initial TS for plants

licensed during the 1980s and 90s.

Additional expectations regarding PASS

capabilities were included in Regulatory

Guide 1.97, ‘‘Instrumentation for LightWater-Cooled Nuclear Power Plants To

Assess Plant and Environs Conditions

During and Following an Accident.’’

Significant improvements have been

achieved since the TMI accident in the

areas of understanding risks associated

with nuclear plant operations and

developing better strategies for

managing the response to potentially

severe accidents at nuclear plants.

Recent insights about plant risks and

alternate severe accident assessment

tools have led the NRC staff to conclude

that some TMI Action Plan items can be

revised without reducing the ability of

licensees to respond to severe accidents.

The NRC’s efforts to oversee the risks

associated with nuclear technology

more effectively and to eliminate undue

regulatory costs to licensees have

prompted the NRC to consider

eliminating the requirements for PASS

in TS and other parts of the licensing

bases of operating reactors.

The staff has completed its review of

the topical reports submitted by the

Combustion Engineering Owners Group

(CEOG) and the Westinghouse Owners

Group (WOG) that proposed the

elimination of PASS. The justifications

for the proposed elimination of PASS

requirements center on evaluations of

the various radiological and chemical

sampling and their potential usefulness

in responding to a severe reactor

accident or making decisions regarding

actions to protect the public from

possible releases of radioactive

materials. As explained in more detail

in the staff’s safety evaluations for the

two topical reports, the staff has

reviewed the available sources of

information for use by decision-makers

in developing protective action

recommendations and assessing core

damage. Based on this review, the staff

found that the information provided by

PASS is either unnecessary or is

effectively provided by other

indications of process parameters or

measurement of radiation levels. The

staff agrees, therefore, with the owners

groups that licensees can remove the TS

requirements for PASS, revise (as

necessary) other elements of the

licensing bases, and pursue possible

design changes to alter or remove

existing PASS equipment.

2.0 Background

In a letter dated May 5, 1999 (as

supplemented by letter dated April 14,

2000), the CEOG submitted the topical

report CE NPSD–1157, Revision 1,

‘‘Technical Justification for the

Elimination of the Post-Accident

Sampling System From the Plant Design

and Licensing Bases for CEOG

Utilities.’’ A similar proposal was

submitted on October 26, 1998 (as

supplemented by letters dated April 28,

1999, April 10 and May 22, 2000), by

the WOG in its topical report WCAP–

14986, ‘‘Post Accident Sampling System

Requirements: A Technical Basis.’’ The

reports provided evaluations of the

information obtained from PASS

samples to determine the contribution

of the information to plant safety and

accident recovery. The reports

considered the progression and

consequences of core damage accidents

and assessed the accident progression

with respect to plant abnormal and

emergency operating procedures, severe

accident management guidance, and

emergency plans. The reports provided

the owners groups’ technical

justifications for the elimination for the

various PASS sampling requirements.

The specific samples and the staff’s

findings are described in the following

evaluation.

The NRC staff prepared this model

safety evaluation (SE) relating to the

elimination of requirements on post

accident sampling and solicited public

comment (65 FR 49271) in accordance

with the consolidated line item

improvement process (CLIIP). The use

of the CLIIP in this matter is intended

to help the NRC to efficiently process

amendments that propose to remove the

PASS requirements from TS. Licensees

of nuclear power reactors to which this

model apply were informed [FR] that

they could request amendments

confirming the applicability of the SE to

their reactors and providing the

requested plant-specific verifications

and commitments.

3.0 Evaluation

The technical evaluations for the

elimination of PASS sampling

requirements are provided in the safety

evaluations dated May 16, 2000, for the

CEOG topical report CE NPSD–1157 and

June 14, 2000, for the WOG topical

report WCAP–14986. The NRC staff’s

safety evaluations approving the topical

reports are located in the NRC’s

Agencywide Documents Access and

Management System (ADAMS)

(Accession Numbers ML003715250 for

CE NPSD–1157 and ML003723268 for

WCAP–14986).

The ways in which the requirements

and recommendations for PASS were

incorporated into the licensing bases of

commercial nuclear power plants varied

as a function of when plants were

licensed. Plants that were operating at

the time of the TMI accident are likely

to have been the subject of confirmatory

orders that imposed the PASS functions

described in NUREG–0737 as

obligations. The issuance of plant

specific amendments to adopt this

change, which would remove PASS and

related administrative controls from TS,

supersede the PASS specific

requirements imposed by post-TMI

confirmatory orders.

As described in its safety evaluations

for the topical reports, the staff finds

that the following PASS sampling

requirements may be eliminated for

plants of Combustion Engineering and

Westinghouse designs:

1. Reactor coolant dissolved gases

2. Reactor coolant hydrogen

3. Reactor coolant oxygen

4. Reactor coolant pH

5. Reactor coolant chlorides

6. Reactor coolant boron

7. Reactor coolant conductivity

8. Reactor coolant radionuclides

9. Containment atmosphere hydrogen

concentration

10. Containment oxygen

11. Containment atmosphere

radionuclides

12. Containment sump pH

13. Containment sump chlorides

14. Containment sump boron

15. Containment sump radionuclides

The staff agrees that sampling of

radionuclides is not required to support

emergency response decision making

during the initial phases of an accident

because the information provided by

PASS is either unnecessary or is

effectively provided by other

indications of process parameters or

measurement of radiation levels.

Therefore, it is not necessary to have

dedicated equipment to obtain this

sample in a prompt manner.

The staff does, however, believe that

there could be significant benefits to

having information about the

radionuclides existing post-accident in

order to address public concerns and

plan for long-term recovery operations.

As stated in the safety evaluations for

the topical reports, the staff has found

that licensees could satisfy this function

by developing contingency plans to

describe existing sampling capabilities

and what actions (e.g., assembling

temporary shielding) may be necessary

to obtain and analyze highly radioactive

samples from the reactor coolant system

(RCS), containment sump, and

containment atmosphere. (See item 4.1

Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices 65021

under Licensee Verifications and

Commitments.) These contingency

plans must be available to be used by a

licensee during an accident; however,

these contingency plans do not have to

be carried out in emergency plan drills

or exercises. The contingency plans for

obtaining samples from the RCS,

containment sump, and containment

atmosphere may also enable a licensee

to derive information on parameters

such as hydrogen concentrations in

containment and boron concentration

and pH of water in the containment

sump. The staff considers the sampling

of the containment sump to be

potentially useful in confirming

calculations of pH and boron

concentrations and confirming that

potentially unaccounted for acid

sources have been sufficiently

neutralized. The use of the contingency

plans for obtaining samples would

depend on the plant conditions and the

need for information by the decisionmakers responsible for responding to

the accident.

In addition, the staff considers

radionuclide sampling information to be

useful in classifying certain types of

events (such as a reactivity excursion or

mechanical damage) that could cause

fuel damage without having an

indication of overheating on core exit

thermocouples. However, the staff

agrees with the topical reports’

contentions that other indicators of

failed fuel, such as letdown radiation

monitors (or normal sampling system),

can be correlated to the degree of failed

fuel. (See item 4.2 under Licensee

Verifications and Commitments.)

In lieu of the information that would

have been obtained from PASS, the staff

believes that licensees should maintain

or develop the capability to monitor

radioactive iodines that have been

released to offsite environs. Although

this capability may not be needed to

support the immediate protective action

recommendations during an accident,

the information would be useful for

decision makers trying to limit the

public’s ingestion of radioactive

materials. (See item 4.3 under Licensee

Verifications and Commitments.)

The staff believes that the changes

related to the elimination of PASS that

are described in the topical reports,

related safety evaluations and this

proposed change to TS are unlikely to

result in a decrease in the effectiveness

of a licensee’s emergency plan. Each

licensee, however, must evaluate

possible changes to its emergency plan

in accordance with 10 CFR 50.54(q) to

determine if the change decreases the

effectiveness of its site-specific plan.

Evaluations and reporting of changes to

emergency plans should be performed

in accordance with applicable

regulations and procedures.

The staff notes that redundant, safetygrade, containment hydrogen

concentration monitors are required by

10 CFR 50.44(b)(1), are addressed in

NUREG–0737 Item II.F.1 and Regulatory

Guide 1.97, and are relied upon to meet

the data reporting requirements of 10

CFR Part 50, Appendix E, Section

VI.2.a.(i)(4). The staff concludes that

during the early phases of an accident,

the safety-grade hydrogen monitors

provide an adequate capability for

monitoring containment hydrogen

concentration. The staff sees value in

maintaining the capability to obtain grab

samples for complementing the

information from the hydrogen monitors

in the long term (i.e., by confirming the

indications from the monitors and

providing hydrogen measurements for

concentrations outside the range of the

monitors). As previously mentioned, the

licensee’s contingency plan (see item

4.1) for obtaining highly radioactive

samples will include sampling of the

containment atmosphere and may, if

deemed necessary and practical by the

appropriate decision-makers, be used to

supplement the safety-related hydrogen

monitors.

[Note 1—Each licensee should specify

a desired implementation period for its

specific amendment request. The

implementation period would be that

period necessary to develop and

implement the items in 4.1 through 4.3

and, as necessary, to make other

changes to documentation or equipment

to support the elimination of PASS

requirements. As an alternative, the

licensee may choose to have a shorter

implementation period and include the

scheduling of items 4.1 through 4.3 as

part of the regulatory commitments

associated with this amendment

request. Amendment requests that

include commitments for

implementation of the items in Section

4 within 6 months of the

implementation of the revised TS will

remain within the CLIIP.]

[Note 2—There may be some

collateral changes to the TS as a result

of the removal of the administrative

controls section for PASS. The

following paragraphs address three

potential changes that the staff is aware

of (editorial changes, mention of PASS

as a potential leakage source outside

containment, and revision of the bases

section for post accident monitoring

instrumentation].

(A) The elimination of the TS and

other regulatory requirements for PASS

would result in additional changes to

TS such as [e.g., the renumbering of

sections or pages or the removal of

references]. [If applicable: The

elimination of PASS requirements

requires the (elimination or

modification) of Condition [2.C.x] in the

operating license.] The changes are

included in the licensee’s application to

revise the TS in order to take advantage

of the CLIIP. The staff has reviewed the

changes and agrees that the revisions are

necessary due to the removal of the TS

section on PASS. The changes do not

revise technical requirements beyond

that reviewed by the NRC staff in

connection with the supporting topical

reports or the preparation of the TS

improvement incorporated into the

CLIIP.

(B) The TS include an administrative

requirement for a program to minimize

to levels as low as practicable the

leakage from those portions of systems

outside containment that could contain

highly radioactive fluids during a

serious transient or accident. The

program includes preventive

maintenance, periodic inspections, and

leak tests for the identified systems.

PASS is specifically listed in TS [5.5.2]

as falling under the scope of this

requirement. The applicability of this

specification depends on whether or not

PASS is maintained as a system that is

a potential leakage path. (Note that

several options (see following) exist for

handling the impact that eliminating

PASS requirements would have on the

specification for the program to control

leakage outside containment.)

(i) The licensee has stated that a plant

change will be implemented such that

PASS will not be a potential leakage

path outside containment for highly

radioactive fluids (e.g., the PASS piping

that penetrates the containment would

be cut and capped). The modification

will be made during the implementation

period for this amendment such that it

is appropriate to delete the reference to

PASS in TS [5.5.2]. Requirements in

NRC regulations (e.g., 10 CFR Part 50,

Appendix J) and other TS provide

adequate regulatory controls over the

licensee’s proposed modification to

eliminate PASS as a potential leakage

path.

(ii) The licensee has stated that a

plant change might be implemented

such that PASS would not be a potential

leakage path outside containment for

highly radioactive fluids (e.g., the PASS

piping that penetrates the containment

might be cut and capped). The

modification will not, however, be made

during the implementation period for

this amendment. The licensee has

proposed to add the following phrase to

the reference to PASS in TS [5.5.2]:

‘‘(until such time as a modification

65022 Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices

eliminates the PASS penetration as a

potential leakage path).’’

The above phrase makes clear that TS

[5.5.2] remains applicable to the PASS

as long as it is a possible leakage path

and reflects that the actual modification

of the piping system may be scheduled

beyond the implementation period for

this amendment. Requirements in NRC

regulations (10 CFR Part 50, Appendix

J) and other TS provide adequate

regulatory controls over the licensee’s

modification to eliminate PASS as a

potential leakage path. Following the

modification to eliminate PASS as a

potential leakage path, the licensee may

elect (in order to maintain clarity and

simplicity of the requirement) to revise

TS [5.5.2] to remove the reference to

PASS, including the phrase added by

this amendment.

(iii) The licensee has stated that the

configuration of the PASS will continue

to be a potential leakage path outside

containment for highly radioactive

fluids (e.g., the PASS piping will

penetrate the containment with valves

or other components in the system from

which highly radioactive fluid could

leak). The licensee has [not proposed to

change TS (5.5.2) or has changed TS

(5.5.2) to revise the reference to this

system from PASS to ( )]. The staff

agrees [that TS 5.5.2 is not affected or

that the change to revise the reference

from PASS to ( )] is acceptable. A

separate amendment request will be

required if the licensee, subsequent to

this amendment, decides to modify the

plant to eliminate this potential leakage

path and proposes to change the

requirements of TS (5.5.2)].

(C) [Note-optional section if licensee

provides markup of affected Bases

pages] The elimination of PASS affects

the discussion in the Bases section for

TS [3.3.3, ‘‘Post Accident Monitoring

Instrumentation’’].

The current Bases mention the

capabilities of PASS as part of the

justification for allowing both hydrogen

monitor channels to be out of service for

a period of up to 72 hours. Although the

licensee’s application included possible

wording for the revised Bases

discussion for TS [3.3.3], the licensee

will formally address the change to the

Bases in accordance with [the Bases

Control Program or its administrative

procedure for revising Bases].

4.0 Verifications and Commitments

As requested by the staff in the notice

of availability for this TS improvement,

the licensee has addressed the following

plant-specific verifications and

commitments.

4.1 Each licensee should verify that

it has, and make a regulatory

commitment to maintain (or make a

regulatory commitment to develop and

maintain), contingency plans for

obtaining and analyzing highly

radioactive samples of reactor coolant,

containment sump, and containment

atmosphere.

The licensee has [verified that it has

or made a regulatory commitment to

develop] contingency plans for

obtaining and analyzing highly

radioactive samples from the RCS,

containment sump, and containment

atmosphere. The licensee has

committed to maintain the contingency

plans within its [specified document or

program]. The licensee has

[implemented this commitment or will

implement this commitment by

(specified date)].

4.2 Each licensee should verify that

it has, and make a regulatory

commitment to maintain (or make a

regulatory commitment to develop and

maintain), a capability for classifying

fuel damage events at the Alert level

threshold (typically this is 300 μCi/ml

dose equivalent iodine). This capability

may utilize the normal sampling system

and/or correlations of sampling or

letdown line dose rates to coolant

concentrations.

The licensee has [verified that it has

or made a regulatory commitment to

develop] a capability for classifying fuel

damage events at the Alert level

threshold. The licensee has committed

to maintain the capability for the Alert

classification within its [specified

document or program]. The licensee has

[implemented this commitment or will

implement this commitment by

(specified date)].

4.3 Each licensee should verify that

it has, and make a regulatory

commitment to maintain (or make a

regulatory commitment to develop and

maintain), the capability to monitor

radioactive iodines that have been

released to offsite environs.

The licensee has [verified that it has

or made a regulatory commitment to

develop] the capability to monitor

radioactive iodines that have been

released to offsite environs. The licensee

has committed to maintain the

capability for monitoring iodines within

its [specified document or program].

The licensee has [implemented this

commitment or will implement this

commitment by (specified date)].

The NRC staff finds that reasonable

controls for the implementation and for

subsequent evaluation of proposed

changes pertaining to the above

regulatory commitments are provided

by the licensee’s administrative

processes, including its commitment

management program. Should the

licensee choose to incorporate a

regulatory commitment into the

emergency plan, final safety analysis

report, or other document with

established regulatory controls, the

associated regulations would define the

appropriate change-control and

reporting requirements. The staff has

determined that the commitments do

not warrant the creation of regulatory

requirements which would require prior

NRC approval of subsequent changes.

The NRC staff has agreed that NEI 99–

04, Revision 0, ‘‘Guidelines for

Managing NRC Commitment Changes,’’

provides reasonable guidance for the

control of regulatory commitments

made to the NRC staff. (See Regulatory

Issue Summary 2000–17, Managing

Regulatory Commitments Made by

Power Reactor Licensees to the NRC

Staff, dated September 21, 2000.) The

commitments should be controlled in

accordance with the industry guidance

or comparable criteria employed by a

specific licensee. The staff may choose

to verify the implementation and

maintenance of these commitments in a

future inspection or audit.

5.0 State Consultation

In accordance with the Commission’s

regulations, the [ ] State official was

notified of the proposed issuance of the

amendments. The State official had [(1)

no comments or (2) the following

comments—with subsequent

disposition by the staff].

6.0 Environmental Consideration

The amendments change a

requirement with respect to the

installation or use of a facility

component located within the restricted

area as defined in 10 CFR Part 20 and

change surveillance requirements. The

NRC staff has determined that the

amendments involve no significant

increase in the amounts and no

significant change in the types of any

effluents that may be released offsite,

and that there is no significant increase

in individual or cumulative

occupational radiation exposure. The

Commission has previously issued a

proposed finding that the amendments

involve no significant hazards

consideration, and there has been no

public comment on such finding (FR).

Accordingly, the amendments meet the

eligibility criteria for categorical

exclusion set forth in 10 CFR

51.22(c)(9). Pursuant to 10 CFR 51.22(b)

no environmental impact statement or

environmental assessment need be

prepared in connection with the

issuance of the amendments.

Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices 65023

7.0 Conclusion

The Commission has concluded,

based on the considerations discussed

above, that: (1) There is reasonable

assurance that the health and safety of

the public will not be endangered by

operation in the proposed manner, (2)

such activities will be conducted in

compliance with the Commission’s

regulations, and (3) the issuance of the

amendments will not be inimical to the

common defense and security or to the

health and safety of the public.

Model No Significant Hazards

Consideration Determination

Description of Amendment Request:

The proposed amendment deletes

requirements from the Technical

Specifications (and, as applicable, other

elements of the licensing bases) to

maintain a Post Accident Sampling

System (PASS). Licensees were

generally required to implement PASS

upgrades as described in NUREG–0737,

‘‘Clarification of TMI [Three Mile

Island] Action Plan Requirements,’’ and

Regulatory Guide 1.97,

‘‘Instrumentation for Light-WaterCooled Nuclear Power Plants to Assess

Plant and Environs Conditions During

and Following an Accident.’’

Implementation of these upgrades was

an outcome of the lessons learned from

the accident that occurred at TMI, Unit

2. Requirements related to PASS were

imposed by Order for many facilities

and were added to or included in the

technical specifications (TS) for nuclear

power reactors currently licensed to

operate. Lessons learned and

improvements implemented over the

last 20 years have shown that the

information obtained from PASS can be

readily obtained through other means or

is of little use in the assessment and

mitigation of accident conditions.

The NRC staff issued a notice of

opportunity for comment in the Federal

Register on August 11, 2000 (65 FR

49271) on possible amendments to

eliminate PASS, including a model

safety evaluation and model no

significant hazards consideration

(NSHC) determination, using the

consolidated line item improvement

process. The NRC staff subsequently

issued a notice of availability of the

models for referencing in license

amendment applications in the Federal

Register on [ ] (65 FR). The licensee

affirmed the applicability of the

following NSHC determination in its

application dated [ ].

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), an

analysis of the issue of no significant

hazards consideration is presented

below:

Criterion 1—The Proposed Change Does

Not Involve a Significant Increase in the

Probability or Consequences of an

Accident Previously Evaluated

The PASS was originally designed to

perform many sampling and analysis

functions. These functions were

designed and intended to be used in

post accident situations and were put

into place as a result of the TMI–2

accident. The specific intent of the

PASS was to provide a system that has

the capability to obtain and analyze

samples of plant fluids containing

potentially high levels of radioactivity,

without exceeding plant personnel

radiation exposure limits. Analytical

results of these samples would be used

largely for verification purposes in

aiding the plant staff in assessing the

extent of core damage and subsequent

offsite radiological dose projections. The

system was not intended to and does

not serve a function for preventing

accidents and its elimination would not

affect the probability of accidents

previously evaluated.

In the 20 years since the TMI–2

accident and the consequential

promulgation of post accident sampling

requirements, operating experience has

demonstrated that a PASS provides

little actual benefit to post accident

mitigation. Past experience has

indicated that there exists in-plant

instrumentation and methodologies

available in lieu of a PASS for collecting

and assimilating information needed to

assess core damage following an

accident. Furthermore, the

implementation of Severe Accident

Management Guidance (SAMG)

emphasizes accident management

strategies based on in-plant instruments.

These strategies provide guidance to the

plant staff for mitigation and recovery

from a severe accident. Based on current

severe accident management strategies

and guidelines, it is determined that the

PASS provides little benefit to the plant

staff in coping with an accident.

The regulatory requirements for the

PASS can be eliminated without

degrading the plant emergency

response. The emergency response, in

this sense, refers to the methodologies

used in ascertaining the condition of the

reactor core, mitigating the

consequences of an accident, assessing

and projecting offsite releases of

radioactivity, and establishing

protective action recommendations to

be communicated to offsite authorities.

The elimination of the PASS will not

prevent an accident management

strategy that meets the initial intent of

the post-TMI–2 accident guidance

through the use of the SAMGs, the

emergency plan (EP), the emergency

operating procedures (EOP), and site

survey monitoring that support

modification of emergency plan

protective action recommendations

(PARs).

Therefore, the elimination of PASS

requirements from Technical

Specifications (TS) (and other elements

of the licensing bases) does not involve

a significant increase in the

consequences of any accident

previously evaluated.

Criterion 2—The Proposed Change Does

Not Create the Possibility of a New or

Different Kind of Accident From Any

Previously Evaluated

The elimination of PASS related

requirements will not result in any

failure mode not previously analyzed.

The PASS was intended to allow for

verification of the extent of reactor core

damage and also to provide an input to

offsite dose projection calculations. The

PASS is not considered an accident

precursor, nor does its existence or

elimination have any adverse impact on

the pre-accident state of the reactor core

or post accident confinement of

radionuclides within the containment

building.

Therefore, this change does not create

the possibility of a new or different kind

of accident from any previously

evaluated.

Criterion 3—The Proposed Change Does

Not Involve a Significant Reduction in

the Margin of Safety

The elimination of the PASS, in light

of existing plant equipment,

instrumentation, procedures, and

programs that provide effective

mitigation of and recovery from reactor

accidents, results in a neutral impact to

the margin of safety. Methodologies that

are not reliant on PASS are designed to

provide rapid assessment of current

reactor core conditions and the

direction of degradation while

effectively responding to the event in

order to mitigate the consequences of

the accident. The use of a PASS is

redundant and does not provide quick

recognition of core events or rapid

response to events in progress. The

intent of the requirements established as

a result of the TMI–2 accident can be

adequately met without reliance on a

PASS.

Therefore, this change does not

involve a significant reduction in the

margin of safety.

Based upon the reasoning presented

above and the previous discussion of

the amendment request, the requested

65024 Federal Register / Vol. 65, No. 211 / Tuesday, October 31, 2000 / Notices

change does not involve a significant

hazards consideration.

Dated at Rockville, Maryland, this 25th day

of October 2000.

For the Nuclear Regulatory Commission.

William D. Beckner,

Chief, Technical Specification Branch,

Division of Regulatory Improvement

Programs, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–27941 Filed 10–30–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Draft Regulatory Guides; Issuance,

Availability

The Nuclear Regulatory Commission

has issued for public comment drafts of

two new guides in its Regulatory Guide

Series. This series has been developed

to describe and make available to the

public such information as methods

acceptable to the NRC staff for

implementing specific parts of the

NRC’s regulations, techniques used by

the staff in evaluating specific problems

or postulated accidents, and data

needed by the staff in its review of

applications for permits and licenses.

Draft Regulatory Guide DG–1102,

‘‘Design, Inspection, and Testing

Criteria for Air Filtration and

Adsorption Units of Post-Accident

Engineered-Safety-Feature Atmosphere

Cleanup Systems in Light-Water-Cooled

Nuclear Power Plants,’’ as a proposed

Revision 3 to Regulatory Guide 1.52, is

being developed to describe methods

acceptable to the NRC staff for

complying with the NRC’s regulations

with regard to the design, inspection,

and testing criteria for air filtration and

iodine adsorption units of engineeredsafety-feature atmosphere cleanup

systems in light-water-cooled nuclear

power plants. This guide applies only to

post-accident atmosphere cleanup

systems that are designed to mitigate the

consequences of postulated accidents.

Draft Regulatory Guide DG–1103,

‘‘Design, Inspection, and Testing

Criteria for Air Filtration and

Adsorption Units of Normal Ventilation

Exhaust Systems in Light-Water-Cooled

Nuclear Power Plants,’’ as a proposed

Revision 2 to Regulatory Guide 1.140, is

being developed to present methods

acceptable to the NRC staff for meeting

the NRC’s regulations with regard to the

criteria for air filtration and adsorption

units installed in the normal ventilation

exhaust systems of light-water-cooled

nuclear power plants.

These draft guides have not received

complete staff approval and do not

represent an official NRC staff position.

Comments may be accompanied by

relevant information or supporting data.

Written comments may be submitted to

the Rules and Directives Branch, Office

of Administration, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555. Copies of comments received

may be examined at the NRC Public

Document Room, 11555 Rockville Pike,

Rockville, MD. Comments will be most

helpful if received by December 29,

2000.

You may also provide comments via

the NRC’s interactive rulemaking

website through the NRC home page

(http://www.nrc.gov). 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; e-mail CAG@NRC.GOV.

Electronic copies of these draft guides,

under Accession Numbers

ML003714744 for DG–8026 and

ML003714764 for DG–8027, are

available in NRC’s Public Electronic

Reading Room, which can also be

accessed through NRC’s web site,

<WWW.NRC.GOV>. For information

about the draft guides, contact Mr. J.

Segala at (301) 415–1858; e-mail

JPS1@NRC.GOV.

Although a time limit is given for

comments on these draft guides,

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.

Regulatory guides are available for

inspection at the Commission’s Public

Document Room, 11555 Rockville Pike,

Rockville, MD. Requests for single

copies of draft or final guides (which

may be reproduced) or for placement on

an automatic distribution list for single

copies of future draft guides in specific

divisions should be made in writing to

the U.S. Nuclear Regulatory

Commission, Washington, DC 20555,

Attention: Reproduction and

Distribution Services Section; or by fax

to (301) 415–2289, or by email to

<DISTRIBUTION@NRC.GOV>.

Telephone requests cannot be

accommodated. Regulatory guides are

not copyrighted, and Commission

approval is not required to reproduce

them.

(5 U.S.C. 552(a)).

Dated at Rockville, Maryland, this 19th day

of October 2000.

For the Nuclear Regulatory Commission.

Clare V. Kasputys,

Deputy Director, Program Management,

Policy Development & Analysis Staff, Office

of Nuclear Regulatory Research.

[FR Doc. 00–27939 Filed 10–30–00; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF MANAGEMENT AND

BUDGET

Cost of Hospital and Medical Care

Treatment Furnished by the United

States; Certain Rates Regarding

Recovery From Tortiously Liable Third

Persons

By virtue of the authority vested in

the President by Section 2(a) of Public

Law 87–693 (76 Stat. 593; 42 U.S.C.

2652), and delegated to the Director of

the Office of Management and Budget

by Executive Order No. 11541 of July 1,

1970 (35 FR 10737), the three sets of

rates outlined below are hereby

established. These rates are for use in

connection with the recovery, from

tortiously liable third persons, of the

cost of hospital and medical care and

treatment furnished by the United States

(Part 43, Chapter I, Title 28, Code of

Federal Regulations) through three

separate Federal agencies. The rates

have been established in accordance

with the requirements of OMB Circular

A–25, requiring reimbursement of the

full cost of all services provided. The

rates are established as follows:

1. Department of Defense

The FY 2001 Department of Defense

(DoD) reimbursement rates for inpatient,

outpatient, and other services are

provided in accordance with Title 10,

United States Code, section 1095. Due to

size, the sections containing the Drug

Reimbursement Rates (section IV.C.)

and the rates for Ancillary Services

Requested by Outside Providers (section

IV.D.) are not included in this package.

Those rates are available from the

TRICARE Management Activity’s

Uniform Business Office website, http:/

/www.tricare.osd.mil/ebc/rm/

rmlhome.html. The medical and

dental service rates in this package

(including the rates for ancillary

services and other procedures requested

by outside providers) are effective

October 1, 2000. Pharmacy rates are

updated on an as needed basis.

2. Health and Human Services

The FY 2001 tortiously liable rates for

Indian Health Service health facilities

are based on Medicare cost reports. The

obligations for the Indian Health Service

hospitals participating in the cost report