NUREG-0737

Supplement No. 1

Clarification of

TMI Action Plan Requirements

Requirements for Emergency Response Capability

U.S. Nuclear Regulatory

Commission

Office of Nuclear Reactor Regulation

06i0l REG,,'(

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Reprinted February 1989

UNITED STATES

NUCLEAR REGULATORY COMMISSION

g ,WASHINGTON, D. C. 20555

June 15, 1982

TO ALL LICENSEES OF OPERATING PLANTS, APPLICANTS FOR AN

OPERATING LICENSE, AND HOLDERS OF CONSTRUCTION PERMITS

Gentlemen:

SUBJECT: NUCLEAR POWER PLANT STAFF WORKING HOURS

(Generic Letter No. 82-12)

On February 8, 1982, we forwarded to you a copy of a policý ,.atement

issued by the Commission entitled "Policy on Factors Causing Fatigue

of Operating Personnel at Nuclear Reactors." That policy statement

has now been slightly revised and the revised version is attached to

this letter. The forwarding letter (Generic Letter 82-02) stated that

actions were underway to incorporate the Commission policy on working

hours into Regulatory Guide 1.33 and into NUREG-0737 (Item I.A.l.3).

Publication of Regulatory Guide 1.33 is now anticipated in June 1982,

following release of ANS 3.2. Revised pages of NUREG-0737 that incorporate the Commission policy on working hours are enclosed.

Our letter of February 8, 1982 requested that you take action as

necessary to revise the administrative section of your technical specifications to assure that your plant administrative procedures follow

the revised working hour guidelines, including a provision for documentation of authorized deviations which should be available for NRC

review. You should review your past actions to assure that they are

consistent with the attached revised policy statement. Note that the

revised guidelines are to be incorporated by October 1, 1982.

Sincerely,

/. Darrell G. Eisenhut, Director Division of Licensing

Office of Nuclear Reactor Regulation

Attachment:

Revision to Item I.A.1.3

of NUREG-0737

I.A.I.3 SHIFT MANNING

Pages I.A.l.3-1 through I.A.l.3-3 are replaced in their entirety by the following.

A bar in the margin indicates changes from the material originally included on

these pages.

Position

This position defines shift manning requirements for normal operation. The letter

of July 31, 1980 from D. G. Eisenhut to all power reactor licensees and applicants

(copy attached) sets forth the interim criteria for shift staffing (to be effective

pending general criteria that will be the subject of future rulemaking). Overtime

restrictions were also included in the July 31, 1980 letter.

Changes to Previous Requirements and Guidance

Errors were discovered in the last column of the table attached to the letter of

July 31, 1980. A corrected table is enclosed; a bar in the margin indicates the

correction. (See p. I.A.1.3-4.)

The overtime requirements have been rewritten to incorporate the provisions of a

Commission policy statement on working hours and operator fatigue.

Clarification

Licensees of operating plants and applicants for operating licenses shall include

in their administrative procedures (required by license conditions) provisions

governing required shift staffing and movement of key individuals about the plant.

These provisions are required to assure that qualified plant personnel to man the

operational shifts are readily available in the event of an abnormal or emergency

situation.
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These administrative procedures shall also set forth a policy, the objective of

which is to prevent situations where fatigue could reduce the ability of operating

personnel to keep the reactor in a safe condition. The controls established

should assure that, to the extent practicable, personnel are not assigned to

shift duties while in a fatigued condition that could significantly reduce their

mental alertness or their decision making ability. The controls shall apply to

the plant staff who perform safety-related functions (e.g., senior reactor

operators, reactor operators, auxiliary operators, health physicists, and key

maintenance personnel).

IE Circular No. 80-02, "Nuclear Power Plant Staff Work Hours", dated February 1,

1980 (copy attached) discusses the concern of overtime work for members of the

plant staff who perform safety-related functions. The guidance contained in

IE Circular No. 80-02 was amended by the July 31, 1980 letter. In turn, the

overtime guidance of the July 31, 1980 letter was revised in Section I.A.l.3

of NURG-0737. The NRC has issued a policy statement (attached) which further

revises the overtime guidance as stated in NUREG-0737. This guidance is as

follows:

Enough plant operating personnel should be employed to maintain adequate shift

coverage without routine heavy use of overtime. The objective is to have operating

personnel work a normal 8-hour day, 40-hour week while the plant is operating.

However, in the event that unforeseen problems require substantial amounts of

overtime to be used, or during extended periods of shutdown for refueling, major

maintenance or major plant modifications, on a temporary basis, the following

guidelines shall be followed:
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a. An individual should not be permitted to work more than 16 hours straight

(excluding shift turnover time).

b. An individual should not be permitted to work more than 16 hours in any

24-hour period, nor more than 24 hours in any 48-hour period, nor more

than 72 hours-in any seven day period (all excluding shift turnover time).

c. A break of at least eight hours should be allowed between work periods

(including shift turnover time).

d. Except during extended shutdown periods, the use of overtime should be

considered on an individual basis and not for the entire staff on shift.

Recognizing that very unusual circumstances may arise requiring deviation from

the above guidelines, such deviation shall be authorized by the plant manager

or his deputy, or higher levels of management. The paramount consideration in

such authorization shall be that significant reductions in the effectiveness of

operating personnel would be highly unlikely. Authorized deviations to the

working hour guidelines shall be documented and available for NRC review.

In addition, procedures are encouraged that would allow licensed operators at

the controls to be periodically relieved and assigned to other duties away from

the control board during their tours of duty.

Operating license applicants shall complete these administrative procedures

before fuel loading. Development and implementation of the administrative

procedures at operating plants will be reviewed by the Office of Inspection

and Enforcement beginning Oct. 1, 1982.
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See Section III.A.l.2 for minimum staffing and augment capabilities for

emergencies.

Applicability

This requirement applies to all licensees of operating reactors and applicants

for operating licenses.

Implementation

(1) Overtime administrative procedures shall be established for operating

reactors by Oct. 1, 1982 and by fuel loading for applicants for operating

license.

(2) Staffing requirements shall be completed by July 1, 1982 for operating

reactors and by fuel load for operating license applicants.

Type of Review

A post-implementation review will be performed on operating reactors.

Applicants for operating licenses will be reviewed prior to implementation.

Documentation Required

The documentation required is as noted in the letter of July 31, 1980.

Technical Specification Changes Required

Changes to Technical Specifications will be required for overtime administrative

procedure and staffinq requirements.
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References

NUREG-0660

IE Circular No. 80-02, "Nuclear Power Plant Staff Work Hours," February 1, 1980

Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees, July 31, 1980

Policy on Factors Causing Fatigue of Operating Personnel at Nuclear Reactors,
POLICY ON FACTORS CAUSING FATIGUE OF OPERATING

PERSONNEL AT NUCLEAR REACTORS

Licensees of operating plants and applicants for operating licenses shall

establish controls to prevent situations where fatigue could reduce the ability

of operating personnel to keep the reactor in a safe condition. The controls

should focus on shift staffing and the use of overtime--key job-related factors

that influence fatigue.

The objective of the controls would be tO assure that, to the extent practicable,

personnel are not assigned to shift duties while in a fatigued condition that

could significantly reduce their mental alertness or their decision making

capability. The controls shall apply to the plant staff who perform safetyrelated functions (e.g., senior reactor operators, reactor operators, health

physicists, auxiliary operators, and key maintenance personnel).

Enough plant operating personnel should be employed to maintain adequate shift

coverage without routine heavy use of overtime. The objective is to have

operating personnel work a normal. 8-hour day, 40-hour week while the plant is

operating. However, in the event that unforeseen problems require substantial

amounts of overtime to be used, or during extended periods of shutdown for

refueling, major maintenance or major plant modifications, on a temporary basis,

the following guidelines shall be followed:

a. An individual should not be permitted to work more than 16 hours

straight (excluding shift turnover time).

b. An individual should not be permitted to work more than 16 hours

in any 24-hour period, nor more than 24 hours in any 48-hour

period, nor more than 72 hours in any seven day period (all

excluding shift turnover time).

c. A break of at least eight hours should be allowed between work

periods (including shift turnover time).

d. Except during extended shutdown periods, the use of overtime

should be considered on an individual basis and not for the

entire staff on a shift.

Recognizing that very unusual circumstances may arise requiring deviation from

the above guidelines, such deviation shall be authorized by the plant manager

or his deputy, or higher levels of management. The paramount consideration

in such authorization shall be that significant reductions in the effectiveness

of operating personnel would be highly unlikely.

In addition, procedures are encouraged that would allow licensed operators at the

controls to be periodically relieved and assigned to other duties away from the

control board during their tour of duty.
NUREG-0737

Supplement No. 1

Clarification of

TMI Action Plan Requirements

Requirements for Emergency Response Capability

Manuscript Completed: December 1982

Date Published: January 1983

Division of Licensing

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

÷o° G

'4t

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ABSTRACT

This document, Supplement I to NUREG-0737, is a letter from D. G. Eisenhut,

Director of the Division of Licensing, NRR, to licensees of operating power

reactors, applicants for operating licenses, and holders of construction

permits forwarding post-TMI requirements for emergency response capability

which have been approved for implementation. On October 30, 1980, the NRC

staff issued NUREG-0737, which incorporated into one document all TMIrelated items approved for implementation by the Commission at that time.

In this NRC report, additional clarification is provided regarding Safety

Parameter Display Systems, Detailed Control Room Design Reviews, Regulatory

Guide 1.97 (Revision 2) - Application to Emergency Response Facilities,

Upgrade of Emergency Operating Procedures, Emergency Response Facilities,

and Meteorological Data.

,'i,

S o UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

December 17, 1982

TO ALL LICENSEES OF OPERATING REACTORS, APPLICANTS FOR OPERATING

LICENSES, AND HOLDERS OF CONSTRUCTION PERMITS

Gentlemen:

SUBJECT: SUPPLEMENT 1 TO NUREG-0737 - REQUIREMENTS FOR EMERGENCY

RESPONSE CAPABILITY (GENERIC LETTER NO. 82-33)

On October 31, 1980, the NRC staff issued NUREG-0737, which incorporated

into one document all TMI-related items approved for implementation by

the Commission at that time. The purpose of this letter is to provide

additional clarification regarding Safety Parameter Display Systems,

Detailed Control Room Design Reviews, Regulatory Guide 1.97 (Revision 2) -

Application to Emergency Response Facilities, Upgrade of Emergency

Operating Procedures, Emergency Response Facilities, e z Meteorological

Data.

The enclosures to this letter are a distillation of the basic requirements

for these topics from the broad range of guidance documents that the NRC

has issued (principally NUREG reports and Regulatory Guides). It is our

intent that the guidance documents themselves, referred to in the enclosures, are not to be used as requirements, but rather that they are to be

used as sources of guidance for NRC reviewers and licensees regarding

acceptable means for meeting the basic requirements.

The following items in NUREG-0737 are affected:

I.C.l Guidance for the Evaluation and Development of Procedures for

Transients and Accidents

I.D.l Control Room Design Reviews

I.D.2 Plant Safety Parameter Display Console

III.A.1.2 Upgrade Emergency Support Facilities.

III.A.2.2 Meteorological Data

The requirements and guidance contained in the enclosure to this letter

replace the corresponding requirements in the affected NUREG-0737 items

and should be used by you in meeting the goals of these action plan items.

You should also note that the staffing levels in table 2 to the enclosure

are only goals, and are not strict requirements.

8212060349
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You will note that the enclosure does not specify a schedule for completing

the requirements. It has become apparent, through discussions with owners'

groups and individual licensees, that our previous schedules did not adequately consider the integration of these related activities. In recognition of this and the difficulty in implementing generic deadlines, the

Commission has adopted a plan to establish realistic plant-specific schedules

that take into account the unique aspects of the work at each plant. By

this plan, each licensee is to develop and submit its own plant-specific

schedule which will be reviewed by the assigned NRC Project Manager. The

NRC Project Manager and licensee will reach an agreement on the final

schedule and in this manner provide for prompt implementation of these

important improvements while optimizing the use of utility and NRC resources.

Applicants for construction permits are expected to comply with the requirements of 10 CFR 50.34(f), and should consider this document to be additional

guidance in meeting these requirements. For holders of construction permits

and applicants for operating licenses, plant-specific schedules for the

implementation of these requirements will be developed in a manner similar

to that being used for operating reactors, taking into consideration the

degree of completion of the power plant.

In order to answer questions you may have regarding the Commission's policy

on these issues and the implementation process to be used by project managers,

regional workshops will be conducted by senior staff members according to the

following schedule:

Region I Washington, D. C. - Week of 2/14/83

Region II Atlanta, Ga. - Week of 2/21/83

Region III Chicago, Ill. - Week of 2/21/83

Region IV & V San Francisco, CA - Week of 2/28/83

You will be notified of specific locations and times for the workshops at

a later time.

Accordingly, pursuant to 50.54(f), operating reactor licensees and holders

of construction permits are requested to furnish, no later than April 15, 1983

a 7-oposed schedule for completing each of the basic requirements for

th. tems identified in the enclosures to this letter. You are encouraged

to ,ork closely with your NRC Project Manager during this process so that

we can reach an agreement on the final schedule as quickly as possible. In

addition, you are requested to submit with it a description of your plans

for phased implementation and integration of the emergency response activities.

Your plans for integration will be reviewed as part of our evaluation of

your proposed schedule. After the staff completes this evaluation, it will

take action, as necessary, to assure that such requirements and commitments

are appropriately enforceable.
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This request for information was approved by the Office of Management

and Budget under clearance number 3150-0065 which expires May 31, 1983.

Comments on burden and duplication may be directed to the Office of

Management and Budget, Reports Management Room 3208, New Executive Office

Building, Washington, D. C. 20503.

\Sincerely,

" '1~a~e •." isenhut, Director

Division of Licensing

Office of Nuclear Reactor Regulation

Enclosure:

Supplement to NUREG-0737

CONTENTS

Pages

1. INTRODUCTION ........ ................ *.. . . . 1

2. USE OF EXISTING DOCUMENTATION ............. ............ 3

3. COORDINATION AND INTEGRATION OF INITIATIVES .... ....... 4

4. SAFETY PARAMETER DISPLAY SYSTEM (SPDS) . . . . . .... . . . 7

o Requirements

o Documentation and NRC Review

o Integration

o Reference Documents

5. DETAILED CONTROL ROOM DESIGN REVIEW ....... ........... 10

o Requirements

o Implementation Scnedule

o Documentation and NRC Review

o Reference Documents

6. REGULATORY GUIDE 1.97 - APPLICATION TO EMERGENCY RESPONSE

FACILITIES ............ ...................... . . . 13

o FRequirements

o Documentation and NRC Review

7. UPGRADE EMERGENCY OPERATING PROCEDURES (EOPs) ......... 15

o Requirements

o DoCumentation and NRC Review

o Reference Documents

8. EMERGENCY RESPONSE FACILITIES ....... .............. 17

o Regulations

o Technical Support Center .......... ................ 19

o Requirements

o Op.: t-tional Support Center ..... ............... .... 21

o Requirements
CONTENTS (Continued)

Pages

o Emergency Operations Facility ..... ........ ....... 22

o Requirements

o Documentation and NRC Review

o Reference Documents.

Table I - Emergency Operations Facility Location Options . . . 26

Table 2 - Minimum Staffing for NRC Licensees for Nuclear Power

Plant Emergencies ..... ................. .... 27
EMERGENCY RESPONSE CAPABILITY

I. INTRODUCTION

This supplement was prepared as a result of a review by the Comnittee

to Review Generic Requirements (CRGR). The supplement represents the

staff's attempt to distill the fundamental requirements for nuclear

plant Emergency Response Capability fronr the wide range of guidance

documents that the NRC has issued. It is not intended that these guidance

documents (NUREG reports and Regulatory Guides) be implemented as written;

rather, they should be regarded as useful sources of guidance for licensees and NRC staff regarding acceptable means for meeting the fundamental

requirements contained in this document. It is also not intended that

either the guidance documents or the fundamental requirements are to be

considered binding legal requi.rements at this time. As indicated below,

however, the fundamental requirements will be translated into binding

legal requirements in the manner specified.

These requirements are a further delineation of the general guidance

issued previously by the Commission in its regulations, orders and policy

statements on emergency planning and TMI issues. It is intended that

these requirements would be applicable to licensees of operating nuclear

power plants. For applicants for a construction permit (CP) or manufacturing license (ML), the requirements described in this document must be

supplemented with the specific provisions in the rule specifying licensing

requirements for pending CP and ML applications. Thus, compliance with

requirements in this document may not be sufficient to meet the related

requirements in 10 CFR 50.34(f) and Appendix E. In this regard, it is

expected that the staff would review CP and ML applications against the

guidance in the current Standard Review Plan (which includes the provisions

of NUREG-0718) and this might lead to more detailed requirements than prescribed in this document in order to satisfy the requirements of 50.34(f)

and Appendix E.

Based on discussions with licensees, the staff has learned that many of the

Commission approved schedules for emergency response facilities probably will

not be met. In recognition of this fact and the difficulty of implementing

generic deadlines, plant-specific schedules will be established which take

into account the unique status of each plant. The following sequence for

developing implementation schedules will be used.

The requirements for emergency response capabilities and facilities are being

transmitted to licensees by this supplement and are being promulgated to NRC

staff. The letter which forwards this supplement requests that licensees submit

a proposed schedule for completing actions to comply with the requirements.
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Each licensee's proposed schedule will then be reviewed by the assigned

NRC Project Manager, who will discuss the subject with the licensee and

mutually agree on schedules and completion dates. The implementation

dates will then be formalized into an enforceable document.

The requirements in this document do not alter previously issued guidance,

which remains in effect. This document does attempt to place that guidance

in perspective by identifying the elements that the NRC staff believes to

be essential to upgrade emergency response capabilities. The proposal to

formalize implementation dates' in an enforceable document reflects the level

of importance which the NRC staff attributes to these requirements. The

Commission does not believe that existing guidance should be imposed in this

manner, but rather that it be used as guidance to be considered in upgrading

emergency response capabilities. This indicates the distinction which the

staff believes should be made between the requirements and guidance.

The following sections describe the requirements, their interrelationhips,

and NRC actions to improve management of emergency response regulations.

Reference documents are cited with a description of content as it relates

to specific initiatives.

The requirements set forth in this document have been reviewed by the Commission and, at a meeting held July 16, 1982, were approved by the Commission as

appropriately clarifying and providing greater detail with respect to related

TMI Action Plan requirements contained in NUREG-0737 for all operating license

applicants. These requirements are, therefore, to be accorded the status of

approved NUREG-0737 items as set forth in the Commission's "Statement of Policy:

Further Commission Guidance for Power Reactor Operating Licenses" (45 FR 85236),

December 24, 1980). In this connection, the provisions for scheduling set forth

herein supersede any schedules with respect to such items contained in NUREG-U7.37.

Accordingly, the requirements should be used by the staff and by adjudicatory

boards as appropriate clarifications and interpretation of the related NUREG-0737

items.

The requirements set forth in this document are believed to be consistent with

the requirements regarding related items for construction permits and manufacturing licenses contained in 10 CFR 50.34(f) and 10 CFR Part 50, Appendix E. Accordingly, no changes to these regulations are required.
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2. USE OF EXISTING DOCUMENTATION

The following NUREG documents are intended to be used as sources of guidance

and information, and the Regulatory Guides are to be considered as guidance

or as an acceptable approach to meeting formal requirements. The items by

virtue of their inclusion in these documents shall not be misconstrued as

requirements to be levied on licensees or as inflexible criteria to be used

by NRC staff reviewers.

NUREG Report Titles

0696 Functional Criteria for Emergency Response Facilities

0700 Guidelines for Control Room Design Reviews

0799 Draft Criteria for Preparation of Emergency Operating

Procedures (to be superseded by NUREG-0899)

0801 - Evaluation Criteria for Detailed Control Room Design

Reviews

0814 - Methodology for Evaluation of Emergency Response Facilities

0818 - Emergency Action Levels for Light Water Reactors

0835 - Human Factors Acceptance Criteria for SPDS

0899 - Guidelines for the Preparation of Emergency Operating

Procedures: Resolution of Comments on NUREG-0799

Regulatory

Guides Titles

1.23 Meteorological Measurement Program for Nuclear Power

(Rev. 1) Plants

1.97 Instrumentation for Light-Water Cooled Nuclear Power

(Rev. 2) Plants to Assess Plant and Environs Conditions During

and Following an Accident

1.101 Emergency Planning for Nuclear Power Plants

(Rev. 2)

1.47 Bypassed and Inoperable Status Indication for Nuclear

Power Plant Safety Systems
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3. COORDINATION AND INTEGRATION OF INITIATIVES

3.1 The design of the Safety Parameter Display System (SPDS), design of

instrument displays based on Regulatory Guide 1.97 guidance, control

room design review, development of function oriented emergency operating

procedures, and operating staff training should be integrated with

respect to the overall enhancement of operator ability to comprehend

plant conditions and cope with emergencies. Assessment of information

needs and display formats and locations should be performed by individual

licensees. The SPDS could affect other control room improvements that

licensees may consider. In some cases, a good SPDS may obviate the need

for large-scale control room modifications. Installation of the SPDS

should not be delayed by slower progress on other initiatives, and should

not be contingent on completion of the control room design review. Nor

should other initiatives, such as upgraded emergency operating procedures,

be impacted by delays in SPDS procurement. While the NRC does not plan

to impose additional requirements on licensees regarding SPDS, the NRC

will work with the industry to assure the development of appropriate industry

standards. for SPDS systems.

3.2 Implementation of part or all of Regulatory Guide 1.97 (Rev. 2) represents

a control room improvement. The implementation of control room improvements is not contingent on implementing Technical Support Center (TSC) and

Emergency Operations Facility (EOF) requirements.

3.3 The Technical Support Center (TSC) and EmergencyOperations Facility (EOF)

are dependent on control -room improvements in terms of communication and

instrumentation needs among the TSC, EOF, and control room. TSC and EOF

facilities are not necessarily dependent on each other. The Operational

Support Center (USC) is independent of TSC and EOF.

3.4 The three groups of initiatives--SPDS, control room improvements, and

emergency response facilities (TSC, EOF, OSC)-- have the following interrelationships:

a. The SPDS is an improvement because it enhances operator ability to

comprehend plant conditions and interact in situations that require

human intervention. The SPDS could affect other control room improvements that licensees may consider. In some cases, a good SPDS could

obviate the need for extensive modifications to control rooms.

b. New instrumentation that may be added to the control room should be

considered a requirement for inclusion in the design of the TSC and

EOF only to the extent that such instrumentation is essential to the

performance of TSC and EOF functions.
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c. The SPDS and control room improvements are essential elements in

operator training programs and the upgraded plant-specific emergency operating procedures.

d. Acquisition, processing, and management of data for SPUS, control

room improvements, and emergency response facilities should be

coordi nated.

3.5 Specific implementation plans and reasonable, achievable schedules for

improvements that will satisfy the requirements will be established by

agreement between the NRC Project Manager and each individual licensee.

The NRC office responsible for implementing each requirement will develop procedures identifying the following.

a. The respective roles of NRR, IE, and Regional Offices in managing

implementation, checking licensee rate of progress, and verifying

compliance, including the extent to which NRC review and inspection

is necessary during implementation.

b. Procedural methods and enforcement measures that could be used to

ensure NRC staff and licensee attention to meeting mutually agreed

upon schedules without significant delays and extensions.

3.6 The NRC Project Manager for each nuclear power plant is assigned program management responsibility for NRC staff actions associated with

implementing emergency response initiatives. The NRC Project Manager

is the principal contact for the licensee regarding these initiatives.

3.7 The NRC will make allowances for work already done by licensees in a

good-faith effort to meet requirements as they understand them. For

each case in which a licensee would have to remove or rip out emergency

response facilities or equipment that was installed in good faith to

meet previous guidance in order to meet the basic requirements described

in this document, the Director of the Office of Nuclear Reactor Regulation

or Inspection and Enforcement will review the circumstances and determine

whether removal is necessary or existing facilities or equipment represent an acceptable alternative. Any regulatory position that would

require the removal or major modification of existing emergency response

facilities or equipment requires the specific approval of the responsible

Office Director.

3.8 The NRC recognizes that acceptable alternative methods of phasing and

integrating emergency response activities may be developed. Each licensee

needs flexibility in integrating these activities, taking into account the

varying degree to which the licensee has implemented past requirements and
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guidance. An example of a way in which these activities could be

integrated is discussed below. Other methods of integration proposed

by licensees would be reviewed considering licensees' progress on

each initiative.

a. SPDS

(1) Review the functions of the nuclear power plant operating

staff that are necessary to recognize and cope with rare

events that (a) pose significant contributions to risk,

(b) could cause operators to make cognitive errors in diagnosing them, and (c) are not included in routine operator

training programs.

(2) Combine the results of this review with accepted human

factors principles to select parameters, data display,

and functions to be incorporated in the SPDS.

(3) Design, build, and install the SPDS in the control room and

train its users.

b. To be done in parallel without delaying SPOS, complete emergency

operating procedure technical guidelines that will be used to

develop plant-specific emergency operating procedures.

c. Using these EOP technical guidelines, the SPDS design, and accepted

human factors principles, conduct a review of the control room

design. Apply the results of this review to:

(1) Verify SPDS parameter selection, data display, and functions.

(2) Develop plant-specific EOPs.

(3) Design control room modifications that correct conditions

adverse to safety (reduce significant contributions to risk),

and add additional instrumentation that may be necessary to

implement Regulatory Guide 1.97.

(4) Train and qualify plant operating staff regarding upgraded EOPs

and modifications.

d. Verify, prior to finalization of designs for modifications and of

procedures and training, that the functions of control room operators

in emergencies can be accomplished (i.e., that the individual initiatives have been integrated sufficiently to meet the needs of control room operators and provide adequate emergency respunse capabilities).

e. Implement EOPs and install control room modifications coincident

with scheduled outages as necessary, and train operators in

advance of these changes as they are phased into operation.
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4. SAFETY PARAMETER DISPLAY SYSTEM (SPDS)

4.1 Requirements

a. The SPDS should provide a concise display of critical plant

variables to the control room operators to aid them'in rapidly

and reliably determining the safety status of the plant.

Although the SPDS will be operated during normal operations

as well as during abnormal conditions, the principal purpose

and function of the SPDS is to aid the control room personnel

during abnormal and emergency conditions in determining the

safety status of the plant and in assessing whether abnormal

conditions warrant corrective action by operators to avoid

a degraded core. This can be particularly important during

anticipated transients and the initial phase of an accident.

b. Each operating reactor shall be provided with a Safety Parameter

Display System that is located convenient to the control room

operators. This system will continuously display information

from which the plant safety status can be readily and reliably

assessed by control room personnel who are responsible for the

avoidance of degraded and damaged core events.

c. The control room instrumentation required (see General Design

Criteria 13 and 19 of Appendix A to 10 CFR 50) provides the

operators with the information necessary for safe reactor

operation under normal, transient, and accident conditions.

The SPDS is used in addition to the basic components and serves

to aid and augment these components. Thus, requirements applicable to control room instrumentation are not needed for this

augmentation (e.g., GOC 2, 3, 4 in Appendix A; 10 CFR Part 100;

single-failure requirements). The SPDS need not meet requirements

of the single-failure criteria and it need not be qualified to

meet Class IE requirements. The SPDS shall be suitably isolated

from electrical or electronic interference with equipment and

sensors that are in use for safety systems. The SPDS need not be

seismically qualified, and additional seismically qualified indication is not required for the sole purpose of being a backup for

SPDS. Procedures which describe the timely and correct safety

status assessment when the SPDS is and is not available, will be

developed by the licensee in parallel with the SPDS. Furthermore,

operators should be trained to respond to accident conditions both

with and without the SPDS available.

d. There is a wide range of useful information that can be provided

by various systems. This information is reflected in such staff

documents as NUREG-0696, NUREG-0835, and Regulatory Guide 1.97
-8-

Prompt implementation of an SPDS can provide an important contribution *to plant safety. The selection of specific information

that should be provided for a particular plant shall be based on

engineering judgement of individual plant licensees, taking into

account the importance of prompt implementation.

e. The SPDS display shall be designed to incorporate accepted human

factors principles so that the displayed information can be

readily preceived and comprehended by SPDS users.

f. The minimum information to be provided shall be sufficient to

provide information to plant operators about:

(i) Reactivity control

(ii) Reactor core cooling and heat removal from the primary

system

(iii) Reactor coolant system integrity

(iv) Radioactivity control..

(v) Containment conditions

The specific paramenters to be displayed shall be determined by

the licensee.

4.2 Documentation and NRC Review

a. The licensee shall prepare a written safety analysis describing

the basis on which the selected parameters are sufficient to

assess the safety status of each identified function for a wide

range of events, which include symptoms of severe accidents.

Such analysis, along with the specific implementation plan for

SPDS shall be reviewed as described below.

b. The licensee's proposed implementation of an SPDS system shall be

reviewed in accordance with the licensee's technical specifications to determined whether the changes involve an unreviewed

safety question or change of technical specifications. If they

do, the shall be processed in the normal fashion with prior NRC

review. If the changes do not involve an unreviewed safety question or a change in the technical specifications, the licensee

may implement such changes without prior approval by NRC or may

request a pre-implementation review and approval. If the changes

are to be implemented without prior NRC approval, the licensee's

analysis shall be submitted to NRC promptly on completion of

review by the licensee's offsite safety review committee. Based

on the results of NRC review, the Director of IE or the Director

of NRR may request or direct the licensee to cease implementation

if a serious safety question is posed by the licensee's proposed

system, or if the licensee's analysis is seriously inadequate.
-9-

4.3 Integration

Prompt implementation of an SPDS is a design goal and of primary

importance. The schedule for implementing SPDS should not be impacted

by schedules for the control room design review and development of

function-oriented emergency operating procedures. For this reason,

licensees should develop and propose an integrated schedule for

implementation in which the SPDS design is an input to the other

initiatives. If reasonable, this schedule will be accepted by NRC.

4.4 Reference Documents

NUREG-0660

NUREG-0737

NUREG-0696

NUREG-0835

Reg. Guide 1.97

(Rev. 2)

-- Need for SPDS identified

-- Specified SPDS

-- Functional Criteria for SPDS

-- Specific acceptance criteria keyed to

NUREG-0696

Instrumentation for Light-Water Cooled

Nuclear Power Plants to Assess Plant and

Environs Conditions During and Following

an Accident
- 10 -

5. DETAILED CONTROL'ROOM DESIGN REVIEW

5.1 Requirements

a. The objective of the control room design reviewis to "improve

the ability of nuclear power plant control room operators to prevent accidents or cope with accidents if they occur by improving

the information provided to them" (from NUREG-0660, Item I.D.l).

As a complement to improvements of plant operating staff capabilities in response.to transients and other abnormal conditions

that will result from implementation of the SPDS and from upgraded emergency operating procedures, this design review will

identify any modifications of control room configurations that

would contribute to a significant reduction of risk and enhancement

in the safety of operation. Decisions to modify the control room

would include consideration of long-term risk reduction and any

potential temporary decline in safety after modifications resulting

from the need to relearn maintenance and operating procedures.

This should be carefully reviewed by persons competent in human

factors engineering and risk analysis.

b. Conduct a control room.design review to identify human engineering

discrepancies. The review shall consist of:

(i) The establishment of a qualified multidisciplinary review

team and a review program incorporating accepted human

engineering principles.

(ii) The use of function and task analysis (that had been used

as the basis for developing emergency operating procedures

Technical Guidelines and plant specific emergency operating

procedures) to identify control room operator tasks and

information and control requirements during emergency

operations. This analysis has multiple purposes and should

also serve as the basis for developing training and staffing

needs and verifying SPDS parameters.

(iii) A comparison of the display and control requirements with a

control room inventory to identify missing displays and

controls.

(iv) A control room survey to identify deviations from accepted

human factors principles. This survey will include, among

other things, an assessment of the control room layout,

the usefulness of audible and visual alarm systmes, the

information recording and recall capability, and the

control room environment.
- 11 -

c. Assess which human engineering discrepancies are significant and

should be corrected. Select design improvements that will correct

those discrepancies. Improvements that can be accomplished with

an enhancement program (paint-tape-label) should be done promptly.

d. Verify that each selected design improvement will provide the

necessary correction, and can be introduced in the control room

without creating any unacceptable human engineering discrepancies

because of significant contribution to increased risk, unreviewed

safety questions, or situations in which a temporary reduction in

safety could occur. Improvements that are introduced should be

coordinated with changes resulting from other improvement programs

such as SPDS, operator training, new instrumentation (Reg. Guide

1.97, Rev. 2), and upgraded emergency operating procedures.

5.2 Documentation and NRC Review

a. All licensees shall submit a program plan within two months of

the start of the control room review that describes how items 1,

2 and 3 above will be accomplished. The staff will review the

program plans as licensees conduct their reviews, and selected

licensee will undergo an in-progress audit by the NRR human

factors staff based on the program plans and advice from resident

inspectors and Project Managers.

b. All licensees shall submit a summary report of the completed review

outlining proposed control room changes, including their proposed

sche,-J.es for implementation. The report will also provide a

summary justification for human engineering discrepancies with

safety significance to be left uncorrected or partially corrected.

c. The staff will review the summary reports, and within two weeks

after receipt of the licensee's summary report, will inform licensee2 whether a pre-implementation onsite audit will be conducted.

The decision will be based on the content of the program plan, the

summary report, and the results of NRR in-progress audits, if any.

The licensee selection for pre-implementation audit may or may not

include licensees selected for in-progress audits under paragraph 1.

d. For control rooms selected for pre-implementation onsite audit,

within one month after receipt of the summary report, the NRC will

conduct:

(i) A pre-implementation audit of proposed modifications (e.g.,

equipment additions, deletions and relocations, and proposed

modifications).
- 12 -

(ii) An audit of the justification for those human engineering

discrepancies of safety significance to be left uncorrected

or only partially corrected.

The audit will consist of a review of the licensee's record of the

control room reviews, discussions'with the licensee review team,

and usually a control room visit. Within a month-after this

onsite audit, NRC will issue its safety evaluation report (SER).

e. For control rooms for which NRC does not perform a preimplementation onsite audit, NRC will conduct a review and issue

its SER within two months after receipt of the licensee's summary

report. The-review shall be similar to that conducted for preimplementation plants under paragraph 4 above, except that it does

not include a specific audit. The-SER shall indicate whether,

based on the review carried out, changes in the licensee's modification plan are needed to assure operational safety. Flexibility

is considered in the control room review, because certain control

board discrepancies can be overcome by techniques not involving

control board changes. These techniques could include improved

procedures, improved training, or the SPDS.

f. The following approach will be used for OL review. For OL applications with SSER dates prior to June 1983, licensing may be

based on either a Preliminary Design Assessment or a Control

Room Design Review (CRDR) at the applicant's option. However,

applicants who choose the Preliminary Design Assessment option

are required to perform a CRDR after licensing. For applications

with SSER dated after June 1983, Control Room Design Review

will be required prior to licensing.

g. After the staff has issued an SER and licensees have addressed any

open issues, they may begin their upgrade according to an approved

schedule that has been negotiated with the staff.

5.3 Reference Documents

NUREG-0585 -- States that licensees should conduct review.

NUREG-0660 -- States that NRR will require reviews for

(Rev. 1) operating reactors and operating licensee

applicants.

NUREG-0700 -- Final guidelines for CRDR.

NUREG-0737 -- States that requirement was issued June,

1980, final guidance not yet issued.

NUREG-0801 -- Staff evaluation criteria.
- 13 -

6. REGULATORY GUIDE 1.97 - APPLICATION TO EMERGENCY RESPONSE FACILITIES

6.1 Requirements

a. Functional Statement

Regulatory Guide 1.97 provides data to assist control room

operators in preventing and mitigating the consequences of

reactor accidents.

b. Control Room

Provide measurements and indication of Type A, B, C, D, E

variables listed in Regulatory Guide 1.97 (Rev. 2). Individual

licensees may take exceptions based on plant-specific design

features. BWR incore thermocouples and continuous offsite dose

monitors are not required pending their further development and

consideration as requirements. It is acceptable to rely on

currently installed equipment if it will measure over the range

indicated in Regulatory Guide 1.97 (Rev. 2), even if the equipment

is presently not environmentally qualified. Eventually, all the

equipment required to monitor the course of an accident would be

environmentally qualified in accordance with the pending Commission

rule on environmental qualification.

Provide reliable indication of the meteorological variables (wind

direction, wind speed, and atmospheric stability) specified in

Regulatory Guide 1.97 (Rev. 2) for site meteorology. No changes

in existing meteorological monitoring systems are necessary if

they have historically provided reliable indication of these variables that are representative of meteorological conditions in the

vicinity (up to about. 10miles) of the plant site. Information on

meteorological conditions for the region in which the site is

located shall be available via communication with the National

Weather Service. These requirements supersede the clarification

of NUREG-0737, Item III.A.2.2.

c. Technical Support Center (TSC)

The Type A, B, C, D and E variables that are essential for performance of TSC functions shall be available in the TSC.

(i) BWR incore thermocouples and continuous offsite dose monitors are not required pending their further development

and consideration as requirements.

(ii) The indicators and associated circuitry shall be of reliable

design but need not meet Class IE, single-failure or seismic

qualification requirements.
- 14 -

d. Emergency Operations Facility (EOF)

(i) Those primary indicators needed to monitor containment

conditions and releases of radioactivity from the plant

shall be available in the EOF.

(ii) The EOF data'indications and associated circuitry shall

be of reliable design but need not meet Class IE, singlefailure or seismic qualification requirements.

6.2 Documentation and NRC Review

NRC review is not a prerequisite for implementation. Staff review

will be in the form of an audit that will include a review of the

licensee's method of implementing Regulatory Guide 1.97 (Rev. 2)

guidance and thellicensee's supporting technical justification of

any proposed alternatives...

The licensee shall submit a report describing how it meets these

requirements. The submittal should include documentation which

may be in the form of a table that includes the following information

for each Type A, B, C, D, E variable shown in Regulatory Guide 1.97

(Rev. 2).

(a) instrument range

(b) environmental qualification (as stipulated in guide or state

criteria)

(c) seismic qualification (as stipulated in guide or state criteria)

(d) quality assurance (as stipulated in guide or state criteria)

(e) redundance and sensor(s) location(s)

(f) power supply (e.g., Class IE, non-Class IE, battery backed)

(g) location of display (e.g., control room board, SPDS, chemical

laboratory)

(h) schedule (for installation or upgrade)

Deviations from the guidance in Regulatory Guide 1.97 (Rev. 2) should

be explicitly shown, and supporting justification or alternatives

should be presented.
- 15 -

7. UPGRADE EMERGENCY OPERATING PROCEDURES (EOPs)

7.1 Requirements

a. The use of human factored, function oriented, emergency operating

procedures will improve human reliability and the ability to

mitigate the consequences of a broad range of initiating events

and subsequent multiple failures or operator errors, without

the need to diagnose specific events.

b. In accordance with NUREG-0737, Item I.C.l, reanalyze transients

and accidents and prepare Technical Guidelines. These analyses

will identify operator tasks, and information and control needs.

The analyses also serve as the basis for integrating upgraded

emergency operating procedures and the control room design review

and verifying the SPDS design.

c. Upgrade EOPs to be consistent with Technical Guidelines and an

appropriate procedure Writer's Guide.

d. Provide appropriate training of operating personnel on the use of

upgraded EOPs prior to implementation of the EOPs.

e. Implement upgraded EOPs.

7.2 Documentation and NRC Review

a. Submit Technical Guidelines to NRC for review. NRC will perform

a pre-implementation review of the Technical Guidelines. Within

two months of receipt of the Technical Guidelines, NRC will

advise the licensees of their acceptability.

b. Each licensee shall submit to NRC a procedures generation package

at least three months prior to the date it plans to begin formal

operator training on the upgraded procedures. NRC approval of the

submittal is not necessary prior to upgrading and implementing

the EOPs. The procedures generation package shall include:

(i) Plant-Specific Technical Guidelines -- plant-specific

guidelines for plants not using generic technical guidelines. For plants using generic technical guidelines,

a description of the planned method for developing plant

specific EOPs from the generic guidelines, including

plant specific information.

(ii) A Writer's Guide that details the specific methods to be

used by the licensee in preparing EOPs based on the

Technical Guidelines.
- 16

(iii) A description of the program for validation of EOPs.

(iv) A brief description of the taining program for the

upgraded EOPs.

c. All procedures generation packages will be reviewed by the staff.

On an audit basis for selected facilities, upgraded EOPs will be

reviewed. The details and extent of this review will be based on

the quality of the procedures generation packages submitted to

NRC. A sampling of upgraded EOPs will be reviewed for technical

adequacy in conjunction with the NRC Reactor Inspection Program.

7.3 Reference Documents

NUREG-0600,

Item I.C.l, I.C.8, I.C.9

NUREG-0799 9-- (Superseded by NUREG-0899)
- 17 -

8. EMERGENCY RESPONSE FACILITIES

8.1 Regulations

10 CFR 50.47(b)(6) (for Operating License applicants) Requirement

for prompt communications among principal response organizations

and to emergency personnel and to the public.

10 CFR 50.47(b)(8) -- Requirement for emergency facilities and equipment to support emergency response.

10 CFR 50.47(b)(9) -- Requirement that adequate methods, systems and

equipment for assessing and monitoring actual or potential offsite

consequences of a radiological emergency condition are in use.

10 CFR 50.54(q) (for Operating Reactors) -- Same requirement as 10 CFR

50.47(b) plus 10 CFR 50, Appendix E.

10 CFR 50, Appendix E, Paragraph IV.E

Requirement for:

"l. Equipment at the site for personnel monitoring;"

"2. Equipment for determining the magnitude of and for continuously assessing the impact of the release of radioactive materials to the environment;"

"3. Facilities and supplies at the site for decontamination

of onsite individuals;"

"4. Facilities and medical supplies at the site for appropriate emergency first aid treatment;"

"5. Arrangements for the services of physicians and other

medical personnel qualified to handle radiation emergencies on site;"

"6. Arrangements for transportation of contaminated injured

individuals from the site to specifically identified

treatment facilities outside the site boundary;"

"7. Arrangements for treatment of individuals injured in

support of licensed activities on the site at treatment facilities outside the site boundary;"

"8. A licensee onsite technical support center and a licensee

near-site emergency operations facility from which effective direction can be given and effective control can be

exercised during an emergency;"

"9. At least one onsite and one offsite communications system;

each system shall have a backup power source.1
- 18

All.communication plans shall have arrangements for emergencies,

including titles and alternates for those in charge at both

ends of the communication links and the primary and backup

means of communication. Where consistent with the function

of the governmental agency, these arrangements will include:

"a. Provision for communications with contiguous State/local

governments within the plume exposure pathway (emergency

planning zone) EPZ. Such communications shall be tested

monthly."

"b. Provisions for communication with Federal emergency

response organizations. Such communication systems

shall be tested annually."

"c. Provision for communications among the nuclear power

reactor control room, the onsite technical support

center, and the near-site emergency operations facility;

and among the nuclear facility, the principal State and

local emergency operations centers, and the field assessment teams. Such communications systems shall be tested

annually."

"d. Provisions for communication by the licensee with NRC

Headquarters and the appropriate NRC Regional Office

Operations Center from the nuclear power reactor control

room, the onsite technical support center, and the nearsite emergency operations facility. Such communications

shall be tested monthly."

Within this section on emergency response facilities, the Technical Support

Center (TSC), Operational Support Center (OSC) and Emergency Operations

Facility (EOF) are addressed separately in terms of their functional statements and recommtnded requirements. The subsections on Documentation and

NRC Review and Reference Documents that follow the EOF discussion apply to

this entire section on emergency response facilities.
- 19 -

8.2 Technical. Support Center (TSC)

8.2.1 Requirements

a. The TSC is the onsite technical support center for

emergency response. WMen activated, the TSC is staffed

by predesignated technical, engineering, senior management,

and other licensee personnel, and five pre-designated NRC

personnel. During periods of activation, the TSC will

operate uninterrupted to provide plant management and

technical support to plant operations personnel, and

to relieve the reactor operators of peripheral duties

and communications not directly related to reactor

system manipulations. The TSC will perform EOF functions

for the Alert Emergency class and for the Site Area

Emergency class and General Emergency cla:_ until the

EOF is functional.

The TSC will be:

b. Located within the site protected area so as to facilitate

necessary interaction with control room, OSC, EOF and

other personnel involved with the emergency.

c. Sufficient to accommodate and support NRC and licensee

predesignated personnel, equipment and documentation

in the center.

d. Structurally built in accordance with the Uniform Building

Code.

e. Environmentally controlled to provide room air temperature,

humidity and cleanliness appropriate for personnel and

equipment.

f. Provided with radiological protection and monitoring equipment necessary to assure that radiation exposure to any

person working in the TSC would not exceed 5 rem whole

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

duration of the accident.

g. Provided with reliable voice and data communications with

the control room and EOF and reliable voice communications

with the OSC, NRC Operations Centers and state and local

operations centers.

h. Capable of reliable data collection, storage, analysis,

display and communication sufficient to determine site

and regional status, determine changes in status, forecast

status and take appropriate actions. The following variables shall be available in the TSC:
- 20 -

(i) the variables in the appropriate Table 1 or 2 of

Regulatory Guide 1.97 (Rev. 2) that are essential

for performance of TSC functions; and

(ii) the meteorological variables in Regulatory Guide

1.97 (Rev. 2) for site vicinity and National Weather

Service data available by voice communication for

the region in which the plant is located.

Principally those data must be available that would

enable evaluating incident sequence, determining

mitigating actions, evaluating damages and determining

plant status during recovery operations.

i. Provided with accurate, complete and current plant records

(drawings, schematic diagrams, etc.) essential for evaluation

of the plant under accident conditions.

j. Staffed by sufficient technical, engineering, and senior

designated licensee officials to provide needed support,

and be fully operational within-approximately I hour

after activation.

k. Designed taking into account good human factors engineering

principles.
- 21

8.3 Operational Support Center (OSC)

8.3.1 Requirements

a. When activated, the OSC will be the onsite area separate

from the control room-where predesignated operations

support personnel will assemble. A predesignated licensee

official shall be responsible for coordinating and

assigning the personnel to tasks designated by control

room, TSC and EOF personnel.

The OSC will be:

b. Located onsite to serve as an assembly point for support

personnel and to facilitate performance of support functions

and tasks.

c. Capable of reliable voice communications with the control

room, TSC and EOF.
- 22

8.4 Emergency Operations Facility (EOF)

8.4.1 Requirements

'a. The EOF is a licensee controlled and operated facility.

The EOF provides for management of overall licensee

emergency response, coordination of radiological and

environmental assessment, development of recomnendations

for public protective actions, and coordination of emergency response activities with Federal, State and local

agencies.

When the EOF is activated, it will be staffed by predesignated emergency personnel identified in the emergency

plan. A designated senior licensee official will manage

licensee activities in the EOF.

Facilities shall be provided in the EOF for the acquisition,

display and evaluation of radiological and meteorological

data and containment conditions necessary to determine

protective measures. These facilities will be used to

evaluate the magnitude and effects of actual or potential

radio-active releases 'from the plant and to determine

dose projections.

The EOF will be:

b. Located and provided with radiation protection features

as described in Table I (previous guidance approved by

the Commission) and with appropriate radiological monitoring systems.

c. Sufficient to accommodate and support Federal, State,

local and licensee predesignated personnel, equipment

and documentation in the EOF.

d. Structurally built in accordance with the Uniform Building

Code.

e. Environmentally controlled to provide room air temperature,

humidity and cleanliness appropriate for personnel and

equipment.

f. Provided with reliable voice and data communications

facilities to the TSC and control room, and reliable

voice communication facilities to OSC and to NRC, State

and local emergency operations centers.
- 23

g. *Capable of reliable collection, storage, analysis, display

and communication of information on containment conditions,

radiological releases and meteorology sufficient to determine site and regional status, determine changes in status,

forecast status and take appropriate actions. Variables

from the following categories that are essential to EOF

functions shall be available in the EOF:

(i) variables from the appropriate Table 1 or 2 of

Regulatory Guide 1.97 (Rev. 2), and

(ii) the meteorological variables in Regulatory Guide

1.97 (Rev. 2) for site vicinity and regional data

available via communication from the National Weather

Service.

h. Provided with up to date plant records (drawings,

schematic diagrams, etc.), procedures, emergency pl,

and environmental information (such as geophysical -:"a)

needed to perform EOF functions.

i. Staffed using Table 2 (previous guidance approved by the

Commission) as a goal. Reasonable exceptions to goals

for the number of additional staff personnel and response

times for their arrival should be justified and will

be considered by NRC staff.

j. Provided with industrial security when it is activated

to exclude unauthorized personnel and when it is idle

to maintain its readiness.

k. Designed taking into account good human factors engineering

principles.

8.4.2 Documentation and NRC Review

The conceptual design for emergency response facilities (TSC,

OSC, and EOF) have been submitted to NRC for review. In

many cases, the lack of detail in these submittals has precluded

an NRC decision of acceptability. Some designs have been

disapproved because they clearly did not meet the intent of

the applicable regulations. NRC does not intend to approve

each design prior to implementation, but rather has provided

in this document those requirements which should be satisfied.

These requirements provided a degree of flexibility within

which licensees can exercise management prerogatives in

designing and building emergency response facilities (ERF)

that satisfy specific needs of each licensee. The foremost

consideration regarding ERFs is that they provide adequate
- 24 -

adequate capabilities of licensees to respond to emergencies.

NUREG guidance on ERFs has been intended to address specific

issues which the Commission believes should be considered in

achieving improved capabilities.

Licensees should assure that the design of ERFs satisfies

these requirements. Exemptions from or alternative methods

of implementing these requirements should be discussed with

NRC staff and in some cases could require Commission approval.

Licensees should continue work on ERFs to complete them according to schedules that will be negotiated on a plant-specific

basis. NRC will conduct appraisals of completed facilities

to verify that these requirements have been satisfied and

that ERFs are capable of performing their intended functions.

Licensees need not document their actions on each specific item

contained in NUREG-0696 or 0814.
- 25 -

8.4.3 Reference Documents (Emergency.Response Facilities)

10 CFR 50.47(b) -- Requirements for emergency facilities and

equipment for OLs.

10 CFR 50.54(q) and Appendix E, Paragraph IV.E -- Requirements

for emergency facilities and equipment for ORs.

NUREG-0660 -- Description of and implementation schedule for

TSC, OSC and EOF.

Eisenhut letter to power reactor licensees 9/13/79 -- Request

for commitment to meet requirements

Denton letter to power reactor licensees 10/30/79 -- Clarification of requirements.

NUREG-0654 -- Radiological Emergency Response Plans

NUREG-0696 -- Functional criteria for emergency response

facilities.

NUREG-0737 -- Guidance on meteorological monitoring and dose

assessment.

Eisenhut letter to power reactor license 2/18/81 -- Commission

approved guidance on location, habitability and staff for

emergency facilities. Request and deadline for submittal

of conceptual design of facilities.

NUREG-0814 (Draft Report for Comment) -- Methodology for evaluation of emergency response facilities.

NUREG-0818 (Draft Report for Comment) -- Emergency Action Levels

Reg. Guide 1.97 (Rev. 2) -- Guidance for variables to be used

in selected emergency-response facilities.

COMJA-80-37, January 21, 1981 -- Commission approval guidance

on EOF location and habitability.

Secretary memorandum S81-19, February 19, 1981 -- Commission

approval of NUREG-0696 as general guidance only.
TADLE I

EMERGENCY OPERATIONS FACILITY

Option 1

Two Facilities

Option 2

One Facility

Close-in Primary: Reduce Habitability*

0

0

0

within 10 miles

protection factor = 5

ventilation isolation

with HEPA (no charcoal)

o At or Beyond 10 miles.

o No special protection factor.

o If beyond 20 miles, specific

approval required by the

Commission, and some provision for NRC site team closer

to site.

o Strongly recommended location

be coordinated with offsite

authorities.

Backup EOF

o between 10-20 miles

o no separate, dedicated

facility

o arrangements for portable

backup equipment

o strongly recommended location

be coordinated with offsite

authorities

o continuity of dose projection

and decision making capability

For both Options:

- located outside security boundary

- space for about 10 NRC employees

- none designated for severe phenomena, e.g., earthquakes

labitability requirements are only for the part of the EOF in which dose assessments communications and

Jecision making take place.

r a utility has begun construction of a new building for an EOF that is located with 5 miles, that new

acility is acceptable (with less than protection factor of 5 and ventilation isolation and HEPA) provided

hat a backup EOF similar to "B" in Option 1 is provided.
TABLE 2

MINIMUM STAFFING REQUIREMENTS FOR NRC LICENSEES

FOR NUCLEAR POWER PLANT EMERGENCIES

Position Title

or ExDertise

Capability for Additions

On

Malor Functional Area Maior Tasks Shift* 30 min. 60 min.

Plant Operations and

Assessment of

Operational Aspects

Emergency Direction and

Control (Emergency

Coordinator)***

Notification/

Communication""**

Shift supervisor (SRO)

Shift foreman (SRO)

Control-room operators

Auxiliary operators

Shift technical advisor,

shift supervisor, or

designated facility

manager

1

1

2

2

n.

1**

Radiological Accident

Assessment and Support

of Operational Accident

Assessment

Nofity licensee, state

local, and federal

personnel & maintain

communication

Emergency operations

facility (EOF) director

Offsite dose

assessment

Offsite surveys

Onsite (out-of-plant)

Inplant surveys

Chemistry/radiochemistry

1 1 2

Senior manager

Senior health physics

(HP) expertise

1

1

HP technicians

Rad/chem technicians

1l

1

2

1

1

2

1

1

1

NOTE: Source of this table is NUREG-0654, "Functional Criteria for Emergency Response Facilities."
TABLE 2 (Continued)

Position Title

or Expertise

Capability for Additions

On

MaJor Functiona.l Area MaJor Tasks Shift* 30 min. 60 min.

Plant System

Engineering, Repair

and Corrective Actions

Technical support Shift technical advisory

Core/thermal hydraulics

Electrical

Mechanical

Mechanical maintenance/

Radwaste operator

Electrical maintenance/

instrument and control

(I&C) technician

1 1

1

1

Repair and corrective

actions

1

1

1

2

1

1l

1R

Protective Actions

(In-Plant)

Radiation protection:

a. Access control

b. HP Coverage for

repair, corrective actions,

search and rescue

first-aid, &

firefibhting

c. Personnel monitoring

d. Dosimetry

HP technicians 2

Fireftighting Fire

brigade

per

technical

specification

Local

support

Rescue Operations

and First-Aid

2 ** Local

support
TAIJ.I.E 2 (ConLlntled)

Pn.iLton TiLle

or ExperLise

capW!i I iLy-f or Addi Lions

fn Shlill.* 0 min. 60 mtU Major FuncLinnal Area M1.iti lrsks rI. N w V I m I J I • v,, n.

Site Access Control

and Persniuml

Accountabi liLy

Set:eiri Ly, rirer i!ihLiiq

coIntlnuil i ca I-. io ns, , ersonnoI accountability

SecuriLy personnel All per

securi t.y

plati

Total 1.0 11 15

" :or each uinafrected nuclear uni t in operation, maintain at least one shirt foreman, one control-room

operator, and one atixiliary operator except that units sharing a control room may share a shift foreman

i r all functions are coverril."

AAMay'be provided by shift persoinnel assirnied other functions.

ARAOverall direction of racili.y response to be assumed by EOF director when all renters are fully manned. Directo

of minuite-Lo-minute.facility operations remains wit.h senior manager in technical support center or control room.

****May be performed by engineering aide to shirt supervisor.

NRC NC,41, FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. NUREG-0737 REPORT NUMBER lAssignedby DDCJ

BIBLIOGRAPHIC DATA SHEET Supplement No. 1

4. TITLE AND SUBTITLE (Add Volume No., dfppropr~arej 2. fLeave blankJ

Clarification of TMI Action Plan Requirements:

Requirements for Emergency Response Capability 3 RECIPIENT'S ACCESSION NO.

7. AUTHORIS) 5. DATE REPORT COMPLETED

MONTH I YEAR December 1 982

9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip CodeJ DATE REPORT ISSUED

Division of Licensing MONTH I YEAR

Office of Nuclear Reactor Regulation January 1983

U.S. Nuclear RegulaLory Commission 6 (Leave biank)

Washington, DC 20555 8. (Leave blank]

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include ZIp Code)

10. PROJECT TASK/WORK UNIT NO

Same as 9 above. 11. FINNO.

13. TYPE OF REPORT PERIOD COVERED llfclus,ve clates)

Technical Report 15. SUPPLEMENTARY NOTES 14. (Leave clank)

16. ABSTRACT 1200 words or less)

This document, Supplement I to NUREG-0737, is a letter from D. G. Eisenhut, Director of

the Division of Licensing, NRR, to licensees of operating power reactors, applicants for

operating licenses, and holders of construction permits forwarding post-TMI requirements

for emergency response capability which have been approved for implementation. On

October 30, 1980, the NRC staff issued NUREG-0737, which incorporated into one document

all TMI-related items approved for implementation by the Commission at that time. In

this NRC report, ad&4itional clarification is provided regarding Safety Parameter Display

Systems, Detailed ;'.1ý-.trol Room Design Reviews, Regulatory Guide 1.97 (Revision 2) - Application to Emergency Response Facilities, Upgraded of Emergency Operating Procedures

Emergency Response Facilities, and Meteorological Data.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17b. IDENTIFIERS.OPEN-ENDED TERMS

18. AVAILABILITY STATEMENT 19. SECURITY CLASS (This leoort) 21 NO OF PAGES

Unclassified .35

Unlimited availability 2JSECTRITYun as S IThis page) 22 PRICE

S

9

NRC FORM 335 I11.11-