NUREG-0737

Clarification of

TMI Action Plan Requirements

Manuscript Completed: November 1980

Date Published: November 1980

Division of Licensing

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

Washington, D.C. 20555
ABSTRACT

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

Division of Licensing, NRR, to licensees of operating power reactors and

applicants for operating licenses forwarding post-TMI requirements which have

been approved for implementation. Following the accident at Three Mile Island

Unit 2, the NRC staff developed the Action Plan, NUREG-0660, to provide a

comprehensive and integrated plan to improve safety at power reactors. Specific

items from NUREG-0660 have been approved by the Commission for implementation

at reactors. In this NRC report, these specific items comprise a single

document which includes additional information about schedules, applicability,

method of implementation review, submittal dates, and clarification of technical

positions. It should be noted that the total set of TMI-related actions have

been collected in NUREG-0660, but only those items that the Commission has

approved for implementation to date are included in this document, NUREG-0737.

i i i
CONTENTS

Page

Abstract...................................................................... iii

Introduction--Letter from D. G. Eisenhut to All Licensees, October 31, 1980... vii

Enclosure 1

Post-TMI Requirements for Operating Reactors ............................. 1-1

Enclosure 2

TMI Action Plan Requirements for Applicants for an Operating License ..... 2-1

Enclosure 3

Clarification of TMI Action Plan Requirements ............................ 3-1

Appendix A

Key to References ........................................................ A-1

Appendix BDesign and Qualifications ................................................ B-1

Appendix C

Nuclear Power Plant Shift Technical Advisor (INPO Report) ................ C-1

v
_° . UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

COCT .1 1 W

TO ALL LICENSEES OF OPERATING PLANTS AND APPLICANTS FOR OPERATING LICENSES

AND HOLDERS OF CONSTRUCTION PERMITS

Gentlemen:

SUBJECT: POST-TMI REQUIREMENTS

On September 5, 1980, the NRC staff sent you a draft clarification letter

regarding approved TMI Action Plan items. During the week of September 22,

1980, four regional meetings were held to provide a more detailed explanation

of these requirements and to obtain industry comments concerning these items.

Based on these discussions and other comments received, the NRC has revised

its requirements regarding these items. It is the purpose of this letter to

set forth those requirements.

This letter incorporates in one document, all TMI-related items approved for

implementation by the Commission at this time. This document is being published

as NUREG-0737. Enclosures 1 and 2 contain an itemized listing of OR and OL

requirements including implementation schedules, applicability, method of

implementation review and licensee submittal dates. Enclosure 3 contains more

detailed clarifications of most of the NRC positions including the identification of any changes from previous requirements and guidance.

Most of the items in the attached document have already been issued as

requirements by previous correspondence. Those items that are being issued

as requirements for the first time by this letter are identified by an

asterisk in Enclosures 1 and 2. Additional guidance on the Emergency

Response Facilities, Section III.A.1.2, will be forwarded separately in the

near future.

Licensees and applicants should note that the set of requirements identified

in the enclosures do not constitute the total set of TMI-related actions in

the TMI-2 Action Plan, NUREG-0660. Rather, as noted above, the enclosures

are a compilation of those items that have been specifically approved by the

Commission for implementation. Upon further staff development of criteria

and planning, additional items will be issued. For example, in the relatively

near future, the staff expects to issue further criteria on emergency

operational facilities (NUREG-0696), auxiliary feedwater system improvements

(derived from NUREG-0667), and instrumentation (Regulatory Guide 1.97,

Revision 2). In general, the implementation of those requirements will

be carefully examined to ensure that they do not unnecessarily impact any

of the requirements in this letter.

vii
- 2 -

The requirements herein (which include the requirements from NUREG-0694) are

applicable to applicants for operating licenses and such applicants are expected

to meet the same schedule of implementation as indicated for operating reactors.

Operating license reviews being finalized over the next few months will be

handled on a case-by-case basis. Any item for which the implementation date

is prior to the expected date of issuance of an operating license will be

considered to be a prerequisite to obtaining that license. For such items,

applicants must submit information or documentation four-months prior to the

staff's scheduled issuance of its Safety Evaluation Report or four months

prior to the listed implementation date, whichever is later.

A large number of post-TMI requirements require the installation of a number

of control room indications. It is important that licensees and applicants

give consideration to human factor engineering considerations in planning

for the installation of such new control room equipment. In the coming

months, the NRC will be requiring human factors engineering reviews of

control room designs as part of Action Plan Item I.D.1, and such an effort

at this time may reduce the potential for later modifications. As an

example of possible considerations, licensees and applicants might well

consider at this time whether some control panel indications are of lesser

safety significance and can be moved to other locations in the control room.

It is expected that the requirements contained herein will be met. However,

it is recognized that licensees have proceeded with implementation of some

of these items prior to issuance of these clarifying criteria. The staff

will consider requests for relief from various aspects of these criteria.

Such requests should explain the need for relief, include a clear description

of design features of the proposed installation, and provide a safety

rationale supporting the adequacy of the proposed installation. A licensee

or applicant seeking relief from any element of our criteria should submit

a request for relief, along with supporting justification, in response to

this letter.

Accordingly, pursuant to §50.54(f) operating reactor licensees are requested

to furnish, within forty-five (45) days of this letter, confirmation that

the implementation dates indicated in Enclosure 1 will be met. For any date

that cannot be met, furnish a proposed revised date, justification for the

delay, and any planned compensating safety actions during the interim. After

our evaluation of your response the NRC staff will take action, as necessary

to assure that such reqirements and commitments are appropriately enforceable.

This may include, as needed, issuance of a Confirmatory or Show-Cause Order.

Sincerely,

S4 iEisen is u

Division Licensing

Office of Nuclear Reactor Regulation

Enclosures:

As stated Viii
ENCLOSURE 1

POST-TMI REQUIREMENTS

FOR OPERATING REACTORS

1-1
ENCLOSURE 1

POST-THI REQUIREMENTS FOR OPERATING REACTORS

(For postimplementation reviews, licensees shall comply with 10 CFR 50.59. If it

that an unreviewed safety question exists or a change to the facility's existing

specifications is required, NRC approval is required before implementation.]

is determined

technical

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

I.A.1.1 Shift technical

advisor

1. On duty 1/1/80

2. Tech specs 12/15/80

3. Trained per 1/1/81

LL Cat B

4. Describe long- 1/1/81

term program

All 9/13/79

All 7/2/80

All 9/13/79

All *

10/30/79

7/2/80

Encl. 3

No

Yes

No

Yes

No

Yes

Yes 1/1/80

Yes 9/1/80

Yes 1/1/81

No 1/1/81

Complete

Encl. 3 No No

I.A.1.2 Shift supervisor

responsibilities

I.A.1.3 Shift manning

Delegate nonsafety duties

1/1/80 All 9/13/79 10/30/80 No Yes No 1/1/80 Complete

1. Limit overtime 11/1/80

2. Min shift crew 7/1/82

All 7/31/80 7/31/80 No

All 7/31/80 7/31/80 No

Yes

Yes

No 11/1/80

Yes 11/1/80 I Amend TS on

shift manning

I.A.2.1 Immediate upgrading

of RO & SRO training

and qualifications

1. SRO exper

2. SROs be ROs

1 yr

3. Three mo

trng on shift

4. Modify

training

5. Facility

certification

5/1/80 All

12/1/80 All

8/1/80 All

3/28/80 3/28/80 No

3/28/80 Encl 3 No

3/28/80 3/28/80 No

3/28/80 3/28/80 No

Yes

Yes

Yes

Yes

Yes

No None

No None

No None

No 8/1/80

No None

Completion to be

verified

Completion by OIE

Completion by OIE

NRR staff to review

OIE verification

8/1/80

5/1/80

All

All 3/28/80 3/28/80 No

I.A.2.3 Administration of

training programs

I.A.3.1 Revise scope &

criteria for

licensing exams

Instructors com- 8/1/80

plete SRO exam.

1. Increase scope 5/1/80

2. Increase pas- 5/1/80

sing grade

3. Simulator 6/1/80

exams

10/1/81

All 3/28/80 3/28/80 No Yes No None NRR to verify

conformance

All 3/28/80 3/28/80 No

All 3/28/80 3/28/80 No

Plants *

having

simulator

All *

None No

No

No

No

No

No None

No None

No None

Encl 3 No No None Plants w/o

simulators

Note: For complete reference citation of NUREG reports, see Appendix A.

*Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

Short-term accident

& procedures

review

1. SB LOCA

2. Inadequate

core cooling

a. Reanalyze

& propose

guidelines

b. Revise

procedures

6/1/80 All 9/13/79 10/30/79 No Yes

3. Transients &

accidents

a. Reanalyze

& propose

guidelines

b. Revise

procedures

1/1/81

First

refueling

outage

after

1/1/82

1/1/81

First

refueling

outage

after

1/1/82

1/1/80

All 9/13/79 Encl 3 Yes No

All 9/13/79 Encl 3 Yes No

All 9/13/79 Encl 3 Yes No

All 9/13/79 Encl 3 Yes No

No None

No 1/1/81

No Not

determined

No 1/1/81

No Not

determined

No 1/1/80

Complete

J.

I.C.2 Shift & relief

turnover procedures

Implement shift

turnover checklist

All 9/13/79 10/30/79 No Yes Complete

I.C.3 Shift-supervisor

responsibility

Clearly define

superv & oper

responsibilities

Establish

authority,

limit access

1/1/80 All 9/13/79 10/30/79 No Yes No 1/1/80 Complete

I.C.4 Control-room access 1/1/80 All 9/13/79 10/30/79 No Yes No 1/1/80 Complete

I.C.5 Feedback of operating

experience

I.C.6 Verify correct

performance of

operating activities

I.D.1 Control-room

design reviews

Licensee to

implement

procedures

Revise

performance

procedures

Preliminary

assessment &

schedule for

correcting

deficiencies

1/1/81 All 5/7/80 Encl 3 No Yes No None

1/1/81 All Encl 3 No Yes No None

TBD All 6/26/80 NUREG/CR-1580

(Draft)

4/82 Final guidance will

be issued 1981

as NUREG-0700

*Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

I.D.2 Plant-safetyparameter display

console

1. Description

2. Installed

3. Fully

implemented

TBD

TBD

TBD

All 6/26/80

All 6/26/80

All 6/26/80

Encl 3

Encl 3

Encl 3

Later Guidance per

NUREG-0696 Rev. 2

II. B.1 Reactor-coolantsystem vents

II.B.2 Plant shielding

II.B.3 Postaccident

sampling

1. Design vents

2. Install vents

(LL Cat B)

3. Procedures

1. Review designs

2. Plant

modifications

(LL Cat B)

3. Equipment

qualification

1. Interim system 2. Plant

modifications

(LL Cat B)

7/1/81

7/1/82

All 9/13/79 10/30/79 No

All 9/13/79 10/30/79 Yes

1/1/82 All 9/13/79 Encl 3

1/1/80

1/1/82

6/30/82

1/1/80

1/1/82

All 9/13/79 10/30/79

All 9/13/79 10/30/79

,(3 7C Encl 3

All CLI-80-21 Encl 3

Yes

No

No

Yes

No

No

Yes

Yes

No Yes

All

All

9/13/79 10/30/79

9/13/79 10/30/79

Encl 3

No

No

Yes

Yes

No 7/1/81

Yes 7/1/81

Yes 1/1/81

No 1/1/80

No 1/1/82

No 11/1/80

No 1/1/80

Yes 1/1/81

submittal

if deviation from

position

No 1/1/81

No

No

Complete

Complete

Complete

U'

II.B.4 Training for

mitigating

core damage

II.D.1 Relief & safetyvalve test

requirements

1. Develop training program

2. Implement

program

a. Initial

b. Complete

1. Submit program

2. RV & SV testing

(LL Cat B)

a. Complete

testing

b. Plantspecific

report

3. Block-valve

testing

1/1/81 All 3/28/80 3/28/80

Encl 3

No

4/1/81 All

10/1/81 All

3/28/80 Encl 3 No

3/28/80 Encl 3 No

Yes

Yes

Yes

1/1/80 All 9/13179 10/30/79 No Yes

No

Yes

7/1/81 All 9/13/79 10/30/79 No

No 1/1/80

No 7/1/81

TBD 1/1/82

TBO 7/1/82

Complete

10/1/81 All

7/1/82 PWR

9/13/79 Encl 3 Yes

* Encl 3 Yes Yes

II.D.3 Valve position

indication

1. Install direct 1/1/80

indications of

valve position

2. Tech specs 12/15A

All 9/13/79 10/30/79 No Yes

No

Yes 1/1/80

Yes 9/1/80

Complete

80 All 7/2/79 7/2/80 Yes

Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

II.E.1.1 Auxiliary feedwater

system evaluation

1. Short term 7/1/81 PWR

1/1/82 PWR

3/10/80 Encl 3 Yes

4/24/80 Encl. 3 Yes

Yes

Yes

Item

specific

Item

specific

Plant

specific

Plant

specific

2. Long term

II.E.1.2 Auxiliary feedwater

system initiation

& flow

I-.E.3.1 a%Emergency. power

for pressurizer

heaters

II.E.4.1 Dedicated hydrogen

penetrations

II.E.4.2 Containment

isolation

dependability

1. Initiation

a. Control

grade

b. Safety grade

2. Flow indication a. Control

grade

b. LL Cat A

tech specs

c. Safety grade

1. Upgrade power

supply.

2. Tech specs

1. Design

2. Install

1-4. Imp diverse

isolation

5. Cntmt pressure

setpoint

a. Specify

pressure

b. Modifications

6. Cntmt purge

valves

7. Radiation

signal on

purge valves

8. Tech specs

6/1/80 PWR 9/13/79 10/30/79

7/1/81 PWR 9/13/80 10/30/79

1/1/80 PWR 9/13/79 10/30/79

12/15/80 PWR 9/13/79 7/2/80

7/1/81 PWR 9/13/79 10/30/79

1/1/80 PWR 9/13/79 10/30/79

12/15/80 PWR 9/13/79 7/2/80

Encl 3

1/1/80 All 9/13/79 10/30/79

7/1/81 All 9/13/79 10/30/79

Encl 3

1/1/80 All 9/13/79 10/30/79

Yes

No

No

Yes

No

No

Yes

Yes

No

No

No

Yes

Yes

No

Yes

Yes

No

No

Yes

Yes

Yes

Yes

Yes

Yes

Yes

Yes

Yes

1/1/80

1/1/81

1/1/80

9/1/80

1/1/81

1/1/81

9/1/80

1/1/81

7/1/81

1/1/81

7/1/81

Complete

Complete

Complete

See II.G.1

Complete

Complete

All

All

All

All

A

*

A

A

Encl 3

Encl 3

Encl 3

Encl 3

No

Yes

No

No

No 1/1/80

No 7/1/81

Yes 1/1/80

No 1/1/81

Yes 1/1/81

Yes 1/1/81

Yes 7/1/81

Yes

No

Yes

Yes

12/15/80 All 9/13/79 7/2/80 Yes No Yes 9/1/80

'Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentatlon Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

II.F.1 Accident-monitoring 1. Noble gas

monitor

2. Iodine/

particulate

sampling

3. Containment

high-range

monitor

1/1/82 All 9/13/79 10/30/79

Encl 3

1/1/82 All 9/13/79 10/30/79

Encl 3

1/1/82 All 9/13/79 10/30/79

Encl 3

No Yes

No Yes

Yes 1/1/81

Submittal

if deviation from

position

Yes 1/1/81

submittal

if deviation from

position

Yes 7/1/81

submittal

if deviation from

position

Yes 1/1/82

Yes 1/1/82

Yes 1/1/82

No Yes

-j

II.F.2 Instrumentation for

detection of

inadequate core

cooling

4. Containment

pressure

5. Containment

water level

6. Containment

hydrogen

1. Subcool meter

2. Tech spec

(LL Cat A)

3. Install level

instruments

(LL Cat B)

1. Upgrade to

emerg sources

2. Tech specs

1/1/82

1/1/82

1/1/82

All

.All

All

9/13/79 10/30/79

Encl 3

9/13/79 10/30/79

Encl 3

9/13/79 10/30/79

Encl 3

9/13/79 10/30/79

7/2/79 7/2/80

No

No

No

Yes

Yes

Yes

1/1/80 PWR

12/15/80 PWR

No

Yes

Yes

No

Yes 1/1/80

Yes 9/1/80

Complete

1/1/82 All 9/13/79 10/30/79

Encl 3

No Yes

Yes

Yes 1/1/81

II.G.1 Power supplies for

pressurizer relief

valves, block valves,

& level indicators

II.K.1 IE Bulletins

II.K.2 Oraers on B&W plants

1/1/80 PWR

12/15/80 PWR

9/13/79 10/30/79 No Yes 1/1/80

Yes 9/1/80

Complete

7/2/80 7/2/80 Yes No See II.E.3.1

79-05, 06, 08

8. Upgrade AFW

system

9. FEMA on ICS

10. Safety-grade

trip

11. Operator

training,

drilling

13. Thermalmechanical

report

Bulletin

specific

See

II. E. 1.1

TBO

7/1/81

Complete

1/1/81

1/1/82

All 4/79 NA No Yes No Bulletin NRR evaluating

specific licensee responses

B&W

B&W

B&W

B&W

B&W

C-E,

W

Per

order

Per

order

Per

order

Per

order

8/21/79

*

8/21/79,

Encl 3

11/7/79,

Encl 3

12/20/79,

Erncl 3

Encl 3

Encl 3

Yes

No

Yes

No

Yes

No

Yes

As re- See

quired I.E.1.1

No Plant specific

No

Yes 1/1/81

No Complete

As re- 1/1/81

quired

As re- 1/1/82

quired

No

No

Yes

Yes

zRequirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

II.K.2 Orders on B&W plants

(continued)

14. Lift

frequency of

PORVs & SVs

15. Effects of

slug flow

on OTSGS

16. RCP seal

damage

17. Voiding in

RCS

See

II.K.3.7

B&W

Complete B&W

Complete B&W

a. Com- B&W

plete

b. 1/1/82 C-E,

W

a. Com- UW -

plete

b. 1/1/82 C-E,

WComplete II&W

9/28/79 NUREG0565

11/21/79 Encl 3

11/21/79 Encl 3

No

No

No

1/9/80

*

Encl 3 No

Encl 3 No

Yes

Yes

Yes

Yes

Yes

Yes

Yes

Yes

No See II.K.3.7

No Complete, under

staff review

No Complete, under

staff review

No Complete, under

staff review

No 1/1/82

No Complete, under

staff review

No 1/1/82

No Complete, under

staff review

19. Benchmark

analysis of

seq AFW flow

20. System re- sponse to SB

7' LOCA co

8/21/79 End 3

Encl 3

No

No

8/21/79 Encl 3 No

II.K.3 Final recommendations,

B&O task force

1. Auto PORV

isolation

a. Design 7/1/81 PWR *

*

Encl 3 Yes

b. Test/ 1st refuel PWR Encl 3 Yes

install 6 mo after

staff

approval

2. Report on 1/1/81 PWR

PORV failures

5/7/80 Encl 3 No

No

No

Yes

Yes

Yes

Yes 7/1/81

Yes 7/1/81

If required by

II.K.3.2

No 1/1/81

3. Reporting SV & 1/1/81

RV failures &

challenges

5. Auto trip of

RCPs

a. Propose 7/1/81

modifications

b. Modify 3/1/82

7. Eval of PORV 1/1/81

opening

probability

All 5/7/80 None No Yes 1/1/81 Initiate data

beginning 4/1/80

PWR 5/7/80 Encl 3 No No 2/15/81

PWR

B&W

5/7/80 Encl 3 Yes No

* Encl 3 No Yes

Yes 7/1/81

No 1/1/81

If required

'Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

II.K.3 Final recommendations,

B&O task force

(continued)

9. PID controller 1/1/81

10. Proposed Plant

anticipatory specif

trip modifications

11. Justify use Plant

of certain specif

PORV

12. Anticipatory

trip on

turbine trip

a. Confirma- 1/1/81

tion or propose modif1-

cations

W

Select

ic W

Plant

ic specific

5/7/80 Encl 3 No Yes

5/7/80 Encl 3 Yes No

* None No

5/7/80 Encl 3 No

b. Modify

ID

13. HPCI & RCIC

init levels

a. Analysis

b. Modify

14. Iso condenser

isol modification

15. Isolation of

HPCI & RCIC

modification

16. Challenges &

failures to

relief valves

a. Study

b. Modify

17. ECC system

outages

1st refuel c

or 6 mo

after staff

approval

1/1/81 BWR

7/1/81 BWR

1/1/82 BWR w/

iso

cond

7/1/81 BWR

4/1/81 BWR

1st refuel BWR

or within

1 yr after

approval

1/1/81 All

5/7/80 Encl 3 Yes

Yes

Yes

No

Yes

No

Yes

Yes

No 12/1/80 Implementation

to be verified.

Yes Plant

specific

No Plant See Sect.

specific 3.2.4.d of

NUREG-0611

No 1/1/81

Yes 1st refuel

after tech

spec amend

request

No 1/1/81

Yes 1/1/81

Yes 7/1/81

Yes 1/1/81

No 4/1/81

Yes 4/1/81

As re- 1/1/81

quired

5/7/80

5/7/80

5/7/80

Encl 3 No

Yes

Encl 3 No

5/7/80 Encl 3 No

5/7/80 Encl 3 No Yes

5/7/80 Encl 3 Yes No

5/7/80 Encl 3 No Yes

'Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

II.K.3 Final recommendations, 18.

B&O task force

(continued)

ADS actuation

a. Study

b. Propose

mods

c. Modifications

4/1/81

4/1/82

BWR

BWR

5/7/80 Encl 3 No

5/7/80 Encl 3 No

Yes

Yes

1st refuel BWR

6 mo after

staff

approval

7/1/81 BWR

5/7/80

Encl 3

Encl 3 Yes No

0

19. Interlock

recirc pump

modification

20. Loss of SVC

water at BRP

21. Restart of CSS

& LPCI

a. Design

b. Modifications

22. RCIC suction

a. Verify

procedures

b. Modifications

24. Space cooling

for HPCI/RCIC

modifications

25. Power on pump

seals

a. Propose

mods

b. Modifications

27. Common ref

level

28. Qual of ADS

accumulators

29: Performance

of isolation

condensers

7/1/81

1/1/81

1st refuel

6 mo

after

staff

approval

1/1/81

1/1/82

1/1/82

7/1/81

1/1/82

1/1/82

7/1/82

7/1/81

1/1/82

4/1/81

Big

Rock

BWR

BWR

5/7/80 Encl 3 No

5/7/80 Encl 3 No

5/7/80 Encl 3 No

5/7/80 Encl 3 Yes

Yes

Yes

Yes

No

Yes

Yes

Yes

No 4/1/81

Yes 4/1/82

Yes Refuel

Yes 7/1/81

TBD 7/1/81

No 1/1/81

Yes 1/1/81

No 1/1/81

Yes 1/1/82

Yes 1/1/82

BWR 5/7/80 Encl 3 No

BWR

BWR

BWR

C-E & W

BWR

C-E & W

BWR

BWR

BWR

w/i so

cond

5/7/80 Encl 3 No

5/7/80 Encl 3 No

5/7/80

*

5/7/80

5/7/80

Encl 3

Encl 3

Encl 3

Encl 3

Encl 3

No Yes

No Yes

Yes No

Yes No

No Yes

No

No

No

No

Yes

7/1/81

1/1/82

1/1/82

7/1/82

1/1/81

5/7/80 Encl 3 No

5/7/80 Encl 3 No

Yes

Yes

Yes 1/1/82

No 4/1/81

*Requirement formally issued by this letter.
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

II.K.3 Final recommendations,

B&O task force

(continued)

30. SB LOCA

methods

a. Schedule 11/15/80 All

outline

b. Model 1/1/82 All

c. New 1/1/83 or All

analyses 1 yr after

staff

approval

31. Compliance 1/1/83 or All

with CFR 50.46 1 yr

after staff

approval

40. RCP seal See

damage II.K.2.16

43. Effects of See

slug flow II.K.2.15

44. Eval tran- 1/1/81 BWR

5/7/80 Encl 3 No Yes

5/7/80 Encl 3 Yes No

5/7/80 Encl 3 Yes No

5/7/80 Encl 3 Yes No

No 11/15/80

No 1/1/82

No 1/1/83 or 1 yr after

staff

approval

TBD 1/1/83

TBD 1/1/81

TBD 1/1/81

I sient with

single failure

45. Manual depres- 1/1/81

surization

46. Michelson Comple'

concerns

57. Manual act TBD

of ADS

5/7/80 Encl 3 No

BWR 5/7/80 Encl 3 No

te BWR

BWR

5/7/80 None No

Yes

Yes

Yes

Yes

Yes

5/7/80 Encl 3 No

No 7/1/80

No TBD

No Complete

NRR to verify

compliance

No licensee action

until guidelines

approved by staff

III.A.1.1 Emergency

preparedness,

short term

Short-term

improvements

Complete All 10/10/79 NUREG0654No

III.A.1.2 Upgrade emergency

support facilities

1. Interim TSC

OSC & EOF

2. Design

3. Modifications

1/1/80 All 9/13/79 No Yes

TBD

TBD

No Complete

TBD TBD

TBD TBD

TBD

TBD

TBD

TBD

TBD

TBD

TBD

TBO

III.A.2 Emergency

preparedness

1. Upgrade emer- 3/1/81

gency plans

to App. E,

10 CFR 50

2. Meteorological 6/1/83

data

All 8/19/80 NUREG0654No

No

Yes

Yes

Yes 1/2/81

Yes 1/2/81

Procedures submitted

3/1/81

Staged implementation

All 8/19/80 NUREG0654

III.D.1.1 Primary coolant

outside containment

1. Leak reduction Complete

2. Tech specs 12/15/80

All 9/13/79 10/30/79 No Yes

All 7/2/79 7/2/80 Yes No

Yes Complete

Yes 9/1/80
ENCLOSURE 1 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Preimple- Postimple- Tech Licensee

cation Shortened tation Applica- ments cation mentation mentation Spec. Submittal

Item Title Description Schedule bility Issued Issued Approval Review Req. Req. by Remarks

III.D.3.3 Inplant radiation 1. Provide means Complete All 9/13/79 10/30/79 No Yes No Complete

monitoring . to determine

presence of

radioiodine

2. Modifications 1/1/81 All 9/13/79 Encl 3 No Yes Yes 1/1/81

to accurately

measure I2

III.D.3.4 Control-room 1. Review 1/1/81 All 5/7/80 Encl 3 No Yes No 1/1/81

habitability 2. Modifications TBD All 5/7/80 Encl 3 No Yes Yes 1/1/81

I

r1
ENCLOSURE 2

TMI ACTION PLAN REQUIREMENTS

FOR APPLICANTS FOR AN OPERATING LICENSE

2-1
ENCLOSURE 2

TMI ACTION PLAN REQUIREMENTS FOR APPLICANTS

[If implementation date is earlier than issuance of operating license,

FOR AN OPERATING LICENSE

the implementation date will be the licensing date.]

1%,

t6

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tatlon Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

I.A.1.1 Shift technical 1. On shift Fuel load All 9/27/79 11/9/79 Yes

advisor 2. Training per LL Cat B Fuel load All 9/27/79 Encl. 3 No

3. Describe long-term Consist- All * Encl. 3 No

program ent with

OL review

schedule

I.A.1.2 Shift supervisor Delegate nonsafety Fuel load All 9/27/79 11/9/79 No

responsibilities duties

I.A.1.3 Shift manning 1. Limit overtime Fuel load All 6/26/80 7/31/80 No

Enc. 3

2. Minimum shift Fuel load All 6/26/80 7/31/80 Yes Case by case

crew Enc. 3

I.A.2.1 Immediate upgrade of 1. SRO experience Fuel load All 3/28/80 None No

RO & SRO training 2. SROs be ROs, 1 yr Initial All 3/28/80 3/28/80 No

and qualifications criticality Encl. 3

3. 3 mo training Fuel load All 3/28/80 None No

on-shift

4. Modify training Fuel load All 3/28/80 None No

5. Facility certification Fuel load All 3/28/80 None No

I.A.2.3 Administration of Instructors complete 2 mo All 3/28/80 Encl. 3 No

training programs SRO exam prior to

issuance of

license

I.A.3.1 Revise scope & 1. Increase scope 10/1/80 All 3/28/80 None No

criteria for 2. Increase passing grade 10/1/80 All 3/28/80 None No

licensing exams 3. Simulator exams

a. Plants with Prior to Plants * Encl. 3 No

simulators fuel load having

simulators

b. All plants Prior to All Encl. 3 No

fuel load or

10/1/81 whichever is later

Note: For complete reference citiation of NUREG reports, see Appendix A.

*Requirement formally issued by this letter.
ENCLOSURE 2 (CONTINUED)

,9

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

I.B.1.2 Evaluation of Organization, resources Fuel load All 6/26/80 None Yes Draft guideline

organization & tng. & qualifications available.

management for operators & accidents

I.C.1 Short-term accident 1. SB LOCA Fuel load All 9/27/79 11/9/79 No

& procedure 2. Inadequate core cooling

review a. Reanalyze & Fuel load All 9/27/79 Encl. 3 No

propose guidelines

b. Revise procedures First refuel- All 9/27/79 Encl. 3 No

ing outage

after 1/1/82

3. Transients & accidents a. Reanalyze & Fuel load All 9/27/79 Encl. 3 No

propose guidelines

b. Revise procedures First refuel- All 9/27/79 Encl. 3 No

ing outage

after 1/1/82

I.C.2 Shift & relief . Revise procedures to Fuel load All 9/27/79 11/9/79 No

turnover procedures assure plant status

known by new shift

I.C.3 Shift supervisor Corporate directive to Fuel load All 9/27/79 11/9/79 Yes

responsibility establish command duties

& revise plant procedures

I.C.4 Control-room access Establish authority Fuel load All 9/27/79 11/9/79 No

& limit access

I.C.5 Feedback of operat- Review & revise 1/1/81 or All 6/26/80 Encl. 3 No

ing experience procedures prior to

issuance of OL

I.C.6 Verify correct Revise performance 1/1/81 or All * Encl. 3 No

performance of procedures prior to

operating activities fuel load

I.C.7 NSSS vendor rev 1. Low-power test program Fuel load All 6/26/80 None No

of proc 2. Power ascension & Full power All 6/26/80 None No

emergency procedures

*Requirement formally issued by this letter.
ENCLOSURE 2 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

I.C.8 Pilot mon of Correct procedure Full power All 6/26/80 None No

selected emergency based on NRC sample

proc for NTOLs audit

I.D.1 Control-room design Preliminary assessment Prior to All 6/26/80 NUREG-1580 No Guidance and

reviews & schedule for correct- issuance of (Draft) schedule being

ing deficiencies OL developed.

I.D.2 Plant-safety- 1. Description TBD All 6/26/80 Encl. 3 No Guidance and

parameter display 2. Installed TBD All 6/26/80 Encl. 3 No schedule being

console 3. Fully implemented TBD All 6/26/80 Encl. 3 No developed in

NUREG-0696.

I.G.1 Training during 1. Propose tests Fuel load All 6/26/80 No

low-power testing 2. Submit analysis Fuel load All 6/26/80 None Yes

and procedures.

3. Training & results Full power All 6/26/80 No

c. II.B.1 Reactor-coolantsystem vents

1. Design & analyses

2. Install

3. Procedures

Full power All

7/1/82 All

1/1/82 or All

prior to

issuance of OL

9/27/79

9/27/79

9/27/79

11/9/79

Encl. 3

Encl. 3

No

Yes

Yes

II.B.2 Plant shielding 1. Radiation & All 9/27/79 Encl. 3 No

shielding review

2. Corrective actions Full power All 9/27/79 Encl. 3 No

to assure access

3. Complete mods 1/1/82 All 9/27/79 Encl. 3 No

4. Equipment qualification A All CLI-80-21 Encl. 3 No

II.B.3 Postaccident sampling 1. Design review A All 9/27/79 Encl. 3 No

2. Corrective actions Full power All 9/27/79 Encl. 3 Yes

3. Procedures Full power All 9/27/79 Encl. 3 Yes

4. Complete actions 1/1/82 All 9/27/79 Encl. 3 Yes

II.B.4 Training for mitigating 1. Develop training program Fuel load All 3/28/80 3/28/80 No core damage 2. Complete training Full power All 6/26/80 None No

AFour months before operating license is issued or 4 months before date indicated.
ENCLOSURE 2 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

II.D.1 Relief & safety- 1. Describe program Fuel load All 9/27/79 11/9/79 No

valve test & schedule

requirements 2. RV & SV tests Fuel load All 9/27/79 11/9/79 To be

determined

3. Block valve tests Fuel load or PWR * 11/9/79

by 7/1/82, Enc. 3

whichever is

later

II.D.3 Valve position Install in control A All 9/27/79 11/9/79 Yes

indication room Encl. 3

II.E.1.1 Auxiliary feedwater 1. Analysis Full power C-E & W 3/10/80 None No See 3/10/80 and

system evaluation B&W 4/24/80 None No 4/24/80 letters.

2. Modification Full power PWR 4/24/80 None As

required

II.E.1.2 Auxiliary feedwater 1. Initiation

system initiation (a) Control grade Fuel load PWR 9/27/79 11/9/79 Yes

and flow (b) Safety grade A PWR 9/27/79 11/9/79 Yes

2. Flow indication

(a) Control grade Fuel load PWR 9/27/79 11/9/79 Yes

(b) Safety grade A PWR 9/27/79 11/9/79 Yes

II.E.3.1 Emergency power for Installed capability 4 mo prior PWR 9/27/79 11/9/79 Yes

pressurizer heaters to issuance Encl. 3

of SER

II.E.4.1 Dedicated hydrogen 1. Design a All 9/27/79 11/9/79 No

penetrations 2. Review & revise Fuel load All 9/27/79 Encl. 3 No

H2 control proc

3. Install 7/1/81 or All 9/27/79 Encl. 3 No

prior to

issuance of OL

M

AFour months before operating license

Requirement formally issued by this I

is issued or 4 months before date indicated.
ENCLOSURE 2 (CONTINUED)

-J

Clarifi- Implemen- Plant Requite- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

II.E.4.2 Containment isolation 1-4 Implement diverse Prior to All 9/27/79 11/9/79 Yes

dependability isolation issuance of OL Encl. 3

5. Containment press 7/1/81 or All * Encl. 3 Yes

setpoint prior to

issuance of OL

6. Containment purge valves 1/1/81 or All * Encl. 3 Yes

prior to

issuance of OL

7. Radiation signal on 7/1/81 or All * Encl. 3 Yes

purge valves prior to

issuance of OL

II.F.1 Accident- 1. Procedures Fuel load All 9/27/79 11/9/79 No

monitoring Encl 3

instrumentation 2. Install instrumentation

a. Noble gas monitor 1/1/82 A All 9/27/79 11/9/79 Yes

Encl. 3

b. Iodine/particulate 1/1/82 A All 9/27/79 11/9/79 Yes

sampling Encl. 3

c. Containment high 1/1/82 A All 9/27/79 11/9/79 Yes

range monitor Encl. 3

d. Containment pressure 6 mo All 9/27/79 11/9/79 Yes

prior to Encl. 3

issuance of OL

e. Containment water 7/1/82 or All 9/27/79 11/9/79 Yes

level prior to Encl. 3

issuance of OL

f. Containment hydrogen 1/1/82 or All 9/27/79 11/9/79 Yes

prior to Encl. 3

issuance of OL

II.F.2 Instrumentation for 1. Procedures Fuel load PWR 9/27/79 11/9/79 No

detection of instruments Encl. 3

inadequate core- 2. Subcooling meter Fuel load PWR 9/27/79 11/9/79 Yes

cooling Encl. 3

3. Describe other Fuel load All 9/27/79 11/9/79 No

instrumentation Encl. 3

4. Install additional 1/1/82 All 9/27/79 11/9/79 Yes

instrumentation Encl. 3

Requirement

AFour months

formally issued by this letter.

before operating license is issued or 4 months before date indicated.
ENCLOSURE 2 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

II.G.1 Power supplies for Power supply from Fuel load PWR 9/27/79 11/9/79 Yes

pressurizer relief emergency buses

valves, block valves,

& level indicators

II.K.1 IE Bulletins 5. Review ESF valves Fuel load All IEB 79-05 None Yes

79-05A

79-06A

79-06B

79-08

6/26/80

10. Operability status Fuel load All IEB 79-05A None No

79-06A

79-06B

79-08

6/26/80

17. Trip per low-level Fuel load W IEB 79-06A None Yes Also see B/S 6/26/80 II. K. 2.10.

20. Prompt manual reactor Fuel load B&W IEB 79-05B None No

trip 6/26/80

21. Auto SG anticipatory Fuel load B&W IEB 79-05B None Yes

reactor trip 6/26/80

22. Aux heat rem Fuel load BWR IEB 79-08 None No

system, proc 6/26/80

23. RV level, procedures Fuel load BWR IEB 79-08 None Yes

6/26/80

IN

II.K.2 Orders on B&W

plants

2. Procedures to control

AFW ind of ICS

9. FMEA on ICS system

10. Safety-grade trip

anticipatory

13. Thermal mechanical

report

14. Lift frequency of

PORV & SVs

15. Effects of slug

flow on OTSGS

A

A

Fuel load

#

Fuel pover

Fuel power A

B&W

B&W

B&W

B&W

B&W

B&W

6/26/80

6/26/80

6/26/80

6/26/80

6/26/80

6/26/80

None

Encl. 3

Encl. 3

Encl. 3

None

Encl. 3

No

As

required

Yes

As

required

No

No

AFour months before operating license

#Six months before full-power license.

is issued or 4 months before date indicated.
ENCLOSURE 2 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

II.K.2 Orders on B&W 16. RCP seal damage Fuel power A B&W 6/26/80 Encl. 3 No

Plants 17. Voiding in RCS 1/1/820 C-E & W * Encl. 3 No

(continued) 19. Benchmark analysis 1/1/820 C-E & W * Encl. 3 No

seq AFW flow

II.K.3 Final recommendations, 1. Auto PORV isolation 1st refuel PWR * Encl. 3 Yes

B&O task force

0I

2. Report on PORV failures

3. Reporting SV & RV

failures & challenges

5. Auto trip of RCPs

a. Propose mods

b. Modify

7. Evaluation of PORV

opening probability

9. PID controller

10. Applicant's propose

anticipatory trip at

high power

11. Justification use of

certain PORVs

12. Confirm anticipatory

trip

a. Propose modifications

b. Modify 13. HPCI & RCIC init levels

a. Analysis

b. Modify

15. Isolation of HPCI and

RCIC

16. Challenges to & failure

of relief valves

a. Study

b. Modify

6 mo. after

staff approval

1/1/81 A

A

Prior to OL

Full power

Full power

A

A

PWR

All

* Encl. 3 No

6/26/80 None Yes

PWR

PWR

PWR

W

Selected

W

*

*

6/26/80

Encl. 3

Encl. 3

None

No

Yes

No

6/26/80 Encl. 3 No

6/26/80 Encl. 3 Yes

See 5/7/80 letter

to ORS.

Fuel load Plant

specific

6/26/80 None No See NUREG-0611,

Sect. 3.2.4.d.

A

A

A

A

A

W

W

BWR

BWR

BWR

6/26/80 Encl. 3 No

* Encl. 3 Yes

*

*

*

*

*

Encl. 3

Encl. 3

Encl. 3

Yes

Yes

Yes

4/1/81

1st refueling

after staff

approval

BWR

BWR

Encl. 3 No

Encl. 3 Yes

AFour months before operating license is issued or 4 months before date indicated. 0

Or 6 months before fuel load.

Requirement formally issued by this letter.
ENCLOSURE 2 (CONTINUED)

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- rents cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

II.K.3 Final recommendations, 17. ECCS outages In accordance All * Encl. 3 As

B&O task force

(continued)

IC18. ADS actuation

a. Study

b. Proposed mods

c. Modify

21. Restart of LPCS & LCPI

a. Design

b. Modification

22. RCIC suction

a. Procedures

b. Modification

24. Space cooling for

HPCI/RCIC,

modifications

25. Power on pump seals

a. Propose mods

b. Modifications

27. Common reference level

28. Qual of ADS accumlators

30. SB LOCA methods

a. Schedule outline

b. Model

with review

schedule for

licensing

1 yr prior

to OL ' 4 mo prior

to OL

1st refuel

6 mo after

staff approval

1/1/81 A

1st refuel

6 mo after

staff approval

1/1/81 A

1/1/82 A

A

7/1/81 or

6 mo prior

to SER

Full power

7/1/81 A

1/1/82 A

In accordance

with review

schedule

In accordance

with review

schedule

BWR *

BWR *

Encl. 3

Encl. 3

Encl. 3

No

Yes

BWR Yes *

BWR

BWR

BWR

BWR

BWR

BWR

C-E & W

BWR

C-E & W

BWR

BWR

All

*

*

A

A

*

*

A

*

*

*

*

required

Encl. 3 No

Encl. 3 Yes

Encl. 3 No

Encl. 3 Yes

Encl. 3 Yes

Encl. 3 No

Encl. 3

Encl. 3

Encl. 3

Encl. 3

Yes

Yes

Yes

As

required

Encl. 3 No

All * Encl. 3 No

Requirement formally issued by this letter.

AFour months before operating license is issued or 4 months before date indicated.
ENCLOSURE 2 (CONTINUED)

V

Clarifi- Implemen- Plant Require- Clarifi- Tech

cation Shortened tation Applica- ments cation Spec.

Item Title Description Schedule bility Issued Issued Req. Remarks

II.K.3 Final recommendations, c. New analyses In accordance All * Encl. 3 No

B&O task force with review

(continued) schedule

31. Plant-specific analysis 1/1/83A All * Encl. 3 No

44. Evaluate transients 1/1/81A BWR * Encl. 3 As

with single failure required

45. Manual depressurization 1/1/81A BWR * Encl. 3 No

46. Michelson concerns Fuel load BWR * Encl. 3 No

III.A.1.1 Emergency Short-term improvements Fuel load All 8/19/80 NUREG-0654 No Use NUREG-0654

preparedness, until Rev. 1 is

short term issued Rdue out

10/80).

III.A.1.2 Upgrade emergency 1. Establish TSC, OSC, TBD All 9/27/79 11/9/79 No

support facilities EOF (interim basis)

2. Design TBD TBD TBD TBD TBD

3. Modifications TBD TBD TBD TBD TBD

III.A.2 Emergency 1. Upgrade emergency plans Fuel load All 8/19/80 NUREG- No

preparedness to App E, 10 CFR 50 0654

2. Meteorological data Fuel load All 6/26/80 NUREG- No

0654

III.D.1.1 Primary coolant Measure leak rates & Full power All 9/27/79 11/9/79 Yes

outside establish program to Encl. 3

containment keep leakage ALARA

III.D.3.3 Inplant 12 radiation 1. Provide means to Fuel load All 9/27/79 11/9/79 Yes

monitoring determine presence Encl. 3

of radioiodine

2. Modifications to 1/1/81 or All 9/27/79 11/9/79 Yes

accurately measure prior to

radioiodine licensing Encl. 3

III.O.3.4 Control-room 1. Identify and evaluate Full power All 6/26/80 Encl. 3 No

habitability potential hazards

2. Schedule for Full power All 6/26/80 Encl. 3 No

modifications

3. Modifications Full power All 6/26/80 Encl. 3 Yes

AFour months before operating license is issued or 4 months before date indicated.

Requirement formally issued by this letter.
ENCLOSURE 3

CLARIFICATION OF

TMI ACTION PLAN REQUIREMENTS

3-1
I.A.1.l SHIFT TECHNICAL ADVISOR

Position

Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical advisor (STA) may serve more than one unit at a

multiunit site if qualified to perform the advisor function for the various

units.

The STA shall have a bachelor's degree or equivalent in a scientific or

engineering discipline and have received specific training in the response and

analysis of the plant for transients and accidents. The STA shall also receive

training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal

duties to the STAs that pertain to the engineering aspects of assuring safe

operations of the plant, including the review and evaluation of operating

experience.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements resulting from NUREG-0660

and the October 30, 1979 letter from H. R. Denton to all operating nuclear

power plants.

Clarification

The letter of October 30, 1979 clarified the short-term STA requirements. That

letter indicated that the STAs must have completed all training by January 1, 1981.

This paper confirms these requirements and requests additional information.

The need for the STA position may be eliminated when the qualifications of the

shift supervisors and senior operators have been upgraded and the man-machine

interface in the control room has been acceptably upgraded. However, until

those long-term improvements are attained, the need for an STA program will

continue.

The staff has not yet established the detailed elements of the academic and

training requirements of the STA beyond the guidance given in its October 30,

1979 letter. Nor has the staff made a decision on the level of upgrading

required for licensed operating personnel and the man-machine interface in the

control room that would be acceptable for eliminating the need of an STA.

Until these requirements for eliminating the STA position have been established,

the staff continues to require that, in addition to the staffing requirements

specified in its July 31, 1980 letter (as revised by item I.A.1.3 of this

enclosure), an STA be available for duty on each operating shift when a plant

is being operated in Modes 1-4 for a PWR and Modes 1-3 for a BWR. At other

times, an STA is not required to be on duty.

Since the October 30, 1979 letter was issued, several efforts have been made

to establish, for the longer term, the minimum level of experience, education,

I.A.1. 1-13 3-3
and training for STAs. These efforts include work on the revision to ANS-3.1,

work by the Institute of Nuclear Power Operations (INPO), and internal staff

efforts.

INPO recently made available a document entitled "Nuclear Power Plant Shift

Technical Advisor--Recommendations for Position Description, Qualifications,

Education and Training." A copy of Revision 0 of this document, dated April 30,

1980, is attached as Appendix C. Sections 5 and 6 of the INPO document describe

the education, training, and experience requirements for STAs. The NRC staff

finds that the descriptions as set forth in Sections 5 and 6 of Revision 0 to

the INPO document are an acceptable approach for the selection and training of

personnel to staff the STA positions. (Note: This should not be interpreted

to mean that this is an NRC requirement at this time. The intent is to refer

to the INPO document as acceptable for interim guidance for a utility in

planning its STA program over the long term (i.e., beyond the January 1, 1981

requirement to have STAs in place in accordance with the qualification requirements specified in the staff's October 30, 1979 letter).)

No later than January 1, 1981, all licensees of operating reactors shall

provide this office with a description of their STA training program and their

plans for requalification training. This description shall indicate the level

of training attained by STAs by January 1, 1981 and demonstrate conformance

with the qualification and training requirements in the October 30, 1979 letter.

Applicants for operating licenses shall provide the same information in their

application, or amendments thereto, on a schedule consistent with the NRC

licensing review schedule.

No later than January 1, 1981, all licensees of operating reactors shall

provide this office with a description of their long-term STA program,

including qualification, selection criteria, training plans, and plans, if

any, for the eventual phaseout of the STA program. (Note: The description

shall include a comparison of the licensee/applicant program with the abovementioned INPO document. This request solicits industry views to assist NRC

in establishing long-term improvements in the STA program. Applicants for

operating licenses shall provide the same information in their application, or

amendments thereto, on a schedule consistent with the NRC licensing review

schedule.)

Applicability

This requirement applies to all licensees of operating reactors and applicants

for operating licenses.

Implementation

(1) Training that meets the lessons-learned requirements shall be completed by

January 1, 1981 or by the time the fuel-loading license is issued, whichever is later.

(2) A description of the current training program and demonstration of conformance with the October 30, 1979 letter shall be submitted

3-4 I.A.1.1-2
(a) no later than January 1, 1981 for licensees of operating reactors;

and

(b) on a schedule consistent with review schedule for applicants for

operating licenses.

(3) A description of the long-term STA program shall be submitted

(a) no later than January 1, 1981 for licensees of operating reactors;

and

(b) on a schedule consistent with review schedule for applicants for

operating licenses.

Type of Review

Operating reactors will undergo postimplementation review.

Applicants for operating licenses will be reviewed as part of the licensing review.

Documentation Required

Documentation will be required as noted above.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.2.1.b

NUREG-0660

INPO Document, see Appendix C

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

Letter from D. G. Eisenhut, NRC, to All Licensees and Applicants, dated July 31,

1980.

I. A. 1.1-3 3-5
I.A.1.3 SHIFT MANNING

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 be more flexible.

Clarification

Page 3 of the July 31, 1980 letter is superseded in its entirety by the following:

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.

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

of which is to operate the plant with the required staff and develop working

schedules such that use of overtime is avoided, to the extent practicable, for

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

operators, reactor operators, health physicists, auxiliary operators, I&C

technicians 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 staff recognizes that there are diverse opinions on the amount of overtime

that would be considered permissible and that there is a lack of hard data on

the effects of overtime beyond the generally recognized normal 8-hour working

day, the effects of shift rotation, and other factors. NRC has initiated

studies in this area. Until a firmer basis is developed on working hours, the

administrative procedures shall include as an interim measure the following

guidance, which generally follows that of IE Circular No. 80-02.

In the event that overtime must be used (excluding extended periods of shutdown

for refueling, major maintenance or major plant modifications), the following

overtime restrictions should be followed:

3-6 I3. A. 1. 3-1
(1) An individual should not be permitted to work more than 12 hours straight

(not including shift turnover time).

(2) There should be a break of at least 12 hours (which can include shift

turnover time) between all work periods.

(3) An individual should not work more than 72 hours in any 7-day period.

(4) An individual should not be required to work more than 14 consecutive

days without having 2 consecutive days off.

However, recognizing that circumstances may arise requiring deviation from the

above restrictions, such deviation shall be authorized by the plant manager or

his deputy, or higher levels of management in accordance with published procedures and with appropriate documentation of the cause.

If a reactor operator or senior reactor operator has been working more than 12

hours during periods of extended shutdown (e.g., at duties away from the

control board), such individuals shall not be assigned shift duty in the

control room without at least a 12-hour break preceding such an assignment.

NRC encourages the development of a staffing policy that would permit the

licensed reactor operators and senior reactor operators to be periodically

assigned to other duties away from the control board during their normal tours

of duty.

If a reactor operator is required to work in excess of 8 continuous hours, he

shall be periodically relieved of primary duties at the control board, such

that periods of duty at the board do not exceed about 4 hours at a time.

The guidelines on overtime do not apply to the shift technical advisor provided he or she is provided sleeping accommodations and a 10-minute availability

is assured.

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 90 days after July 31, 1980.

See section III.A.1.2 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 November 1, 1980 and by fuel loading for applicants for

operating licenses.

I.A.1.3-2 3-7
(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 postimplementation 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 RequiredChanges to technical specifications will be required for minimum shift crew

manning.

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.

3-8 I.A.1.3-3
NEW GUIDANCE FOR INTERIM REQUIRED SHIFT STAFFING

One Unit, Two Units, Two Units, Three Units,

One Control One Control Two Control Two Control

Operating Status Room Room Rooms Rooms

One Unit Operating* 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SS (SRO)

1 SRO 1 SRO 1 SRO I SRO

2 R0 3 R0 3 R0 4 R0

2 AO 3 AO 3 AO 4 AO

Two Units Operating* NA 1 SS (SRO) 1 SS (SRO) 1 SS (SRO)

1 SRO 2 SRO 2 SRO )Only I SRO & 4 ROs

3 RO 4 RO 5 RO ) required if both

3 AO 4 AO ) units are operated

) from one control

) room

5 AO

All Units Operating* NA 1 SS (SRO) 1 SS (SRO) I SS (SRO)

I SRO 2 SRO 2 SRO

3 R0 4 R0 5 R0

3 AO 4 AO 5 AO

All Units Shut Down 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SS (SRO)

I RO 2 R0 2 R0 3 R0

I AO 3 AO 3 AO 5 AO

!A3

;a

I

I

SS - shift supervisor

SRO - licensed senior

RO - licensed reactor operator

reactor operator AO - auxiliary operator

NOTE: (1) In order to operate or supervise the operation of more than one unit, an operator (SRO or RO)

must hold an appropriate, current license for each such unit.

(2) In addition to the staffing requirements indicated in the table, a licensed senior operator

will be required to directly supervise any core alteration activity.

(3) See item I.A.1.1 for shift technical advisor requirements.

* Modes 1 through 4 for PWRs.

Modes 1 through 3 for BWRs.
SSINS No.: 6830

Accession No.:

UNITED STATES 7912190657

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, P.C. 20555

February 1, 1980

IE Circular No. 80-02

NUCLEAR POWER PLANT STAFF WORK HOURS

Description of Circumstances:

Studies indicate that with fatigue, especially because of loss of sleep, an

individual's detection of visual signals deteriorates markedly, the time it

takes for a person to make a decision increases and more errors are made, and

reading rates decrease. Other studies show that fatigue results in personnel

ignoring some signals because they develop their own subjective standards as

to what is important, and as they become more fatigued they ignore more signals.

Inspections of personnel performance and training since the accident at Three

Mile. Island, have shown that in certain situations facility personnel are

either required or allowed to remain on duty for extended periods of time.

Also, complaints have been received from some licensed nuclear power plant

operators-concerning the number of continuous hours they have been on duty.

Licensee management is responsible for providing a sufficient number of

trained personnel who are in the proper physical condition to operate and

maintain the plant. Licensee management should review their administrative

procedures covering the working hours of nuclear power plant staff. These

procedures should establish a sound policy covering working hours for plant

staff who perform safety related functions (e.g., senior reactor operators,

reactor operators, health physicists, auxiliary operators, I&C technicians,

key maintenance personnel, etc.)

Subcommittee ANS-3 is currently developing criteria to address the subject of

operator work hours. These guidelines will become a part of ANSI N18.7. The

NRC is also considering issuing requirements for administrative procedures

that would control staff overtime. Until either the ANSI Standard is issued

and endorsed by NRC (via a Regulatory Guide) or separate requirements are

issued by NRC, it is recommended that the following guidance be used. The

guidance should be applied to all personnel performing a safety r'elated

function:

1. Scheduled work should be limited to the following maximum work hours:

a. An individual should not be permitted to work more than 12 hours

straight.

3-10 I ,A.1.3-5
IE Circular No. 80-02 February 1, 1980

Page 2 of 2

b. There should be at least a 12-hour break between all work periods.

c. An individual should not work more than 72-hours in any 7-day period.

d. An individual should not work more than 14 consecutive days without

having 2 consecutive days off.

2. In the event that special circumstances arise that require deviation from

the above, such deviations should be authorized by the Station Manager

with appropriate documentation of the cause. Plants should be staffed

and schedules developed to operate such that exceptions are not required.

3. If an operator is required to work in excess of 12 continuous hours, his

duties should be carefully selected. It is preferable that he not be

assigned any task that affects core reactivity or could possibly endanger

the safe operation of the plant.

No written response to this Circular is required. If you desire additional

information regarding this matter, contact the Director of the appropriate NRC

Regional Office.

I.A.1.3-6 3-11
UNITED STATES

NUCLEAR REGULATORY COMMItvSSION

WASHINGTON, 0. C. 20555

July 31, 1980

TO ALL LICENSEES OF OPERATING PLANTS AND APPLICANTS FOR OPERATING LICENSES

AND HOLDERS OF CONSTRUCTION PERMITS

SUBJECT: INTERIM CRITERIA FOR SHIFT STAFFING

This is to provide you with the shift manning requirements as indicated in

item (1) of our letter of May 7, 1980. Pending completion of the long-term

development of criteria for shift staffing and administrative controls, the

NRC staff has developed interim criteria for licensees of operating plants

and applicants for operating licenses. Except for senior reactor operators,

these interim criteria for shift staffing shall remain as described in the

Standard Review Plan, Section 13.1.2, NUREG 75/087. Special requirements

regarding the utilization and qualifications of an on-shift technical advisor

to the shift supervisor were provided in our letter of October 30, 1979.

We have changed the previous requirements for senior reactor operators and

now require that there be one licensed senior reactor operator in the control room

at all times, other than during cold shutdown conditions. This will therefore

require that there be a minimum of two senior reactor operators at each site

at all times, other than during cold shutdown conditions, to assure the

availability of one senior reactor operator in the control room without affecting

the freedom of the shift supervisor to move about the site as needed. The

criteria for reactor and auxiliary operators are stated below and the required

staffing levels for selected station configurations and various plant operating

modes are summarized in the enclosed table.

At any time a licensed nuclear unit is being operated in Modes 1-4 for a PWR

(Power Operation, Startup, Hot Standby, or Hot Shutdown respectively) or in

,Modes 1-3 for a BWR (Power Operation, Startup, or Hot Shutdown respectively),

the minimum shift crew shall include two licensed senior reactor operators

(SRO), one of whom shall be designated as the shift supervisor, two licensed

reactor operators (RO) and two unlicensed auxiliary operators (AO). For a

multi-unit station, depending upon the station configuration, shift staffing

may be adjusted to allow credit for licensed senior reactor operators (SRO)

and licensed reactor operators (RO) to serve as relief operators on more than

one unit; however, these individuals must be properly licensed on each such

unit. At all other times, for a unit loaded with fuel, the minimum shift crew

shall include one shift supervisor who shall be a licensed senior reactor operator

(SRO), one licensed reactor operator (RO) and one unlicensed auxiliary operator.

Adjunct requirements to the shift staffing criteria stated above are as follows:

a. A shift supervisor with a senior reactor operator's license, who is also

a member of the station supervisory staff, shall be onsite at all times

when at least one unit is loaded with fuel.

3-12 I .A .1.3-7
- 2 -

b. A licensed senior reactor operator (SRO) shall, at all times, be in the

control roomjfrom which a reactor is being operated. The shift supervisor

may from time-to-time act as relief operator for the licensed senior reactor

operator assigned to the control room.

c. For any station with more than one reactor containing fuel, the number of

licensed senior reactor operators onsite shall, at all times, be at least

one more than the number of control rooms from which the reactors are being

operated.

d. In addition to the licensed senior reactor operators specified in a., b.,

and c. above, for each reactor containing fuel, a licensed reactor operator

(RO) shall be in the control room at all times.

e. In addition to the operators specified in a., b., c., and d. above, for

each control room from which a reactor is being operated, an additional

licensed reactor operator (RO) shall be onsite at all times and available

to serve as relief operator for that control room. As noted above, this

individual may serve as relief operator for each unit being operated from

that control room, provided he holds a current license for each unit.

f. Auxiliary (non-licensed) operators shall be properly qualified to support

the unit to which assigned.

g. In addition to the staffing requirements stated above, shift crew assignments during periods of core alterations shall include a licensed senior

reactor operator (SRO) to directly supervise the core alterations. This

licensed senior reactor operator may have fuel handling duties but shall

not have other concurrent operational duties.

These criteria do not relieve licensees of any special requirements for additional

operators which may have been imposed for individual units.

General application of revised shift staffing criteria will be the subject of

a rulemaking proceeding. However, these interim criteria will be effective

for plants receiving operating licenses during the interim period (including

TMI-l). Licensees of plants already holding operating licenses shall examine

their current staffing practices and capabilities in light of these interim

criteria and advise this office within 90 days of receipt of this letter of

the date by which their shift staffing could be in compliance with these

criteri a. Licensees of operating plants shall take steps to meet the revised

criteria as soon as practical, but no later than July 1, 1982. In your response

to this letter, you are requested to discuss your plans, schedules and commitments

to meet these staffing criteria. Holders of construction permits who have not

as yet applied for an operating license should factor these criteria into their

recruitment and crew training plans.

I .A.1.3-8 3-13
- 3 -

in addition, 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.

The administrative procedures shall also set forth a policy concerning overtime work for the senior reactor operators, reactor operators, and shift

technical advisor required by these interim criteria. These procedures shall

stipulate that overtime shall not be routinely scheduled to compensate for

an inadequate number of personnel to meet the shift crew staffing requirements.

In the event that overtime must be used, due to unanticipated or unavoidable

circumstances, the following overtime restrictions shall be followed:

(I individual shall not be permitted to work more than 12 hours strai

(no ding shift turnover time).

(2) An individua not be permitted to an 24 hours in any

48 hour period S Q_ ^T6 S Et 49 br)

(3) An individual work an 72 hours in any 7 day period.

individual shall not work more than 1 secutive days without having

two consecutive days off.

However, recognizing that circumstances may arise requiring d on from the

above restrictions, such deviation may be authorized by the plant manager or highei

levels of management in accordance with published procedures and with appropriate

documentation of the cause.

The limitations on overtime follow the guidance provided in IE Circular 80-02,

except for the requirement noted above on the restriction on use of overtime

in circumstances that are unavoidable.

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 90 days after the date of this letter.

Sincerely,

Division o Licensing

Enclosures:

As stated

cc: OR Licensees, and OL Applicants

CP Holders Service Lists

3-14 I .A.1.3-9
INTERIM REQUIRED SHIFT STAFFING

One Unit

One Control

Room

Two Units

One Control

Room

Two Units

Two Control Rooms

Three Uni s

Two Co ol Rooms

One Unit Operating* 1 SS (SRO)

1 SRO

0

2A

1 SS (SRO) 1 SS (SRO) 5 (SRO)

1 SRO A 1 SRO 1 SRO

3 RO 3 RO 4 RO

3 AO 3 AO 4 AO

Two Units Operat-ing* NA 1 SS (SRO) 1 SS (SRO) I SS (SRO)

1 SRO 2 SRO 2 SRO Only one SRO

3 RO 4 RO required if both

AO 4 AO units are operated

from one control

room)

W \ ] /4 RO

to 11 4R5AO

All Units Operating*. NA I SS (SRO) /I XSS GI",!I (SRO) 1SS(SRO) Ss(ro 1SO S2 SRO

3R /4R5 RO

4AO \4 AO

All Units Shu~down 1 SS (SRO) I SS (SRO) (SRO)\1SS(l0 I81SS 

I RO /2 RO _ 2 RO \3 RO

AO /3 AO 3AO 5 AO

SS - Shift Supervisor RO - Licensed Reactor Operator

SRO - Licensed Senior Reactor erator AO - Auxiliary Operator

NOTE: 1. In order o operate or supervise the operation of more than one unit, an operator (SRO or 10O) nitist

2 hold appropriate, current license for each such unit.

2. ddition to the staffing requirements indicated in the table a licensed senior operdtor will'h.

/ required to directly supervise any core alteration activity.

odes 1 through 4 for PWRs.

/ pModes 1 through 3 for BWRs.
I.A.2.1 IMMEDIATE UPGRADING OF REACTOR OPERATOR AND SENIOR REACTOR OPERATOR

TRAINING AND QUALIFICATIONS

Position

Effective December 1, 1980, an applicant for a senior reactor operator (SRO)

license will be required to have been a licensed operator for 1 year.

Changes to Previous Requirements

Changes to the previous requirements will permit various paths to provide

experience equivalent to 1 year's experience as a licensed operator.

Clarification

Applicants for SRO either come through the operations chain (C operator to B

operator to A operator, etc.) or are degree-holding staff engineers who obtain

licenses for backup purposes.

In the past, many individuals who came through the operator ranks were administered SRO examinations without first being an operator. This was clearly a

poor practice and the letter of March 28, 1980 requires reactor operator

experience for SRO applicants.

However, NRC does not wish to discourage staff engineers from becoming licensed

SROs. This effort is encouraged because it forces engineers to broaden their

knowledge about the plant and its operation.

In addition, in order to attract degree-holding engineers to consider the

shift supervisor's job as part of their career development, NRC should provide

an alternate path to holding an operator's license for 1 year.

The track followed by a high-school graduate (a nondegreed individual) to

become an SRO would be 4 years as a control room operator, at least one of

which would be as a licensed operator, and participation in an SRO training

program that includes 3 months on shift as an extra person.

The track followed by a degree-holding engineer would be, at a minimum, 2 years

of responsible nuclear power plant experience as a staff engineer, participation

in an SRO training program equivalent to a cold applicant training program,

and 3 months on shift as an extra person in training for an SRO position.

Holding these positions assures that individuals who will direct the licensed

activities of licensed operators have had the necessary combination of education,

training, and actual operating experience prior to assuming a supervisory role

at that facility.

The staff realizes that the necessary knowledge and experience can be gained

in a variety of ways. Consequently, credit for equivalent experience should

be given to applicants for SRO licenses.

3-16 I.A.2.1-1
Applicants for SRO licenses at a facility may obtain their 1-year operating

experience in a licensed capacity (operator or senior operator) at another

nuclear power plant. In addition, actual operating experience in a position

that is equivalent to a licensed operator or senior operator at military

propulsion reactors will be acceptable on a one-for-one basis. Individual

applicants must document this experience in their individual applications in

sufficient detail so that the staff can make a finding regarding equivalency.

Applicants for SRO licenses who possess a degree in engineering or applicable

sciences are deemed to meet the above requirement, provided they meet the

requirements set forth in sections A.J.a and A.2 in enclosure 1 in the letter

from H. R. Denton to all power reactor applicants and licensees, dated March

28, 1980, and have participated in a training program equivalent to that of a

cold senior operator applicant.

NRC has not imposed the 1-year experience requirement on cold applicants for

SRO licenses. Cold applicants are to work on a facility not yet in operation;

their training programs are designed to supply the equivalent of the experience

not available to them.

Applicability

This requirement applies to all operating reactors and applicants for operating

licenses (after initial criticality).

Implementation

This requirement applies to applicants for senior reactor operator licenses

received after December 1, 1980.

Type of Review

A postimplementation review will be performed.

Documentation Required

No documentation is required from the facility. Information will be contained

in individual applications.

Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

Letter from H. R. Denton, NRC, to All Power Reactor Applicants and Licensees,

dated March 28, 1980.

I.A.2.1-2 3-17
d-tal REac,

0 UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

MAR 2 8 1980

AIL MMHER REACTOR APPLICANTS AND LICENSEES

Gentlenmen:

SUBJECT: QUALIFICATIONS OF REACTOR OPERATORS

In a letter dated September 13, 1979, we informed you of NRR requirements

established as of that date based on our review of the IMI-2 accident.

*Enclosure 9 to the letter outlined the staff recommendations concerning

improvements in the area of operator training for your information.

Since that time, the Comnission has acted on the staff recommendations.

It is the purpose of this letter to set forth the revised criteria to be

used by the staff in evaluating reactor operator training and licensing

that can be irrplamented under the current regulations and to establish an

effective date for their implementation. Other criteria that will be

established require additional staff work are also addressed. However,

inplementation dates cannot be provided at this time. Commission review

in the area of operator training and qualification is continuing and can

be expected to result in additional criteria. Finally, requirements will

be established through rule making proceedings.

Enclosure 1 details the revised criteria and the effective date for their

implementation. Your attention is specifically directed to Sections A,

B and C of Enclosure 1 since these call out new criteria that will be

imrplemented in the near future; therefore, your plans regarding training

and licensing activities should be promptly revised to conform to these

criteria.

Enclosures 2 and 3 provide guidance for establishing training programs in

heat transfer, fluid flow and thermodynamics; and mitigating core damage.

Enclosure 4 details control manipulations for requalification programs.

Based on our understanding of the industry's reasons for establishing the

Institute of Nuclear Power Operations and our review of the latest

revisions to applicable ANSI standards, we believe you share our desire

to significantly upgrade the requirements for operations personnel.

3-18 I.A.2.1-3
- 2 -

Therefore, we urge you to effect the implementation of the criteria as

soon as practicable rather than wait for the stated implementation

date. Also, we urge you to start planning for the long range requirements so that they may be rapidly imrplemented upon completion of the

rule making procedure.

Harold R. Denton, Director

Office of Nuclear Reactor Regulation

Enclosures

1. Requirements for Reactor Operator

Training and Licensing

2. Training in Heat Transfer, Fluid

Flow and Thermodynardcs

3. Training Criteria For Mitigating

Core Darage

4. Control Manipulations

I.A.2.1-4 3-19
ENCLOSURE 1

CRITERIA FOR REACTOR OPERATOR

TRAINING AND LICENSING

A. Eligibility Requirerents to be AdmLnistered an Examination.

1. Experience*

a. Applicants for senior operator licenses shall have 4

years of responsible power plant experience. Responsible

power plant experience should be that obtained as a

control room operator (fossil or nuclear) or as a

power plant staff engineer involved in the day-to-day

activities of the facility, comnencing with the final

year of construction. A maximLu of 2 years power

plant experience may be fulfilled by academic or

related technical training, on a one-for-one tirme

basis. ,,o years shall be nuclear power plant experience.

At least 6 months of the nuclear power plant experience

shall be at the plant for which he seeks a license.

Effective date: Applications received on or after

May 1, 1980.

*Precritical applicants will be required to meet unique qualifications

designed to acconmodate the fact that their facility has not yet been in

operation.

3-20 I.A.2.1-5
-2-

b. .n'!.icants for senior operator licenses shall have held an

operator's license for 1 year.

Flsective date: Applications received after December 1,

).i80.

2. Training

a. Senior operator*: Applicants shall have 3 months of

shift training as an extra man on shift.

b. Control rocm operator*: Applicants shall have 3

nmths training on shift as an extra person in the

control rom.

Effective date: Applications received after August

1, 1980.

c. Training programs shall be rmdified, as necessary, to

provide:

1) Training in heat transfer, fluid flow and thermodyrianics.

2) Training in the uose of installed plant systeus

to control or mitigate an accident in which the

core is severely damaged.

3) Increased emphasis on reactor and plant

transients.

*Precritical applicants will be required to meet unique qualifications

desly:ied to -4ec-v.odate the fact that their facility has not yet been in

lmertitA-t. I.A.2.1-6 3-21
-3-

Effective date: Present programs htave been modified

in response to Bulletins and Orders. Revised programs

should be submitted for OLB review by August 1, 1980.

d. Training center and facility instructors who teach

system, integrated responses, transient and simulator

courses shall demonstrate their competence to NRC by

successful completion of a senior operator examination.

Effective date: Applications should be submitted no

later than August 1, 1980 for individuals who do not

already hold a senior operator license.

e. Instructors shall be enrolled in appropriate requalification programs to assure they are cognizant of

current operating history, probles, and changes to

procedures and administrative limitations.

Effective date: Programx should be initiated May 1,

1980. Programs should be submitted to OLB for review

by August 1, 1980.

3. Facility Certifications

Certifications coipleted pursuant to Sections 55.10(a)(6)

and 55.33a(4) and (5) of 10 CFR Part 55 shall be signed by

the highest level of corporate nanagement for plant operation

(for example, Vice President for Operations).

3-22 I.A.2,1-7
-4-

Effective I'3te: Applications received on or after "ay 1,

1980.

B. NRC Examinations

1. Increased Scope of Fxaminations

a. A new category shall be added to the operator written

examination entitled, Principles of Heat Transfer and

Fluid Mechanics."

b. A new category shall be added to the senior operator

written examination entitled, "Theory of Fluids and

Thermodynamics."

c. Time limits shall be imposed for completion of the written

examinations:

1. Operator: 9 hours.

2. Senior Operator: 7 hours.

d. The passing grade for the written examination shall be 80%

overall and 70% in each category.

e. All applicants for senior operator licenses shall be

required to be administered an operating test as well as

the written examination.

f. Applicants will grant permission to NRC to inform their

facility management regarding the results of the eMzminations for purposes of enrollment in requalification

program.

Effective date: Examinations administered on or after May

1, 1980 for items a. through e. Applications received on

or after May 1, 1980 for Iten f.

I.A.2.1-8 3-23
-5-

C. Requalification Program

1. Content of the licensed operator requalification programs shall

be mrrdified to include instruction in heat transfer, fluid

flow, thermodynamics and mitigation of accidents involving a

degraded core.

Effective date: May 1, 1980.

2. The criteria for requiring a licensed individual to participate

in accelerated requalification shall be modified to be consistent

with the new passing grade for issuance of a license; 80Z

overall and 70Z each category.

Effective date: Concurrent with the next facility administered

annual requalification examination after the issue date of this

letter.

3. Programs should be modified to require the control manipulations

listed in Enclosure 4. Normal control manipulations, such as

plant or reactor startups, must be performed. Control manipulations during abnormal or energency operations nmst be

walked through with, and evaluated by, a mmber of the training

staff at a rrdninu. An appropriate simulator may be used to

satisfy the requirements for control manipulations.

Effective date: Programs modified by August 1, 1980. Renewal

applications received after Noverber 1, 1980 mist reflect

compliance with the program.

3-24 I.A.2.1-9
-6-

D. Lorn Range Criteria and/or Requireeants

The following require additional staff work and/oa rulemakbig prior

to their ieplaemntation.

1. Qualifications

a. Shift supervisors shall have an engineering degree or

equivalent qualifications.

b. Senior operators shall have successfully completed a

course in appropriate ergineering and scientific subject

equal to 60 credit hours of college level subjects.

2. Training

a. All applicants shall attend simulator training program.

Required control manipulations and exercises to be

performed shall be the same for "cold' and 'hot"

applicants.

b. Eligibility requirements shall be developed for instructors,

in addition to that listed in A.2 above.

3. NRC Examinations

a. NRC shall admdnister the certification examinations that

are presently administered at the conclusion of the off-site

portion of the cold training programs.

b. All applicants shall be required to be administered a

simulator examination in addition to the written examinations and plant oral tests.

c. NRC. shall administer the requalification program annual

examination.

I.A.2.1-10 3-25
-7-

4. Reaualification Proprans

All licensees shall participate in simulator programs as part of the

requalification program. Control manipulations shall be performed

pursuant to Enclosure 4.

3-26 I.A.2.1-11
ENCLOSURE 2

TRAINING IN HEAT TRuANSFER, FLUID FLOW AND TMNERMODYNAMICS

1. Basic Properties of Fluids and Matter. This section should cover a

basic introduction to matter and its properties. This section should

include such concepts as temperature teasurements and effects, density

and its effects, specific weight, buoyancy, viscosity and other properties

of fluids. A working knowledge of steam tables should also be included.

Energy mDvement should be discussed including such fundamentals as heat

exchange, specific heat, latent heat pf vaporization and sensible heat.

2. Fluid Statics. This section should cover the pressure, temperature and

volume effects on fluids. Fxample of these parametric changes should be

illustrated by the instructor and related calculations should be performed

by the students and discussed in the training sessions. Causes and

effects of pressure and temperature changes in the various components and

systems should be discussed as applicable to the facility with particular

emphasis on safety significant features. The characteristics of force

and pressure, pressure in liquids at rest, principles of hydraulics,

saturation pressure and temperature and subcooling should also be included.

3. Fluid Dynamics. This section should cover the flow of fluids and such

concepts as Bernoulli's principle, energy in moving fluids, flow measure

theory and devices and pressure losses due to friction and orificing.

Other concepts and terms to be discussed in this section are NPSH, carry

over, carry under, kinetic energy, head-loss relationships and troo phase

I.A.2.1-12 3-27
flow fundamentals. Practical applications relating to the reactor coolant

system and steam generators should also be included.

4. Heat Transfer by Conduction, Convection and Radiation. This section

should cover the fundamentals of heat transfer by conductions. This

section should include discussions on such concepts and terms as specific

heat, heat flux and atomic action. Heat transfer characteristics of fuel

rods and heat exchangers should be included in this section.

This section should cover the fundamentals of heat transfer by convection. Natural and forced circulation should be discussed as applicable

to the various systems at the facility. The convection current patterns

created by expanding fluids in a confined area should be included in this

section. Heat transport and fluid flow reductions or stoppage should be

discussed due to steam and/or noncondensible gas formation during normal

and accident: conditions.

This section should cover the fundamentals of heat transfer by thermal

radiation in the form of radiant energy. The eletromagnetic energy

emitted by a body as a result of its temperature should be discussed and

illustrated by the use of equations and sample calculations. Comparisons

should be nude of a black body absorber and a white body emitter.

5. Change of Phase - Boiling. This section should include descriptions of

the state of natter, their inherent characteristics and therrmdynamic

properties such as enthalpy and entropy. Calculations should be performed

involving steam quality and void fraction properties. The types

- 2 -

3-28 I.A.2.1-13
of boiling should be discussed as applicable, to the facility during

normal evolutions and accident conditions.

6. Birnout and Flow Instability. This section should cover descriptions

and mechanisns for calculating such terms as critical flux, critical

power, DME ratio and hot channel factors. This section should also

include instructions for preventing and monitoring for clad or fuel

damage and flow instabilities. Sanple calculations should be illustrated

by the instructor and calculations should be performed by the students

and discussed in the training sessions. 1-1ethods and procedures for using

the plant ccoputer to determine quantitative values of various factors

during plant operation and plant heat balance determinations should also

be covered in this section.

7. Reactor Heat Transfer Limits. This section should include a discussion

of heat transfer limits by examiniing fuel rod and reactor design and

limitations. The basis for the limits should be covered in this section

along with recoamended methods to ensure that limits are not approached

or exceeded. This section should cover discussions of peaking factors,

radial and axial power distributions and changes of these factors due to

the influence of other variables such as roderator temperature, xenon and

control rod position.

-3-

I.A.2.1-14 3-29
ENCLOSURE 3

TRAINING CRITERIA FOR MITIGATING 0YRE DAMAGE

A program is to be developed to ensure that all operating personnel are training

in the use of installed plant systems to control or mitigate an accident in

i.hich the core is severely damaged. The training program should include the

following topics.

A. Incore Instrutentation

1. Use of fixed or movable incore detectors to determine extent of core

damage and geometry changes.

2. Use of thermocouples in determining peak temperatures; methods for

extended range readings; methods for direct readings at terminal

junctions.

3. Methods for calling up (printing) incore data from the plant computer.

B. Excore Nuclear Instrumentation (NIS)

1. Use of NIS for determination of void formation; void location basis

for NIS response as a function of core temperatures and density

changes.

C. Vital Instrurnentation

1. Instrumentation response in an accident environment; failure sequence

(time to failure, rethod of failure); indication reliability (actual

vs indicated level).

3-30 I.A.2.1-15
2. Alternative methods for measuring flows, pressures, levels, and

temperatures.

a. Determination of pressurizer level if all level transmitters

fail.

b. Determination of letdown flow with a clogged filter (low flow).

c. Determination of other Reactor Coolant System parameters if the

primary Tethod of ueasurement has failed.

D. Primary Chenistry

1. Expected chemistry results with severe core damage; consequences of

transferring small quantities of liquid outside containment; importance

of using leak tight systemns.

2. Expected isotopic breakdown for core damage; for clad damage.

3. Corrosion effects of extended -nTersion in primary waters time to

failure.

E. 'Radiation Tbnitoring

1. Response of Process and Area Monitors to severe damages; behavior of

detectors when saturated; method for detecting radiation readings by

direct measur-emnt at detector output (overranged detector); expected

accuracy of detectors at different locations; use of detectors to

determine extent of core damage.

2. Methods of determining dose rate inside containment from measurerents

taken outside containment.

-2-

I.A.2.1-16 3-31
F. Gas Generation

1. ?Methods of 12 generation during an accident; other sources of gas

(Xe, Ke); techniques for venting or disposal of non-condensibles.

2. E1 flawrability and explosive limit; sources of 02 in containment or

Reactor Coolant System.

- 3 -

3-32 I.A.2.1-17
Suggested References:

Collier, J.,

1972.

Eckert, E.

New York:

El -Wakil, M.

1971.

Gebhart, B.

Mboney, D.

G. Convection Boiling and Condensation. New York: McGraw-Hill,

R. G. and Drake, R. M., Jr. Analysis of Beat and Mass Transfer.

McGraw-Hill, 1973.

M. Nuclear Beat Transport. Scranton, PA: International,

Heat Transfer. 2nd ed. New York: McGraw-Hill, 1971.

Mechanical Engineering Theriodynamics. Prentice Hall, 1953.

I.A.2.1-18 3-33
ENCLOSURE 4

COJ.-ROL TNIPTLIATIONS

The following control manipulations and plant evolutions Where applicable to

the plant design are acceptable for meeting the reactivity control manipulations required by Appendix A, Paragraph 3.a. of 10 CFR Part 55. The starred

items shall be perforired on an annual basis; all other items shall be

performed on a two-year cycle. Hokwever, the requalification programs shall

contain a comnitment that each individual shall perform or participate in a

combination of reactivity control manipulations based on the availability of

plant equipment and system. Those control manipulations which are not

performed at the plant may be performed on a simulator. The use of the Technical

Specifications should be maximized during the simulator control manipulations.

Personnel with senior licenses are credited with these activities if they

direct or evaluate control manipulations as they are performed.

PWR/BWR/HTGR

*(1) Plant or reactor startups to include a range that reactivity

feedback from nuclear heat addition is noticeable and heatup rate is

established.

(2) Plant shutdown.

*(3) Manual control of steam generators and/or feedwater during startup

and shutdown.

(4) Boration and or dilution during power operation.

*(5) Any significant (> 10%) power changes in manual rod control or

recirculation flow.

(6) Any reactor power change of 10% or greater where load change is

performed with load limit control or where flux, temperature, or

speed control is on ramial (for HTGR).

I.A.2.1-19
*(7)

(8)

(9)

*(10)

(11)

(12)

(13)

(14)

(15)

*(16)

(17)

Loss of coolant including:

1. significant WRI steam generator leaks

2. inside and outside primary containment

3. large and small, including leak-rate determination

4. saturated Peactor Coolant response (PIR).

Loss of instrument air (if silmulated plant specific).

Loss of electrical power (and/or degraded power sources).

Loss of core coolant flow/natural circulation.

Loss of condenser vacuu.

Loss of service water if required for safety.

Loss of shutdown cooling.

loss of ca-rponent cooling system or cooling to an individual

component.

Loss of normal feedwater or nonral feedwater system failure.

Loss of all feedwater (normal and emergency).

Loss of protective system channel.

(18)

(19)

(20)

(21)

(22)

(23)

(24)

(25)

(26)

(27)

1ispositioned control rod or rods (or rod drops).

Inability to drive control rods.

Conditions requiring use of emergency boration or standby liquid

control system.

Fuel cladding failure or high activity in reactor coolant or offgas.

TArbine or generator trip.

Malfunction of automatic control system(s) which affect reactivity.

IMalfunction of reactor coolant pressure/volume control system.

Reactor trip.

Main steam line break (inside or outside contairnent).

nuclear instrumentation failure(s).

-2-

I .A.2.1-20 3.35
I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS

Position

Pending accreditation of training institutions, licensees and applicants for

operating licenses will assure that training center and facility instructors

who teach systems, integrated responses, transient, and simulator .courses

demonstrate senior reactor operator (SRO) qualifications and be enrolled in

appropriate requalification programs.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements included in the letter of

March 28, 1980 from H. R. Denton to all power reactor applicants and licensees.

Clarification

The above position is a short-term position. In the future, accreditation of

training institutions will include review of the procedure for certification

of instructors. The certification of instructors may, or may not, include

successful completion of an SRO examination.

The purpose of the examination is to provide NRC with reasonable assurance

during the interim period, that instructors are technically competent.

The requirement is directed to permanent members of training staff who teach

the subjects listed above, including members of other organizations who routinely

conduct training at the facility. There is no intention to require guest

lecturers who are experts in particular subjects (reactor theory, instrumentation, thermodynamics, health physics, chemistry, etc.) to successfully complete

an SRO examination. Nor is it intended to require a system expert, such as

the instrument and control supervisor teaching the control rod drive system,

to sit for an SRO examination.

Applicability

This requirement applies to all operating reactors and operating license

applicants.

Implementation

The requirements for operating reactors have been completed. Applications for

SRO examinations should be submitted. All applicants for operating license

should submit documentation 2 months prior to the expected issuance of an

operating license.

Type of Review

A postimplementation review will be performed.

Documentation Required

No documentation is required.

3-36 I.A.2.3-1
Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

Letter from H. R. Denton, NRC, to All Power Reactor Applicants and Licensees,

dated March 28, 1980.

I.A.2.3-2 3-37
I.A.3.1 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS--SIMULATOR

EXAMS (ITEM 3)

Position

Simulator examinations will be included as part of the licensing examinations.

Changes to Previous Requirements and Guidance

The administration of simulator examinations will be deferred for applicants

whose facilities do not have simulators on site as of October 1, 1980. These

deferred simulator examinations will be initiated by October 1, 1981.

Clarification

The clarification does not alter the staff's position regarding simulator

examinations.

The clarification does provide additional preparation time for utility companies and NRC to meet examination requirements as stated. A study is under way to

consider how similar a nonidentical simulator should be for a valid examination.

In addition, present simulators are fully booked months in advance.

Application of this requirement was stated on June 1, 1980 to applicants where

a simulator is located at the facility. Starting October 1, 1981, simulator

examinations will be conducted for applicants of facilities that do. not have

simulators at the site.

NRC simulator examinations normally require 2 to 3 hours. Normally, two

applicants are examined during this time period by two examiners.

Utility companies should make the necessary arrangements with an appropriate

simulator training center to provide time for these examinations. Preferably

these examinations should be scheduled consecutively with the balance of the

examination. However, they may be scheduled no sooner than 2 weeks prior to

and no later than 2 weeks after the balance of the examination.

Applicability

This requirement applies to all applicants for operator and senior operator

licenses at power reactors.

Implementation

The schedule for operating reactors is October 1, 1981 for licensees without

simulators and June 1980 for licensees with simulators.

The schedule for applicants for operating license without simulators is October 1,

1981 or prior to fuel load, whichever is later, including cold examinations.

The schedule for applicants for operating license with simulators is prior to

full load including cold examination.

3-38 I.A.3.1-1
Type of Review

No review will be performed. Arrangements will be made during the normal

scheduling of examinations.

Documentation Required

No documentation is required. Arrangements will be made during the normal

scheduling of examinations.

Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

Letter from H. R. Denton, NRC, to All Power Reactor Applicants and Licensees,

dated March 28, 1980.

I. A. 3.1-2 3-39
I.B.1.2 INDEPENDENT SAFETY ENGINEERING GROUP

Position

Each applicant for an operating license shall establish an onsite independent

safety engineering group (ISEG) to perform independent reviews of plant operations.

The principal function of the ISEG is to examine plant operating characteristics,

NRC issuances, Licensing Information Service advisories, and other appropriate

sources of plant design and operating experience information that may indicate

areas for improving plant safety. The ISEG is to perform independent review

and audits of plant activities including maintenance, modifications,.operational

problems, and operational analysis, and aid in the establishment of programmatic

requirements for plant activities. Where useful improvements can be achieved,

it is expected that this group will develop and present detailed recommendations

to corporate management for such things as revised procedures or equipment

modifications.

Another function of the ISEG is to maintain surveillance of plant operations

and maintenance activities to provide independent verification that these

activities are performed correctly and that human errors are reduced as far as

practicable. ISEG will then be in a position to advise utility management on

the overall quality and safety of operations. ISEG need not perform detailed

audits of plant operations and shall not be responsible for sign-off functions

such that it becomes involved in the operating organization.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements, however further guidance is

provided in the "Clarification" section that follows.

Clarification

The new ISEG shall not replace the plant operations review committee (PORC)

and the utility's independent review and audit group as specified by current

staff guidelines (Standard Review Plan, Regulatory Guide 1.33, Standard Technical

Specifications). Rather, it is an additional independent group of a minimum

of five dedicated, full-time engineers, located onsite, but reporting offsite

to a corporate official who holds a high-level, technically oriented position

that is not in the management chain for power production. The ISEG will

increase the available technical expertise located onsite and will provide

continuing, systematic, and independent assessment of plant activities.

Integrating the shift technical advisors (STAs) into the ISEG in some way

would be desirable in that it could enhance the group's contact with and

knowledge of day-to-day plant operations and provide additional expertise.

However, the STA on shift is necessarily a member of the operating staff and

cannot be independent of it.

It is expected that the ISEG may interface with the quality assurance (QA)

organization, but preferably should not be an integral part of the QA organization.

3-40 I.B.1.2-1
The functions of the ISEG require daily contact with the operating personnel

and continued access to plant facilities and records. The ISEG review functions

can, therefore, best be carried out by a group physically located onsite.

However, for utilities with multiple sites, it may be possible to perform

portions of the independent safety assessment function in a centralized location

for all the utility's plants. In such cases, an onsite group still is required,

but it may be slightly smaller than would be the case if it were performing

the entire independent safety assessment function. Such cases will be reviewed

on a case-by-case basis.

At this time, the requirement for establishing an ISEG is being applied only to

applicants for operating licenses in accordance with Action Plan item I.B.1.2.

The staff intends to review this activity in about a year to determine its

effectiveness and to see whether changes are required. Applicability to

operating plants will be considered in implementing long-term improvements in

organization and management for operating plants (Action Plan item I.B.1.1).

Applicability

This requirement applies to all applicants for operating license.

Implementation

This requirement shall be implemented prior to issuance of an operating license

(or fuel-loading license).

Type of Review

A preimplementation review will.be performed.

Documentation Required

Each applicant for an operating license shall document in its application or

amendments thereto, its plan for establishing and staffing the ISEG, including

the qualifications of and the training to be given the ISEG staff.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0660

NUREG-0694, Item I.B.1.1 and Item I.B.1.2

I.B.1.2-2 3-41
I.C.1 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR

TRANSIENTS AND ACCIDENTS

Position

In letters of September 13 and 27, October 10 and 30, and November 9, 1979,

the Office of Nuclear Reactor Regulation required licensees of operating

plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency

procedure guidelines, upgrade emergency procedures, including procedures for

operating with natural circulation conditions, and to conduct operator retraining

(see also item I.A.2.1). Emergency procedures are required to be consistent

with the actions necessary to cope with the transients and accidents analyzed.

Analyses of transients and accidents were to be completed in early 1980 and

implementation of procedures and retraining were to be completed 3 months

after emergency procedure guidelines were established; however, some difficulty

in completing these requirements has been experienced. Clarification of the

scope of the task and appropriate schedule revisions are being develop6d. In

the course of review of these matters on Babcock and Wilcox (B&W)-designed

plants, the staff will follow up on the bulletin and orders matters relating

to analysis methods and results, as listed in NUREG-0660, Appendix C (see

Table C.1, items 3, 4, 16, 18, 24, 25, 26, 27; Table C.2, items 4, 12, 17,

18, 19, 20; and Table C.3, items 6, 35, 37, 38, 39, 41, 47, 55, 57).

Changes to Previous Requirements and Guidance

A. Modification to Clarification

(1) Addresses owners' group and vendor submittals.

(2) References to task action plan items I.C.8 and I.C.9.

(3) Scope of procedures review is explained.

(4) Establishes configuration control of guidelines for emergency

procedures.

B. Modification to Implementation

(1) Deleted reference to NUREG-0578, Recommendation 2.1.9 for item

I.C.1(a)2, inadequate core cooling.

Clarification

The letters of September 13 and 27, October 10 and 30, and November 9, 1979,

required that procedures and operator training be developed for transients and

accidents. The initiating events to be considered should include the events

presented in the final safety analysis report (FSAR) loss of instrumentation

buses, and natural phenomena such as earthquakes, floods, and tornadoes. The

purpose of this paper is to clarify the requirements and add additional requirements for the reanalysis of transients and accidents and inadequate core

cooling.

3-42 I.C.1-1
Based on staff reviews to date, there appear to be some recurring deficiencies

in the guidelines being developed. Specifically, the staff has found a lack

of justification for the approach used (i.e., symptom-, event-, or functionoriented) in developing diagnostic guidance for the operator and in procedural

development. It has also been found that although the guidelines take implicit

credit for operation of many systems or components, they do not address the

availability of these systems under expected plant conditions nor do they

address corrective or alternative actions that should be performed to mitigate

the event should these systems or components fail.

The analyses conducted to date for guideline and procedure development contain

insufficient information to assess the extent to which multiple failures are

considered. NUREG-0578 concluded that the single-failure criterion was not

considered appropriate for guideline development and called for the consideration

of multiple failures and operator errors. Therefore, the analyses that support

guideline and procedure development should consider the occurrences of multiple

and consequential failures. In general, the sequence of events for the transients

and accidents and inadequate core cooling analyzed should postulate multiple

failures such that, if the failures were unmitigated, conditions of inadequate

core cooling would result.

Examples of multiple failure events include:

(1) Multiple tube ruptures in a single steam generator and tube rupture in

more than one steam generator;

(2) Failure of main and auxiliary feedwater;

(3) Failure of high-pressure reactor coolant makeup system;

(4) An anticipated transient without scram (ATWS) event following a loss of

offsite power, stuck-open relief valve or safety/relief valve, or loss of

main feedwater; and

(5) Operator errors of omission or commission.

The analyses should be carried out far enough into the event to assure that

all relevant thermal/hydraulic/neutronic phenomena are identified (e.g., upper

head voiding due to rapid cooldown, steam generator stratification). Failures

and operator errors during the long-term cooldown period should also be

addressed.

The analyses should support development of guidelines that define a logical

transition from the emergency procedures into the inadequate core cooling

procedure including the use of instrumentation to identify inadequate core

cooling conditions. Rationale for this transition should be discussed.

Additional information that should be submitted includes:

(1) A detailed description of the methodology used to develop the guidelines;

(2) Associated control function diagrams, sequence-of-event diagrams, or

others, if used;

I. C. 1-2 3-43
(3) The bases for multiple and consequential failure considerations;

(4) Supporting analysis, including a description of any computer codes used;

and

(5) A description of the applicability of any generic results to plant-specific

applications.

Owners' group or vendor submittals may be referenced as appropriate to support

this reanalysis. If owners' group or vendor submittals have already been

forwarded to the staff for review, a brief description of the submittals and

justification of their adequacy to support guideline development is all that

is required.

Pending'staff approval of the revised analysis and guidelines, the staff will

continue the pilot monitoring of emergency procedures described in Task Action

Plan item I.C.8 (NUREG-0660). For PWRs, this will involve review of the loss of

coolant, steam-generator-tube rupture, loss of main feedwater, and inadequate

core cooling procedures. The adequacy of each PWR vendor's guidelines will be

identified to each NTOL during the emergency-procedure review. Since the

analysis and guidelines submitted by the General Electric Company (GE) owners'

group that comply with the requirements stated above have been reviewed and

approved for trial 'implementation on six plants with applications for operating

licenses pending, the interim program for BWRs will consist of trial implementation on these six plants.

Following approval of analysis and guidelines and the pilot monitoring of

emergency procedures, the staff will advise all licensees of the adequacy of

the guidelines for application to their plants. Consideration will be given

to human factors engineering and system operational characteristics, such as

information transfer under stress, compatibility with operator training and

control-room design, the time required for component and system response,

clarity of procedural actions, and control-room-personnel interactions. When

this determination has been made by the staff, a long-term plan for emergency

procedure review, as described in task action plan item I.C.9, will be made

available. At that time, the reviews currently being conducted on NTOLs under

item I.C.8 will be discontinued, and the review required for applicants for

operating licenses will be as described in the long-term plan. Depending on

the information submitted to support development of emergency procedures for

each reactor type or vendor, this transition may take place at different

times. For example, if the GE guidelines are shown to be effective on the six

plants chosen for pilot monitoring, the long-term plan for BWRs may be complete

in early 1981. Operating plants and applicants will then have the option of

implementing the long-term plan in a manner consistent with their operating

schedule, provided they meet the final date required for implementation. This

may require a plant that was reviewed for an operating license under item

I.C.8 to revise its emergency procedures again prior to the final implementation

date for Item I.C.9. The extent to which the long-term program will include

review and approval of plant-specific procedures for operating plants has not

been established. Our objective, however, is to minimize the amount of plantspecific procedure review and approval required. The staff believes this

objective can be acceptably accomplished by concentrating the staff review and

approval on generic guidelines. A key element in meeting this objective is

3-44 I.C.1-3
the use of staff-approved generic guidelines and guideline revisions by

licensees to develop procedures. For this approach to be effective, it is

imperative that, once the staff has issued approval of a guideline, subsequent

revisions of the guideline should not be implemented-by licensees until reviewed

and approved by the staff. Any changes in plant-specific procedures based on

unapproved guidelines could constitute an unreviewed safety issue under

10 CFR 50.59. Deviations from this approach on a plant-specific basis would

be acceptable provided the basis is submitted by the licensee for staff review

and approval. In this case, deviations from generic guidelines should not be

implemented until staff approval is formally received in writing. Interim

implementation of analysis and procedures for small-break loss-of-coolant

accident and inadequate core cooling should remain on the.schedule contained

in NUREG-0578, Recommendation 2.1.9.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

Reanalysis of transients and accidents and inadequate core cooling and preparation of guidelines.for development of emergency procedures should be completed

and submitted to the NRC for review. by January 1, 1981. The NRC staff will

review the analyses and guidelines and determine their acceptability by July 1,

1981, and will issue guidance to licensees on preparing emergency procedures

from the guidelines. Following NRC approval of the guidelines, licensees and

applicants for operating licenses issued prior to January 1, 1982, should

revise and implement their emergency procedures at the first refueling outage

after January 1, 1982. Applicants for operating licenses issued after January 1,

1982 should implement the procedures prior to operation. This schedule supersedes the implementation schedule included in NUREG-0578, Recommendation 2.1.9

for item I.C.1(a)3, Reanalysis of Transients and Accidents. For those licensees

and/or owners groups that will have difficulty in attaining the January 1,

1981 due date for submittal of guidelines, a comprehensive program plan,

proposed schedule, and a detailed justification for all delays and problems

shall be submitted in lieu of the guidelines.

Type of Review

A preimplementation review of guidelines will be performed.

A preimplementation review of procedures will be performed.

Documentation Required

See above, "Implementation."

Technical Specification Changes Required

Changes to technical specifications will not be required.

I.C.1-4 3-45
Reference

NUREG-0578, Recommendation 2.1.9

NUREG-0660, Item I.C.8 and Appendix C, Tables C.1, C.2, C.3

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,

dated September 27, 1979.

Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees, dated

October 10, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,

dated November 9, 1979.

3-46 I. C. 1-5
I.C.5 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF

Position

In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to Plant Staff (NUREG-0660), each applicant for an operating

license shall prepare procedures to assure that operating information pertinent

to plant safety originating both within and outside the utility organization

is continually supplied to operators and other personnel and is incorporated

into training and retraining programs. These procedures shall:

(1) Clearly identify organizational responsibilities for review of operating

experience, the feedback of pertinent information to operators and other

personnel, and the incorporation of such information into training and

retraining programs;

(2) Identify the administrative and technical review steps necessary in

translating recommendations by the operating experience assessment group

into plant actions (e.g., changes to procedures; operating orders);

(3) Identify the recipients of various categories of information from operating

experience (i.e., supervisory personnel, shift technical advisors, operators,

maintenance personnel, health physics technicians) or otherwise provide

means through which such information can be readily related to the job

functions of the recipients;

(4) Provide means to assure that affected personnel become aware of and

understand information of sufficient importance that should not wait for

emphasis through routine training and retraining programs;

(5) Assure that plant personnel do not routinely receive extraneous and

unimportant information on operating experience in such volume that it

would obscure priority information or otherwise detract from overall job

performance and proficiency;

(6) Provide suitable checks to assure that conflicting or contradictory

information is not conveyed to operators and other personnel until

resolution is reached; and,

(7) Provide periodic internal audit to assure that the feedback program

functions effectively at all levels.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Each utility shall carry out an operating experience assessment function that

will involve utility personnel having collective competence inmall areas

important to plant safety. In connection with this assessment function, it is

important that procedures exist to assure that important information on operating

experience originating both within and outside the organization is continually

I.C.5-1 3-47
provided to operators and other personnel and that it is incorporated into

plant operating procedures and training and retraining programs.

Those involved in the assessment of operating experience will review information from a variety of sources. These include operating information from the

licensee's own plant(s), publications such as IE Bulletins, Circulars, and

Notices, and pertinent NRC or industrial assessments of operating experience.

In some cases, information may be of sufficient importance that it must be

dealt with promptly (through instructions, changes to operating and emergency

procedures, issuance of special changes to operating and emergency procedures,

issuance of special precautions, etc.) and must be handled in such a manner to

assure that .operations management personnel would be directly involved in the

process. In many other cases, however, important information will become

available which should be brought to the attention of operators and other

personnel for their general information to assure continued safe plant operation. Since the total volume of information handled by the assessment group

may be large, it is important that assurance be provided that high-priority

matters are dealt with promptly and that discrimination is used in the feedback

of other information so that personnel are not deluged with unimportant and

extraneous information to the detriment of their overall proficiency. It is

important, also, that technical reviews be conducted to preclude premature

dissemination of conflicting or contradictory information.

Applicability

This requirement applies to all operating reactor and applicants for operating

license.

Implementation

Procedures governing feedback of operating experience to plant staff shall be

completed and the procedures put into effect on or before January 1, 1981 or

prior to issuance of an operating license, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

No documentation is required.

Technical Specificaton Changes Required

Changes to technical specifications will not be required.

References

NUREG-0660, Item I.C.5

Letter from D. G. Eisenhut, NRC, to All Licensees, dated May 7, 1980.

3-48 I.C.5-2
I.C.6 GUIDANCE ON PROCEDURES FOR VERIFYING CORRECT PERFORMANCE OF

OPERATING ACTIVITIES

Position

It is required (from NUREG-0660) that licensees' procedures be reviewed and

revised, as necessary, to assure that an effective system of verifying the

correct performance of operating activities is provided as a means of reducing

human errors and improving the quality of normal operations. This will reduce

the frequency of occurrence of situations that could result in or contribute

to accidents. Such a verification system may include automatic system status

monitoring, human verification of operations and maintenance activities independent of the people performing the activity (see NUREG-0585, Recommendation 5),

or both.

Implementation of automatic status monitoring if required will reduce the

extent of human verification of operations and maintenance activities but will

not eliminate the need for such verification in all instances. The procedures

adopted by the licensees may consist of two phases--one before and one after

installation of automatic status monitoring equipment, if required, in accordance with item I.D.3.

Changes to Previous Requirements and Guidance

Proposed requirement in NUREG-0660; this requirement is formally issued by

this letter.

Clarification

Item I.C.6 of the U.S. Nuclear Regulatory Commission Task Action Plan

(NUREG-0660) and Recommendation 5 of NUREG-0585 propose requiring that

licensees' procedures be reviewed and revised, as necessary, to assure that an

effective system of verifying the correct performance of operating activities

is provided. An acceptable program for verification of operating activities

is described below.

The American Nuclear Society has prepared a draft revision to ANSI Standard

N18.7-1972 (ANS 3.2) "Administrative Controls and Quality Assurance for the

Operational Phase of Nuclear Power Plants." A second proposed revision to

Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation),"

which is to be issued for public comment in the near future, will endorse the

latest draft revision to ANS 3.2 subject to the following supplemental

provisions:

(1) Applicability of the guidance of Section 5.2.6 should be extended to

cover surveillance testing in addition to maintenance.

(2) In lieu of any designated senior reactor operator (SRO), the authority to

release systems and equipment for maintenance or surveillance testing or

return-to-service may be delegated to an on-shift SRO, provided provisions

are made to ensure that the shift supervisor is kept fully informed of

system status.

I.C.6-1 3-49
(3) Except in cases of significant radiation exposure, a second qualified

person should verify correct implementation of equipment control measures

such as tagging of equipment.

(4) Equipment control procedures should include assurance that control-room

operators are informed of changes in equipment status and the effects of

such changes.

(5) For the return-to-service of equipment important to safety, a second

qualified operator should verify proper systems alignment unless functional

testing can be performed without compromising plant safety, and can prove

that all equipment, valves, and switches involved in the activity are

correctly aligned.

NOTE: A licensed operator possessing knowledge of the systems involved and

the relationship of the systems to plant safety would be a "qualified"

person. The staff is investigating the level of qualification necessary

for other operators to perform these functions.

For plants that have or will have automatic system status monitoring as discussed

in Task Action Plan item I.D.3, NUREG-0660, the extent of human verification

of operations and maintenance activities will be reduced. However, the need

for such verification will not be eliminated in all instances.

Applicability

This requirement applies to all operating reactor and operating license

applicants.

Implementation

Licensees/applicants must review and revise procedures as necessary to reflect

this position by January 1, 1981 or prior to fuel load, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

No documentation is required.

Technical Specification-Changes Required

Changes to technical specifications will not be required.

References

NUREG-0585, Recommendation 5

NUREG-0660, Item I.C.6, I.D.3

3-50 I.C.6-2
I.D.1 CONTROL-ROOM DESIGN REVIEWS

Position

In accordance with Task Action Plan I.D.1, Control Room Design Reviews

(NUREG-0660), all licensees and applicants for operating licenses will be

required to conduct a detailed control-room design review to identify and

correct design deficiencies. This detailed control-room design review is

expected to take about a year. Therefore, the Office of Nuclear Reactor

Regulation (NRR) requires that those applicants for operating licenses who are

unable to complete this review prior to issuance of a license make preliminary

assessments of their control rooms to identify significant human factors and

instrumentation problems and establish a schedule approved by NRC for correcting

deficiencies. These applicants will be required to complete the more detailed

control room reviews on the same schedule as licensees with operating plants.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

NRR is presently developing human engineering guidelines to assist each licensee

and applicant in performing detailed control-room review. A draft of the

guidelines has been published for public comment as NUREG/CR-1580, "Human

Engineering Guide to Control Room Evaluation." The due date for comments on

this draft document was September 29, 1980. NRR will issue the final version

of the guidelines as NUREG-0700, by February 1981, after receiving, reviewing,

and incorporating substantive public comments from operating reactor licensees,

applicants for operating licenses, human factors engineering experts, and

other interested parties. NRR will issue evaluation criteria, by July 1981,

which will be used to judge the acceptability of the detailed reviews performed and the design modifications implemented.

Applicants for operating licenses who will be unable to complete the detailed

control-room design review prior to issuance of a license are required to

perform a preliminary control-room design assessment to identify significant

human factors problems. Applicants will find it of value to refer to the

draft document NUREG/CR-1580, "Human Engineering Guide to Control Room Evaluation," in performing the preliminary assessment. NRR will evaluate the applicants'

preliminary assessments including the performance by NRR of onsite review/audit.

The NRR onsite review/audit will be on a schedule consistent with licensing

needs and will. emphasize the following aspects of the control room:

(1) The adequacy of information presented to the operator to reflect plant

status for normal operation, anticipated operational occurrences, and

accident conditions;

(2) The groupings of displays and the layout of panels;

(3) Improvements in the safety monitoring and human factors enhancement of

controls and control displays;

I. D. 1-1 3-51
(4) The communications from the control room to points outside the control

room, such as the onsite technical support center, remote shutdown panel,

offsite telephone lines, and to other areas within the plant for normal

and emergency operation.

(5) The use of direct rather than derived signals for the presentation of

process and safety information to the operator;

(6) The operability of the plant from the control room with multiple failures

of nonsafety-grade and nonseismic systems;

(7) The adequacy of operating procedures-and operator training with respect

to limitations of instrumentation displays in the control room;

(8) The categorization of alarms, with unique definition of safety alarms.

(9) The physical location of the shift supervisor's office either adjacent to

or within the control-room complex.

Prior to the'onsite review/audit, NRR will require a copy of the applicant's

preliminary assessment and additional information which will be used in formulating the details of the onsite review/audit.

Applicability

This requirement applies to all operating reactors and operating license

applicants.

Implementation

(1) Operating reactors and applicants for OLs:

Complete review, using NRC guidelines (NUREG-0700) issued in 1981, on a

schedule that will be determined upon issuance of the guidelines.

(2) Applicants for OLs whose schedules do not permit a full review prior to

licensing: Preliminary-review complete and approved by NRC prior to

issuance of the operating license.

Type of Review

Type of review for operating reactors will be determined upon issuance of

the guidance. A preimplementation review will be performed for operating

license applicants.

Documentation Required

Operating Reactors--To be determined upon issuance of the guidance.

Applicants for OLs with impacted schedules should report on results of

preliminary review prior to licensing.

3-52 3I.D.1-2
Technical Specification Changes Required

Changes to technical specifications will not be required unless there are

modifications to the control room.

References

NUREG-0660, Item I.D.1

NUREG/CR-1580 (Draft)

NUREG-0700

I.D.1-3 3-53
I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE

Position

In accordance with Task Action Plan 1.D.2, Plant Safety Parameter Display

Console (NUREG-0660), each applicant and licensee shall install a safety

parameter display system (SPDS) that will display to operating personnel a

minimum set of parameters which define the safety status of the plant. This

can be attained through continuous indication of direct and derived variables

as necessary to assess plant safety status.

Changes to Previous Requirements and Guidance

There are no changes to previous guidance.

Clarification

These requirements for the SPDS are being developed in NUREG-0696, which is

scheduled for issuance in November 1980.

Applicability

This requirement applies to all operating reactors and operating license

applications.

Implementation

Schedules for implementation-will be issued in conjunction with issuance of

NUREG-0696.

Type of Review

To be determined in conjunction with issuance of NUREG-0696.

Documentation Required

To be determined in conjunction with issuance of NUREG-0696.

Technical Specification Changes Required

To be determined in conjunction with issuance of NUREG-0696.

References

NUREG-0660, Item 1.D.2

NUREG-0696

3-54 I.D.2-1
II.B.1 REACTOR COOLANT SYSTEM VENTS

Position

Each applicant and licensee shall install reactor coolant system (RCS) and

reactor. vessel head high point vents remotely operated from the control room.

Although the purpose of the system is to vent noncondensible gases from the

RCS which may inhibit core cooling during natural circulation, the vents must

not lead to an unacceptable increase in the probability of a loss-of-coolant

accident (LOCA) or a challenge to containment integrity. Since these vents

form a part of the reactor coolant pressure boundary, the design of the events

shall conform to the requirements of Appendix A to 10 CFR Part 50, "General

Design Criteria." The vent system shall be designed with sufficient redundancy

that assures a low probability of inadvertent or irreversible actuation.

Each licensee shall provide the following information concerning the design

and operation of the high point vent system:*

(1) Submit a description of the design, location, size, and power supply for

the vent system along with results of analyses for loss-of-coolant accidents

initiated by a break in the vent pipe. The results of the analyses

should demonstrate compliance with the acceptance criteria of 10 CFR 50.46.

(2) Submit procedures and supporting analysis for operator use of the vents

that also include the information available to the operator for initiating

or terminating vent usage.

Changes to Previous Requirements and Guidance

(1) The probability of a valve failing to close, once opened, should be

minimized.

(2) Establishes environmental qualification (Commission Order, May 23, 1980).

(3) Establishes provisions for testing.

(4) Delete requirements of September 27, 1979 letter from Vassallo to applicants stating that vents shall satisfy single-failure criteria of IEEE-279.

Vent systems are not required to have redundant paths. A degree of

redundancy should be provided by powering different vents from different

emergency buses.

(5) Documentation date changed to July 1, 1981 and implementation date to

July 1, 1982.

Clarification does not change NRC concept of requirement, but provides more

detail on scope. The dates have been revised to provide time for procurement

and installation.

sIt was the intent of the October 30, 1979 letter to delete the requirement

to meet the criteria of 10 CFR 50.44 and SRP 6.2.5 for beyond-design-basis

events. The analysis requirements of Position 2 in the September 13, 1979

letter are therefore unnecessary.

II. B. 1-1 . 3-55
Clarification

A. General

(1) The important safety function enhanced by this venting capability is core

cooling. For events beyond the present design basis, this venting capability

will substantially increase the plant's ability to deal with large quantities

of noncondensible gas which could interfere with core cooling.

(2) Procedures addressing the use of the reactor coolant system vents should

define the conditions under which the vents should be used as well as the

conditions under which the vents should not be used. The procedures

should be directed toward achieving a substantial increase in the plant

being able to maintain core cooling without loss of containment integrity

for events beyond the design basis. The use of vents for accidents

within the normal design basis must not result in a violation of the

requirements of 10 CFR 50.44 or 10 CFR 50.46.

(3) The size of the reactor coolant vents is not a critical issue. The

desired venting capability can be achieved with vents in a fairly broad

spectrum of sizes. The criteria for sizing a vent can be developed in

several ways.. One approach, which may be considered, is to specify a

volume of noncondensible gas to be vented and in a specific venting time.

For containments particularly vulnerable to failure from large hydrogen

releases over a short period of time, the necessity and desirability for

contained venting outside the containment must be considered (e.g., into

a decay gas collection and storage system).

(4) Where practical, the reactor coolant system vents should be kept smaller

than the size corresponding to the definition of LOCA (10 CFR 50, Appendix

A). This will minimize the challenges to the emergency core cooling

system (ECCS) since the inadvertent opening of a vent smaller than the

LOCA definition would not require ECCS actuation, although it may result

in leakage beyond technical specification limits. On PWRs, the use of

new or existing lines whose smallest orifice is larger than the LOCA

definition will require a valve in series with a vent valve that can be

closed from the control room to terminate the LOCA that would result if

an open vent valve could not be reclosed.

(5) A positive indication of valve position should be provided in the. control

room.

(6) The reactor coolant vent system shall be operable from the control room.

(7) Since the reactor coolant system vent will be part of the reactor coolant

system pressure boundary, all requirements for the reactor pressure

boundary must be met, and, in addition, sufficient redundancy should be

incorporated into the design to minimize the probability of an inadvertent

actuation of the system. Administrative procedures, may be a viable

option to meet the single-failure criterion. For vents larger than the

3-56 II.B.1-2
LOCA definition, an analysis is required to demonstrate compliance with

10 CFR 50.46.

(8) The probability of a vent path failing to close, once opened, should be

minimized; this is a new requirement. Each vent must have its power

supplied from an emergency bus. A single failure within the power and

control aspects of the reactor coolant vent system should not prevent

isolation of the entire vent system when required. On BWRs, block valves

are not required in lines with safety valves that are used for venting.

(9) Vent paths from the primary system to within containment should go to

those areas that provide good mixing with containment air.

(10) The reactor coolant vent system (i.e., vent valves, block valves, position

indication devices, cable terminations, and piping) shall be seismically

and environmentally qualified in accordance with IEEE 344-1975 as supplemented by Regulatory Guide 1.100, 1.92 and SEP 3.92, 3.43, and 3.10.

Environmental qualifications are in accordance with the May 23, 1980

Commission Order and Memorandum (CLI-80-21).

(11) Provisions to test for operability of the reactor coolant vent system

should be a part of the design. Testing should be performed in accordance

with subsection IWV of Section XI of the ASME Code for Category B valves.

(12) It is important that the displays and controls added to the control room

as a result of this requirement not increase the potential for operator

error. A human-factor analysis should be performed taking into consideration:

(a) the use of this information by an operator during both normal and

abnormal plant conditions,

(b) integration into emergency procedures,

(c) integration into operator training, and

(d) other alarms during emergency and need for prioritization of alarms.

B. BWR Design Considerations

(1) Since the BWR owners' group has suggested that the present BWR designs

have an inherent capability to vent, a question relating to the capability

of existing systems arises. The ability of these systems to vent the RCS

of noncondensible gas generated during an accident must be demonstrated.

Because of differences among the head vent systems for BWRs, each licensee

or applicant should address the specific design features of this plant

and compare them with the generic venting capability proposed by the BWR

owners' group. In addition, the ability of these systems to meet the

same requirements as the PWR vent system must be documented.

(2) In addition to RCS venting, each BWR licensee should address the ability

to vent other systems, such as the isolation condenser which may be

II.B.1-3 3-57
required to maintain adequate core cooling. If the production of a large

amount of noncondensible gas would cause the loss of function of such a

system, remote venting of that system is required. The qualifications of

such a venting system should be the same as that required for PWR venting

systems.

C. PWR Vent Design Considerations

(1) Each PWR licensee should provide the capability to vent the reactor

vessel head. The reactor vessel head vent should be capable of venting

noncondensible gas from the reactor vessel hot legs (to the elevation of

the top of the outlet nozzle) and cold legs (through head jets and other

leakage paths).

(2) Additional venting capability is required for those portions of each hot

leg that cannot be vented through the reactor vessel head vent or pressurizer. It is impractical to vent each of the many thousands of tubes

in a U-tube steam generator; however, the staff believes that a procedure

can be developed that assures sufficient liquid or steam can enter the

U-tube region so that decay heat can be effectively removed from the RCS.

Such operating procedures should incorporate this consideration.

(3) Venting of the pressurizer is required to assure its availability for

system pressure and volume control. These are important considerations,

especially during natural circulation.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

Installation should take place by July 1, 1982. Until staff approval is

obtained, installation may proceed; but operating procedures should not be

implemented and valves should be placed in a condition so as to minimize the

potential for inadvertent actuation (e.g., remove power).

Type of Review

A preimplementation review will be performed prior to authorizing use of the

vent.

Documentation Required

By July 1, 1981, the licensee shall provide the following information on the

reactor coolant vent system for staff review:

(1) The information requested in items 1 and 2 under "Position";

3-58 II. B. 1-4
(2) A discussion of the design with respect to conformance to the design

criteria discussed under "Clarification," including deviations, if any,

with adequate justification for such deviations; and,

(3) Supporting information including logic diagrams, electrical schematics,

piping and instrumentation diagrams, test procedures, and technical

specifications.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0660

Commission Orders, May 23, 1980 (CLI-80-21)

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,

dated September 27, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

II. B. 1-5 .3-59
II.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF

EQUIPMENT FOR SPACES/SYSTEMS WHICH MAY BE USED IN POSTACCIDENT

OPERATIONS

Position

With the assumption of a postaccident release of radioactivity equivalent to

that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50%

of the core radioiodine, 100% of the core noble gas inventory, and 1% of the

core solids are contained in the primary coolant), each licensee shall perform

a radiation and shielding-design review of the spaces around systems that may,

as a result of an accident, contain highly radioactive materials. The design

review should identify the location of vital areas and equipment, such as the

control room, radwaste control stations, emergency power supplies, motor

control centers, and instrument areas, in which personnel occupancy may be

unduly limited or safety equipment may be unduly degraded by the radiation

fields during postaccident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding,

or postaccident procedural controls. The design review shall determine which

types of corrective actions are needed for vital areas throughout the facility.

Changes to Previous Requirements and Guidance

This-requirement was originally issued by letters to all operating nuclear

power plants, dated September 13 and October 30, 1979, and was incorporated

into NUREG-0660. Significant changes in requirements or guidance are:

(1) Adds several areas to be evaluated for access to ensure that these areas

are not overlooked.

(2) Specifies that the source term for recirculated depressurized coolant

need not be assumed to contain noble gas since this gas will be released

from the liquid when it is depressurized.

(3) Specifies that certain systems be considered as potential sources and

that leakage from systems outside of containment need not be considered

as potential sources.

(4) Allows averaging over 30 days of the dose rate criteria for areas requiring

continuous occupancy and that the control room and technical support

center should be considered areas requiring continuous occupancy. This

ensures that the dose rate criteria is applied correctly to these areas.

(5) Specifies source terms to be used in conjunction with Commission Order

and Memorandum dated May 23, 1980 (CLI-80-21) on equipment qualification,

and specifies schedule in above order.

(6) Because of difficulty in obtaining equipment (e.g., remote-operated

valves), the implementation date is moved to January 1, 1982, or the

first outage of sufficient duration thereafter, but no later than July 1,

1982.

3-60 II. B. 2-1
Clarification

The purpose of this item is to ensure that licensees examine their plants to

determine what actions can be taken over the short-term to reduce radiation

levels and increase the capability of operators to control and mitigate the

consequences of an accident. These actions should be taken pending conclusions

resulting in the long term degraded core rulemaking, which may result in a

need to consider additional sources.

Any area which will or may require occupancy to permit an operator to aid in

the mitigation of or recovery from an accident is designated as a vital area.

For the purposes of this evaluation, vital areas and equipment are not necessarily

the same vital areas or equipment defined in 10 CFR 73.2 for security purposes.

The security center is listed as an area to be considered as potentially

vital, since access to this area may be necessary to take action to give

access to other areas in the plant.

The control room, technical support center (TSC), sampling station and sample

analysis area must be included among those areas where access is considered

vital after an accident. (See Item III.A.1.2 for discussion of the TSC and

emergency operations facility.) The evaluation to determine the necessary

vital areas should also include, but not be limited to, consideration of the

post-LOCA hydrogen control system, containment isolation reset control area,

manual ECCS alignment area (if any), motor control centers, instrument panels,

emergency power supplies, security center, and radwaste control panels. Dose'

rate determinations need not be for.these areas if they are determined not to

be vital.

As a minimum, necessary modifications must be sufficient to provide for vital

system operation and for occupancy of the control room, TSC, sampling station,

and sample analysis area.

In order to assure that personnel can perform necessary postaccident operations

in the vital areas, the following guidance is to be used by licensees to

evaluate the adequacy of radiation protection to the operators:

(1) Source Term

The minimum radioactive source term should be equivalent to the source terms

recommended in Regulatory Guides 1.3, 1.4, 1.7 and Standard Review' Plan 15.6.5

with appropriate'decay times based on plant design (i.e., you may assume the

radioactive decay that occurs before fission products can be transported to

various systems).

(a) Liquid-Containing Systems: 100% of the core equilibrium noble gas

inventory, 50% of the core equilibirum halogen inventory, and 1% of

all others are assumed to be mixed in the reactor coolant and liquids

recirculated by residual heat removal (RHR), high- pressure coolant

injection (HPCI), and low-pressure coolant injection (LPCI), or the

equivalent of these systems. In determining the source term for

recirculated, depressurized cooling water, you may assume that the

water contains no noble gases.

II.B.2-2

3-61
(b) Gas-Containing Systems: 100% of the core equilibrium noble gas

inventory and 25% of the core equilibrium halogen activity are

assumed to be mixed in the containment atmosphere. For vaporcontaining lines connected to the primary system (e.g., BWR steam

lines), the concentration of radioactivity shall be determined

assuming the activity is contained in the vapor space in the primary

coolant system.

(2) Systems Containing the Source

Systems assumed in your analysis to contain high levels of radioactivity in a

postaccident situation should include, but not be limited to, containment,

residual heat removal system, safety injection systems, chemical and volume

control system (CVCS), containment spray recirculation system, sample lines,

gaseous radwaste systems, and standby gas treatment systems (or equivalent of

these systems). If any of these systems or others that could contain high

levels of radioactivity were excluded, you should explain why such systems

were excluded. Radiation from leakage of systems located outside of containment need not be considered for this analysis. Leakage measurement and reduction

is treated under Item III.D.1.1, "Integrity of Systems Outside Containment

Likely To Contain Radioactive Material for PWRs and BWRs." Liquid waste

systems need not be included in this analysis. Modifications to liquid waste

systems will be considered after completion of Item III.D.1.4, "Radwaste

System Design Features To Aid in Accident Recovery and Decontamination."

(3) Dose Rate Criteria

The design dose rate for personnel in a vital area should be such that the

guidelines of GDC 19 will not be exceeded during the course of the accident.

GDC 19 requires that adequate radiation protection be provided such that the

dose to personnel should not be in excess of 5 rem whole body, or its equivalent

to any part of the body for the duration of the accident. When determining

the dose to an operator, care must be taken to determine the necessary occupancy

times in a specific area. For example, areas requiring continuous occupancy

will require much lower dose rates than areas where minimal occupancy is

required. Therefore, allowable dose rates will be based upon expected occupancy,

as well as the radioactive source terms and shielding. However, in order to

provide a general design objective, we are providing the following dose rate

criteria with alternatives to be documented on a case-by-case bases. The

recommended dose .rates are average rates in the area. Local hot spots may

exceed the dose rate guidelines. These doses are design objectives and are

not to be used to limit access in the event of an accident.

(a) Areas Requiring Continuous Occupancy: <15 mrem/hr (averaged over 30

days). These areas will require full-time occupancy during the

course of the accident. The control room and onsite technical

support center are areas where continuous occupancy will be required.

The dose rate for these areas is based on the control room occupancy

factors contained in SRP 6.4.

II.B.2-3

3-62
(b) Areas Requiring Infrequent Access: GDC 19. These areas may require

access on an irregular basis, not continuous occupancy. Shielding

should be provided to allow access at a frequency and duration

estimated by the licensee. The plant radiochemical/chemical analysis

laboratory, radwaste panel, motor control center, instrumentation

locations, and reactor coolant and containment gas sample stations

are examples of sites where occupancy may be needed often, but not

continuously.

(4) Radiation Qualification of Safety-Related Equipment

The review of safety-related equipment which may be unduly degraded by radiation

during postaccident operation of this equipment relates to equipment inside

and outside of the primary containment. Radiation source terms calculated to

determine environmental qualification of safety-related equipment consider the

following:

(a) LOCA events which completely depressurize the primary system should

consider releases of the source term (100% noble gases, 50% iodines,

and 1% particulates) to the containment atmosphere.

(b) LOCA events in which the primary system may not depressurize should

consider the source term (100% noble gases, 50% iodines, and 1%

particulate) to remain in the primary coolant. This method is used

to determine the qualification doses for equipment in close proximity

to recirculating fluid systems inside and outside of containment.

Non-LOCA events both inside and outside of containment should use

10% noble gases, 10% iodines, and 0% particulate as a source term.

The following table summarizes these considerations:

Containment LOCA Source Term Non-LOCA

(Noble Gas/Iodine/ High-Energy Line Break Source Term

Particulate) (Noble Gas/Iodine/Particulate)

Outside (100/50/1) (10/10/0)

in RCS in RCS

Inside Larger of (10/10/0)

(T1007501) in RCS

in containment

or

(100/50/1)

in RCS

II.B.2-4

3-63
Applicability

This requirement applies to all operating reactors and applicants for an

operating license.

Implementation

(1) For Vital Area Access

By January 1, 1982 modifications should be completed: For operating plants,

documentation should be completed by January 1, 1982. For OL applicants,

documentation of the evaluation should be completed at least four months

before the operating license is issued.

(2) For Equipment Qualification

All safety-related electrical equipment must be fully qualified by June 30,

1982. Documentation in. accordance with:

(a) Operating Reactors and NTOL (operating license expected by February

1981): submittal to be received no later than November 1, 1980.

(b) Operating Licenses (operating license expected by June 30, 1982):

submittal no later than 4 months before issuance of operating license.

Operating licenses in accordance with review schedule.

Type of Review

A postimplementation review will be performed.

Documentation Required

For Vital Area Access--For operating license applicants provide a summary of

the shielding design review, a description of the results of this review, and

a description of the modifications made or to be made to implement the result

of the review. Include in your submittal:

(1) Specification of source terms used in the evaluation; including time

after shutdown that was assumed for source terms in systems;

(2) Specification of systems assumed in your analysis to contain high levels

of radioactivity in a postaccident situation. If any of the systems

listed in "Clarification," item 2, were excluded, explain why such systems

are excluded from review;

(3) Specification of areas where access is considered necessary for vital

system operation after an accident. If any of the areas listed in the

"Clarification" section above were not considered to be areas requiring

access after an accident, explain why they were excluded;

II.B.2-5

3-64
(4) The projected doses to individuals for necessary occupancy times in vital

areas and a dose rate map for potentially occupied areas.

Documentation Required

For Operating Reactors--By January 1, 1981, have available for review the

final design details of the implementation of the above position and clarifications. If deviations to the above position or clarification are necessary,

provide detailed explaination and justification for the deviations by January 1,

1981.

For Equipment Qualification--Provide the information required by the Commission

Memorandum and Order on equipment qualification (CLI-80-21).

Technical Specification Changes Required

Technical specifications will not be required.

References

NUREG-0578, Recommendation 2.1.6.b

NUREG-0660, Item II.B.2

Commission Order and Memorandum, May 23, 1980 (CLI-80-21)

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees, dated

April 25, 1980.

Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees, dated

May 7, 1980.

II.B.2-6

3-65
II.B.3 POSTACCIDENT SAMPLING CAPABILITY

Position

A design and operational review of the reactor coolant and containment atmosphere

sampling line systems shall be performed to determine the capability of personnel

to promptly obtain (less than 1 hour) a sample under accident conditions

without incurring a radiation exposure to any individual in excess of 3 and

18-3/4 rem to the whole body or extremities, respectively. Accident conditions

should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If

the review indicates that personnel could not promptly and safely obtain the

samples, additional design features or shielding should be provided to meet

the criteria.

A design and operational review of the radiological spectrum analysis facilities

shall be performed to determine the capability to promptly quantify (in less

than 2 hours) certain radionuclides that are indicators of the degree of core

damage. Such radionuclides are noble gases (which indicate cladding failure),

iodines and cesiums (which indicate high fuel temperatures), and nonvolatile

isotopes (which indicate fuel melting). The initial reactor coolant spectrum

should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should

also consider the effects of direct radiation from piping and components in

the auxiliary building and possible contamination and direct radiation from

airborne effluents. If the review indicates that the analyses required cannot

be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary

for monitoring reactor conditions. Procedures shall be provided to perform

boron and chloride chemical analyses assuming a highly radioactive initial

sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be

capable of being completed promptly (i.e., the boron sample analysis within an

hour and the chloride sample analysis within a shift).

Changes to Previous Requirements and Guidance

This requirement was originally issued to all operating plants by letters

dated September 13 and October 30, 1979. Significant changes in requirements

or guidance are:

(1) Allows combined time of 3 hours or less for sampling and analysis.

(2) Specifies that licensee may use online sampling and analysis to meet the

3-hour time requirement but must provide capability to remove grab samples

of reactor coolant and containment atmosphere for separate analysis.

(3) Implementation date has been changed to January 1, 1982.

(4) Provides design guidance for sampling and analytical capability.

Clarification

The following items are clarifications of requirements identified in NUREG-0578,

NUREG-0660, or the September 13 and October 30, 1979 clarification letters.

3-66 II.B.3.1-1
(1) The licensee shall have the capability to promptly obtain reactor coolant

samples and containment atmosphere samples. The combined time allotted

for sampling and analysis should be 3 hours or less from the time a

decision is made to take a sample.

(2) The licensee shall establish an onsite radiological and chemical analysis

capability to provide, within the 3-hour time frame established above,

quantification of the following:

(a) certain radionuclides in the reactor coolant and containment atmosphere

that may be indicators of the degree of core damage (e.g., noble

gases; iodines and cesiums, and nonvolatile isotopes);

(b) hydrogen levels in the containment atmosphere;

(c) dissolved gases (e.g., H2), chloride (time allotted for analysis

subject to discussion below), and boron concentration of liquids.

(d) Alternatively, have inline monitoring capabilities to perform all or

part of the above analyses.

(3) Reactor coolant and containment atmosphere sampling during postaccident

conditions shall not require an isolated auxiliary system [e.g., the

letdown system, reactor water cleanup system (RWCUS)) to be placed in

operation in order to use the sampling system.

(4) Pressurized reactor coolant samples are not required-if the licensee can

quantify the amount of dissolved gases with unpressurized reactor coolant

samples. The measurement of either total dissolved gases or H2 gas in

reactor coolant samples is considered adequate. Measuring the 02 concentration is recommended, but is not mandatory.

(5) The time for a chloride analysis to be performed is dependent upon two

factors: (a) if the plant's coolant water is seawater or brackish water

and (b) if there is only a single barrier between primary containment

systems and the cooling water. Under both of the above conditions the

licensee shall provide for a chloride analysis within 24 hours of the

sample being taken. For all other cases, the licensee shall provide for

the analysis to be completed within 4 days. The chloride analysis does

not have to be done onsite.

(6) The design basis for plant equipment for reactor coolant and containment

atmosphere sampling and analysis must assume that it is possible to

obtain and analyze a sample without radiation exposures to any individual

exceeding the criteria of GDC 19 (Appendix A, 10 CFR Part 50) (i.e., 5

rem whole body, 75 rem extremities). (Note that the design and operational

review criterion was changed from the operational limits of 10 CFR Part 20

(NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H. R.

Denton to all licensees).)

(7) The analysis of primary coolant samples for boron is required for PWRs.

(Note that Revision 2 of Regulatory Guide 1.97, when issued, will likely

specify the need for primary coolant boron analysis capability at BWR

plants.)

II.B.3.1-2 3-67
(8) If inline monitoring is used for any sampling and analytical capability

*specified herein, the licensee shall provide backup sampling through grab

samples, and shall demonstrate the capability of analyzing the samples.

Established planning for analysis at offsite facilities is acceptable.

Equipment provided for backup sampling shall be capable of providing at

least one sample per day for 7 days following onset of the accident and

,at least one sample per week until the accident condition no longer

exists.

(9) The licensee's radiological and chemical sample analysis capability shall

include provisions to:

(a) Identify and quantify the isotopes of the nuclide categories discussed

above to levels corresponding to the source terms given in Regulatory

Guide 1.3 or 1.4 and 1.7. Where necessary and practicable, the

ability to dilute samples to provide capability for measurement and

reduction of personnel exposure should be provided. Sensitivity of

onsite liquid sample analysis capability should be such as to permit

measurement of nuclide concentration in the range from approximately

1 pCi/g to 10 Ci/g.

(b) Restrict background levels of radiation in the radiological and

chemical analysis facility from sources such that the sample analysis

will provide results with.an acceptably small error (approximately a

factor of 2). This can be accomplished through the use of sufficient

shielding around samples and outside sources, and by the use of

ventilation system design which will control the presence of airborne

radioactivity.

(10) Accuracy, range, and sensitivity shall be adequate to provide pertinent

data to the operator in order to describe radiological and chemical

status of the reactor coolant systems.

(11) In the design of the postaccident sampling and analysis capability,

consideration should be given to the following items:

(a) Provisions for purging sample lines, for reducing plateout in sample

lines, for minimizing sample loss or distortion, for preventing

blockage of sample lines by loose material in the RCS or containment,

for appropriate disposal of the samples, and for flow restrictions

to limit reactor. coolant loss from a rupture of the sample line.

The postaccident reactor coolant and containment atmosphere samples

should be representative of the.reactor coolant in the core area and

the containment atmosphere following a transient or accident. The

sample lines should be as short as possible to minimize the volume

of fluid to be taken from containment. The residues of sample

collection should be returned to containment or to a closed system.

(b) The ventilation exhaust from the sampling station should be filtered

with charcoal adsorbers and high-efficiency particulate air (HEPA)

filters.

3-68 II. B. 3.1-3
Applicability

This requirement applies to all operating reactors and applicants for operating

licenses.

Implementation

Installation should take place by January 1, 1982.

Type of Review

A postimplementation review will be performed.

Documentation Required

Operating Reactors--By January 1, 1982 have available for review the final

design details of the-implementation of the above position and clarifications.

The final design includes piping and instrumentation diagrams (P&IDs), together

with either (a) a summary description of procedures for sample collection,

sample transfer or transport, and sample analysis, or (b) copies of procedures

for sample collection, sample transfer or transport, and sample analysis. If

deviations to the above position or clarification are necessary, provide

detailed explanation and justification for the deviations by January 1, 1982.

Operating License Applicants--Provide a description of the implementation of

the position and clarification including P&IDs, together with either (a) a

summary description of procedures for sample collection, sample transfer or

transport, and sample analysis, or (b) copies of procedures for sample collection,

sample transfer or transport, and sample analysis, in a accordance with the

proposed review schedule but in no case less than 4 months prior to the issuance

of an operating license.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.8.a

NUREG-0660, Item II.B.3

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

I.B. 3.1-4 3-69
II.B.4 TRAINING FOR MITIGATING CORE DAMAGE

Position

Licensees are required to develop a training program to teach the use of

installed equipment and systems to control or mitigate accidents in which the

core is severely damaged. They must then implement the training program.

Changes to Previous Requirements and Guidance

Persons who must participate in the training program are to be defined.

The implementation schedule has been revised to reflect the TMI Action Plan

schedule.

Clarification

Shift technical advisors and operating personnel from the plant manager through

the operations chain to the licensed operators shall receive all the training

indicated in Enclosure 3 to H. R. Denton's March 28, 1980 letter.

Managers and technicians in the Instrumentation and Control (I&C), health

physics, and chemistry departments shall receive training commensurate with

their responsibilities.

Applicability

This requirement applies to all operating reactors and operating license

applicants.

Implementation

Licensees with operating reactors will develop a training program by January 1,

1981 and initiate the training program by April 1, 1981. The initial program

should be complete by October 1, 1981. Applicants for operating licenses

should develop a training program prior to fuel loading and complete the

program prior to full-power operation.

Type of Review

A postimplementation review will be performed.

Documentation Required

Programs shall be available for review by January 1, 1981.

Technical Specification Changes Required

Changes to technical specifications will not be required.

3-70 II. B. 4-1
References

NUREG-0660, Item II.B

Letter from H. R. Denton, NRC, to All Power Reactor Applicants and Licensees,

dated March 28, 1980.

II.B.4-2 3-71
II.D.1 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER

REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2)

Position

Pressurized-water reactor and boiling-water reactor licensees and applicants

shall conduct testing to qualify the reactor coolant system relief and safety

valves under expected operating conditions for design-basis transients and

accidents.

Changes to Previous Requirements and Guidance

A. Safety and Relief Valves and Piping--The types of documentation required

for safety and relief valves and piping and the specific submittal dates

are considered to be a clarification of item II.D.1 as described in

NUREG-0660. The submittal of information was implied but not explicitly

discussed in that report.

B. Block Valves--Qualification of PWR block valves is a new requirement.

Since block valves must be qualified to ensure that a stuck-open relief

valve can be isolated, thereby terminating a small loss-of-coolant accident

due to a stuck-open relief valve. Isolation of a stuck-open power-operated

relief valve (PORV) is not required to ensure safe plant shutdown.

However isolation capability under all fluid conditions that could be

experienced under operating and accident conditions will result in a

reduction in the number of challenges to the emergency core-cooling

system. Repeated unnecessary challenges to these system are undesirable.

C. ATWS Testing--Testing of anticipated transients without scram (ATWS) for

later phases of the valve qualification program was noted in item II.D.1

of NUREG-0660. The clarification below provides updated information on

PWR ATWS temperature and pressure conditions and clarifies that ATWS

testing need not be accomplished by July 1981.

Clarification

Licensees and applicants shall determine the expected valve operating conditions

through the use of analyses of accidents and anticipated operational occurrences

referenced in Regulatory Guide 1.70, Revision 2. The single failures applied

to these analyses shall be chosen so that the dynamic forces on the safety and

relief valves are maximized. Test-pressures shall be the highest predicted by

conventional safety analysis procedures. Reactor coolant system relief and

safety valve qualification shall include qualification of associated control

circuitry, piping, and supports, as well as the valves themselves.

A. Performance Testing of Relief and Safety Valves--The following information

must be provided in report form by October 1, 1981:

(1) Evidence supported by test of safety and relief valve functionability for

expected operating and accident (non-ATWS) conditions must be provided to

NRC. The testing should demonstrate that the valves will open and reclose

under the expected flow conditions.

3-72 II.3D. 1-1
(2) Since it is not planned to test all valves on all plants, each licensee

must submit to NRC a correlation or other evidence to substantiate that

the valves tested in the EPRI (Electric Power Research Institute) or

other generic test program demonstrate the functionability of as-installed

primary relief and safety valves. This correlation must show that the

test conditions used are equivalent to expected operating and accident

conditions as prescribed in the final safety analysis report (FSAR). The

effect of as-built relief and safety valve discharge piping on valve

operability must also be accounted for, if it is different from the

generic test loop piping.

(3) Test data including criteria for success and failure of valves tested

must be provided for NRC staff review and evaluation. These test data

should include data that would permit plant-specific evaluation of

discharge piping and supports that are not directly tested.

B. Qualification of PWR Block Valves--Although not specifically listed as a

short-term lessons-learned requirement in NUREG-0578, qualification of

PWR block valves is required by the NRC Task Action Plan NUREG-0660 under

task item II.D.1. It is the understanding of the NRC that testing of several

commonly used block valve designs is already included in the generic EPRI

PWR safety and relief valve testing program to be completed by July 1,

1981. By means of this letter, NRC is establishing July 1, 1982 as the

date for verification of block valve functionability. By July 1, 1982,

each PWR licensee, for plants so equipped, should provide evidence supported

by test that the block or isolation valves between the pressurizer and

each power-operated relief valve can be operated, closed, and opened for

all fluid conditions expected under operating and accident conditions.

C. ATWS Testing--Although ATWS testing need not be completed by July 1,

1981, the test facility should be designed to accommodate ATWS conditions

of approximately 3200 to 3500 (Service Level C pressure limit) psi and

700'F with sufficient capacity to enable testing of relief and safety

valves of the size and type used on operating pressurized-water reactors.

Applicability

This requirement applies to all operating reactors and operating license

applicants.

Implementation

See implementation schedules in the "Documentation Required" section.

Type of Review

Preimplementation review will be performed for EPRI and BWR test programs with

respect to qualification of relief and safety valves. Also, the applicants'

proposal for functional testing or qualification of PWR valves will be reviewed.

Postimplementation review will also be performed of the test data and test

results as applied to plant-specific situations.

II. D. 1-2 3-73
Documentation Required

Preimplementation review will be based on EPRI, BWR, and applicant submittals

with regard to the various test programs. These submittals should be made on

a timely basis as noted below, to allow for adequate review and to ensure that

the following valve qualification dates can be met:

Final PWR (EPRI) Test Program--July 1, 1980

Final BWR Test Program--October 1, 1980

Block Valve Qualification Program--January 1, 1981

Postimplementation review will be based on the applicants' plant-specific

submittals for qualification of safety relief valves and block valves. To

properly evaluate these plant-specific applications, the test data and results

of the various programs will also be required by the following dates:

PWR (EPRI)/BWR Generic Test Program Results--July 1, 1981

Plant-specific submittals confirming adequacy of safety and relief valves

based on licensee/applicant preliminary review of generic test program

results--July 1, 1981

Plant-specific reports for safety and-relief valve qualification--

October 1, 1981

Plant-specific submittals for piping and support evaluations--January 1,

1982

Plant-specific submittals for block valve qualification--July 1, 1982

Technical Specification Changes Required

No technical specification changes are required.

References

NUREG-0578

NUREG-0660, Item II.D.1

3-74 II.D.1-3
II.D.3 DIRECT INDICATION OF RELIEF-AND SAFETY-VALVE POSITION

Position

Reactor coolant system relief 'and safety valves shall be provided with a

positive indication in the control room derived from a reliable valve-position

detection device or a reliable indication of flow in the discharge pipe.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

(1) The basic requirement is to provide the operator with unambiguous indication of valve position (open or closed) so that appropriate operator

actions can be taken.

(2) The valve position should be indicated in the control room. An alarm

should be provided in conjunction with this indication.

(3) The valve position indication may be safety grade. If the position

indication is not safety grade, a reliable single-channel direct indication

powered from a vital instrument bus may be provided if backup methods of

determining valve position are available and are discussed in the emergency

procedures as an aid to operator diagnosis of an action.

(4) The valve position indication should be seismically qualified consistent

with the component or system to which it is attached.

(5) The position indication should be qualified for its appropriate environment

(any transient or accident which would cause the relief or safety valve

to lift) and in accordance with Commission Order, May 23rd, 1980 (CLI-20-81).

(6) It is important that the displays and controls added to the control room

as a result of this requirement not increase the potential for operator

error. A human-factor analysis should be performed taking into consideration:

(a) the use of this information by an operator during both normal and

abnormal plant conditions,

(b) integration into emergency procedures,

(c) integration into operator training, and

(d) other alarms during emergency and need for prioritization of alarms.

Applicability

This requirement applies to all reactor licenses and applicants for operating

license. (Operating reactor licensees completed this requirement by January

1980.)

II.D.3-1 3-75
Implementation

Implementation will be completed prior to the issuance of a fuel-loading

license.

Type of Review

A preimplementation review will be performed.

Documentation Required

Documentation should be provided that discusses each item of the clarification,

as well as electrical schematics and proposed test procedures in accordance

with the proposed review schedule, but in no case less than 4 months prior to

the scheduled issuance of the staff safety evaluation report.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.3.a

NUREG-0660, Item II.D.3

NUREG-0694, Part 1

Commission Order and Memorandum, May 23, 1980 (CLI-20-81)

Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,

dated September 27, 1979.

Letter from D. B. Vassallo, NRC, to All Pending Operating License Applicants,

dated November 9, 1979.

3-76 II. D. 3-2
II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION

Position

The Office of Nuclear Reactor Regulation is requiring reevaluation of the

auxiliary feedwater (AFW) systems for all PWR operating plant licensees and

operating license applications. This action includes:

(1) Perform a simplified AFW system reliability analysis that uses event-tree

and fault-tree logic techniques to determine the potential for AFW system

failure under various loss-of-main-feedwater-transient conditions.

Particular emphasis is given to determining potential failures that could

result from human errors, common causes, single-point vulnerabilities,

and test and maintenance outages;

(2) Perform a deterministic review of the AFW system using the acceptance

criteria of Standard Review Plan Section 10.4.9 and associated Branch

Technical Position ASB 10-1 as principal guidance; and

(3) Reevaluate the AFW system flowrate design bases and criteria.

Changes to Previous Requirements and Guidance

Short-term requirements will be implemented by July 1, 1981. The date for

implementation of short-term requirements has been slipped because staff review

of submittals is not complete.

Clarification

Operating Plant Licenses--Items 1 and 2 above have been completed for Westinghouse (W), Combustion Engineering (C-E), and two Babcock and Wilcox (B&W)

operating plants (Rancho Seco, short-term only, and TMI-1). As a result of

staff review of items 1 and 2, letters were issued to these plants that

required the implementation of certain short- and long-term AFW system upgrade

requirements. Included in these letters was a request for additional information regarding item 3 above. The staff is'now in the process of evaluating

licensees' responses and commitments to these letters.

The remaining B&W operating plants (Oconee 1-3, Crystal River 3, ANO-1, and

Davis-Besse 1) have submitted the analysis described in item 1 above. The

analysis'is presently undergoing staff review. When the results of the staff

reviews are complete, each of the remaining B&W plants will receive a letter

specifying the short- and long-term AFW system upgrade requirements based on

item 1 above. Included in these letters will be a request for additional

information regarding items 2 and 3 above.

Operating License Applicants--Operating license applicants have been requested

to respond to staff letters of March 10, 1980 (W and C-E) and April 24, 1980

(B&W). These responses will be reviewed during the normal review process for

these applications.

II.E.1. 1-13 3-77
Applicability

This requirement applies to all PWR operating plants and applicants for

operating licenses.

Implementation

For operating reactors, the NRC staff will review and evaluate operating plant

licensee responses to staff recommendations for improving AFW system reliability

and requested information on AFW system flowrate design basis in time to

support licensee implementation of the short-term requirements by July 1, 1981

.and long-term requirements by January 1982.

Applicants for operating license should refer to letters of March 10, 1980 (W

and C-E) and April 24, 1980 (B&W) for implementation schedule.

Type of Review

A preimplementation review will be performed.

Documentation Required

Licensees and applicants will be required to submit the information indicated

above.

Technical Specification Changes Required

Changes to technical specifications will be determined by specific item.

Reference

NUREG-0660, Item II.E.1.1

Letter from D. F. Ross, Jr., NRC, to All Pending W and C-E License Applicants,

dated March 10, 1980.

Letter from D. F. Ross, Jr., NRC, to All Pending B&W License Applicants, dated

April 24, 1980.

3-78 II.E.1.1-2
II.E.1.2 AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION

PART 1: Auxiliary Feedwater System Automatic Initiation

Position

Consistent with satisfying the requirements of General Design Criterion 20 of

Appendix A to 10 CFR Part 50 with respect to the timely initiation of the

auxiliary feedwater system (AFWS), the following requirements shall be implemented

in the short term:

(1) The design shall provide for the automatic initiation of the AFWS.

(2) The automatic initiation signals and circuits shall be designed so that a

single failure will not result in the loss of AFWS function.

(3) Testability of the initiating signals and circuits shall be a feature of

the design. .

(4) The initiating signals and circuits shall be powered from the emergency

buses.

(5) Manual capability to initiate the AFWS from the control room shall be

retained and shall be implemented so that a single failure in the manual

circuits will not result in the loss of system function.

(6) The ac motor-driven pumps and valves in the AFWS shall be included in the

automatic actuation (simultaneous and/or sequential) of the loads onto

the emergency buses.

(7) The automatic initiating signals and circuits shall be designed so that

their failure will not result in the loss of manual capability to initiate

the AFWS from the control room.

In the long term, the automatic initiation signals and circuits shall be

upgraded in accordance with safety-grade requirements.

Changes to Previous Requirements and Guidance

There are no changes to the previous guidance issued in the H. R. Denton

letter to licensees, dated October 30, 1979.

Clarification

The intent of this recommendation is to assure a reliable automatic initiation

system. This objective can be met by providing a system which meets all the

requirements of IEEE Standard 279-1971.

II. E. 1.2-1 3-79
The staff has determined that the following salient paragraphs of IEEE 279-1971

should be addressed as a minimum:

IEEE 279-1971, Paragraph

4.1*

4.2*

4.3, & 4.4

4.6

4.7

4.9* & 4.10*

4.11

4.12

4.13

4.17*

General Functional Requirements

Single Failure

Qualification

Channel Independence

Control and Protection System Interaction

Capability for Testing

Channel Bypass

Operating Bypass

Indication of Bypass

Manual Initiation

Applicability

This requirement applies to all PWR operating reactors and

operating license.

Implementation

applicants for

Final design information should be submitted by January. 1, 1981.

grade system will be installed by July 1, 1981.

The safetyAll applicants for operating license should submit documentation 4 months

prior to the expected issuance of the staff safety evaluation report for an

operating license or 4 months prior to the listed implementation date, whichever

is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

Each licensee shall provide by January 1, 1981 sufficient documentation to

support a reasonable assurance finding by the NRC that the above requirements

are met. The documentation should include as a minimum

(1) A discussion of the design with respect to the above paragraphs of IEEE

279-1971; and

(2) Supporting information including system design description, logic diagrams,

electrical schematics, piping and instrument diagrams, test procedures,

and technical specifications.

*These requirements were part of the short-term, control-grade requirements.

3-80 II.E.1.2-2
Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.7.a

NUREG-0660, Item II.E.1.2

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

II.E.1.2-3 3-81
PART 2: Auxiliary Feedwater System Flowrate Indication

Position

Consistent with satisfying the requirements set forth in General Design

Criterion 13 to provide the capability in the control room to ascertain the

actual performance of the AFWS when it is called to perform its intended

function, the following requirements shall be implemented:

1. Safety-grade indication of auxiliary feedwater flow to each steam

generator shall be provided in the control room.

2. The auxiliary feedwater flow instrument channels shall be powered

from the emergency buses consistent with satisfying the emergency

power diversity requirements of the auxiliary feedwater system set

forth in Auxiliary Systems Branch Technical Position 10-1 of the

Standard Review Plan, Section 10.4.9.

Changes to Previous Requirements and Guidance

The requirements for Westinghouse (W) and Combustion Engineering (C-E) plants

have been relaxed to require only a single-channel flow indication, instead of

redundant channels. This single channel need not be seismically qualified

nor need it be powered from a Class 1E power source.

The auxiliary feedwater flow indication requirements have been relaxed for

PWRs with U-tube steam generators because flow indication is of secondary

importance in assuring steam generator cooling capability for steam generators

of this design.

Clarification

The intent of this recommendation is to assure a reliable indication of AFWS

performance. This objective can be met by providing an overall indication

system that meets the following appropriate design principles:

(1) For Babcock and Wilcox Plants

(a) To satisfy these requirements, B&W plants must provide as a minimum

two auxiliary feedwater flowrate indicators for each steam generator.

(b) The flow indication system should conform to the following salient

paragraphs of IEEE 279-1971:

3-82 II. E.1.2-4
IEEE 279-1971, PARAGRAPH

4.1* General Functional Requirements

4.2* Single Failure

4.3 & 4.4 Qualification

4.6 Channel Independence

4.7 Control and Protection System Interaction

4.9* & 4.10* Capability for Testing

(2) For Westinghouse and Combustion Engineering Plants

(a) To satisfy these requirements, W and C-E plants must provide as a

minimum one auxiliary feedwater flowrate indicator and one wide-range

steam-generator level indicator for each steam generator or two flowrate indicators.

(b) The flow indication system should be:

(i) environmentally qualified

(ii) powered from highly reliable, battery-backed non-Class IE power

source

(iii) periodically testable

(iv) part of plant quality assurance program

(v) capable of display on demand

It is important that the displays and controls added to the control room as a

result of this requirement not increase the potential for operator error. A

human-factor analysis should be performed taking into consideration:

(a) the use of this information by an operator during both normal and abnormal

plant conditions,

(b) integration into emergency procedures,

(c) integration into operator training, and

(d) other alarms during emergency and need for prioritization of alarms.

Applicability

This requirement applies to all PWR operating reactors and applicants for

operating license.

Implementation

Final design information should be submitted by January 1, 1981. The system

will be installed by July 1, 1981. All applicants for operating license

should submit documentation 4 months prior to the expected issuance of the

staff safety evaluation report for an operating license or 4 months prior to

the listed implementation date, whichever is later.

*These requirements were part of the short-term, control-grade requirements.

II. E. 1.2-5 3-83
Type of Review

A postimplementation review will be performed.

Documentation Required

By January 1, 1981 each licensee shall provide sufficient documentation to

support a reasonable assurance finding by the NRC that the above-specified

requirements have been met. The documentation should include as a minimum:

(1) A discussion of the design with respect to each of the requirements

specified above; and

(2) Supporting information including system design description, logic diagrams,

electrical schematics, piping and instrument diagrams, test procedures,

and technical specifications.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.7.b

NUREG-0660, Item II.E.1.2

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979

- - - 3-84 _II.E.1.2-6 E .
II.E.3.1 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS

Position

Consistent with satisfying the requirements of General Design Criteria 10, 14,

15, 17, and 20 of Appendix A-to 10 CFR Part 50 for the event of loss of offsite

power, the following positions shall be implemented:.

(1) The pressurizer heater power supply design shall provide the capability

to supply, from either the offsite power source or the emergency power

source (when offsite power is not available), a predetermined number of

pressurizer heaters and associated controls necessary to establish and

maintain natural circulation at hot standby conditions. The required

heaters and their controls shall be connected to the emergency buses in a

manner that will provide redundant power supply capability.

(2) Procedures and training shall be established to make the operator aware

of when and how the required pressurizer heaters shall be connected to

the emergency buses. If required, the procedures shall identify under

what conditions selected emergency loads can be shed from the emergency

power source to provide sufficient capacity for the connection of the

pressurizer heaters.

(3) The time required to accomplish the connection of the preselected pressurizer heater to the emergency buses shall be consistent with the timely

initiation and maintenance of natural circulation conditions.

(4) Pressurizer heater motive and control power interfaces with the emergency

buses shall be accomplished through devices that have been qualified in

accordance with safety-grade requirements.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements in October 30, 1979 letter

from H. R. Denton to all licensees.

Clarification

(1) Redundant heater capacity must be provided, and each redundant heater or

group of heaters should have access to only one Class lE division power

supply.

(2) The number of heaters required to have access to each emergency power

source is that number required to maintain natural circulation in the hot

standby condition.

(3) The power sources need not necessarily have the capacity to provide power

to the heaters concurrently with the loads required for loss-of-coolant

accident.

HI ._- E-3-1-1_ _ --3-85 --.-- ----
(4) Any changeover of the heaters from normal offsite power to emergency

onsite power is to be accomplished manually in the control room.

(5) In establishing procedure to manually load the pressurizer heaters onto

the emergency power sources, careful consideration must be given to:

(a) which ESF loads may be appropriately shed for a given situation;

(b) reset of the safety injection actuation signal to permit the operation of the heaters; and

(c) instrumentation and criteria for operator use to prevent overloading

a diesel generator.

(6) The Class IE interfaces for main power and control power are to be protected by safety-grade circuit breakers (see also Regulatory Guide 1.75).

(7) Being non-Class IE loads, the pressurizer heaters must be automatically

shed from the emergency power sources upon the occurrence of a safety

injection actuation signal (see item 5.b. above).

Applicability

This requirement applies to all PWR operating reactors and applicants for

operating license.

Implementation

Implementation is complete for operating reactors.

All applicants for operating license should submit documentation 4 months

prior to the expected issuance of the staff safety evaluation report for an

operating license or 4 months prior to the listed implementation date, whichever is later.

Type of Review

A review will be performed as part of the licensing review process.

Documentation Required

Each applicant shall provide sufficient documentation to support a reasonable

assurance finding by the NRC that each of the subparts of the position stated

above are met. The documentation should include as a minimum, supporting

information including system design description, logic diagrams, electrical

schematics, test procedures, and technical specifications.

Technical Specification Changes Required

Changes to technical specifications will be required.

II.E.3.1-2 3-86
References

NUREG-0578, Recommendation 2.1.1

NUREG-0660, Item II.E.3.1

NUREG-0694, Part 2

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

*II.E.3.1-3 3-87
II.E.4.1 DEDICATED HYDROGEN PENETRATIONS

Position

Plants using external recombiners or purge systems for postaccident combustible

gas control of the containment atmosphere should provide containment penetration

systems for external recombiner or purge .systems that are dedicated to that

service only, that meet the redundancy and single-failure requirements of

General Design Criteria 54 and 56 of Appendix A to 10 CFR 50, and that are

sized to satisfy the flow requirements of the recombiner or purge system.

The procedures for the use of combustible gas control systems following an

accident that results in a degraded core and release of radioactivity to the

containment must be reviewed and revised, if necessary.

Changes to Previous Requirements and Guidance

Changes in the implementation date have been made because of equipment procurement problems and to minimize the number of plant shutdowns necessary must

make to install equipment related to the TMI Action Plan.

Clarification

(1) An acceptable alternative to the dedicated penetration is a combined

design that is single-failure proof for containment isolation purposes

and single-failure proof for operation of the recombiner or purge system.

(2) The dedicated penetration or the combined single-failure proof alternative

shall be sized such that the flow requirements for the use of the recombiner

or purge system are satisfied. The design shall be based on 10 CFR 50.44

requirements.

(3) Components furnished to satisfy this requirement shall be safety grade.

(4) Licensees that rely on purge systems as the primary means for controlling

combustible gases following a loss-of-coolant accident should be aware of

the'positions taken in SECY-80-399, "Proposed Interim Amendments to

10 CFR Part 50 Related to Hydrogen Control and Certain Degraded Core

Considerations." This proposed rule, published in the Federal Register

on October 2, 1980, would require plants that do not now have recombiners

to have the capacity to install external recombiners by January 1, 1982.

(Installed internal recombiners are an acceptable alternative to the

above.)

(5) Containment atmosphere dilution (CAD) systems are considered to be purge

systems for the purpose of implementing the requirements of this TMI Task

Action item.

3-88 II. E. 4. 1-1
Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

For operating reactors, design modifications shall be completed by July 1,

1981.

Operating license applicants must have design changes completed by July 1,

1981 or prior to issuance of an operating license, whichever is later.

Type of Review

For operating reactors review will take place after implementation.

Documentation Required

The licensees shall inform the NRC when the required design modifications have

been completed.

Technical Specification Changes Required

Changes to technical specifications will be required for plants that need to

make modifications.

References

NUREG-0578

Letter from H. R. Denton, NRC, to All Operating Reactor Plants, dated

October 30, 1979.

II. E.4.1-2 3-89
II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY

Position

(1) Containment isolation system designs shall comply with the recommendations

of Standard Review Plan Section 6.2.4 (i.e., that there be diversity in

the parameters sensed for the initiation of containment isolation).

(2) All plant personnel shall give careful consideration to the definition of

essential and nonessential systems, identify each system determined to be

essential, identify each system determined to be nonessential, describe

the basis for selection of each essential system, modify their containment

isolation designs accordingly, and report the results of the reevaluation

to the NRC.

(3) All nonessential systems shall be automatically isolated by the containment

isolation signal.

(4) The design of control systems for automatic containment isolation valves

shall be such that resetting the isolation signal will not result in the

automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.

(5) The containment setpoint pressure that initiates containment isolation

for nonessential penetrations must be reduced to the minimum compatible

with normal operating conditions.

(6) Containment purge valves that do not satisfy the operability criteria set

forth in Branch Technical Position CSB 6-4 or the Staff Interim Position

of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item

II.3.f during operational conditions 1, 2, 3, and 4. Furthermore, these

valves must be verified to be closed at least every 31 days. (A copy of

the Staff Interim Position is enclosed as Attachment 1.)

(7) Containment purge and vent isolation valves must close on a high radiation

signal.

Changes to Previous Requirements and Guidance

Although there has been no change in the requirements since NUREG-0660 was

issued, positions 5, 6, and 7 have not been previously transmitted to licensees.

These three positions were not part of the original NUREG-0578 requirements of

Recommendation 2.1.4; however they were added to item II.E.4.1 of NUREG-0660

as a result of further staff evaluation of features needed to improve containment

isolation dependability. The schedule for implementing positions 5, 6, and 7

on operating plants has been changed from NUREG-0660. The design for position 5

shall be completed by January 1, 1981 with modifications completed by July 1,

1981. Position 6 shall be implemented by January 1, 1981. Position 7 shall

be implemented by July 1, 1981 or during the following outage of sufficient

duration, but no later than January 1, 1982.

3-90 II. E.4.2-1
Clarification

(1) The reference to SRP 6.2.4 in position 1 is only to the diversity requirements set forth in that document.

(2) For postaccident situations, each nonessential penetration (except instrument lines) is required to have two isolation barriers in series that

meet the requirements of General Design Criteria 54, 55, 56, and 57, as

clarified by Standard Review Plan, Section 6.2.4. Isolation must be

performed automatically (i.e., no credit can be given for operator action).

Manual valves must be sealed closed, as defined by Standard Review Plan,

Section 6.2.4, to qualify as an isolation barrier. Each automatic

isolation valve in a nonessential penetration must receive the diverse

isolation signals.

(3) Revision 2 to Regulatory Guide 1.141 will contain guidance on the classification of essential versus nonessential systems and is due to be issued

by June 1981. Requirements for operating plants to review their list of

essential and nonessential systems will be issued in conjunction with

this guide including an appropriate time schedule for completion.

(4) Administrative provisions to close all isolation valves manually before

resetting the isolation signals is not an acceptable method of meeting

position 4.

(5) Ganged reopening of containment isolation valves is not acceptable:

Reopening of isolation valves must be performed on a valve-by-valve

basis, or on a line-by-line basis, provided that electrical independence

and other single-failure criteria continue to be satisfied.

(6) The containment pressure history during normal operation should be used

as a basis for arriving at an appropriate minimum pressure setpoint for

initiating containment isolation. The pressure setpoint selected should

be far enough above the maximum observed (or expected) pressure inside

containment during normal operation so that inadvertent containment

isolation does not occur during normal operation from instrument drift or

fluctuations due to the accuracy of the pressure sensor. A margin of

1 psi above the maximum expected containment pressure should be adequate

to account for instrument error. Any proposed values greater than

1 psi will require detailed justification. Applicants for an operating

license and operating plant licensees that have operated less than one

year should use pressure history data from similar plants that have

operated more than one year, if possible, to arrive at a minimum containment setpoint pressure.

(7) Sealed-closed purge isolation valves shall be under administrative control

to assure that they cannot be inadvertently opened. Administrative

control includes mechanical devices to seal or lock the valve closed, or

to prevent power from being supplied to the valve operator. Checking the

valve position light in the control room is an adequate method for verifying

every 24 hours that the purge valves are closed.

II. E.4. 2-2 3-91
Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

As part of Category "A" lessons-learned requirements, all operating plants

were required to be in conformance with positions 1 through 4 by January 1,

1980.

Each licensee will provide, and justify, the minimum containment pressure that

will be used to initiate containment isolation as stated in position 5 by

January 1, 1981. By July 1, 1981, all operating plants must be in compliance

with position 5. All operating plants must be in compliance with position 6

by January 1, 1981. All operating plants must be in compliance with position 7

by July 1, 1981.

Applicants for an operating license must be in compliance with positions 1

through 4 before receiving an operating license. Applicants must be in compliance with positions 5 and 7 by July 1, 1981, and position 6 by January 1,

1981 or before they receive their operating license, whichever is later for

each position.

Applicants must provide, and justify, the minimum containment pressure that

will be used for initiating containment isolation as stated in position 5.

Type of Review

A postimplementation review will be performed for operating reactors.

Documentation Required

The type and dates of documentation required are as previously stated.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.4

NUREG-0660, Item II.E.4.2

Standard Review Plan, Section 6.2.4

3-92 II.E.4.2-3
II.E.4.2, ATTACHMENT 1, OCTOBER 23, 1979* INTERIM POSITION FOR CONTAINMENT

PURGE AND VENT VALVE OPERATION PENDING RESOLUTION OF ISOLATION

VALVE OPERABILITY

Once the conditions listed below are met, restrictions on use of the containment

purge and vent system isolation valves will be revised based on our review of

your responses to the November 1978 letter on this subject justifying your

proposed operational mode. The November 1978 letters to all licensees identified

certain events related to containment purging of concern to the NRC and requested

commitments to either cease purging or justify purging operations. The revised

restrictions can be established separately for each system.

(1) Whenever the containment integrity is required, emphasis should be placed

on operating the containment in a passive mode as much as possible and on

limiting all purging and venting times to as low as achievable. To

justify venting or purging, there must be an established need to improve

working conditions to perform a safety-related surveillance or safetyrelated maintenance procedure. (Examples of improved working conditions

would include deinerting, reducing temperature,** humidity, and airborne

activity sufficiently to permit efficient performance or to significantly

reduce occupational radiation exposures.)

(2) Maintain the containment purge and vent isolation valves closed whenever

the reactor is not in the cold shutdown or refueling mode until such time

as you can show that:

(a) All isolation valves greater than 3-in. nominal diameter used for

containment purge and venting operations are operable under the most

severe design-basis-accident (DBA) flow-condition loading and can

close within the time limit stated in the technical specifications,

design criteria, or operating procedures. The operability of butterfly

valves may, on an interim basis, be demonstrated by limiting the

valve to be no more than 300 to 500 open (900 being full open). The

maximum opening shall be determined in consultation with the valve

supplier. The valve opening must be such that the critical valve

parts will not be damaged by DBA-LOCA (loss-of-coolant accident)

loads and that the valve will tend to close when the fluid dynamic

forces are introduced, and

(b) Modifications, as necessary, have been made to segregate the containment ventilation isolation signals to ensure that, as a minimum, at

least one of the automatic safety injection actuation signals is

uninhibited and operable to initiate valve closure when any other

isolation signal may be blocked, reset, or overridden.

*Previously referred to as DOE Interim Position.

**Only when temperature and humidity controls are not in the present design.

II. E.4.2-4 3-93
II.F.1 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION

Introduction

Item II.F.1 of NUREG-0660 contains the following subparts:

(1) Noble gas effluent radiological monitor;

(2) Provisions for continuous sampling of plant effluents for postaccident

releases of radioactive iodines and particulates and onsite laboratory

capabilities (this requirement was inadvertently omitted from NUREG-0660;

see Attachment 2 that follows, for position);

(3) Containment high-range radiation monitor;

(4) Containment pressure monitor;

(5) Containment water level monitor; and

(6) Containment hydrogen concentration monitor.

NUREG-0578 provided the basic requirements associated with items (1) through

(3) above. Letters issued to all operating nuclear power plants dated

September 13, 1979 and October 30, 1979 provided clarification of staff requirements associated with items (1) through (6) above. Attachments 1 through 6

present the NRC position on these matters.

It is important that the displays and controls added to the control room as a

result of this requirement not increase the potential for operator error. *A

human-factor analysis should be performed taking into consideration:

(a) the use of this information by an operator during both normal and

abnormal plant conditions,

(b) integration into emergency procedures,

(c) integration into operator training, and

(d) other alarms during emergency and need for prioritization of alarms.

References

NUREG-0660, item II.F.1

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

3-94 II.F.1-1
II.F.1, ATTACHMENT 1, NOBLE GAS EFFLUENT MONITOR

Position

Noble gas effluent monitors shall be installed with an extended range designed

to function during accident conditions as well as during normal operating

conditions. Multiple monitors are considered necessary to cover the ranges of

interest.

(1) Noble gas effluent monitors with an upper range capacity of 105 pCi/cc

(Xe-133) are considered to be practical and should be installed in all

operating plants.

(2) Noble gas effluent monitoring shall be provided for the total range of

concentration extending from normal condition (as low as reasonably

achievable (ALARA)) concentrations to a maximum of 105 pCi/cc (Xe-133).

Multiple monitors-are considered to be necessary to cover the ranges of

interest. The range capacity of individual monitors should overlap by a

factor of ten.

Changes to Previous Requirements and Guidance

This requirement was originally issued by letters to all operating power

plants dated September 13 and October 30, 1979. Significant changes in requirements or guidance are:

(1) Deletion of specific range overlap requirement.

(2) Specifies that offline monitoring is not required for safety valve and

dump valve discharge lines.

(3) Implementation date changed from January 1, 1981 to January 1, 1982.

(4) Specifies that inline sensors are acceptable for concentrations between

102 pCi/cc to 105 pCi/cc of noble gases.

Clarification

(1) Licensees shall provide continuous monitoring of high-level, postaccident

releases of radioactive noble gases from the plant. Gaseous effluent

monitors shall meet the requirements specified in the enclosed Table II.F.1-1.

Typical plant effluent pathways to be monitored are also given in the

table.

(2) The monitors shall be capable of functioning both during and following an

accident. System designs shall accommodate a design-basis release'and

then be capable of following decreasing concentrations of noble gases.

(3) Offline monitors are not required for the PWR secondary side main steam

safety valve and dump valve discharge lines. For this application,

externally mounted monitors viewing the main steam line upstream of the

valves are acceptable with procedures to correct for the low energy

gammas the external monitors would not detect. Isotopic identification

is not required.

II.F.1-2 3-95
(4) Instrumentation ranges shall overlap to cover the entire range of effluents

from normal (ALARA) through accident conditions.

The design description shall include the following information.

(a) System description, including:

(i) instrumentation to be used, including range or sensitivity,

energy dependence or response, calibration frequency and technique,

and vendor's model number, if applicable;

(ii) monitoring locations (or points of sampling), including description of methods used to assure representative measurements and

background correction;

(iii) location of instrument readout(s) and method of recording,

including description of the method or procedure for transmitting

or disseminating the information or data;

(iv) assurance of the capability to obtain readings at least every

15 minutes during and following an accident; and,

(v) the source of power to be used.

(b) Description of procedures or calculational methods to be used for

converting instrument readings to release rates per unit time, based

on exhaust air flow and considering radionuclide spectrum distribution

as a function of time after shutdown.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

Implementation must be completed by January 1, 1982.

Type of Review

A postimplementation review will be performed.

Documentation Required

Licensees and licensing applicants should have available for review the final

design description of the as-built system, including piping and instrument

diagrams together with either (1) a description of procedures for system

operation and calibration, or (2) copies of procedures for system operation

and calibration. Operating Reactors--By January 1, 1981 operating reactors

should have available for review the final design details of the implementation

of the above position and clarifications. If deviations to the above position

or clarification are necessary, provide detailed explanation and justification

for the deviations by January 1, 1981.

3-96 II. F. 1-3
---

License applicants will submit the above details in accordance with the proposed

review schedule, but in no case less than 4 months prior to the issuance of an

operating license.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.8.b

American National Standard ANSI N13.1-1969, February 1969

Letter from D. G. Eisenhut, NRC, to all Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

II.F.1-4 3-97
TABLE II.F.1-1

HIGH-RANGE NOBLE GAS EFFLUENT MONITORS

REQUIREMENT

PURPOSE

- Capability to detect and measure concentrations of noble

gas fission products in plant gaseous effluents during and

following an accident. All potential accident release

paths shall be monitored.

- To provide the plant operator and emergency planning agencies with information on plant releases of noble gases during and following an accident.

DESIGN BASIS MAXIMUM RANGE

Design range values may be expressed in Xe-133 equivalent values for monitors

employing gamma radiation detectors or in microcuries per cubic centimeter of

air at standard temperature and pressure (STP) for monitors employing beta

radiation detector (Note: lR/hr @1 ft = 6.7 Ci Xe-133 equivalent for point

source). Calibrations with a higher energy source are acceptable. The decay

of radionuclide noble gases after an accident (i.e., the distribution of noble

gases changes) should be taken into account.

105 pCi/cc - Undiluted containment exhaust gases

reactor building purge, PWR drywell

the standby gas treatment system).

(e.g., PWR

purge through

- Undiluted PWR condenser air removal system exhaust.

104 pCi/cc - Diluted containment exhaust gases (e.g., > 10:1

dilution, as with auxiliary building exhaust air).

- BWR reactor building (secondary containment) exhaust air.

- PWR secondary containment exhaust air.

103 pCi/cc - Buildings with systems containing primary coolant

or primary coolant offgases (e.g., PWR auxiliary

buildings, BWR turbine buildings).

- PWR steam safety valve discharge, atmospheric steam dump

valve discharge.

102 pCi/cc - Other release points (e.g., radwaste buildings,

fuel handling/storage buildings).

3-98 II.F.1-5
TABLE II.F.1-1

(CONTINUED)

REDUNDANCY

SPECIFICATIONS

POWER SUPPLY

- Not required; monitoring the final release point of several

discharge inputs is acceptable.

- (None) Sampling design criteria per ANSI N13.1.

- Vital instrument bus or dependable backup power supply to

normal ac.

CALIBRATION

DISPLAY

QUALIFICATION

- Calibrate monitors using gamma detectors to Xe-133 equivalent

(1 R/hr @ 1 ft = 6.7 Ci Xe-133 equivalent for point source).

Calibrate monitors using beta detectors to Sr-90 or similar

long-lived beta isotope of at least 0.2 MeV.

- Continuous and recording as equivalent Xe-133 concentrations

or pCi/cc of actual noble gases.

- The instruments shall provide sufficiently accurate responses

to perform the intended function in the environment to

which they will be exposed during accidents.

DESIGN

CONSIDERATIONS

Offline monitoring is acceptable for all ranges of noble

gas concentrations.

Inline (induct) sensors are acceptable for 102 pCi/cc to

105 pCi/cc noble gases. For less than 102 pCi/cc, offline

monitoring is recommended.

Upsteam filtration (prefiltering to remove radioactive

iodines and particulates) is not required; however, design

should consider all alternatives with respect to capability

to monitor effluents following an accident.

For external mounted monitors (e.g., PWR main steam line),

the thickness of the pipe should be taken into account in

accounting for low-energy gamma radiation.

II. F. 1-6 3-99
II.F.1, ATTACHMENT 2 SAMPLING AND ANALYSIS OF PLANT EFFLUENTS

Position

Because iodine gaseous effluent monitors for the accident condition are not

considered to be practical at this time, capability for effluent monitoring of

radioiodines for the accident condition shall be provided with sampling conducted

by adsorption on charcoal or other media, followed by onsite laboratory analysis.

Changes to Previous Requirements and Guidance

This requirement was originally issued by letters to all operating power

plants dated September 13, 1979 and October 30, 1979. This requirement was

inadvertently omitted from NUREG-0660. Significant changes in requirements or

guidance are:

(1) Changes implementation date to January 1, 1982.

(2) Specifies a shielding basis design envelope for design of samplers and

sample transport devices.

(3) Specifies provisions for isokinetic sampling.

(4) Specifies representative sampling per criteria of ANSI N131-1969.

(5) Allows use of gamma radiation measurement and shielding/distance factors

in lieu of analysis of highly radioactive samples.

Clarification

(1) Licensees shall provide continuous sampling of plant gaseous effluent for

postaccident releases of radioactive iodines and particulates to meet the

requirements of the enclosed Table II.F.1-2. Licensees shall also provide

onsite laboratory capabilities to analyze or measure these samples. This

requirement should not be construed to prohibit design and development of

radioiodine and particulate monitors to provide online sampling and

analysis for the accident condition. If gross gamma radiation measurement

techiques are used, then provisions shall be made to minimize noble gas

interference.

(2) The shielding design basis is given in Table II.F.1-2. The sampling

system design shall be such that plant personnel could remove samples,

replace sampling media and transport the samples to the onsite analysis

facility with radiation exposures that are not in excess of the criteria

of GDC 19 of 5-rem whole-body exposure and 75 rem to the extremities

during the duration of the accident.

(3) The design of the systems for the sampling of particulates and iodines

should provide for sample nozzle entry velocities which are approximately

isokinetic (same velocity) with expected induct or instack air velocities.

For accident conditions, sampling may be complicated by a reduction in

stack or vent effluent velocities to below design levels, making it

necessary to substantially reduce sampler intake flow rates to achieve

the isokinetic condition. Reductions in air flow may well be beyond the

3-100 II. F.1-7
capability of available sampler flow controllers to maintain isokinetic

conditions; therefore, the staff will accept flow control devices which

have the capability of maintaining isokinetic conditions with variations

in stack or duct design flow velocity of + 20%. Further departure from

the isokinetic condition need not be considered in design. Corrections

for non-isokinetic sampling conditions, as provided in Appendix C of ANSI

13.1-1969 may be considered on an ad hoc basis.

(4) Effluent streams which may contain air with entrained water, e.g. air

ejector discharge, shall have provisions to ensure that the adsorber is

not degraded while providing a representative sample, e.g., heaters.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

This requirement will be implemented by January 1, 1982.

Type of Review

A postimplementation review will be performed.

Documentation Required

By January 1, 1981 operating reactors should have available for review the

final design details of the implementation of the above position and clarifications. If deviations to the above position or clarification are necessary,

provide detailed explanation and justification for the deviations by January 1,

1981.

License applicants will submit the above details in accordance with the proposed

review schedule, but in no case less than 4 months prior to the issuance of an

operating license.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.8.b

American National Standard ANSI N13.1-1969, February 1969

Letter from D. R. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

II.F.1-8 3^101
TABLE II.F.1-2

SAMPLING AND ANALYSIS OR MEASUREMENT OF HIGH-RANGE RADIOIODINE AND

PARTICULATE EFFLUENTS IN GASEOUS EFFLUENT STREAMS

EQUIPMENT

PURPOSE

DESIGN BASIS

SHIELDING

ENVELOPE

- Capability to collect and analyze or measure representative

samples of radioactive iodines and particulates in plant

gaseous effluents during and following an accident. The

capability to sample and analyze for radioiodine and

particulate effluents is not required for PWR secondary

main steam safety valve and dump valve discharge lines.

- To determine quantitative release of radioiodines and

particulates for dose calculation and assessment.

- 102 pCi/cc of gaseous radioiodine and particulates, deposited

on sampling media; 30 minutes sampling time, average gamma

energy CE) of 0.5 MeV.

SAMPLING MEDIA -

- Iodine > 90% effective adsorption for all forms of gaseous iodine.

- Particulates > 90% effective retention for 0.3 micron (p) diameter particles.

SAMPLING CONSIDERATIONS

- Representative sampling per ANSI N13.1-1969.

- Entrained moisture in effluent stream should not degrade adsorber.

- Continuous collection required whenever exhaust flow occurs.

- Provisions for limiting occupational dose to personnel incorporated in

sampling systems, in sample handling and transport, and in analysis of

samples.

ANALYSIS

- Design of analytical facilities and preparation of analytical procedures

shall consider the design basis sample.

Highly radioactive samples may not be compatible with generally accepted

analytical procedures; in such cases, measurement of emissive gamma

radiations and the use of shielding and distance factors should be considered in design.

3-102 II.F.1-9
II.F.1, ATTACHMENT 3, CONTAINMENT HIGH-RANGE RADIATION MONITOR

Position

In containment radiation-level monitors with a maximum range of 108 rad/hr

shall be installed. A minimum of two such monitors that are physically separated

shall be provided. Monitors shall be developed and qualified to function in

an accident environment.

Changes to Previous Requirements and Guidance

This requirement was originally issued by letters to all operating power

plants dated September 13 and October 30, 1979 and was incorporated into

NUREG-0660. Significant changes in requirements or guidance are:

(1) Specifies a lower range so that the monitor can follow the radiation

increase from lower levels of radiation for personnel safety up to the

maximum expected in major accidents;

(2) Specifies that monitors be located in containment to view a large segment

of the containment atmosphere which will more accurately reflect and

monitor accident conditions;

(3) Requires monitors in both primary containment (drywell) and secondary

containment for BWR Mark III, because under certain accident conditions

the drywell and secondary containment are interconnected through the

suppression pool resulting in high radiation in both containments following

an accident;

(4) Specifies accuracy and energy response in order to ensure accurate measurements independent of the energy spectrum of an accident (this specification

was referenced in the letter of October 30; 1979 in referencing Regulatory

Guide 1.97, Rev. 2);

(5) Specifies design and qualification criteria to ensure that the monitor

will function in an accident environment;

(6) Specifies that electronic calibration is acceptable for higher dose rate

ranges because such methods are sufficient to provide acceptable accuracy;

(7) Deletes the requirement for NRR (Office of Nuclear Reactor Regulation)

preimplementation review if the monitors meet the listed specifications

because the monitor specifications ensure that adequate monitors will be

installed;

(8) Moves the implementation date to January 1, 1982 because of the potential

unavailability of appropriate equipment and because the qualification of

monitors is incomplete; _

(9) Requires documentation by July 1, 1981, of alternative proposals for

monitors that do not meet the requirements of Table II.F.1-3.

II. F. 1-10 3-103
Clarification

(1) Provide two radiation monitor systems in containment which are documented

to meet the requirements of Table II.F.1-3.

(2) The specification of 108 rad/hr in the above position was based on a

calculation of postaccident containment radiation levels that included

both particulate (beta) and photon (gamma) radiation. A radiation detector

that responds to both beta and gamma radiation cannot be qualified to

post-LOCA (loss-of-coolant accident) containment environments but gammasensitive instruments can be so qualified. In order to follow the course

of an accident, a containment monitor that measures only gamma radiation

is adequate. The requirement was revised in the October.30, 1979 letter

to provide for a photon-only measurement with an upper range of 107 R/hi.

(3) The monitors shall be located in containment(s) in a manner as to provide

a reasonable assessment of area radiation conditions inside containment.

The monitors shall be widely separated so as to provide independent

measurements and shall "view" a large fraction of the containment volume.

Monitors should not be placed in areas which are protected by massive

shielding and should be reasonably accessible for replacement, maintenance,

or calibration. Placement high in a reactor building dome is not recommended

because of potential maintenance difficulties.

(4) For BWR Mark III containments, two such monitoring systems should be

inside both the primary containment (drywell) and the secondary containment.

(5) The monitors are required to respond to gamma photons with energies as

low as 60 keV and to provide an essentially flat response for gamma

energies between 100 keV and 3 MeV, as specified in Table II.F.1-3.

Monitors that use thick shielding to increase the upper range will underestimate postaccident radation levels in containment by several orders of

magnitude because of their insensitivity to low energy gammas and are not

acceptable.

Applicability

This requirement applies to all operating reactors and all applicants for

operating licenses.

Implementation Date

Implementation for operating reactors must be completed by January 1, 1982.

License applicants will submit the required documentation in accordance with

the appropriate review schedule, but in no case less than 4 months prior to

the issuance of the staff evaluation report for an operating license.

3-104 II. F.1-11
Type of Review

A postimplementation review will be performed.

Documentation Required

For operating licenses applicants, provide a description of the installed

high-range containment monitors and specify the locations of these monitors

inside containment. The description of the monitors should include:

(1) The description of or name of manufacturer and model number of the monitors;

(2) Verification that the monitors meet the specifications of Table II.F.1-3;

(3) Verification that the monitors will be operable on January 1, 1982; and,

(4) A plant layout drawing showing the location of the monitors.

Operating Reactors--By July 1, 1981 have available for review the final design

details of the implementation of the above position and clarifications. If

deviations to the above position or clarifications are necessary, provide a

detailed explanation of and justification for the deviations by July 1, 1981.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.8.b

NUREG-0660

Regulatory Guide 1.97, Revision 2

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

II. F.1-12 3-105
TABLE II.F.1-3

CONTAINMENT HIGH-RANGE RADIATION MONITOR

REQUIREMENT

RANGE

RESPONSE

REDUNDANT

DESIGN AND

QUALIFICATION

SPECIAL

CALIBRATION

SPECIAL

ENVIRONMENTAL

QUALIFICATIONS

The capability to detect and measure the radiation level

within the reactor containment during and following an

accident.

1 rad/hr to 108 rads/hr (beta and gamma) or alternatively

1 R/hr to 107 R/hr (gamma only).

60 keV to 3 MeV photons, with linear energy response

+ 20%) for photons of 0.1 MeV to 3 MeV. Instruments must

be accurate enough to provide usable information.

A minimum of two physically separated monitors (i.e.,

monitoring widely separated spaces within containment).

Category 1 instruments as described in Appendix A, except

as listed below.

In situ calibration by electronic signal substitution is

acceptable for all range decades above 10 R/hr. In situ

calibration for at least one decade below 10 R/hr shall be

by means of calibrated radiation source. The original

laboratory calibration is not an acceptable position due

to the possible differences after in situ installation.

For high-range calibration, no adequate sources exist, so

an alternate was provided.

Calibrate and type-test representative specimens of detectors

at sufficient points to demonstrate linearity through all

scales up to 106 R/hr. Prior to initial use, certify calibration of each detector for at least one point per decade

of range between 1 R/hr and 103 R/hr.

3-106 II. F.1-13
II.F.1, ATTACHMENT 4, CONTAINMENT PRESSURE MONITOR

Position

A continuous indication of containment pressure shall be provided in the

control room of each operating reactor. Measurement and indication capability

shall include three times the design pressure of the containment for concrete,

four times the design pressure for steel, and -5 psig for all contaiments.

Changes to Previous Requirements And Guidance

Regulatory Guide 1.97, Rev. 2 has been referenced since the October 30, 1979

letter as the guide for the design and qualification criteria for the containment pressure monitor. However, there have been many changes made to this

proposed revision and it has not yet been made final. Therefore, the appropriate sections of the latest version of Regulatory Guide 1.97 has been added

to this letter, Appendix A, and this is to be considered a new requirement.

The implementation date has been changed because of the new requirements and

because of equipment procurement problems. The new implementation schedule is

intended to allow licensees enough time to complete design modifications with

a minimum number of plant shutdowns.

Clarification

(1) Design and qualification criteria are outlined in Appendix A.

(2) Measurement and indication capability shall extend to 5 psia for subatmospheric containments.

(3) Two or more instruments may be used to meet requirements. However,

instruments that need to be switched from one scale to another scale to

meet the range requirements are not acceptable.

(4) Continuous display and recording of the containment pressure over the

specified range in the control room is required.

(5) The accuracy and response time specifications of the pressure monitor

shall be provided and justified to be adequate for their intended function.

APPLICABILITY

This requirement applies to all operating reactors and all applicants for

operating licenses.

Implementation

For operating reactors, design modifications should be completed by January 1,

1982.

Operating license applicants with an operating license dated before January 1,

1982 must have design changes completed by January 1, 1982; those applicants

with license dated after January 1, 1982 must have all design modifications

completed before they can receive their operating license.

II. F.1-14 3-107
Type of Review

A postimplementation review will be performed for operating reactors.

Documentation Required

The licensees shall inform the NRC when the required design modifications have

been completed. Applicants with operating license dates beyond January 1,

1982 shall provide the required design information at least 6 months before

the expected date of operation.

Technical Specification Changes Required

Changes to technical specification will be required.

References

NUREG-0660

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

3-108 II. F. 1-15
II.F.1, ATTACHMENT 5, CONTAINMENT WATER LEVEL MONITOR

Position

A continuous indication of containment water level shall be provided in the

control room for all plants. A narrow range instrument shall be provided for

PWRs and cover the range from the bottom to the top of the containment sump.

A wide range instrument shall also be provided for PWRs and shall cover the

range from the bottom of the containment to the elevation equivalent to a

600,000 gallon capacity. For BWRs, a wide range instrument shall be provided

and cover the range from the bottom to 5 feet above the normal water level of

the suppression pool.

Changes to Previous Requirements and Guidance

Regulatory Guide 1.97, Rev. 2 was referenced in the October 30, 1979 letter as

the guide for the design and qualification criteria for the wide range containment water level monitor. However, there have been many changes made to this

proposed revision and it has not yet been made final. Therefore, the appropriate sections of the latest version of Regulatory Guide 1.97 has been added

to this letter (Appendix A) and this is to be considered a new requirement.

The implementation date has been changed because of the new requirements and

because of equipment procurement problems. The new implementation schedule is

intended to allow licensees enough time to complete design modifications with

a minimum number of plant shutdowns.

Clarification

(1) The containment wide-range water level indication channels shall meet the

design and qualification criteria as outlined in Appendix A. The narrowrange channel shall meet the requirements of Regulatory Guide 1.89.

(2) The measurement capability of 600,000 gallons is based on recent plant

designs. For older plants with smaller water capacities, licensees may

propose deviations from this requirement based on the available water

supply capability at their plant.

(3) Narrow-range water level monitors are required for all sizes of sumps but

are not required in those plants that do not contain sumps inside the

containment.

(4) For BWR pressure-suppression containments, the emergency core cooling

system (ECCS) suction line inlets may be used as a starting reference

point for the narrow-range and wide-range water level monitors, instead

of the bottom of the suppression pool.

(5) The accuracy requirements of the water level monitors shall be provided

and justified to be adequate for their intended function.

II. F. 1-16 3-109
Applicability

This requirement applies to all operating reactors and all operating licenses

for applicants

Implementation

For operating reactors, design modifications should be completed by January 1,

1982.

Operating license applicants with an operating license date before July 1, 1981

must have design changes completed by July 1, 1981, whereas those applicants

with license dates past July 1, 1981 must have all design modifications completed

before they can receive their operating license.

Type of Review

A postimplementation review will be performed for operating reactors and

applicants for an operating license prior to January 1, 1982.

A preimplementation review will be performed for applicants for an operating

license after January 1, 1982.

Documentation Required

Submittals from operating reactors licensees and applicants for operating

licenses (with an operating license date before January 1, 1982) shall be

provided by January 1, 1982. Applicants with operating license dates beyond

January 1, 1982 shall provide the required design information at least 6

months before the expected date of operation.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0660

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

3-110 II. F.1-17
II.F.1, ATTACHMENT 6, CONTAINMENT HYDROGEN MONITOR

Position

A continuous indication of hydrogen concentration in the containment atmosphere

shall be provided in the control room. Measurement capability shall be provided

over the range of 0 to 10% hydrogen concentration under both positive and

negative ambient pressure.

Changes to Previous Requirements and Guidance

Regulatory Guide 1.97, Rev. 2 was referenced in the October 30, 1979 letter as

the guide for the design and qualification criteria for the containment hydrogen.

monitor. However, there have been many changes made to this proposed revision

and it has not yet been made final. Therefore, the appropriate sections of

the latest version of Regulatory Guide 1.97 have been added to this letter

(Appendix A) and, therefore, this is to be considered a new requirement.

The implementation date has been changed due to equipment procurement problems.

The new implementation schedule is intended to allow licensees enough time to

complete design modifications with a minimum number of plant shutdowns.

Clarification

(1) Design and qualification criteria are outlined in Appendix A.

(2) The continuous indication of hydrogen concentration is not required

during normal operation.

If an indication is not available at all times, continuous indication and

recording shall be functioning within 30 minutes of the initiation of

safety injection.

(3) The accuracy and placement of the hydrogen monitors shall be provided and

justified to be adequate for their intended function.

Applicability

This requirement applies to all operating reactors and all applicants for

operating licenses.

Implementation

For operating reactors, design modifications should be completed by January 1,

1982.

Operating license applicants with an operating license date before January 1,

1982 must have design changes completed by January 1, 1982, whereas those

applicants with license dates past January 1, 1982 must have all design modifications completed before they can receive their operating license.

II.F.1-18 3-111
Type of Review

A postimplementation review for operating reactors and applicants for an

operating license prior to January 1, 1982 will be performed.

A preimplementation review for applicants for an operating license after

January 1, 1982 will be performed.

Documentation Required

Operating reactors and applicants for operating license receiving an operating

license before January 1, 1982 will submit documentation before January 1, 1982.

Applicants with operating license issued after January 1, 1982 shall provide

the required design information at least 6 months prior to the expected date

of operation.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0660

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

3-112 II. F. 1-19
II.F.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING

Position

Licensees shall provide a description of any additional instrumentation or

controls (primary or backup) proposed for the plant to supplement existing

instrumentation (including primary coolant saturation monitors) in order to

provide an unambiguous, easy-to-interpret indication of inadequate core cooling

(ICC). A description of the functional design requirements for the system

shall also be included. A description of the procedures to be.used with the

proposed equipment, the analysis used in developing these procedures, and a

schedule for installing the equipment shall be provided.

Changes to Previous Requirements and Guidance

(1). Specify the "Design and Qualification Criteria" for the final ICC monitoring

system in section, "Clarification" (items 7, 8, and 9), Attachment 1, and

Appendix A.

(2) Specify complete documentation package to allow NRC evaluation of the

final ICC monitoring systems to begin on January 1, 1981.

(3) No preimplementation review is required but postimplementation review of

installation and preimplementation review before use as a basis for

operator decisions are required.

(4) Installation of additional instrumentation is now required by January 1,

1982.

(5) Clarification item (6) has been expanded to provide licensees/applicants

with more flexibility and diversity in meeting the requirements for

determining liquid level indication by providing possible examples of

alternative methods.

Previous guidance on the design and qualification criteria for upgrading of

existing instrumentation was based on Regulatory Guide 1.97, which is still

being developed. Detailed design requirements for incore thermocouples and

additional instrumentation were not specified. The pertinent portions of

draft Regulatory Guide 1.97 have now been included as Appendix A. Design

requirements for incore thermocouples used in the ICC monitoring system are

specified in Attachment 1. The only significant change in design requirements

involves a relaxation of qualification requirements for display systems amenable

to computer processing. This facilitates procurement of computer systems and

makes feasible the use of cathode ray tube (CRT) displays that may be needed

for proper interpretation of some reactor-water-level systems under development.

This relaxation can be accomplished without compromise of ICC monitoring

reliability by requiring 99% availability for the display systems, by requiring

postaccident maintenance accessibility for nonredundant portions of the system,

and by relying on diverse methods of ICC monitoring that include completely

qualified display systems.

II.F.2-1 3-113
The staff has concluded that the previous installation requirement of January 1,

1981 for additional instrumentation is unrealistic for most licensees, due to

procurement and development problems associated with proposed measurement

methods. Further, the staff cannot find the proposed methods acceptable for

use until development programs have been completed.

Clarification

(1) Design of new instrumentation should provide an unambiguous indication of

ICC. This may require new measurements or a synthesis of existing measurements which meet design criteria (item 7).

(2) The evaluation is to include reactor-water-level indication.

(3) Licensees and applicants are required to provide the necessary design

analysis to support the proposed final instrumentation system for inadequate

core'cooling and to evaluate the merits of various instruments to monitor

water'level and to monitor other parameters indicative of core-cooling

conditions.

(4) The indication of ICC must be unambiguous in that it should have the

following properties:

(a) It must indicate the existence of inadequate core cooling caused by

various phenomena (i.e., high-void fraction-pumped flow as well as

stagnant boil-off); and,

(b) It must not erroneously indicate ICC because of the presence of an

unrelated phenomenon.

(5) The indication must give advanced warning of the approach of ICC.

(6) The indication must cover the full range from normal operation to complete

core uncovery. For example, water-level instrumentation may be chosen to

provide advanced warning of two-phase level drop to the top of the core

and could be supplemented by other indicators such as incore and core-exit

thermocouples provided that the indicated temperatures can be correlated

to provide indication of the existence of ICC and to infer the extent of

core uncovery. Alternatively, full-range level instrumentation to the

bottom of the core may be employed in conjunction with other diverse

indicators such as core-exit thermocouples to preclude misinterpretation

due to any inherent deficiencies or inaccuracies in the measurement

system selected.

(7) All instrumentation in the final ICC system must be evaluated for conformance to Appendix A, "Design and Qualification Criteria for Accident

Monitoring Instrumentation," as clarified or modified by the provisions

of items 8 and 9 that follow. This is a new requirement.

(8) If a computer is provided to process liquid-level signals for display,

seismic qualification is not required for the computer and associated

3-114 II. F. 2-2
hardware beyond the isolator or input buffer at a location accessible for

maintenance following an accident. The single-failure criteria of item 2,

Appendix A, need not apply to the channel beyond the isolation device if

it is designed to provide 99% availability with respect to functional

capability for liquid-level display. The display and associated hardware

beyond the isolation device need not be Class lE, but should be energized

from a high-reliability power source which is battery backed. The quality

assurance provisions cited in Appendix A, item 5, need not apply to this

portion of the instrumentation system. This is a new requirement.

(9) Incore thermocouples located at the core exit or at discrete axial levels

of the ICC monitoring system and which are part of the monitoring system

should be evaluated for conformity with Attachment 1, "Design and Qualification Criteria for PWR Incore Thermocouples," which is a new requirement.

(10) The types and locations of displays and alarms should be determined by

performing a human-factors analysis taking into consideration:

(a) the use of this information by an operator during both normal and

abnormal plant conditions,

(b) integration into emergency procedures,

(c) integration into operator training, and

(d) other alarms during emergency and need for prioritization of alarms.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

This requirement must be implemented by January 1, 1982.

Type of Review

A postimplementation review will be performed for installation, and a preimplementation review will be performed prior to use.

Documentation Required

By January 1, 1981, the licensee shall provide a report detailing the planned

instrumentation system for monitoring of ICC. The report should contain the

necessary information, either by inclusion or by reference to previous submittals

including pertinent generic reports, to satisfy the requirements which follow:

(1) A description of the proposed final system including:

(a) a final design description of additional instrumentation and

displays;

II.F.2-3 3-115
(b) a detailed description of existing instrumentation systems (e.g.,

subcooling meters and incore thermocouples), including parameter

ranges and displays, which provide operating information pertinent

to ICC considerations; and

(c) a description of any planned modifications to the instrumentation

systems described in item 1.b above.

(2) The necessary design analysis, including evaluation of various instruments

to monitor water level, and available test data to support the design

described in item 1 above.

(3) A description of additional test programs to be conducted for evaluation,

qualification, and calibration of additional instrumentation.

(4) An evaluation, including proposed actions, on the conformance of the ICC

instrument system to this document, including Attachment 1 and Appendix A.

Any deviations should be justified.

(5) A description of the computer functions associated with ICC monitoring

and functional specifications for relevant software in the process computer and other pertinent calculators. The reliability of nonredundant

computers used in the system should be addressed.

(6) A current schedule, including contingencies, for installation, testing

and calibration, and implementation of any proposed new instrumentation

or information displays.

(7) Guidelines for use of the additional instrumentation, and analyses used

to develop these procedures.

(8) A summary of key operator action instructions in the current emergency

procedures for ICC and a description of how these procedures will be

modified when the final monitoring system is implemented.

(9) A description and schedule commitment for any additional submittals which

are needed to support the acceptability of the proposed final instrumentation system and emergency procedures for ICC.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.3.b

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

3-1 16 II.F.2-4
II.F.2, ATTACHMENT 1, DESIGN AND QUALIFICATION CRITERIA FOR PRESSURIZED-WATER

REACTOR INCORE THERMOCOUPLES

(1) Thermocouples located at the core exit for each core quadrant, in conjunction with core inlet temperature data, shall be of sufficient number

to provide indication of radial distribution of the coolant enthalpy

(temperature) rise across representative regions of the core. Power

distribution symmetry should be considered when determining the specific

number and location of thermocouples to be provided for diagnosis of

local core problems.

(2) There should be a primary operator display (or displays) having the

capabilities which follow:

(a) A spatially oriented core map available on demand indicating the

temperature or temperature difference across the core at each core

exit thermocouple location.

(b)' A selective reading of core exit temperature, continuous on demand,.

which is consistent with parameters pertinent to operator actions in

connecting with plant-specific inadequate core cooling procedures.

For example, the action requirement and the displayed temperature

might be either the highest of all operable thermocouples or the

average of five highest thermocouples.

(c) Direct readout and hard-copy capability should be available for all

thermocouple temperatures. The range should extend from 200'F (or

less) to 1800F (or more).

(d) Trend capability showing the temperature-time history of representative core exit temperature values should be available on demand.

(e) Appropriate alarm capability should be provided consistent with

operator procedure requirements.

(f) 'The operator-display device interface shall be human-factor designed

to provide rapid access to requested displays.

(3) A backup display (or displays) should be provided with the capability for

selective reading of a minimum of 16 operable thermocouples, 4 from each

core quadrant, all within a time interval no greater than 6 minutes. The

range should extend from 200F (or less) to 2300'F (or more).

(4) The types and locations of displays and alarm's should be determined by

performing a human-factors analysis taking into consideration:

(a) the use of this information by an operator during both normal and

abnormal plant conditions.

II. F. 2-5 3-117
(b) integration into emergency procedures,

(c) integration into operator training, and

(d) other alarms during emergency and need for prioritization of alarms.

(5) The instrumentation must be evaluated for conformance to Appendix B,

"Design and Qualification Criteria for Accident Monitoring Instrumentation,"

as modified by the provisions of items 6 through 9 which follow.

(6) The primary and backup display channels should be electrically independent,

energized from independent station Class IE power sources, and physically

separated in accordance with Regulatory Guide 1.75 up to and including

any isolation device. The primary display and associated hardware beyond

the isolation device need not be Class IE, but should be energized from a

high-reliability power source, battery backed, where momentary interruption

is not tolerable. The backup display and associated hardware should be

Class IE.

(7) The instrumentation should be environmentally qualified as described-in

Appendix B, item 1, except that seismic qualification is not required for

the primary display and associated hardware beyond the isolater/input

buffer at a location accessible for maintenance following an accident.

(8) The primary and backup display channels should be design to provide 99%

availability for each channel with respect to functional capability to

display a minimum of four thermocouples per core quadrant. The availability

shall be addressed in technical specifications.

(9) The quality assurance provisions cited in Appendix B, item 5, should be

applied except for the primary display and associated hardware beyond the

isolation device.

3-118 II.F.2-6
II.G.1 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT

Position

Consistent with satisfying the requirements of General Design Criteria 10, 14,

15, 17, and 20 of Appendix A to 10 CFR Part 50 for the event of loss-of-offsite

power, the following positions shall be implemented:

Power Supply for Pressurizer Relief and Block Valves and Pressurizer

Level Indicators

(1) Motive and control components of the power-operated relief valves (PORVs)

shall be capable of being supplied from either the offsite power source

or the emergency power source when the offsite power is not available.

(2) Motive and control components associated with the PORV block valves shall

be capable of being supplied from either the offsite power source or the

emergency power source when the offsite power is not available.

(3)- Motive and control power connections to the emergency buses for the PORVs

and their associated block valves shall be through devices that have been

qualified in accordance with safety-grade requirements.

(4) The pressurizer level indication instrument channels shall be powered

from the vital instrument buses. The buses shall have the capability of

being supplied from either the offsite power source or the emergency

power source when offsite power is not available.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

(1) Although the primary concern resulting from lessons learned from the

accident at TMI is that the PORV block valves must be closable, the

design should retain, to the extent practical, the capability to also

open these valves.

(2) The motive and control power for the block-valve should be supplied from

an emergency power bus different from the source supplying the PORV.

(3) Any- changeover of the PORV and block-valve motive and control power from

the normal offsite'power to the emergency onsite power is to be

accomplished manually in the control room.

(4) For those designs in which instrument air is needed for operation, the

electrical power supply should be required to have the capability to be

manually connected to the emergency power sources.

II.G.1-1 3 -119
Applicability

This requirement applies to all PWR operating reactors and all applicants for a

PWR operating license.

Implementation

Implementation is complete for operating reactors. This requirement shall be

implemented by applicants for operating license prior to the issuance of a

fuel-loading license.

Type of Review

No further review for operating reactors is necessary.

Documentation Required

Each applicant shall provide sufficient documentation to support a reasonable

assurance finding by the NRC that each of the positions stated above are met.

The documentation should include, as a minimum, supporting information including

system design description, logic diagrams, electrical schematics, test procedures, and technical specifications.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.1

NUREG-0660, Item II.G.1

NUREG-0694, Part 1

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1980.

3-120 II. G. 1-2
II.K.2.2 CONTROL OF AUXILIARY FEEDWATER INDEPENDENT OF THE INTEGRATED CONTROL

SYSTEM

Position

For Babcock and Wilcox (B&W)-designed reactors, provide procedures and training

to initiate and control auxiliary feedwater independent of the integrated

control system (ICS).

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

No further clarification is required at this time.

Applicability

This requirement applies to all operating license applicants of B&W-designed

reactors.

Implementation

All applicants for operating license should submit documentation 4 months

prior to the expected issuance of the staff safety evaluation report for an

operating license or 4 months prior to the listed implementation date, whichever

is later.

Type of Review

A review is not applicable at this time.

Documentation Required

Applicants shall provide sufficient documentation at least 4 months prior to

the issuance of the staff safety evaluation report for a full power license to

support a reasonable assurance finding by the NRC that the position specified

above has been met.

Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

NUREG-0660, Item II.K.2, Table C.2, Item 2

II.K.2.2-1 3-121
II.K.2.8 AUXILIARY FEEDWATER SYSTEM UPGRADING

Position

All operating Babcock and Wilcox (B&W) plants were ordered to be shut down

shortly after the TMI-2 accident. The orders included both short-term and

long-term actions. The NRR Bulletins and Orders Task Force reviewed the

licensees' compliance with the short-term actions of the orders and issued

safety evaluation reports which served as the basis for plant restart.

Additional items were identified in the review of the long-term actions which

require further work by the licensees.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

The licensees were required to comply with the Commission Orders regarding

certain short-term and long-term auxiliary feedwater system (AFWS) modifications. The staff evaluated the short-term actions, and safety evaluations

were prepared before the plants were allowed to return to operation. The

staff evaluation of the additional (long-term) items will be performed in

conjunction with item II.E.1.1 in NUREG-0660, Auxiliary Feedwater System

Evaluation and item II.E.1.2, AFWS Automatic Initiation and Flow Indicator.

Applicability

This requirement applies to all B&W operating reactors.

Implementation

No separate implementation is required for this item. All AFWS upgrade

modifications for B&W plants are being reviewed as part of Section II.E.l.l

and Section II.E.1.2 in NUREG-0660.

Type of Review

See Section II.E.l.l and Section II.E.1.2 in NUREG-0660.

Documentation Required

See Section II.E.i.l and Section II.E.1.2 in NUREG-0660.

Technical Specification Changes Required

Changes to technical specifications will be made as required.

3-122 II.K.2.8-1
References

NUREG-0645, Volume 1, Section 2.4.6

NUREG-0660, Items II.E.1.l, II.E.1.2, and II.K.2.

II.K.2.8-2 3-123
II.K.2.9 FAILURE MODE EFFECTS ANALYSIS ON THE INTEGRATED CONTROL SYSTEM

Position

For Babcock and Wilcox (B&W)-designed reactors provide a failure-mode-and-effects

analysis (FMEA) of the integrated control system (ICS).

Changes to Previous Requirements and Guidance

There are no changes from those issued in the November 7, 1979 letter from

R. W. Reid, NRC.

Clarification

A generic failure-mode-and-effects analysis of the ICS (BAW-1564) was submitted

on August 17, 1979 by the operating plant licensees. This report was reviewed

by the staff and Oak Ridge National Laboratory (ORNL). Requests for additional

information, regarding the recommendations contained in the report, were sent

to the licensees on November 7, 1979. The responses to the November 7, 1979

letter have been received and are under review.

Applicability

This requirement applies to all B&W operating reactors and operating license

applicants.

Implementation

Operating Reactors--Staff recommendations are pending completion of staff

review.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

Operating Reactors--Documentation has been completed.

Operating License Applicants--B&W applicants should provide the following:

(1) Identify whether the previous generic submittal (BAW-1564) is applicable

to your plant, and

(2) Specify what actions have been taken at your facility to comply with the

recommendations listed in BAW-1564.

3-124 II.K.2.9-1
Technical Specification Changes Required

Changes to technical specifications will be determined following staff review.

References

NUREG-0645, Volume 1, Section 2.4.6

NUREG-0694, Part 2

Commission Orders on B&W Plants

Babcock & Wilcox Co., "Integrated Control System Reliability Analysis," report

BAW-1564.

Letter from R. W. Reid, NRC, to All B&W Operating Plants, dated November 7,

1979.

II.K.2.9-2 3-125
II.K.2.10 SAFETY-GRADE ANTICIPATORY REACTOR TRIP

Position

For Babcock and Wilcox (B&W)-designed reactors, install safety-grade, anticipatory

reactor trip (ART) on loss-of-feedwater and turbine trip.

Changes to Previous Requirements and Guidance

New request for final design submittal. Extension of date for submittal of

design from October 1, 1980 to January 1, 1981.

Clarification

Operating Reactors

(1) IE Bulletin 79-05B, Item 5, issued on April 21, 1979, directed B&W licensees

to provide a design and schedule for implementation of a safety-grade

reactor trip upon:

(a) loss of feedwater;

(b) turbine trip; and

(c) significant reduction in steam generator level.

(2) In accordance with IE Bulletin 79-05B, the B&W.licensees submitted a

conceptual design for a safety-grade, anticipatory.reactor trip which

would be initiated upon turbine trip and loss of feedwater only. Included

in the licensees' responses was a generic evaluation prepared by B&W

which proposed that the anticipatory reactor trip on low steam generator

level was not necessary.

(3) Staff review of these submittals resulted in a preliminary design approval

for the safety-grade anticipatory reactor trip being issued to the B&W

licensees on December 20, 1979. However, the approval letters also

specified the additional information which would be required to be submitted prior to final staff approval of the design.

(4) The staff will complete its review of the generic evaluation by B&W which

indicates that the proposed anticipatory trip on low steam generator

level is unnecessary. Further clarification will be provided bn this

matter, if required, following completion of the staff review.

Operating License Applicants--Compliance with item II.K.1 of NUREG-0694 (C.1.21)

satisfies this requirement.

Applicability

This requirement applies to all B&W operating reactors and applicants for

operating license.

3-126 II. K.2.10-1
Implementation

Operating Reactors--Final design information will be submitted by January 1, 1981.

Safety-grade trip will be installed by July 1, 1981.

Operating License Applicants--Implementation of NUREG-0694, II.K.1 (C.1.21)

prior to the issuance of the fuel load satisfies this requirement.

Type of Review

A postimplementation review will be performed.

Documentation Required

The following information was identified as required by the staff:

(1) The final design submittal should include the final logic diagrams,

electrical schematic diagrams, piping and instrumentation diagrams, and

location layout drawings.

(2) For sensors located in nonseismic areas which have not previously contained reactor protection systems (RPS) inputs, perform and submit an

analysis which shows that the installation (including circuit routing) is

designed such that the effects of credible faults (i.e., grounding,

shorting, application of high voltage, or electromagnetic interference)

or failures in these areas could not be propagated back to the RPS and

degrade the RPS performance or operability.

(3) Submit "Seismic and Environmental Qualification Summary Reports" for the

equipment which have not been previously submitted. In addition, demonstrate that the environmental test conditions bound the actual worst-case

accident conditions expected at the installed locations.

(4) Assure that the anticipatory reactor trip (ART) testability includes

provisions to perform channel functional tests at power. Testing of this

circuitry is to be included in the RPS monthly surveillance tests.

(5) Include in the final design submittal the RPS checkout procedure which

will demonstrate both the operability of the new trip circuitry and the

continued operability of the previous RPS.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0645, Volume 1, Section 2.4.6

NUREG-0694, Item II.K.1 (C.1.21)

II.K.2.10-2 3-127
Commission Orders on B&W Plants

IE Bulletin 79-05B, Item 5, April 21, 1979

Letter from R. W. Reid, NRC, to B&W Licensees, dated December 20, 1979.

3-128 II.K.2.10-3
II.K.2.13 THERMAL MECHANICAL REPORT--EFFECT OF HIGH-PRESSURE INJECTION ON

VESSEL INTEGRITY FOR SMALL-BREAK LOSS-OF-COOLANT ACCIDENT WITH

NO AUXILIARY FEEDWATER

Position

A detailed analysis shall be performed of the thermal-mechanical conditions in

the reactor vessel during recovery from small breaks with an extended loss of

all feedwater.

Changes to Previous Requirements and Guidance

Licensees of Babcock and Wilcox (B&W) operating reactors shall submit the

results of their evaluations by January 1, 1981. The completion schedule has

been changed to allow time to complete the results of the evaluation. Also,

this requirement has been changed to include all operating pressurized-water

reactors (PWRs) and applicants.

Clarification

The position deals with the potential for thermal shock of reactor vessels

resulting from cold safety injection flow. One aspect that bears heavily on

the effects of safety injection flow is the mixing of safety injection water

with reactor coolant in the reactor vessel. B&W provided a report on July 30,

1980 that discussed the mixing question and the basis for a conservative

analysis of the potential for thermal shock to the reactor vessel. Other PWR

vendors are also required to address this issue with regard to recovery from

small breaks with an extended loss of-all feedwater. In particular, demonstration shall be provided that sufficient mixing would occur of the cold highpressure injection (HPI) water with reactor coolant so that significant thermal

shock effects to the vessel are precluded.

Applicability

This requirement applies to all PWR operating reactors and applicants for an

operating license.

Implementation,

Implementation of any modifications will be determined by to the results of

NRC staff review of the report.

Type of Review

A postimplementation review will be performed.

Documentation Required

Licensees of B&W operating reactors shall submit the results of their evaluations by January 1, 1981. Other PWR licensees shall submit the results of

their evaluation by January 1, 1982. Applicants for operating license shall

submit the results of their evaluations at least 6 month5 prior to the issuance

of the staff safety evaluation report for a full-power license.

II. K. 2.13-1 3-129
Technical Specification Changes Required

Changes to technical specifications will be determined following staff review.

References

NUREG-0645, Volume 1, Section 2.4.5

Letter from D. F. Ross Jr., NRC, to All B&W Operating Plants, dated August 21,

1979.

Letter from D. G. Eisenhut, NRC, to All Licensees, dated May 7, 1980.

3-130 3 II.K.2.13-2
II.K.2.15 EFFECTS OF SLUG FLOW ON STEAM GENERATOR TUBES

Position

Although the staff believed that the potential for slug flow was not great in

Babcock and Wilcox (B&W) plants because of the venting path provided by the

internal vent valves, the staff required that a confirmatory evaluation of the

effects of slug flow on steam generator tubes be performed by the licensees to

assure that the tubes could withstand any mechanical loading which could

result from slug flow.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

The request for this information was originally sent to the B&W licensees in a

letter from R. W. Reid, NRC, to all B&W operating plants, dated November 21,

1979.

The results of this analysis have been submitted by the licensees and is

presently undergoing NRC staff review.

Applicability

This requirement applies to all B&W operating reactors and operating license

applicants.

Implementation

Implementation of any modifications will be determined by to the results of

NRC staff review of the evaluation.

Type of Review

A postimplementation review will be performed.

Documentation Required

No additional documentation is required at this time from licensees. Applicants

must supply the requested information at least 4 months before the staff

safety evaluation report for.a full-power license is scheduled to be issued.

Technical Specification Changes Required

Changes to technical specifications will not be required.

II.K.2.15-1 3-131
References

NUREG-0565, Recommendation 2.6.2.1

NUREG-0645, Volume 1, Section 2.4.6

NUREG-0694, Part 2

Letter from R. W. Reid, NRC, to All B&W Operating Plants, dated November 21,

1979.

3-132 II.K.2.15-2
II.K.2.16 REACTOR COOLANT PUMP SEAL DAMAGE

Position,

Evaluate the impact of reactor coolant pump seal damage and leakage due to

loss-of-seal cooling upon loss of offsite power. If damage cannot be precluded,

licensees should provide an analysis of the limiting small-break loss-of-coolant

accident (LOCA) with subsequent reactor coolant pump (RCP) seal damage.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

The request for this information was originally sent to the Babcock and Wilcox

(B&W) licensees in a letter from R. W. Reid, NRC, to all B&W operating plants,

dated November 21, 1979.

The results of these evaluations have been submitted by the licensees and are

presently undergoing NRC staff review.

Applicability

This requirement applies to all B&W operating reactors and operating license

applicants.

Implementation

Implementation of any modifications will be determined by to the results of

NRC staff review of the evaluations.

Type of Review

A postimplementation review will be performed.

Documentation Required

No additional documentation is required at this time from licensees. Applicants

shall submit the requested information at least 4 months before the staff

safety evaluation report for a full-power license is scheduled to be issued.

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0565, Recommendation 2.6.2.f

NUREG-0645, Volume 1, Section 2.4.6

II.K.2.16-1 3-133
NUREG-0694, Part 2

Letter from R. W. Reid, NRC, to All B&W Operating Plants, dated November 21,

1979.

3-1 34 II. K.2.16-2
II.K.2.17 POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS

Position

Analyze the potential for voiding in the reactor coolant system (RCS) during

anticipated transients.

Changes to Previous Requirements and Guidance

The previous requirement has been changed to include all PWR operating reactors

and applicants.

Clarification

The background for this concern and a request for this analysis was originally

sent to the Babcock and Wilcox (B&W) licensees in a letter from R. W. Reid,

NRC, to all B&W operating plants, dated January 9, 1980.

The results of this evaluation have been submitted by the B&W licensees and is

presently undergoing staff review.

Applicability

This requirement applies to all PWR operating reactors and operating license

applicants.

Implementation

Implementation of any modifications will be determined by to the results of

NRC staff review of the licensees/applicants evaluation.

The analysis for all but B&W licenses should be submitted by January 1, 1982

or 6 months before the expected issuance date of the staff safety evaluation

report for the license, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

No additional documentation is required at this time from B&W licensee's. All

others should submit analyses as indicated in "Implementation."

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0660, Item II.K.2 (C.17)

Letter from R. W. Reid, NRC, to All B&W Operating Plants, dated January 9,

1980.

II.K.2.17-1 3-135
II.K.2.19 SEQUENTIAL AUXILIARY FEEDWATER FLOW ANALYSIS

Position

Provide a benchmark analysis of sequential auxiliary feedwater (AFW) flow to

the steam generators following a loss of main feedwater.

Changes to Previous Requirements and Guidance

The previous requirement has been changed to include all operating pressurizedwater reactors (PWRs) and applicants for operating license.

Clarification

This requirement was originally sent to the Babcock and Wilcox (B&W) licensees

in a letter from D. F. Ross, Jr., NRC, to all B&W operating plants, dated

August 21, 1979.

The results of this analysis has been submitted by the B&W licensees and is

presently undergoing staff review.

Applicability

This requirement applies to all PWR operating reactors and applicants for

operating licenses.

Implementation

Implementation of any modifications will be determined by the results of NRC

staff review of this analysis.

The analysis for all but B&W licensees should be submitted by January 1, 1982

or 6 months before the expected issuance date of the staff safety evaluation

report for a license, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required.

No additional documentation is required at this time from B&W licensees. All

others should submit analyses as indicated in "Implementation."

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0645, Volume 1, Section 2.4.6

Letter from D. F. Ross, Jr., NRC, to All B&W Operating Plants, dated August 21,

1979.

3-136 II.K.2.19-1
II.K.2.20 SMALL-BREAK LOSS-OF-COOLANT ACCIDENT WHICH REPRESSURIZES THE

REACTOR COOLANT SYSTEM TO THE POWER-OPERATED RELIEF VALVE SETPOINT

Position

Provide an analysis which shows the plant response to a small-break loss-ofcoolant accident (LOCA) during which the reactor coolant system (RCS) is

repressurized to the power-operated relief valve (PORV) setpoint with subsequent

failure of the PORV to close.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

This requirement was originally sent to the Babcock and Wilcox (B&W) licensees

in a letter from D. F. Ross, Jr., NRC, to all B&W operating plants, dated

August 21, 1979.

The results of this analysis has been submitted by the B&W licensees and is

presently undergoing staff review.

Applicability

This requirement applies to all B&W operating reactors.

Implementation

Implementation of any modifications will be determined by the results of NRC

staff evaluation of this analysis.

Type of Review

A postimplementation review will be performed.

Documentation Required

No additional documentation is required at this time.

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0565, Recommendation 2.6.2.c

NUREG-0645, Volume 1, Section 2.4.6

Letter from 0. F. Ross, Jr., NRC, to All B&W Operating Plants, dated August 21,

1979.

II. K. 2. 20-1 3-137
II.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC POWER-OPERATED RELIEF VALVE

ISOLATION SYSTEM

Position

All PWR licensees should provide a system that uses the PORV block valve to

protect against a small-break loss-of-coolant accident. This system will

automatically cause the block valve to close when the reactor coolant system

pressure decays after the PORV has opened. Justification should be provided

to assure that failure of this system would not decrease overall safety by

aggravating plant transients and accidents.

Each licensee shall perform a confirmatory test of the automatic block valve

closure system following installation.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Implementation of this action item was modified in the May 1980 version of

NUREG-0660. The change delays implementation of this action item until after

the studies specified in TMI Action Plan item II.K.3.2 have been completed, if

such studies confirm that the subject system is necessary.

Applicability

This requirement applies to all PWR operating reactors and applicants for

operating license.

Implementation

If required by action plan item II.K.3.2, licensees shall implement modifications and perform confirmatory tests at the next refueling outage following

staff approval of the design, unless this outage is scheduled within 6 months

of the approval date. In this event, modifications will be completed during

the following refueling outage.

Type of Review

A preimplementation review will be performed.

Documentation Required

If modifications are required, documentation shall include piping and instrumentation diagrams, electrical schematics of design modifications, and an

analysis of conformance to IEEE 279-1971 requirements. Documentation shall be

submitted by July 1, 1981.

Technical Specification Changes Required

Changes to technical specifications will be required.

3-138 II.K.3.1-1
References

NUREG-0565, Recommendation 2.1.2.a

NUREG-0611, Recommendations 3.2.4.e and 3.2.4.f

NUREG-0635, Recommendations 3.2.4.a and 3.2.4.b

NUREG-0660

II.K.3.1-2 3-139
II.K.3.2 REPORT ON OVERALL SAFETY EFFECT OF POWER-OPERATED RELIEF VALVE

ISOLATION SYSTEM

Position

(1) The licensee should submit a report for staff review documenting the

various actions taken to decrease the probability of a small-break lossof-coolant accident (LOCA) caused by a stuck-open power-operated relief

valve (PORV) and show how those actions constitute sufficient'improvements

in reactor safety.

(2) Safety-valve failure rates based on past history of the operating plants

designed by the specific nuclear steam supply system (NSSS) vendor should

be included in the report submitted in response to (1) above.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Based on its review of feedwater transients and small LOCAs for operating

plants, the Bulletins and Orders Task Force in the Office of Nuclear Reactor

Regulation recommended that a report be prepared and submitted for staff

review which documents the various actions that have been taken to reduce the

probability of a small-break LOCA caused by a stuck-open PORV and show how

these actions constitute sufficient improvements in reactor safety. Action

Item II.K.3.2 of NUREG-0660, published in May 1980, changed the implementation

of this recommendation as follows: In addition to modifications already

implemented on PORVs, the report specified above should include safety

examination of an automatic PORV isolation system identified in Task Action

Plan item II.K.3.1.

Modifications to reduce the likelihood of a stuck-open PORV will be considered

sufficient improvements in reactor safety if they reduce the probability of a

small-break LOCA caused by a stuck-open PORV such that it is not a significant

contributor to the probability of a small-break LOCA due to all causes.

(According to WASH-1400, the median probability of a small-break LOCA S2 with

a break diameter between 0.5 in. and 2.0 in. is 10-3 per reactor-year with a

variation ranging from 10-2. to 10-4 per reactor-year.)

The above-specified report should also include *an analysis of safety-valve

failures based on the operating experience of the pressurized-water-reactor

(PWR) vendor designs. The licensee has the option of preparing and submitting

either a plant-specific or a generic report. If a generic report is submitted,

each licensee should document the applicability of the generic report to his

own plant.

Based on the above guidance and clarification, each licensee should perform an

analysis of the probability of a small-break LOCA caused by a stuck-open PORV

or safety valve. This analysis should consider modifications which have been

made since the TMI-2 accident to improve the probability. This analysis shall

evaluate the effect of an automatic PORV isolation system specified in Task

3-140 II. K. 3. 2-1
Action Plan item II.K.3.1. In evaluating the automatic PORV isolation system,

the potential of causing a subsequent stuck-open safety valve and the overall

effect on safety (e.g., effect on other accidents) should be examined.

Actual operational data may be used in this analysis where appropriate. The

bases for any assumptions used should be clearly stated and justified.

The results of the probability analysis should then be used to determine

whether the modifications already implemented have reduced the probability of

a small-break LOCA due to a stuck-open PORV or safety valve a sufficient

amount to satisfy the criterion stated above, or whether the automatic PORV

isolation system specified in Task Action item II.K.3.1 is necessary.

In addition to the analysis described above, the licensee should compile

operational data regarding pressurizer safety valves for PWR vendor designs.

These data should then be used to determine safety-valve failure rates.

The analyses should be documented in a report. If this requirement is implemented on a generic basis, each licensee should review the appropriate generic

report and document its applicability to his own plant(s). The report and the

documentation of applicability (where appropriate) should be submitted for NRC

staff review by the specified date.

Applicability

This requirement applies to all operating PWRs and operating license applicants.

Implementation

The report documenting the specified analyses and the licensee's documentation

of applicability (where appropriate) should be submitted for staff review by

January 1, 1981.

All applicants for operating license should submit documentation 4 months

prior to the expected issuance of the staff safety evaluation report for an

operating.license or 4 months prior to the listed implementation date, whichever

is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

The licensee is to submit for staff review a report on the probability of

small-break LOCA and safety-valve failure rates along with documentation of

applicability (where appropriate) by January 1, 1981.

Technical Specification Changes Required

Changes to technical specifications will not be required.

II.K.3.2-2 3-141
References

WASH-1400 (NUREG-75/014)

NUREG-0565, Recommendation 2.1.2.d

NUREG-0611, Recommendations 3.2.4.g and 3.2.4.i

NUREG-0635, Recommendation 3.2.4.c

NUREG-0660, Items II.K.3.1, II.K.3.2

3-142 II. K. 3.2-3
II.K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT

Position

Tripping of the reactor coolant pumps in case of a loss-of-coolant accident

(LOCA) is not an ideal solution. Licensees should consider other solutions to

the small-break LOCA problem (for example, an increase in safety injection

flow rate). In the meantime, until a better solution is found, the reactor

coolant pumps should be tripped automatically in case of a small-break LOCA.

The signals designated to initiate the pump trip'are discussed in NUREG-0623.

Changes to Previous Requirements and Guidance

Implementation dates are changed to be consistent with test schedule for LOFT

test (L3-6) and to provide for blind posttest analysis.

Clarification

This action item has been revised in the May 1980 version of NUREG-0660 to

provide for continued study of criteria for early reactor coolant pump trip.

Implementation, if any is required, will be delayed accordingly. As part of

the continued study, all holders of approved emergency core cooling (ECC)

models have been required to analyze the forthcoming LOFT test (L3-6). The

capability of the industry models to correctly predict the experimental

behavior of this test will have a strong input on the staff's determination of

when and how the reactor coolant pumps should be tripped.

Applicability

This requirement applies to all PWR operating reactors and operating license

applicants.

Implementation

(1) Document models are to be used for analysis prior to December 3, 1980.

(2) DOE/NRC is to run the LOFT test (L3-6) from December 3, 1980 to December 17,

1980.

(3) NRC will distribute initial conditions approximately 4 weeks after the

test.

(4) Prediction results will be submitted approximately 4 weeks after receipt

of initial conditions.

(5) NRC determination of model acceptability is due April 1, 1981.

(6) Proposed design modifications (if necessary) are due by July 1, 1981.

(7) Modification (if necessary) is due by March 1, 1982.

II.K.3.5-1 3-143
Type of Review

An NRC preimplementation review will be performed (if any modifications are

required).

Documentation Required

Prediction by vendor analysis of LOFT test (L3-6) is required. Additional

information needed will depend upon prediction results.

Technical Specification Changes Required

Changes to technical specifications are to be determined.

References

NUREG-0565, Recommendation 2.3.2.a

NUREG-0611, Recommendation 3.2.2.a

NUREG-0623.

NUREG-0635, Recommendation 3.2.2.a

NUREG-0660

3-144 II.K.3.5-2
II.K.3.7 EVALUATION OF POWER-OPERATED RELIEF VALVE OPENING PROBABILITY

DURING OVERPRESSURE TRANSIENT

Position

Most overpressure transients should not result in the opening of the poweroperated relief valve (PORV). Therefore, licensees should document that the

PORV will open in less than 5% of all anticipated overpressure transients

using the revised setpoints and anticipatory trips for the range of plant

conditions which might occur during a fuel cycle.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Based on its review of best-estimate calculations performed by Babcock and

Wilcox (B&W), the NRC staff believes that the frequency of PORV challenges has

been reduced using the revised PORV and high-pressure reactor trip setpoints

and assuming that the anticipatory reactor trips function as designed. At

this time, however, the staff is unable to make a quantitative judgment of the

expected frequency. Therefore, licensees with B&W-designed plants should

perform additional analyses of anticipated transients which indicate'the

sensitivity of PORV challenges to (1) the variation in core physics parameters

which may occur in the plant cycle; (2) single failures in mitigating systems;

and (3) transients which do not actuate the anticipatory reactor trips.

Analytical assumptions should include those specified in the plant final

safety analysis reports (FSARs). The results of these more-detailed and

extensive analyses should be used to determine the expected frequency of PORV

openings for overpressure transients. This frequency should be less than 5%

of the total number of overpressure transients, thereby confirming the findings

of the staff's review.

The results of this study should be documented and submitted for staff review

by the scheduled date.

Applicability

This requirement applies to all B&W operating reactors.

Implementation

The licensee's report documenting the specified analyses should be submitted

for staff review by January 1, 1981.

Type of Review

A postimplementation review will be performed.

II.K.3.7-1 3-145
Documentation Required

The licensee should perform the specified analyses and submit the documentation

of the results for staff review by the scheduled date.

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0565, Recommendation 2.1.2.b

NUREG-0645, Recommendation 2.4.5, Item 27

Letter from D. F. Ross, Jr., NRC, to All B&W Operating Plants, dated August 21,

1979.

3-146 II.K.3.7-2
II.K.3.9 PROPORTIONAL INTEGRAL DERIVATIVE CONTROLLER MODIFICATION

Position

The Westinghouse-recommended modification to the proportional integral derivative

(PID) controller should be implemented by affected licensees.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

The Westinghouse-recommended modification is to raise the interlock bistable

trip setting to preclude derivative action from opening the power-operated

relief valve (PORV). Some plants have proposed changing the derivative action

setting to zero, thereby eliminating it from consideration. Either modification is acceptable to the staff. This represents a newly available option.

Applicability

This requirement applies to all Westinghouse operating reactors and operating

license applicants.

Implementation

Operating Reactors--For operating reactors, modifications will be completed by

January 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of an operating

license or 4 months prior to the listed implementation date, whichever is

later.

Type-of Review

A postimplementation review will be performed.

Documentation Required

The licensee and applicant shall inform the NRC when the modification has been

completed.

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

JNUREG-0611, Recommendation 3.2.4.b

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II.K.3.9-1 3-147
II.K.3.10 PROPOSED ANTICIPATORY TRIP MODIFICATION

Position

The anticipatory trip modification proposed by some licensees to confine the

range of use to high-power levels should not be made until it has been shown

on a plant-by-plant basis that the probability of a small-break loss-of-coolant

accident (LOCA) resulting from a stuck-open power-operated relief valve (PORV)

is substantially unaffected by the modification.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

This evaluation is required for only those licensees/applicants who propose

the modification.

Applicability

This requirement applies to selected Westinghouse operating reactors and

operating license applicants.

Implementation

Operating Reactors--Completion date for meeting requirements will be dictated

by plant schedule for proposed modification.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A preimplementation review will be performed.

Documentation Required

(1) The licensee is to submit the required analysis and document-proposed

change for staff approval prior to implementation. Documentation is to

be submitted as proposed by the licensee.

(2) Modification schedule is to be determined on a plant-specific basis.

Technical Specification Changes Required

Changes to technical specifications will be required.

3-148 II.K.3.10-1
References

NUREG-0611, Recommendation 3.2.4.c

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II.K.3.10-2 3-149
II.K.3.12 CONFIRM EXISTENCE OF ANTICIPATORY REACTOR TRIP UPON TURBINE TRIP

Position

Licensees with Westinghouse-designed operating plants should confirm that

their plants have an anticipatory reactor trip upon turbine trip. The licensee

of any plant where this trip is not present should provide a conceptual design

and evaluation for the installation of this trip.

Changes to Previous Requirements and Guidance

The date for submittal of design has been extended from July 1, 1980 to January 1,

1981.

Clarification

No further clarification is required.

Applicability

This requirement applies to all Westinghouse operating reactors and operating

license applicants.

Implementation

Operating Reactors--Confirmation or proposed modification is to be completed

by January 1, 1981. The modifications should be completed by the first refueling

outage following staff approval of the design unless this outage is scheduled

within 6 months of the approval date. In this event, modifications will be

completed during the next refueling outage.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evalution report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A preimplementation review will be performed (if design modifications are

required).

Documentation Required

(1) Licensees should submit confirmation of existence of anticipatory reactor

trip upon turbine trip or submit proposed design changes and schedule for

implementation by January 1, 1981.

(2) A commitment to implement modifications should be provided (if required).

(3) All applicants for operating license should submit documentation 4 months

prior to the expected issuance of the staff safety evaluation report for

an operating license or 4 months prior to the listed implementation date,

whichever is later.

3-150 II.K.3.12-1
Technical Specification Changes Required

Changes 'to technical specifications will be required.

References

NUREG-0611, Recommendation 3.2.4.a

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II. K. 3.12-2 3-151
II.K.3.13 SEPARATION OF HIGH-PRESSURE COOLANT INJECTION AND REACTOR CORE

ISOLATION COOLING SYSTEM INITIATION LEVELS--ANALYSIS AND

IMPLEMENTATION

Position

Currently, the reactor core isolation cooling (RCIC) system and the high-pressure

coolant injection (HPCI) system both initiate on the same low-water-level

signal and both isolate on the same high-water-level signal. The HPCI system

will restart on low water level but the RCIC system will not. The RCIC system

is a low-flow system when compared to the HPCI system. The initiation levels

of the HPCI and RCIC system should be separated so that the RCIC system initiates

at a higher water level than the HPCI system. Further, the initiation logic

of the RCIC system should be modified so that the RCIC system will restart on

low water level. These changes have the potential to reduce the number of

challenges to the HPCI system and could result in less stress on the vessel

from cold water injection. Analyses should be performed to evaluate these

changes. The analyses should be submitted to the NRC staff and changes should

be implemented if justified by the analyses.

Changes to Previous Requirements and Guidance

(1) Analysis and proposed modifications are required by Janaury 1, 1981.

(2) Implementation of modifications are required by July 1, 1981 (if applicable).

Clarification

No further clarification is required.

Applicability

This requirement applies to all operating BWRs and operating license applicants

with RCIC and HPCI systems.

Implementation

Analysis and proposed modifications are required by January 1, 1981. Implementation of modifications is required by July 1, 1981 (if applicable)

Type of Review

A preimplementation review will be performed if required.

Documentation Required

(1) The licensee is to provide results of evaluation and proposed modifications

(if necessary) to NRC staff by January 1, 1981. The licensee is to

provide sufficient supporting analysis to demonstrate that the systems,

as modified, would not degrade proper system functions.

3 -152 II. K. 3.13-1
(2) The licensee is to implement modifications (if necessary) by July 1,

1981.

(3) All applicants for operating license should submit documentation 4 months

prior to the expected issuance of the staff safety evaluation report for

an operating license or 4 months prior to the listed implementation date,

whichever is later.

Technical Specification Changes Required

Changes to technical specifications will be required.

Reference

NUREG-0626, Recommendation A.1

II.K.3.13-2 3-153
II.K.3.14 ISOLATION OF ISOLATION CONDENSERS ON HIGH RADIATION

Position

Isolation condensers have radiation monitors on their vents. These monitors

provide alarms in the control room but do not isolate the isolation condenser.

The isolation condensers are currently isolated on a high-radiation signal in

the steam line leading to the isolation condensers. The design should be

modified such that the isolation condensers are automatically isolated upon

receipt of a high-radiation signal at the vent rather than at the steam line.

The purpose of the change is to increase the availability of the isolation

condensers as heat sinks.

Changes to Previous Requirements and Guidance

There has been no change in the requirements for this task action item from

the final recommendations of the Bulletins and Orders (B&O). Task Force. The

schedule has been extended to allow completion of design and procurement.

Clarification

No further clarification is required.

Applicability

This requirement applies to all operating BWRs that have isolation condensers.

Implementation

Design modifications shall be completed by January 1, 1982.

Type of Review

A postimplementation review will be performed.

Documentation Required

The licensee shall inform the NRC when the required design modifications have

been completed.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0626, Recommendation A.2.

Letter from D. G. Eisenhut, NRC, to All Operatiqg Reactor Licensees, dated

May 7, 1980.

3-154 II.K.3.14-1
II.K.3.15 MODIFY BREAK-DETECTION LOGIC TO PREVENT SPURIOUS ISOLATION OF

HIGH-PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING

Position

The high-pressure coolant injection (HPCI) and reactor core isolation cooling

(RCIC) systems use differential pressure sensors on elbow taps in the steam

lines to their turbine drives to detect and isolate pipe breaks in the systems.

The pipe-break-detection circuitry has resulted in spurious isolation of the

HPCI and RCIC systems due to the pressure spike which accompanies startup of

the systems. The pipe-break-detection circuitry should be modified so that

pressure spikes resulting from HPCI and RCIC system initiation will not cause

inadvertent system isolation.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

No further clarification is required.

Applicability

This requirement applies to all operating boiling water reactors (BWRs) and

applicants for operating license with HPCI and RCIC systems.

Implementation

Operating Reactors--For operating reactors, these requirements will be completed

by July 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the-expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

Submit sufficient documentation to support a reasonable assurance finding by

the NRC that the modifications, as implemented, have resulted in satisfying

the concerns expressed in the "Position" statement above.

Technical Specifications Changes Required

Changes to technical specifications will be required.

Reference

NUREG-0626, Recommendation A.3

II.K.3.15-1 3-155
II.K.3.16 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF VALVES--FEASIBILITY

STUDY AND SYSTEM MODIFICATION

Position

The record of relief-valve failures to close for all boiling-water reactors

(BWRs) in the past 3 years of plant operation is approximately 30 in 73 reactoryears (0.41 failures per reactor-year). This has demonstrated that the failure

of a relief valve to close would be the most likely cause of a small-break

loss-of-coolant accident (LOCA). The high failure rate is the result of a

high relief-valve challenge rate and.a relatively high failure rate per

challenge (0.16 failures per challenge). Typically, five valves are challenged

in each event. This results in an equivalent failure rate per challenge of

0.03. The challenge and failure rates can be reduced in the following ways:

(1) Additional anticipatory scram on loss of feedwater,

(2) Revised relief-valve actuation setpoints,

(3) Increased emergency core cooling (ECC) flow,

(4) Lower operating pressures,

(5) Earlier initiation of ECC systems

(6) Heat removal through emergency condensers,

(7) Offset valve setpoints to open fewer valves per challenge,

(8) Installation of additional relief vales with a block- or isolation-valve

feature to eliminate opening of the safety/relief valves (SRVs), consistent

with the ASME Code,

(9) Increasing the high steam line flow setpoint for main steam line isolation

valve (MSIV) closure,

(10) Lowering the pressure setpoint for MSIV closure,

(11) Reducing the testing frequency of the MSIVs,

(12) More-stringent valve leakage criteria, and

(13) Early removal of leaking valves.

An investigation of the feasibility and contraindications of reducing challenges

to the relief valves by use of the aforementioned methods should be conducted.

Other methods should also be included in the feasibility study. Those changes

which are shown to reduce relief-valve challenges without compromising the

performance of the relief valves or other systems should be implemented.

Challenges to the relief valves should be reduced substantially (by an order

of magnitude).

3-156 II.K.3.16-1
Changes to Previous Requirements and Guidance

The schedule for plant modifications has been changed to allow time for staff

review of evaluation and purchase of required hardware.

Clarification

Failure of the power-operated relief valve (PORV) to reclose during the TMI-2

accident resulted in damage to the reactor core. As a consequence, relief

valves in all plants, including BWRs, are being examined with a view toward

their possible role in a small-break LOCA.

The safety/relief valves (SRV) are dual-function pilot-operated relief valves

that use a spring-actuated pilot for the safety function and an external.

air-diaphragm-actuated pilot for the relief function.

The operating history of the SRV has been poor. A new design is used in some

plants but the operational history is too brief to evaluate the effectiveness

of the new design. Another way of improving the performance of the valves is

to reduce the number of challenges to the valves. This may be done by the

methods described above or by other means. The feasibility and contraindications of reducing the number of challenges to the valves by the various methods

should be studied. Those changes which are shown to decrease the number of

challenges without compromising the performance of the valves or other systems

should be implemented.

The failure of an SRV to reclose will be the most probable cause of a smallbreak LOCA. Based on the above guidance and clarification, results of a

detailed evaluation should be submitted to the staff. The licensee shall

document the proposed system changes for staff approval before implementation.

Applicability

This requirement applies to all operating BWRs and BWR operating license

applicants.

Implementation

Results of the evaluation shall be submitted by April 1, 1981 for staff review.

The actual modification shall be accomplished during the next scheduled refueling

outage following staff approval or no later than 1 year following staff approval.

Modification to be implemented should be documented at the time of implementation.

Type of Review

A preimplementation review will be performed.

Documentation Required

By April 1, 1981, licensees must submit the results of the feasibility study

for reducing SRV challenges and propose any necessary modifications for reducing

SRV challenges.

II. K. 3.16-2 3-157
Technical Specification Changes Required

Modification may include testing frequency or leakage criteria which may

require technical specification changes.

Reference

NUREG-0625, Recommendations A-2.8, F-3.4

3-158 II.K.3.16-3
II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE-COOLING SYSTEMS LICENSEE REPORT

AND PROPOSED TECHNICAL SPECIFICATION CHANGES

Position

Several components of the emergency core-cooling (ECC) systems are permitted

by technical specifications to have substantial outage times (e.g., 72 hours

for one diesel-generator; 14 days for the HPCI system). In addition, there

are no cumulative outage time limitations for ECC systems. Licensees should

submit a report detailing outage dates and lengths of outages for all ECC

systems for the last 5 years of operation. The report should also include the

causes of the outages (i.e., controller failure, spurious isolation).

Changes to Previous Requirements and Guidance

This clarification adds the requirement to propose changes that will improve

and control availability.

Clarification

The present technical specifications contain limits on allowable outage times

for ECC systems and components. However, there are no cumulative outage time

limitations on these same systems. It is possible that ECC equipment could

meet present technical specification requirements but have a high unavailability

because of frequent outages within the allowable technical specifications.

The licensees should submit a report detailing outage dates and length of

outages for all ECC systems for the last 5 years of operation, including

causes of the outages. This report will provide the staff with a quantification of historical unreliability due to test and maintenance outages, which

will be used to determine if a need exists for cumulative outage requirements

in the technical specifications.

Based on the above guidance and clarification, a detailed report should be

submitted. The report should contain (1) outage dates and duration of outages;

(2) cause of the outage; (3) ECC systems or components involved in the outage;

and (4) corrective action taken. Test and maintenance outages should be

included in the above listings which are to cover the last 5 years of operation. The licensee should propose changes to improve the availability of ECC

equipment, if needed.

Applicant for an operating license shall establish a plan to meet these requirements.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

Licensees should submit detailed report by January 1, 1981.

II.K.3.17-1 3-159
Applicants for operating license should submit their plan for data collection

in accordance with the review schedule for licensing.

Type of Review

A postimplementation review will be performed.

Documentation Required

(1) Licensees shall submit a report containing the items noted in the above

sections.

(2) Licensees shall submit suggested changes to improve the availability of

ECC equipment, if needed.

Technical Specification Changes Required

Changes depend on results of the licensee study.

References

NUREG-0626, Recommendation A.6

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

3-160 II.K.3.17-2
II. K. 3.18 MODIFICATION OF AUTOMATIC DEPRESSURIZATION SYSTEM LOGIC--FEASIBILITY

FOR INCREASED DIVERSITY FOR SOME EVENT SEQUENCES

Position

The automatic depressurization system (ADS) actuation logic should be modified

to eliminate the need for manual actuation to assure adequate core cooling. A

feasibility and risk assessment study is required to determine the optimum

approach. One possible scheme that should be considered is ADS actuation on

low reactor-vessel water level provided no high-pressure coolant injection

(HPCI) or high-pressure coolant system (HPCS) flow exists and a low-pressure

emergency core cooling (ECC) system is running. This logic would complement,

not replace, the existing ADS actuation logic.

Changes to Previous Requirements and Guidance

The schedule has been changed to accommodate the vendor-projected completion

date and staff review.of a very complex change.

Clarification

No further clarification is required.

Applicability

This requirements applies to all operating boiling-water reactors (BWRs) and

BWR applicants for operating license.

Implementation

Operating Reactors--The feasibility study shall be completed by April 1, 1981.

Proposed modifications shall be submitted by April 1, 1982. The licensee will

implement modifications at the next refueling outage following staff approval

of the design unless this outage is scheduled within 6 months of the approval

date. In this event, modifications will be completed during the following

refueling outage.

Operating License Applicants--All applicants for operating

submit documentation 1 year prior to the expected issuance

license or 1 year prior to the listed implementation date,

license should

of an operating

whichever is later.

Type of Review

A preimplementation review of modifications will be performed.

Documentation Required

Operating

NRC staff

tions for

Reactors--The Licensee shall provide results of feasibility study to

by April 1, 1981. Licensee shall describe the proposed modificastaff approval by April 1, 1982.

Operating License Applicants--Applicants for operating license shall provide

results of feasibility study 1 year prior to issuance of operating license. A

II.K.3.18-1 3-161
description of the proposed modification for staff approval is required 4

months prior to issuance of an operating license.

Technical Specification Changes Required

Changes to technical specifications will be required.

Reference

NUREG-0626, Recommendation A.7

Letter from D. G. Eisenhut, NRC, to All Licensees, dated May 7, 1980.

3-162 II.K.3.18-2
II.K.3.19 INTERLOCK ON RECIRCULATION PUMP LOOPS

Position

Interlocks should be installed on nonjet pump plants (other than Humboldt Bay)

to assure that at least two recirculation loops are open for recirculation

flow for modes other than cold shutdown. This is to assure that the level

measurements in the downcomer region are representative of the level in the

core region.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

No clarification is required.

Applicability

This requirement applies to all operating nonjet-pump boiling-water reactors

(BWRs), except for Humboldt Bay.

Implementation

For operating reactors, these requirements will be completed by July 1, 1981.

Type of Review

A postimplementation review of modifications will be performed.

Documentation Required

Licensees shall submit sufficient documentation by July 1, 1981 to support a

reasonable assurance finding by the NRC that the modifications, as implemented,

have resulted in satisfying the "Position" statement above.

Technical Specification Changes Required

Changes to technical specifications will be required.

Reference

NUREG-0626, Recommendation A.8

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II.K.3.19-1 3-163
II.K.3.20 LOSS OF SERVICE WATER FOR BIG ROCK POINT

Position

The service water system for Big Rock Point has only one cooling train and is

powered from normal alternating current power. The Big Rock Point licensee

should verify the acceptability of the consequences of a loss-of-service-water

supply to the essential plant components in the event of a loss of offsite

power.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Licensee will be required to submit an evaluation showing the acceptability of

the consequences of a loss of service water to the essential plant components

in the event of a los.s of offsite power. The staff will review the licensee's

submittal in order to determine whether plant modifications or procedural

modifications will be required.

Applicability

This requirement applies only to Big Rock Point.

Implementation

For Big Rock Point this requirement will be completed by July 1, 1981.

Type of Review

A postimplementation review will be performed.

Documentation Required

The licensee shall submit an evaluation documenting the acceptability of the

consequences of a loss of service water to the essential plant components in

the event of a loss of offsite power.

Technical Specification Changes Required

Changes to technical specifications may be required, depending upon modifications

(if any).

Reference

NUREG-0626, Recommendation A.9

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

3-164 II.K.3.20-1
II.K.3.21 RESTART OF CORE SPRAY AND LOW-PRESSURE COOLANT-INJECTION SYSTEMS

Position

The core-spray and low-pressure, coolant-injection (LPCI) system flow may be

stopped by the operator. These systems will not restart automatically on loss

of water level if an initiation signal is still present. The core spray and

LPCI system logic should be modified so that these systems will restart, if

required, to assure adequate core cooling. Because this design modification

affects several core-cooling modes under accident conditions, a preliminary

design should be submitted for staff review and approval prior to making the

actual modification.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Modification of system design should be made in accordance with those requirements set forth in Sections 4.12, 4.13, and 4.16 of IEEE Standard 279-1971

with regard to protective function bypasses and completion of protective

action once initiated.

Applicability

This requirement applies to all BWR operating reactors and applicants for BWR

operating license.

Implementation

Operating Reactors--Analysis and proposed design modifications shall be completed

by January 1, 1981. Licensee shall implement modifications at the next refueling

outage following staff approval of the design, unless this outage is scheduled

within 6 months of the approval date. In this event, modifications will be

completed during the following refueling outage.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of an operating

license or 4 months prior to the listed implementation date, whichever is

later.

Type of Review

A preimplementation review will be performed.

Documentation Required

Each licensee or applicant for operating license shall submit proposed design

modifications and supporting analysis that will contain sufficient information

to support a reasonable assurance finding by the NRC that the above position

is met. The documentation should include as a minimum:

II. K.3.21-1 3-165
(1) A discussion of the design with respect to the above paragraphs of

IEEE 279-1971;

(2) Support information including system design description, logic diagrams,

electrical schematics, piping and instrument diagrams, test procedures,

and technical specifications; and

(3) Sufficient documentation to demonstrate that the systems, as modified,

would not degrade proper system functions.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0626, Recommendation A-10

IEEE Standard 279-1971, Section 4.12, 4.13, and 4.16.

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 2980.

3-166 II.K.3.21-2
II.K.3.22 AUTOMATIC SWITCHOVER OF REACTOR .CORE ISOLATION COOLING SYSTEM SUCTION--

VERIFY PROCEDURES AND MODIFY DESIGN

Position

The reactor core isolation cooling (RCIC) system takes suction from the condensate storage tank with manual switchover to the suppression pool when the

condensate storage tank level is low. This switchover should be made automatically.

Until the automatic switchover is implemented, licensees should verify that

clear and cogent procedures exist for the manual. switchover of the RCIC system

suction from the condensate storage tank to the suppression pool.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements issued in the letter of

May 7, 1980.

Clarification

No further clarification is required at this time.

Applicability

This requirement applies to all operating boiling-water reactors (BWRs) and

applicants for operating license with a reactor core isolation cooling (RCIC)

system.

Implementation

Operating Reactors--Procedures shall be verified by January 1, 1981. Design

shall be modified by January 1, 1982.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Verify procedures - January 1, 1981

Modify design - January 1, 1982

Type of Review

A postimplementation review of modifications will be performed.

Documentation Required

Operating Reactors--Licensee shall document procedure verification by January 1,

1981. Licensee shall submit supporting analysis and implemented design changes

by January 1, 1982 and provide sufficient supporting evaluation to demonstrate

that the system, as modified, will not degrade proper system function.

II.K.3.22-1 3-167
Operating License Applicants--Submit appropriate verification in accordance

with the review schedule for licensing.

Technical Specification Changes Required

Changes to technical specifications will be required.

Reference

NUREG-0626, Recommendation B.1

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

3-168 II.K.3.22-2
II.K.3.24 CONFIRM ADEQUACY OF SPACE COOLING FOR HIGH-PRESSURE COOLANT

INJECTION AND REACTOR CORE ISOLATION COOLING SYSTEMS

Position

Long-term operation of the reactor core isolation cooling (RCIC) and highpressure coolant injection (HPCI) systems may require space cooling to maintain

the pump-room temperatures within allowable limits. Licensees should verify

the acceptability of the consequences of a complete loss of alternating-current

power. The RCIC and HPCI systems should be designed to withstand a complete

loss of offsite alternating-current power to their support systems, including

coolers, for at least 2 hours.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

No clarification is required.

Applicability

This requirement applies to all operating boiling-water reactors (BWRs) and

BWR operating license applicants with RCIC and HPCI systems.

Implementation

Operating Reactors--For operating reactors, these requirements will be completed

by January 1, 1982.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A postimplementation review of modifications for operating reactors will be

performed.

Documentation Required

Operating Reactors--Licensee should submit results of verification tests and

modifications (if needed) by January 1, 1982.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

.safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

II. K. 3.24-1 3-169
Applicants for operating licenses shall submit the evaluation and proposals by

January 1, 1982 or no later than 6 months prior to expected issuance of the

staff safety evaluation report in support of license issuance whichever is

later.

Technical Specification Changes Requied

Changes to technical specifications will be required.

Reference

NUREG-0626, Recommendation B.3

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

3-170 II.K.3.24-2
II.K.3.25 EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON PUMP SEALS

Position

The licensees should determine, on a plant-specific basis, by analysis or

experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed to withstand a

complete loss of alternating-current (ac) power for at least 2 hours. Adequacy

of the seal design should be demonstrated.

Changes to Previous Requirements and Guidance

The evaluation and proposed modifications shall be submitted by July 1, 1981.

The May 7, 1980 letter called for modifications by January 1, 1982. This

clarification adds a documentation requirement for the evaluation to be submitted by July 1, 1981. The modification date remains unchanged. Additionally,

this task has changed to include Westinghouse and Combustion Engineering

operating reactors and operating reactor applicants.

Clarification

The intent of this position is to prevent excessive loss of reactor coolant

system (RCS) inventory following an anticipated operational occurrence. Loss

of ac power for this case is construed to be loss of offsite power. If seal

failure is the consequence of loss of cooling water to the reactor coolant

pump (RCP) seal coolers for 2 hours, due to loss of offsite power, one acceptable solution would be to supply emergency power to the component cooling

water pump. This topic is addressed for Babcock and Wilcox (B&W) reactors in

Section II.K.2.16.

Application

This requirement applies to all BWR, Westinghouse and Combustion Engineering

operating reactors and applicants for operating license.

Implementation

For BWR operating reactors the evaluation and proposed modifications shall be

submitted by July 1, 1981 and modifications shall be completed.by January 1,

1982. Westinghouse and Combustion Engineering operating reactors shall submit

the evaluation and proposed modifications by January 1, 1982 and complete

modifications by July 1, 1982.

Type of Review

A preimplementation review of modifications will be performed.

Documentation Required

BWR licensees and Westinghouse and Combustion Engineering licensees shall

provide results of evaluation and proposed modifications by July 1, 1981 and

January 1, 1982, respectively.

II. K. 3.25-1 3-171
Applicants for operating licenses shall submit the evaluation and proposals by

January 1, 1982 or no later than 6 months prior to expected issuance of the

staff safety evaluation report in support of license issuance, whichever is

later.

Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

NUREG-0626, Recommendation B.4

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

3-172 II.K.3.25-2
II.K.3.27 PROVIDE COMMON REFERENCE LEVEL FOR VESSEL LEVEL INSTRUMENTATION

Position

Different reference points of the various reactor vessel water level instruments

may cause operator confusion. Therefore, all level instruments should be

referenced to the same point. Either the bottom of the vessel or the top of

the active fuel are reasonable reference points.

Changes to Previous Requirements and Guidance

The submittal date has been extended from October 1, 1980 to January 1, 1981.

Clarification

No further clarification is required at this time.

Applicability

This requirement applies to all operating BWRs and applicants for operating

license.

Implementation

Operating Reactors--These requirements will be completed by July 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A postimplementation review for operating reactors will be performed.

Documentation Required

Operating Reactors--The licensee shall implement actions and submit documentation of the modifications By January 1, 1981.

Operating License Applicants--All applicants for an operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0626, Recommendation B.6

Letter from 0. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II. K. 3.27-1 3-173
II.K.3.28 VERIFY QUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION

SYSTEM VALVES

Position

Safety analysis reports claim that air or nitrogen accumulators for the automatic depressurization system (ADS) valves are provided with sufficient capacity

to cycle the valves open five times at design pressures. GE has also stated

that the emergency core cooling (ECC) systems are designed to withstand a

hostile environment and still perform their function for 100 days following an

accident. Licensee should verify that the accumulators on the ADS valves meet

these requirements, even considering normal leakage. If this cannot be demonstrated, the licensee must show that the accumulator design is still acceptable.

Changes to Previous Requirements and Guidance

No changes have been made to the previous requirement as specified in the

letter from D. G. Eisenhut dated May 7, 1980 to all operating reactor licensees

and in NUREG-0626.

Clarification

The ADS valves, accumulators, and associated equipment and instrumentation

must be capable of performing their functions during and following exposure to

hostile environments and taking no credit for nonsafety-related equipment or

instrumentation. Additionally, air (or nitrogen) leakage through valves must

be accounted for in order to assure that enough inventory of compressed air is

available to cycle the ADS valves.

Applicability

This requirement applies to all operating BWR plants and all applicants for

operating license.

Implementation

This requirement shall be completed by January 1, 1982.

Type of Review

A review of evaluation results or a postimplementation review of any accumulator

design changes in operating reactors will be performed.

Documentation Required

All operating reactor licensees shall submit evaluation results for staff

review to show that accumulators are qualified and shall implement actions, as

required, by January 1, 1982. All applicants for operating license shall

submit documentation 4 months before the expected issuance of the staff safety

evaluation report for an operating license or 4 months before the listed

implementation date, whichever is later.

3-174 II.K.3.28-1
Technical Specification Changes Required

Changes to technical specifications will be required.

Reference

NUREG-0626, Section A.2-15

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II.K.3.28-2 3-175
II.K.3.29 STUDY TO DEMONSTRATE PERFORMANCE OF ISOLATION CONDENSERS

WITH NONCONDENSIBLES

Position

If natural circulation plays an important role in depressurizing the system

(e.g., in the use of isolation condensers), then the various modes of two-phaseflow natural circulation, including noncondensibles, which may play a significant

role in plant response following a small-break loss-of-coolant accident (LOCA)

should be demonstrated.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Licensees should provide confirmatory verification, using applicable experimental

data, of the analysis models used to calculate the various modes of single and

two-phase natural circulation predicted to occur in their plants during transient

and accident events.

Applicability

This requirement applies to all operating boiling water reactors (BWRs) with

isolation condensers.

Implementation

For operating reactors, these requirements will be completed by April 1, 1981.

Type of Review

A postimplementation review will be performed.

Documentation Required

Licensees shall provide results of evaluation to NRC staff by April 1, 1981.

Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

NUREG-0626, Recommendation B.13

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

3-176 II.K.3.29-1
II.K.3.30 REVISED SMALL-BREAK LOSS-OF-COOLANT-ACCIDENT METHODS TO SHOW

COMPLIANCE WITH 10 CFR PART 50, APPENDIX K

Position

The analysis methods used by nuclear steam supply system (NSSS) vendors and/or

fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for

compliance with Appendix K to 10 CFR Part 50 should be revised, documented,

and submitted for NRC approval. The revisions should account for comparisons

with experimental data, including data from the LOFT Test and Semiscale Test

facilities.

Changes to Previous Requirements and Guidance

The changed requirement (1) allows for justification of acceptability of

present small-break LOCA models by comparison with test date, and (2) requests

each licensee to outline scope and schedule for model revision or comparison

with test data by late fall, 1980. The original requirement did not allow

provision for showing acceptability of present models by comparison with plant

data.

Clarification

As a result of the accident at TMI-2, the Bulletins and Orders Task Force was

formed within the Office of Nuclear Reactor Regulation. This task force was

charged, in part, to review the analytical predictions of feedwater transients

and small-break LOCAs for the purpose of assuring the continued safe operation

of all operating reactors, including a determination of acceptability of

emergency guidelines for operators.

As a result of the task force reviews, a number of concerns were identified

regarding the adequacy of certain features of small-break LOCA models,

particularly the need to confirm specific model features (e.g., condensation

heat transfer rates) against applicable experimental data. These concerns, as

they applied to each light-water reactor (LWR) vendor's models, were documented

in the task force reports for each LWR vendor. In addition to the modeling

concerns identified, the task force also concluded that, in light of the TMI-2

accident, additional systems verification of the small-break LOCA model as

required by II.4 of Appendix K to 10 CFR 50 was needed. This included providing

predictions of Semiscale Test S-07-1OB, LOFT Test (L3-1), and providing experimental verification of the various modes of single-phase and two-phase natural

circulation predicted to occur in each vendor's reactor during small-break

LOCAs.

Based on the cumulative staff requirements for additional small-break LOCA

model verification, including both integral system and separate effects

verification, the staff considered model revision as the appropriate method

for reflecting any potential upgrading of the analysis methods.

The purpose of the verification was to provide the necessary assurance that

the small-break LOCA models were acceptable to calculate the behavior and

consequences of small primary system breaks. The staff believes that this

assurance can alternatively be provided, as appropriate, by additional

II. K.3.30-1 3-177
justification of the acceptability of present small-break LOCA models with

regard to specific staff concerns and recent test data. Such justification

could supplement or supersede the need for model revision.*

The specific staff concerns regarding small-break LOCA models are provided in

the analysis sections of the B&O Task Force reports for each LWR vendor,

(NUREG-0635, -0565, -0626, -0611, and -0623). These concerns should be reviewed

in total by each holder of an approved emergency core cooling system (ECCS)

model and addressed in the evaluation as appropriate.

The recent tests include the entire Semiscale small-break test series and LOFT

Tests (L3-1) and (L3-2). The staff believes that the present small-break LOCA

models can be both qualitatively and quantitatively assessed against these

tests. Other separate effects tests (e.g., ORNL core uncovery tests) and

future tests, as appropriate, should also be factored into this assessment.

Based on the preceding information, a detailed outline of the proposed program

to address this issue should be submitted. In particular, this submittal

should identify (1) which areas of the models, if any, the licensee intends to

upgrade, (2) which areas the licensee intends to address by further justification of acceptability, (3) test data to be used as part of the overall verification/upgrade effort, and (4) the estimated schedule for performing the necessary

work and submitting this information for staff review and approval.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

Detailed outline of the scope and schedule for meeting this requirement should

be submitted by each licensee and applicant by November 15, 1980. This submittal

will form the basis for a meeting with the staff to review and approve the

overall plan. Meetings with the staff to review this submittal are expected

for late fall 1980.

The additional information requested should be submitted by January 1, 1982.

The plant-specific analyses using the revised models should be submitted by

January 1, 1983, or one year after any model revisions are approved.

Type of Review

A postimplementation review of the schedule will be performed. A preimplementation review will be performed by the staff to approve the model and analyses.

*As an example, a model that presently does not properly account for horizontal

countercurrent two-phase flow in the hot leg piping should either be revised

to properly account for the phenomenon, or demonstrated to produce a conservative

result for the entire spectrum of small breaks considered.

3-178 II.K.3.30-2
Documentation Required

(1) Licensees shall submit outline of program for model justification/revision

by November 15, 1980.

(2) Licensees shall submit additional information for model justification

and/or revised analysis model for staff approval by January 1, 1982.

(3) Licensees shall submit their plant-specific analyses using the revised

models by January 1, 1983 or one year after any model revisions are

approved.

(4) Applicants shall submit appropriate information in accordance with the

licensing review schedule.

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0565, Recommendation 2.2.2a

NUREG-0611, Recommendation 3.2.la

NUREG-0623

NUREG-0626, Recommendation A.12

NUREG-0635, Recommendation 3.2.1.a and 3.2.5.a

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II.K.3.30-3 3-179
II.K.3.31 PLANT-SPECIFIC CALCULATIONS TO SHOW COMPLIANCE WITH 10 CFR PART 50.46

Position

Plant-specific calculations using NRC-approved models for small-break loss-ofcoolant accidents (LOCAs) as described in item II.K.3.30 to show compliance

with 10 CFR 50.46 should be submitted for NRC approval by all licensees.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

See "Clarification" for item II.K.3.30.

Applicability

This requirement applies to all operating reactors and applicants for operating

license.

Implementation

Calculations shall be submitted by January 1, 1983 or 1 year after staff

approval of LOCA analysis models, whichever is later, only if model changes

have been made.

Type of Review

A review for conformance with 10 CFR 50.46 limits will be performed.

Documentation Required

Operating Reactors--Licensee shall provide results of evaluation to staff, in

accordance with the schedule as indicated above.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Technical Specification Changes Required

Changes to technical specifications are to be determined.

References

NUREG-0565, Recommendation 2.2.2.b

NUREG-0611, Recommendation 3.2.1.b

3 -180 II.K.3.31-1
NUREG-0626, Recommendations A.13 and B.10

NUREG-0635, Recommendation 3. 2.1. b

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II.K.3.31-2 3-181
II.K.3.44 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO

VERIFY NO FUEL FAILURE

Position

For anticipated transients combined with the worst single failure and assuming

proper operator actions, licensees should demonstrate that the core remains

covered or provide analysis to show that no significant fuel damage results

from core uncovery. Transients which result from a stuck-open relief valve

should be included in this category.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

No further clarification is required at this time.

Applicability

This requirement applies to all operating boiling-water reactors (BWRs) and

BWR license applicants.

Implementation

Operating Reactors--For operating reactors, these requirements will be completed

by January 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior .to the expected issuance of the staff

safety evaluation report for an operating license or 4 months prior to the

listed implementation date, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

Operating Reactors--Licensee shall provide results of evaluation to staff by

January 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of an the staff

safety evaluation report for an operating license or 4 months prior to the listed

implementation date, whichever is later.

Technical Specification Changes Required

Changes to technical specifications will be determined following review of

evaluation.

3-182 II.K.3.44-1
Reference

NUREG-0626, Recommendation A.14

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees, dated

May 7, 1980.

II. K. 3.44-2 3-183
II.K.3.45 EVALUATION OF DEPRESSURIZATION WITH OTHER THAN AUTOMATIC

DEPRESSURIZATION SYSTEM

Position

Analyses to support depressurization modes other than full actuation of the

automatic depressurization system (ADS) (e.g., early blowdown with one or two

safety relief valves (SRVs)) should be provided. Slower depressurization

would reduce the possibility of exceeding vessel integrity limits by rapid

cooldown.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

No further clarification is required at this time.

Applicability

This requirement applies to all operating boiling-water reactors (BWRs) and

BWR license applicants.

Implementation

Operating Reactors--For operating reactors, these requirements will be completed

by January 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of an the staff

safety evaluation report for an operating license or 4 months prior to the listed

implementation date, whichever is later.

Type of Review

A postimplementation review will be performed.

Documentation Required

Operating Reactors--Licensee shall provide results of evaluation to staff by

January 1, 1981.

Operating License Applicants--All applicants for operating license should

submit documentation 4 months prior to the expected issuance of the staff

safety evaluation report for an operating license or 4 months.prior to the

listed implementation date, whichever is later.

Technical Specification Changes Required

Changes to technical specifications will be determined following review of

evaluation.

Reference

NUREG-0626, Recommendation A.15

II. K. 3.45-1 3-184
II.K.3.57 IDENTIFY WATER SOURCES PRIOR TO ACTUATION OF AUTOMATIC

DEPRESSURIZATION SYSTEM

Position

Emergency procedures should include verification that a source of cooling

water, such as the core spray, low-pressure coolant injection (LPCI), or

condensate systems, is available prior to manual actuation of the automatic

depressurization system (ADS). Alternate water sources should be identified

in the procedures, and reference should be made to procedures for startup and

operation of systems that provide these sources. This is being implemented

through the guidelines being developed to assure adequate core cooling.

Changes to Previous Requirements and Guidance

There are no changes to the previous position or requirements.

Clarification

Exceptions to the requirement that a source of cooling water be available

prior to manual actuation of the ADS should be identified and justified.

Symptomatic guidelines have been developed by the BWR owners' group and are

being implemented for trial use of the near-term operating licenses. Implementation of the symptomatic approach for operating reactors will be accomplished

on a schedule compatible with that identified under item I.C.1.

Applicability

This requirement applies to all operating boiling-water reactors.

Implementation

Guidelines have been submitted by the BWR owners' group. Implementation will

be consistent with item I.C.1. No additional licensee action is required

until guidelines are approved by the staff.

Type of Review

A postimplementation review will be performed.

Documentation Required

The BWR owners' group has submitted guidelines that are being reviewed by the

staff.

Technical Specification Changes Required

Changes to technical specifications will not be required.

Reference

NUREG-0626, Recomnmiendation A.5

II.K.3.57-1 3-185
III.A.1.2 UPGRADE EMERGENCY SUPPORT FACILITIES

Additional clarification will be provided in the near future.

3 -186 III .Al-1,2-1
III.A.2 IMPROVING LICENSEE EMERGENCY PREPAREDNESS--LONG-TERM

Position

Each nuclear facility shall upgrade its emergency plans to provide reasonable

assurance that adequate protective measures can and will be taken in the event

of a radiological emergency. Specific criteria to meet this requirement is

delineated in NUREG-0654 (FEMA-REP-1), "Criteria for Preparation and Evaluation

of Radiological Emergency Response Plans and Preparation in Support of Nuclear

Power Plants."

Changes to Previous Requirements and Guidance

The final regulations on emergency planning (45 FR 55401-55413) which become

effective on November 3, 1980, require the submittal and implementation of the

radiological emergency response plans of licensees and state and local entities

within the plume exposure and ingestion emergency planning zones (EPZ) by

January 2, 1981.

NUREG-0654 has been revised to include changes developed from team reviews and

comments obtained during the comment period.

The revised NUREG-0654 establishes the schedule for installation of meteorological equipment to meet a prescribed implementation date (also see proposed

Revision 1 to Regulatory Guide 1.23). The NRC rule establishes July 1, 1981

as the date when the prompt notification capability is to be functional.

Item III.A.1.2 establishes dates when emergency response facilities must be

functional.

Clarification

In accordance with Task Action Plan item III.A.1.1, "Upgrade Emergency Preparedness," each nuclear power facility was required to immediately upgrade its

emergency plans with criteria provided October 10, 1979, as revised by NUREG-0654

(FEMA-REP-1, issued for interim use and comment, January 1980). New plans

were submitted by January 1, 1980, using the October 10, 1979 criteria.

Reviews were started on the upgraded plans using NUREG-0654. Concomitant to

these actions, amendments were developed to 10 CFR Part 50 and Appendix E to

10 CFR Part 50, to provide the long-term implementation requirements. These

new rules were issued in the Federal Register on August 19, 1980, with an

effective date of November 3, 1980. The revised rules delineate requirements

for emergency preparedness at nuclear reactor facilities.

NUREG-0654 (FEMA-REP-1), "Criteria for Preparation and Evaluation of Radiological

Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,"

provides detailed items to be included in the upgraded emergency plans and,

along with the revised rules, provides for meteorological criteria, means for

providing for a prompt notification to the population, and the need for emergency response facilities (see Item III.A.1.2).

Implementation of the new rules levied the requirement for the licensee to

provide procedures implementing the upgraded emergency plans to the NRC for

review. Publication of Revision 1 to NUREG-0654 (FEMA-REP-1) which incorporates

the many public comments received is expected in October 1980. This is the

III.A.2-1 3-187
document that will be used by NRC and FEMA in their evaluation of emergency

plans submitted in accordance with the new NRC rules.

NUREG-0654, Revision 1; NUREG-0696, "Functional Criteria for Emergency

Response Facilities;" and the amendments to 10 CFR Part 50 and Appendix E to

10 CFR Part 50 regarding emergency preparedness, provide more detailed criteria for emergency plans, design, and functional criteria for emergency

response facilities and establishes firm dates for submission of upgraded

emergency plans for installation of prompt notification systems. These

revised criteria and rules supersede previous Commission guidance for the

upgrading of emergency preparedness at nuclear power facilities.

Revision 1 to NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power

Plants," provides meteorological criteria to fulfill, in part, the standard

that "Adequate methods, systems, and equipment for assessing and monitoring

actual or potential offsite consequences of a radiological emergency condition

are in use" (see 10 CFR §50.47). The position in Appendix 2 to NUREG-0654

outlines four essential elements that can be categorized into three functions:

measurements, assessment, and communications.

Proposed Revision 1 to Regulatory Guide 1.23, "Meteorological Measurements

Programs in Support of Nuclear Power Plants," has been adopted to provide

guidance criteria for the primary meteorological measurements program consisting of a primary system and secondary system(s) where necessary, and a

backup system. Data collected from these systems are intended for use in the

assessment of the offsite consequences of a radiological emergency condition.

Appendix 2 to NUREG-0654 delineates two classes of assessment capabilities to

provide input for the evaluation of offsite consequences of a radiological

emergency condition. Both classes of capabilities provide input to decisions

regarding emergency actions. The Class A capability should provide information to determine the necessity for notification, sheltering, evacuation, and,

during the initial phase of a radiological emergency, making confirmatory

radiological measurements. The Class B capability should provide information

regarding the placement of supplemental meteorological monitoring equipment,

and the need to make additional confirmatory radiological measurements. The

Class B capability shall identify the areas of contaminated property and

foodstuff requiring protective measures and may also provide information to

determine the necessity for sheltering and evacuation.

Proposed Revision 1 to Regulatory Guide 1.23 outlines the set of meteorological

measurements that should be accessible from a system that can be interrogated;

the meteorological data should be presented in the prescribed format. The

results of the assessments should be accessible from this system; this information should incorporate human-factors engineering in its display to convey the

essential information to the initial decision makers and subsequent management

team. An integrated system should allow the eventual incorporation of effluent

monitoring and radiological monitoring information with the environmental.

transport to provide direct dose consequence assessments.

3-188 III.A. 2-2
Requirements of the new emergency-preparedness rules under paragraphs 50.47

and 50.54 and the revised Appendix E to Part 50 taken together with NUREG-0654

Revision 1 and NUREG-0696, when approved for issuance, go beyond the previous

requirements for meteorological programs. To provide a realistic time frame

for implementation, a staged schedule has been established with compensating

actions provided for interim measures.

Applicability

This requirement applies to all operating reactors and applicants for operating

licenses.

Implementation

Schedule for Operating Reactors -- For operating reactors the following implementation milestones shall be met to address the four basic elements of the

introduction to Appendix 2 to NUREG-0654.

Milestones are numbered and tagged with the following code; a-date, b-activity,

c-minimum acceptance criteria. They are as follows:

(1) a. January 2, 1981

b. Submittal of radiological emergency response plans

c. A description of the plan to include elements of NUREG-0654, Revision

1, Appendix 2

(2) a. March 1, 1981

b. Submittal of implementing procedures

c. Methods, systems, and equipment to assess and monitor actual or

potential offsite consequences of a radiological emergency condition

shall be provided

(3) a. April 1, 1981

b. Implementation of radiological emergency response plans

c. Four elements of Appendix 2 to NUREG-0654 with the exception of the

Class B model of element 3, or

Alternative to item (3) requiring compensating actions:

A meteorological measurements program which is consistent with the existing

technical specifications as the baseline or an element 1 program and/or

element 2 system of Appendix 2 to NUREG-0654, or two independent element

2 systems shall provide the basic meteorological parameters (wind direction

and speed and an indicator of atmospheric stability) on display in the

control room. An operable dose calculational methodology (DCM) shall be

in use in the control room and at appropriate emergency response facilities.

III.A.2-3 3-189
The following compensating actions shall be taken by the licensee for

this alternative:

(i) if only element 1 or element 2 is in use:

o The licensee (the person who will be responsible for making

offsite dose projections) shall check communications with the

* cognizant National Weather Service (NWS) first order station

and NWS forecasting station on a monthly basis to ensure that

routine meteorological observations and forecasts can be accessed.

° The licensee shall calibrate the meteorological measurements

program at a frequency no less than quarterly and identify a

readily available source of meteorological data (characteristic -

of site conditions) to which they can gain access during calibration

periods.

o During conditions of measurements system unavailability, an

alternate source of meteorological data which is characteristic

of site conditions shall be identified to which the licensee

can gain access.

o The licensee shall maintain a site inspection schedule for

evaluation of the meteorological measurements program at a

frequency no less than weekly.

° It shall be a reportable occurrence if the meteorological data

unavailability exceeds the goals outlined in Proposed Revision

1 to Regulatory Guide 1.23 on a quarterly basis.

(ii) The portion of the DCM relating to the transport and diffusion

of gaseous effluents shall be consistent with the characteristics

of the Class A model outlined in element 3 of Appendix 2 to

NUREG-0654.

(iii) Direct telephone access to the individual responsible for

making offsite dose projections (Appendix E to 10 CFR Part

50(IV)(A)(4)) shall be-available to the NRC in the event of a

radiological emergency. Procedures for establishing contact

and identification of contact individuals shall be provided as

part of the implementing procedures.

This alternative shall notbe exercised after July 1, 1982. Further, by

July 1, 1981, a functional description of the upgraded programs (four elements)

and schedule for installation and full operational capability shall be provided

(see milestones 4 and 5).

(4) a. March 1, 1982

b. Installation of Emergency Response Facility hardware and software

c. Four elements of Appendix 2 to NUREG-0654, with exception of the

Class B model of element 3.

3-190 III.A.2-4
(5) a. July 1, 1982

b. Full operational capability of milestone 4.

C. The Class A model (designed to be used out to the plume exposure

EPZ) may be used in lieu of a Class B model out to the ingestion

EPZ. Compensating actions to be taken for extending the application

of the Class A model out to the ingestion EPZ include access to

supplemental information (meso and synoptic scale) to apply judgment

regarding intermediate and long-range transport estimates. The

distribution of meteorological information by the licensee should be

as follows by July 1, 1982:

NRC and Emergency

Meteorological Response OrganizaInformation CR TSC EOF tions

Basic Met. Data X X X X (NRC)

(e.g., 1.97 Parameters)

Full Met. Data X X X

(1.23 Parameters)

DCM (for Dose X X X X

Projections)

Class A Model (to X X X X

Plume Exposure EPZ)

Class B Model or X X X

Class A Model

(to Ingestion EPZ)

(6) a. July 1, 1982 or at the time

whichever is sooner..

of the completion of milestone 5,

b. Mandatory review of the DCM by the licensee

c. Any DCM in use should be reviewed to ensure consistency with the

operational Class A model. Thus, actions recommended during the

initial phases of a radiological emergency would be consistent with

those after the TSC and EOF are activated.

(7) a. September 1, 1982

b. Description of the Class B model provided to the NRC

c. Documentation of the technical bases and justification for selection

of the type Class B model by the licensee with a discussion of the

site-specific attributes.

(8) a. June 1, 1983

b. Full operational capability of the Class B model

III.A.2-5 3-191
c. Class B model of element 3 of Appendix 2 to NUREG-0654, Revision 1

Schedule for Near-Term Operating Licenses--For applicants for an operating

license, at least milestones 1, 2, and 3 shall be met prior to the issuance of

an operating license. Subsequent milestones shall be met by the same dates

indicated for operating reactors. for the alternative to milestone 3, the

meteorological measurements program shall be consistent with the NUREG-75/087,

"Standard Review Plan for the Review of Safety Analysis Reports for Nuclear

Power Plants," Section 2.3.3 program as the baseline or element 1 and/or

element 2 systems..

Type of Review

A postimplementation review will be performed for the April 1, 1981 requirement.

Documentation Required

Complete updated emergency plans shall be provided by January 2, 1981 and

complete implementing procedures shall be submitted by March 1, 1981.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-75/087

NUREG-0654 (FEMA-REP-1), Revision 1

NUREG-0696

Regulatory Guide 1.23, Proposed Revision 1

3-192 III. A. 2-6
III.D.1.1 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN

RADIOACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND

BOILING-WATER REACTORS,

Position

Applicants shall implement a program to reduce leakage from systems outside

containment that would or could contain highly radioactive fluids during a

serious transient or accident to as-low-as-practical levels. This program shall include the following:

(1) Immediate leak reduction

(a) Implement all practical leak reduction measures for all systems that

could carry radioactive fluid outside of containment.

(b) Measure actual leakage rates with system in operation and report

them to the NRC.

(2) Continuing Leak Reduction -- Establish and implement a program of preventive

maintenance to reduce leakage to as-low-as-practical levels. This program

shall include periodic integrated leak tests at intervals not to exceed each

refueling cycle.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Applicants shall provide a summary description, together with initial leak-test

results, of their program to reduce leakage from systems outside containment

that would or could contain primary coolant or other highly radioactive fluids

or gases during or following a serious transient or accident.

(1) Systems that should be leak tested are as follows (any other plant

system which has similar functions or postaccident characteristics even

though not specified herein, should be included):

Residual heat removal (RHR)

Containment spray recirculation

High-pressure injection recirculation

Containment and primary coolant sampling

Reactor core isolation cooling

Makeup and letdown (PWRs only)

III.D.l.1-1 3-193
Waste gas (includes headers and cover gas system outside of containment in

addition to decay or storage system)

Include a list of systems containing radioactive materials which are

excluded from program and provide justification for exclusion.

(2) Testing of gaseous systems should include helium leak detection or equivalent testing methods.

(3) Should consider program to reduce leakage potential release paths due to

design and operator deficiencies as discussed in our letter to all operating

nuclear power plants regarding North Anna and related incidents, dated

October 17, 1979.

Applicability

This requirement applies to all operating license applicants. Operating reactors

satisfied the requirements of NUREG-0578, Recommendation 2.1.6a (Systems Integrity),

by providing information required by January 1, 1980.

Implementation

This requirement shall be implemented by applicants for operating license prior

to issuance of a full-power license.

Documentation Required

Applicants shall submit the information requested in the "Clarification" section

of this position at least 4 months prior to issuance of a fuel-loading license.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0578, Recommendation 2.1.6.a

NUREG-0660, Item III.D.1.1

NUREG-0694, Part 2

Letter-from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

October 17, 1979.

3-194 III. D. 1.1-2
III.D.3.3 IMPROVED INPLANT IODINE INSTRUMENTATION UNDER ACCIDENT CONDITIONS

Position

(1) Each licensee shall provide equipment and associated training and procedures

for accurately determining the airborne iodine concentration in areas

within the facility where plant personnel may be present during an accident.

(2) Each applicant for a fuel-loading license to be issued prior to January 1,

1981 shall provide the equipment, training, and procedures necessary to

accurately determine the presence of airborne radioiodine in areas within

the plant where plant personnel may be present during an accident.

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

Effective monitoring of increasing iodine levels in the buildings under accident

conditions must include the use of portable instruments using sample media

that will collect iodine selectively over xenon (e.g., silver zeolite) for the

following reasons:

(1) The physical size of the auxiliary and/or fuel handling building precludes

locating stationary monitoring instrumentation at all areas where airborne

iodine concentration data might be required.

(2) Unanticipated isolated "hot spots" may occur in locations where no stationary

monitoring instrumentation is located.

(3) Unexpectedly high background radiation levels near stationary monitoring

instrumentation after an accident may interfere with filter radiation

readings.

(4) The time required to retrieve samples after an accident may result in

high personnel exposures if these filters are located in high-dose-rate

areas.

After January 1, 1981, each applicant and licensee shall have the capability

to remove the sampling cartridge to a low-background, low-contamination area

for further analysis. Normally, counting rooms in auxiliary buildings will

not have sufficiently low backgrounds for such analyses following an accident.

In the low background area, the sample should first be purged of any entrapped

noble gases using nitrogen gas or clean air free of noble gases. The licensee

shall have the capability to measure accurately the iodine concentrations

present on these samples under accident conditions. There should be sufficient

samplers to sample all vital areas.

For applicants with fuel-loading dates prior to January 1, 1981, provide by

fuel loading (until January 1, 1981) the capability to accurately detect the

presence of iodine in the region of interest following an accident. This can

III. D. 3.3-1 3-195
be accomplished by using a portable or cart-mounted iodine sampler with attached

single-channel analyzer (SCA). The SCA window should be calibrated to the

365 KeV of iodine-131 using theSCA. This will give an initial conservative

estimate of presence of iodine and can be used to determine if respiratory

protection is required. Care must be taken to assure that the counting system

is not saturated as a result of too much activity collected on the sampling

cartridge.

APPLICABILITY

This requirement applies to all operating reactors and all applicants for an

operating license.

IMPLEMENTATION

Applicants for fuel-loading license prior to January 1, 1981 shall meet position 2

prior to fuel loading. Licensees and applicants shall meet position 1 by

January 1, 1981, or prior to licensing, whichever is later.

-Type of Review

A postimplementation review will be performed.

Documentation Required

For applicants for an operating license, provide a description of the in-plant

airborne radioiodine sampling and analysis systems specifying the number and

types of samplers, sample media, sample flushing methods, and sample analysis

equipment type and location.

Technical Specification Changes Required

Changes to technical specifications will not be required.

References

NUREG-0578 Recommendation 2.1.8.c

NUREG-0660, Item III.D.3.3

Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants, dated

September 13, 1979.

Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants, dated

October 30, 1979.

3-196 III.D.3.3-2
III. D. 3.4 CONTROL-ROOM HABITABILITY REQUIREMENTS

Position

In accordance with Task Action Plan item III.0.3.4 and control room habitability,

licensees shall assure that control room operators will be adequately protected

against the effects of accidental release of toxic and radioactive gases and

that the nuclear power plant can be safely operated or shut down under design

basis accident conditions (Criterion 19, "Control Room," of Appendix A, "General

Design Criteria for Nuclear Power Plants," to 10 CFR Part 50).

Changes to Previous Requirements and Guidance

There are no changes to the previous requirements.

Clarification

(1) All licensees must make a submittal to the NRC regardless of whether or

not they met the criteria of the referenced Standard Review Plans (SRP)

sections. The new clarification specifies that licensees that meet the

criteria of the SRPs should provide the basis for their conclusion that

SRP 6.4 requirements are met. Licensees may establish this basis by

referencing past submittals to the NRC and/or providing new or additional

information to supplement past submittals.

(2) All licensees with control rooms that meet the criteria of the following

sections of the Standard Review Plan:

2.2.1-2.2.2

2.2.3

6.4

Identification of Potential Hazards in Site Vicinity

Evaluation of Potential Accidents;

Habitability Systems

shall report their findings regarding the specific SRP sections as explained

below. The following documents should be used for guidance:

(a) Regulatory Guide 1.78, "Assumptions for Evaluating the Habitability

of Regulatory Power Plant Control Room During a Postulated Hazardous

Chemical Release";

(b) Regulatory Guide 1.95,

Room Operators Against

"Protection of Nuclear Power Plant Control

an Accident Chlorine Release"; and,

(c) K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room

Ventilation System Design for Meeting General Design Criterion 19,"

13th AEC Air Cleaning Conference, August 1974.

Licensees shall submit the results of their findings as well as the basis

for those findings by January 1, 1981. In providing the basis for the

habitability finding, licensees may reference their past submittals.

Licensees should, however, ensure that these submittals reflect the

current facility design and that the information requested in Attachment 1

is provided.

III. D. 3.4-1 3-197
(3) All licensees with control rooms that do not meet the criteria of the

above-listed references, Standard Review Plans, Regulatory Guides, and

other references.

These licensees shall perform the necessary evaluations and identify appropriate

modifications.

Each licensee submittal shall include the results of the analyses of control

room concentrations from postulated accidental release of toxic gases and

control room operator radiation exposures from airborne radioactive material

and direct radiation resulting from design-basis accidents. The toxic gas

accident analysis should be performed for all potential hazardous chemical

releases occurring either on the site or within 5 miles of the plant-site

boundary. Regulatory Guide 1.78 lists the chemicals most commonly encountered

in the evaluation of control room habitability but is not all inclusive.

The design-basis-accident (DBA) radiation source term should be for the loss-ofcoolant accident LOCA containment leakage and engineered safety feature (ESF)

leakage contribution outside containment as described in Appendix A and B of

Standard Review Plan Chapter 15.6.5. In addition, boiling-water reactor (BWR)

facility evaluations should add any leakage from the main steam isolation

valves (MSIV) (i. e., valve-stem leakage, valve seat leakage, main steam

isolation valve leakage control system release) to the containment leakage and

ESF leakage following a LOCA. This should not be construed as altering the

staff recommendations in Section D of Regulatory Guide 1.96 (Rev. 2) regarding

MSIV leakage-control systems. Other DBAs should be reviewed to determine

whether they might constitute a more-severe control-room hazard than the LOCA.

In addition to the accident-analysis results, which should either identify the

possible need for control-room modifications or provide assurance that the

habitability systems will operate under all postulated conditions to permit

the control-room operators to remain in the control room to take appropriate

actions required by General Design Criterion 19, the licensee should submit

sufficient information needed for an independent evaluation of the adequacy of

the habitability systems. Attachment 1 lists the information that should be

provided along with the licensee's evaluation.

Applicability

This requirement applies to all operating reactors and operating license

applicants.

Implementation

Licensees shall submit their responses to this request on or before January 1,

1981. Applicants for operating licenses shall submit their responses prior to

issuance of a full-power license. Modifications needed for compliance with

the control-room habitability requirements specified in this letter should be

identified, and a schedule for completion of the modifications should be

provided. Implementation of such modifications should be started without

awaiting the results of the staff review. Additional needed modifications, if

any, identified by the staff during its review will be specified to licensees

3-198 III. D. 3.4-2
Type of Review

A postimplementation review will be performed.

Documentation Required

By January 1, 1981 licensees shall provide the information described in

Attachment 1. Applicants for an operating license shall submit their responses

prior to full-power licensing.

Technical Specification Changes Required

Changes to technical specifications will be required.

References

NUREG-0660, Item III.D.3.4.

Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees,- dated

May 7, 1980.

III.D.3.4-3 3-199
III.D.3.4, ATTACHMENT 1, INFORMATION REQUIRED FOR CONTROL-ROOM

HABITABILITY EVALUATION

(1) Control-room mode of operation, i.e., pressurization and filter

recirculation for radiological accident isolation or chlorine release

(2) Control-room characteristics

(a) air volume control room

(b) control-room emergency zone (control room, critical files, kitchen,

washroom, computer room, etc.)

(c) control-room ventilation system schematic with normal and emergency

air-flow rates

(d) infiltration leakage rate

(e) high efficiency particulate air (HEPA) filter and charcoal adsorber

efficiencies

(f) closest distance between containment and air intake

(g) layout of control room, air intakes, containment building, and

chlorine, or other chemical storage facility with dimensions

(h) control-room shielding including radiation streaming from

penetrations, doors, ducts, stairways, etc.

(i) automatic isolation capability-damper closing time, damper leakage

and area

(j) chlorine detectors or toxic gas (local or remote)

(k) self-contained breathing apparatus availability (number)

(1) bottled air supply (hours supply)

Cm) emergency food and potable water supply (how many days and how many

people)

(n) control-room personnel capacity (normal and emergency)

(o) potassium iodide drug supply

(3) Onsite storage of chlorine and other hazardous chemicals

(a) total amount and size of container

(b) closest distance from control-room air intake

3-200 III.D.3.4-4
(4) Offsite manufacturing, storage, or transportation facilities of hazardous

chemicals

(a) identify facilities within a 5-mile radius;

(b) distance from control room

(c) quantity of hazardous chemicals in one container

(d) frequency of hazardous chemical transportation traffic (truck, rail,

and barge)

(5) Technical specifications (refer to standard technical specifications)

(a) chlorine detection system

(b) control-room emergency filtration system including the capability to

maintain the control-room pressurization at 1/8-in. water gauge,

verification of isolation by test signals and damper closure times,

and filter testing requirements.

3-201 III. D.3.4-5
APPENDIX A

KEY TO REFERENCES

The final paragraph of each clarification item lists the reference materials

related to that item. Those listed as NUREG-XXXX are NRC documents available

for purchase from: GPO Sales Program, U.S. Nuclear Regulatory Commission,

Washington, D.C. 20555 and the National Technical Information Service,

Springfield, Virginia 22161. They are also available for inspection and

copying for a fee in the NRC Public Document Room at 1717 H Street, N.W.,

Washington, D. C. to avoid frequent repetition within this document, the NUREG

reports are listed only by number. A complete list with title and date of

publication follows:

WASH-1400 (NUREG-75/014), "Reactor Safety Study - An Assessment of Accident

Risks in U.S. Commercial Nuclear Power Plants," Executive Summary, Main

Report, Appendices 1-11, U.S. Nuclear Regulatory Commission, December 1975.

NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports

for Nuclear Power P-lants - LWR Edition," U.S. Nuclear Regulatory Commission,

1975 (available only from the National Technical Information Service,

Springfield, Virginia 22161).

NUREG-0565, "Staff Report on the Generic Evaluation of Small-Break Loss-ofCoolant Accident Behavior for Babcock and Wilcox Operating Plants," U.S.

Nuclear Regulatory Commission, January 1980.

NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term

Recommendations," U.S. Nuclear Regulatory Commission, July 1979.

NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report," U.S. Nuclear

Regulatory Commission, August 1979.

NUREG-0588, "Interim Staff Position on Environmental Qualification of SafetyRelated Electrical Equipment," U.S. Nuclear Regulatory Commission,

December 1979.

NUREG-0611, "Generic Evaluation of Feedwater Transients and Small-Break Loss-ofCoolant Accidents in Westinghouse Designed Operating Plants," U.S. Nuclear

Regulatory Commission, January 1980.

NUREG-0623, "Generic Assessment of Delayed Reactor Coolant Pump Trip During

Small-Break Loss-of-Coolant Accidents in Pressurized Water Reactors," U.S.

Nuclear Regulatory Commission, November 1979.

NUREG-0625, "Report of the Siting Policy Task Force," U.S. Nuclear Regulatory

Commission, August 1979.

A-1
NUREG-0626, "Staff Report on the Generic Assessment of Feedwater Transients

and Small-Break Loss-of-Coolant Accidents in Boiling Water Reactors Designed

by the General Electric Company," U.S. Nuclear Regulatory Commission,

January 1980.

NUREG-0635, "Generic Assessment of Small-Break Loss-of-Coolant Accidents in

Combustion Engineering Designed Operating Plants," U.S. Nuclear Regulatory

Commission, January 1980.

NUREG-0645, "Final Report of Bulletins and Orders Task Force of the Office of

Nuclear Reactor Regulation," Vols. 1 and 2, U.S. Nuclear Regulatory

Commission, January 1980.

NUREG-0654 (FEMA-REP-1), "Criteria for Preparation and Evaluation of Radiological

Emergency Response Plans and Preparedness in Support of Nuclear Power

Plants," U.S. Nuclear Regulatory Commission, January 1980.

NUREG-0660, Vols. 1 and 2, "NRC Action Plan Developed as a Result of the TMI-2

Accident," U.S.'Nuclear Regulatory Commission, May 1980; Revision 1,

August 1980.

NUREG-0667, "Transient Response of Babcock & Wilcox Designed Reactors," U.S.

Nuclear Regulatory Commission, May 1980.

NUREG-0694, "TMI-Related Requirements for New Operating Licenses," U.S. Nuclear

Regulatory Commission, June 1980.

NUREG-0696 (draft), "Functional Criteria for Emergency Response Facilities,"

U.S. Nuclear Regulatory Commission, July 1980.

NUREG-0700, "Guidelines for the Design Review of Nuclear Power Plant Control

Rooms," U.S. Nuclear Regulatory Commission, to be published.

NUREG-CR-1580 (draft), "Human Engineering Guide for Control Room Evaluation,"

Essex Corporation, July 1980.

The following NRC letters are available for inspection and copy for a fee in

the NRC Public Document Room at 1717 H Street, N.W., Washington, D. C.:

8/21/79 Letter from D. F. Ross, Jr., NRC, to All B&W Operating Plants, dated

August 21, 1979, Subject: Identification and Resolution of Long-Term

Generic Issues Related to the Commission Orders of May 1979.

9/13/79 Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power

Plants, dated September 13, 1979, Subject: Followup Actions

Resulting from the NRC Staff Reviews Regarding the Three Mile Island

Unit 2 Accident.

9/27/79 Letter from D. B. Vassallo, NRC, to All Pending Operating License

Applicants, dated September 27, 1979, Subject: Followup Actions

Resulting from the NRC Staff Reviews Regarding the Three Mile Island

Unit 2 Accident.

A-2
10/10/79 Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,

dated October 10, 1979, Subject: Emergency Planning.

10/17/79 Letter from D. G. Eisenhut, NRC, to All Operating Nuclear Power Plants,

dated October 17, 1979, Subject: Radioactive Release at North Anna

Unit 1 and Lessons Learned.

10/30/79 Letter from H. R. Denton, NRC, to All Operating Nuclear Power Plants,

dated October 30, 1979, Subject: Discussion of Lessons Learned

Short-Term Requirements.

11/7/79 Letter from R. W. Reid, NRC, to All B&W Operating Plants, dated

Novemeber 7, 1979, Subject: Request for Additional Information -

BAW Report 1564, "Integrated Control System Reliability Analysis."

7/31/80 Letter from D. G. Eisenhut, NRC, to All Licensees and Applicants,

dated July 31, 1980, Subject: Interim Criteria for Shift Staffing.

11/9/79 Letter from D. B. Vassallo, NRC, to All Pending Operating License

Applicants, dated November 9, 1979, Subject: Discussion of Lessons

Learned Short-Term Requirements.

11/21/79 Letter from R. W. Reid, NRC, to All B&W Operating Plants, dated

November 21, 1979, Subject: Request for Additional Information on

Small-Break Loss-of-Coolant Accident.

12/20/79 Letter from R. W. Reid, NRC, to All B&W Licensees, dated December 20,

1979, Subject: Preliminary Design Approval for the Safety-Grade

Anticipatory Reactor Trip (ART) on Loss-of-Feedwater and Turbine

Trip.

1/9/80 Letter from R. W. Reid, NRC to All B&W Operating Plants, dated

January 9, 1980, Subject: Concern for Voiding During Transients on

B&W Plants.

3/10/80 Letter from D. F. Ross, Jr., NRC, to All Pending W and C-E License

Applicants, dated March 10, 1980, Subject: Actions Required from

Operating License Applicants of Nuclear Steam Supply Systems Designed

by W and C-E Resulting from the NRC Bulletins and Orders Task Force

Review Regarding TMI-2 Accident.

3/28/80 Letter from H. R. Denton, NRC, to All Power Reactor Applicants and

Licensees, dated March 28, 1980, Subject: Qualifications of Reactor

Operators.

4/24/80 Letter from D. F. Ross, Jr., NRC, to All Pending B&W License Applicants,

dated April 24, 1980, Subject: Actions Required from Operating

License Applicants of Nuclear Steam Supply Systems Designed by B&W

Resulting from the NRC Bulletins and Orders Task Force Review Regarding

TMI-2 Accident.

A-3
4/25/80 Letter from D. G. Eisenhut, NRC, to All Power Reactor Licensees,

dated April 25, 1980, Subject: Clarification of NRC Site Requirements

for Emergency Response Facilities at Each Site.

5/7/80 Letter from D. G. Eisenhut, NRC, to All Operating Reactor Licensees,

dated May 7, 1980, Subject: Five Additional TMI-2 Related Requirements

to Operating Reactors.

Documents with the following types of designation and other miscellaneous

documents are available for inspection and copying for a fee in the NRC Public

Document Room at 1717 H Street., N.W., Washington, D.C.:

Commission Order (CLI-80-21)

NRC Regulation (45 FR 55401-55413)

Inspection and Enforcement documents

Regulatory Guides

Standard Review Plan

Technical Specifications

Branch Technical Position

Staff Interim Position

Other documents that are national technical standards are available for inspection

from public technical libraries:

ANSI Standards

IEEE Standards

Code of Federal Regulations

A-4
APPENDIX B

DESIGN AND QUALIFICATION CRITERIA FOR ACCIDENT MONITORING INSTRUMENTATION

Applicability

To the extent feasible and practical (in conformance with the stipulations of

Appendix A and ancillary requirements), equipment is to be installed by the

specified implementation dates. Where equipment is unavailable, precluding

conformance with equipment qualification and schedular requirements, the

implementation dates are to be met by installation of best available equipment.

In such cases, deviations are to be described and a schedule for the feasible

installation of equipment in conformance with the stipulations of Regulatory

Guide 1.97 (when the guide is used) is to be provided. I

Appendix A is consistent with our current draft version of Regulatory

Guide 1.97. We expect no further revisions to our requirements.

Criteria

(1) The instrumentation should be environmentally qualified in accordance

with Regulatory Guide 1.89 (NUREG-0588). Qualification applies to the

complete instrumentation channel from sensor to display where the display

is a direct-indicating meter or recording device. Where the instrumentation channel signal is to be used in a computer-based display, recording

and/or diagnostic program, qualification applies to and includes the

channel isolation device. The location of the isolation device should be

such that it would be accessible for maintenance during accident conditions.

The seismic portion of environmental qualification should be in accordance

with Regulatory Guide 1.100. The instrumentation should continue to read

within the required accuracy following, but not necessarily during, a

safe shutdown earthquake. Instrumentation, whose ranges are required to

extend beyond those ranges calculated in the most severe design basis

accident event for a given variable, should be qualified using the following guidance.

The qualification environment shall be based on the design basis accident

events, except the assumed maximum of the value-of the monitored variable

shall be the value equal to the maximum range for the variable. The

monitored variable shall be assumed to approach this peak by extrapolating

the most severe initial ramp associated with the design basis accident

events. The decay for this variable shall be considered proportional to

the decay for this variable associated with the design basis accident

events. No additional qualificaton margin needs to be added to the

extended range variable. All environmental envelopes except that pertaining to the variable measured by-the information display channel shall

be those associated with the design basis accident events.

The above environmental qualification requirement does not account for

steady-state elevated levels that may occur in other environmental parameters associated with the extended range variables. For example, a

sensor measuring containment pressure must be qualified for the measured

process variable range, but the corresponding ambient temperature is not

mechanistically linked to that pressure. Rather, the ambient temperature

value is the bounding value for design basis accident events analyzed in

B-1
Chapter 15 of the final saety analysis report (FSAR). The extended range

requirement is to ensure that the equipment will continue to provide

information should conditions degrade beyond those postulated in the

safety analysis. Since variable ranges are nonmechanistically determined,

extension of associated parameter levels is not justifiable and has,

therefore, not been required.

(2) No single failure within either the accident-monitoring instrumentation,

its auxiliary supporting features or its power sources concurrent with

the failure that are a condition or result of a specific accident should

prevent the operator from being presented the information necessary for

him to determine the safety status of the plant and to bring the plant to

a safe condition and maintain it in a safe condition following that

accident. Where failure of one accident-montoring channel results in

ambiguity (that is, the redundant displays disagree) which could lead the

operator to defeat or fail to accomplish a required safety function,

additional information should be provided to allow the operator to deduce

the actual conditions in the plant. This may be accomplished by: (a) prov.iding additional independent channels of information of the same variable

(addition of an identical channel), or (b) providing an independent

channel which monitors a different variable bearing a known relationship

to the multiple channels (addition of a diverse channel), or (c) providing

the capability, if sufficient time is available, for the operator to

perturb the measured variable and determine which channel has failed by

observation of. the response on each instrumentation channel. Redundant

or diverse channels should be electrically independent, energized from

station Class lE power source, and physically separated in accordance

with Regulatory Guide 1.75 up to and including any isolation device. At

least one channel should be displayed on a direct-indicating or recording

device. (NOTE: Within each redundant division of a safety system,

redundant monitoring channels are not required.)

(3) The instrumentation should be energized from station Class lE power

sources.

(4) An instrumentation channel should be available prior to an accident

except as provided in Paragraph 4.11, "Exemption," as defined in IEEE Std

279 or as specified in technical specifications.

(5) The recommendations of the following regulatory guides pertaining to

quality assurance should be followed:

1.28 "Quality Assurance Program Requirements (Design.&

Construction)

1.30 "Quality Assurance Requirements for the Installation,

Inspection, and Testing of Instrumentation and Electric

Equipment"

1.38 "Quality Assurance Requirements for Packaging, Shipping,

Receiving, Storage, and Handling of Items for Water-Cooled

Nuclear Power Plants"

B-2
1.58 "Qualification of Nuclear Power Plant Inspection,

Examination, and Testing Personnel"

1.64 "Quality Assurance Requirements for the Design of Nuclear

Power Plants"

1.74 "Quality Assurance Terms and Definitions"

1.88 "Collection, Storage, and Maintenance of Nuclear Power

Plant Quality Assurance Records"

1.123 "Quality Assurance Requirements for Control of Procurement

of Items and Services for Nuclear Power Plants"

1.144 "Auditing of Quality Assurance Programs for Nuclear Power

Plants"

Task RS 810-5 "Qualification of Quality Assurance Program Audit Personnel

for Nuclear Power Plants" (Guide number to be inserted.)

Reference to the above regulatory guides (except Regulatory Guides 1.30

and 1.38) are being made pending issuance of a regulatory guide endorsing

NQA-1 (Task RS 002-5), now in progress.

(6) Continuous indication (it may be by recording) display should be provided

at all times. Where two or more instruments are needed to cover a particular

range, overlapping of instrument span should be provided.

(7) Recording of instrumentation readout information should be provided.

Where trend or transient information is essential for operator information

or action, the recording should be analog stripchart or stored and displayed continuously on demand. Intermittent displays, such as data

loggers and scanning recorders, may be used if no significant transient

response information is likely to be lost by such devices.

(8) The instruments should be specifically identified on the control panels

so that the operator can easily discern that they are intended for use

under accident conditions.

(9) The transmission of signals from the instrument or associated sensors for

other use should be through isolation devices that are designated as part

of monitoring instrumentation and that meet the provisions of the document.

(10) Means should be provided for checking, with a high degree of confidence,

the operational availability of each monitoring channel, including its

input sensor, during reactor operation. This may be accomplished in

various ways; for example:

(a) By perturbing the monitored variable

(b) By introducing and varying, as appropriate, a substitute input to

the sensor of the same nature as the measured variable

B-3
(c) By cross-checking between channels that bear a known relationship to

each other and that have readouts available.

(11) Servicing, testing, and calibrating programs should be specified to

maintain the capability of the monitoring instrumentation. For those

instruments where the required interval between testing will be less than

the normal time interval between generating station shutdowns, a capability

for testing during power operation should be provided.

(12) Whenever means for removing channels from service are included in the

design, the design should facilitate administrative conrol of the access

to such removal means.

(13) The design should facilitate administrative control of the access to all

setpoint adjustments, module calibration adjustments, and test points.

(14) The monitoring instrumentation design should minimize the development of

conditions that would cause meters, annunciators, recorders, alarms,

etc., to give anomalous indications potentially confusing to the operator.

(15) The instrumentation should be designed to facilitate the recognition,

location, replacement, repair, or adjustment of malfunctioning components

or modules.

(16) To the extent practical, monitoring instrumentaton inputs should be from

sensors that directly measure the desired variables.

(17) To the extent practical, the same instruments should be used for accident

monitoring as are used for the normal operations of the plant to enable

the operator to use, during accident situations, instruments with which

the operator is most familiar. However, where the required range of

monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments should be

used.

(18) Periodic testing should be in accordance with the applicable portions of

Regulatory Guide 1.118 pertaining to testing of instruments channels.

B-4
APPENDIX C

N U C L E A R P O W E R

P L A N T S H I F T T E C H N I C A L A D V I S O R

Recommendations for Position Description,

Qualifications, Education and Training

THE

INSTITUTE OF NUCLEAR

POWER OPERATIONS

Revision 0

April 30, 1980

C-1
FOREWORD

The Shift Technical Advisor position is generally accepted

by the industry and the NRC as being an interim position. Long

range criteria (three to five years) require that the qualifications of shift supervisors and senior operators be upgraded

with the shift supervisor required to have an engineering

degree or equivalent qualifications.

In developing recommendations for the STA position and

giving consideration to the current shortage of qualified

engineering graduates to fill the interim positions, the working

groups attempted to identify those areas of education and levels

of experience considered necessary to effectively accomplish

the position's most important function - accident assessment.

Recognizing that many engineering or scientific degree programs

do not normally include the range and depth of technical subjects

required for accident assessment, the recommendations included

identify the subject areas and depth of study necessary but do

not specify through what programs they should be acquired.

The user is cautioned to ensure that the recommended

education and training is conducted in a professional manner by

competent instructors and at the proper level. Institutions

and programs accredited by recognized agencies such as ECPD/ABET

or others ensure that adequate standards are met.

The program identified should provide the technical depth

necessary to meet long-term qualification requirements of both

the Senior Reactor Operator and the Shift Supervisor at the

time when the STA position is eliminated. Since the shift

supervisor position normally is involved in a broader range of

i Rev. 0

C-3
managerial responsibilities, additional training in nontechnical

subjects such as technical writing, oral communication, and

decision making is recommended.

Development of the technical and language skills at the

level recommended along with the applied fundamentals and

practical training recommended is considered an acceptable

equivalent to an engineering degree insofar as qualifications

for Shift Supervisor are concerned.

ii Rev. 0

C-4
TABLE OF CONTEINTS

Section Page

1. Definitions ............................................. 1

2. Introduction ....................................... 4

3. Objective ..................................... 6

4. Position Description ..... ..................... 7

5. General Education and-Experience .............. 10

5.1 Education and Training .................... 10

5.2 Experience ..... ........................... 10

5.3 Absences from STA Duties .................. 11

6. Education and Training Requirements ........... 12

6.1 Education ........................................ ... 12

6.1.1 Prerequisites Beyond High

School Diploma .. ........ .......... 12

6.1.2 College Level Fundamental

Education ........ *. ........... 13

6.2 Applied Fundamentals - Plant Specific ..... 15

6.3 Management/Supervisory Skills . ............ 15

6.4 Plant Systems .... .... ........... .. 16

6.5 Administrative Controls .................. 16

6.6 General Operating Procedures ............. 17

6.7 Transient/Accident Analysis

and Emergency-Procedures .. ............... 17

6.8 Simulator Training ....................... 17

6.9 Annual Requalification Program ............ 19

iii Rev. 0

C-5
1. DEFINITIONS

The definitions given below are of a restricted

nature for the purpose of these recommendations.

Academic Training - Successfully completed college-level

work which may or may not lead to a recognized degree in

a discipline related to the position.

ExDerience - Applicable work in design, construction,

preoperational and startup testing activities, operation,

maintenance, or technical services. Observation of others

performing these functions shall not be considered acceptable

experience.

Licensed Operator - Any individual who possesses an

operator's license pursuant to Title 10, Code of Federal

Regulations, Part 55, "Operators' Licenses".

Licensed Senior Operator - Any individual who possesses

a senior operator's license pursuant to 10 CFR Part 55.

Manager of Nuclear Power - The individual in the utility

organization who is directly responsible for the operation

of that utility's nuclear power plants and will usually

be the person to whom the Plant Manager reports.

Nuclear Power Plant - Any plant using a nuclear reactor to

produce electric power, process steam or space heating.

-1- Rev. 0

C-7
Nuclear Power Plant Experience- Experience acquired in the

preoperational and startup testing activities or operation

of nuclear power plants. Experience in design, construction, maintenance, and instructing may be considered

applicable nuclear power plant experience and should be

evaluated on a case-by-case basis.

(1) Experience acquired at military or production

nuclear plants may qualify as equivalent nuclear

power plant experience.

(2) Nuclear power plant systemis and operations training

(classroom, on-the-job or simulator) may qualify as

nuclear power plant experience if it applies to the

plant at which the position is to be filled or a

similar plant.

Nuclear Reactor - Any assembly of fissionable material which

is designed to achieve a controlled, self-sustaining neutron

chain reaction.

On-The-Job Training - Participation in nuclear power plant

startup, operation, maintenance, or technical services- under

the direction of experienced personnel.

Related Technical Training - Formal training beyond the high

school level in technical subjects associated with the

position in question, such as acquired in training schools

or programs conducted by the military, industry, utilities,

universities, vocational schools, or others. Such traininZ

programs shall be of a scheduled and planned length and

include text material and lectures.

Shall, Should and May - The word "shall" is used to denote

a requirement; the word "should" to denote a recommendation;

and the word "may" to denote permission - neither a requirement

nor a recommendation.

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C-8
STA - Shift Technical Advisor - That position at a nuclear

power plant established to evaluate plant conditions and

provide advice to the Shift St)pervisor during plant

transients and accidents. Inherent in this function is

the detection and reporting of potential safety problems.

Utility (Owner Organization) - The organization, including

the on-site operating organization, which has overall

legal, financial and technical responsibility for the

operation of one or more nuclear power plants. This

shall include contracted personnel (vendors, consultants,

etc.).

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C-9
2. IN;TRODCCTION

After the the accident at Three Mile Island, investigations

by several committees and the Nuclear Regulatory Commission

concluded that certain deficiencies may exist in the level

of technical expertise generally available to the shift

operating staff prior to, during, and immediately after an

accident or severe plant transient. Although adequate

expertise may be available some time later, the lack of

skilled analytical capability during such occurances may

contribute to equipment damage or danger to the plant staff

and the public. Subsequent recommendations and regulations

require that additional technical expertise be made available

to each operating shift. Current regulatory requirements

identify those individuals providing this expertise on

shift as Shift Technical Advisors (STAs).

The purpose of this document is to describe the position

and identify specific areas of formal education, plant

training and experience necessary to assure an advanced

level of analytical ability on shift. These recommendations

will provide a level of technical ability that is essential

to improved operational safety and are consistent with

regulatory requirements. This Institute position was developed

in conjunction with representatives of utilities, equipment

vendors and engineering educators, giving consideration tospecific contributions the function must make to shift

operations.

For convenience, the necessary contributions are

identified in the form of a position description. Although

this format suggests that the function will be performed

by a new position, it is not intended to pre-empt management's

prerogative to accomplish the function through other qualified

individuals within an existing organizational structure.

-4- Rev. 0

C-10
It should be noted that the areas of formal education

identified are not normally included in any one course

or in the courses for any one established engineering or

related scientific degree program. Rather, the areas

and depth of study are those needed to effectively perform

the function. The areas identified do provide a basis

for either exempting certain subject areas for qualified

engineering graduates or for establishing developmental

programs for non-graduates or graduates of a degree

program that does not include the requisite subject

areas.

-5- Rev. 0

C-11
3. OBJECTIVE

The objective of creating the STA position is to

improve the quality of plant technical management

and operation by providing additional on-shift expertise in the area of operational safety, thus reducing

the probability of abnormal or emergency condition

occurrences and mitigating the consequences of

these conditions if they do occur.

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C-12
4. POSITION DESCRIPTION

The responsibilities of the Shift Technical Advisor

should be documented in such a way that the incumbent

clearly understands the duties and responsibilities of

the position. The following position description is a

suitable method for describing the work to be performed

and the measures of incumbent performance.

Function

Provide advanced technical assistance to the operating

shift complement during normal and abnormal operating

conditions.

General Qualifications

(1) That combination of educating, training and nuclear

plant experience identified in Sections 5 and 6.

(2) An in-depth understanding of nuclear plant equipment,

systems and operating practices and procedures.

(3) Well developed analytical skills and the ability to

make sound judgements under stressful conditions.

General Duties

(1) During assigned tour of duty be cognizant of plant

and equipment status.

(2) Maintain independence from normal plant operations

as necessary to make objective evaluations of plant

operations and to advise or assist plant supervision

in correcting conditions that may compromise the

safety of operations.

(3) Be readily available to provide appropriate assistance

to the normal shift complement.

7- Rev. 0

C-13
Typical Responsibilities

(1) During transients and accidents, compare existing

critical narameters,j(i.e. neutron power level;

reactor coolant systemf level, pressure and temperature;

containment pressure, temperature, humidity and

radiation level; and plant radiation levels) with

those predicted in the Plant Transient and Accident

Analysis, to ascertain whether the plant is responding

to the incident as predicted.

Report any abnormalities to the Shift Supervisor

immediately and provide assistance in formulating a

plan for appropriate corrective action.

(2) Make a qualitative assessment of plant parameters

during and following an accident in order to ascertain

whether core damage has occurred.

(3) During emergencies be observant of critical parameters,

ascertain that there is adequate core cooling including

availability of a heat sink for the coolant system,

and, in the event that critical parameters become

unavailable due to instrument failure, perform

calculations or through other means determine

approximate values for the parameters in question.

(4) Investigate the cause(s) of abnormal or unusual

events occurring on assigned shift and assess any

adverse affects therefrom. Recommend changes to

procedures or equipment as necessary to prevent

recurrence.

(5) Evaluate the effectiveness of plant procedures in

terms of terminating or mitigating accidents and

make recommendations to the Shift Supervisor when

changes are needed.

(6) Assist the operations staff in interpreting and

applying the requirements of Technical Specifications.

(7) Perform an early review of the planned activities

for the upcoming shift to ascertain whether special

considerations or precautions are warranted and

make appropriate recommendations to the Shift Supervisor.

This review should include scheduled surveillance

tests and major maintenance items.

-8- Rev. 0

C-14
(8) Evaluate effectiveness of plant instructions and

recc.-nmend needed changes to the appropriate Supervisor.

(9) Evaluate core power distribution during and following

load changes. Perform hot channel factor and/or

rod program analyses as required.

(10) Review abnormal and emergency procedures.

(11) Prepare special reports when requested by the Operations Superintendent.

(12) Provide an engineering evaluation of Licensee Event

Reports from other plants as assigned.

Accountability

The STA is accountable for the following end results:

(1) Contributes to maximizing safety of operations by

independently observing plant status and advising

shift supervision of conditions that could compromise

plant safety.

(2) Contributes to maximizing plant safety during transient

or accident situations by independently assessing

plant conditions and by providing the technical

assistance necessary to mitigate the incident and

minimize the effect on personnel, the environment,

and plant equipment.

-9- Rev. O

C-15
5. GENERAL EDUCATION AND EXPERIENCE

5.1 EDUCATION AND TRAINING

The Shift Technical Advisor shall meet the

education and training requirements of Section 6.

5.2 EXPERIENCE

The Shift Technical Advisor shall have a minimum

of 16 months of nuclear power plant experience., at

least two months of which shall be at an operating

nuclear plant.

A maximum of six months of this experience may

be obtained in the military or at a production

nuclear plant and should be evaluated on a case-bycase basis.

A maximum of three months of systems and operations training may be applied toward these experience

requirements.

At least 12 months of this experience shall be

at the station at which the position is to be filled.

This may be waived in part when two essentially

identical plants are involved.

Experience gained at a nuclear station prior to

initial fuel loading is acceptable, if the individual

actively participates in preparation and review of

plant procedures and test programs, and is on-site

for at least one year during the preoperational

test phase.

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C-16
5.3 ABSENCES FROM STA DUTIES

Persons not actively performing the STA

functions for a period of thirty (30) days or longer

shall, prior to assuming responsibilities of the

position, as a minimum receive training sufficient

to ensure he is cognizant of facility/procedure

changes that occured during his absence.

Persons not performing the STA function for a

period of six (6) months or longer shall, prior to

assuming the responsibilities of the position,

receive the annual requalification training described

in this document.

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C-17
6. EDUCATIO1N AN4D TRAINING REQUIREMENTS

A waiver for any of the required education or training

shall be granted only by the Manager of Nuclear Power and

should be evaluated on a case-by-case basis. Such waivers

may be considered when a candidate has documented accredited

college courses or can demonstrate an acceptable level of

knowledge through comprehensive examinations in the area to

be waived.

For courses completed at an accredited college, a

semester credit hour shall be considered equivalent to

approximately 15 contact hours in a full-time training

program.

.When courses prescribed in Sections 6.1.2 and 6.2 are

not administered by an accredited college or university the

curriculum and instructor shall be certified by the INPO.

6.1 EDUCATION

6.1.1 Prerequisites Beyond High School Diploma It is

assumed that many candidates may have received

previous training and are qualified to begin

the coursework prescribed in 6.1.2. Prerequisite

education considered necessary for successful.

completion of the advanced coursework is

identified below. This coursework may be

waived without formal documentation of specific

course completion.

Contact Hours

Mathematics

Trigonometry, Analytical 90

Geometry, College Algebra

Chemistry

Inorganic Chemistry 30

-12- Rev. 0

C-18
Physics

Engineering Physics (heat, 150

mechanics, likht sound,

electricity and magnetism)

TOTAL 270

6.1.2 College Level Fundamental Education

Contact Hours

Mathematics 90

Engineering mathematics through

the introduction to ordinary

differential equations and the

utilization of Laplace

transforms to interpret control

response.

Reactor Theory 100

Atomic and Nuclear Physics

Statics, through 2-group

Diffusion Theory

Dynamics, Point Kinetics,

Reactivity Feedback

Reactor Chemistry 30

Inorganic Chemistry (as related

to reactor systems)

Corrosion - Reaction Rates

Nuclear Materials 40

Strength of Materials

Reactor Material Properties

(phase diagrams, fuel densification)

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C-19
Ther.mal Sciences (for nuclear systems) 1 20

Thermodynamics

Laws of Thermodynamics

Properties of Water and Steam

Steam Cycles and Efficiency

Fluid Dynamics

Bernoulli's Equation

Fluid Friction and Head Loss

Elevation Head

Pump and System Characteristics

Two Phase Flow

Heat Transfer

Methods of Heat Transfer

Boiling Heat Transfer

Heat Exchangers

Electrical ScienceS 60

Electronics (Circuit theory, digital

electronics)

Motors, Generators, Transformers,

Switchgear

Instrumentation and Control Theory

Nuclear Instrumentation and Control 40

Radiation Detectors

Reactor. Instrumentation

Reactivity Control and Feedback

Nuclear Radiation Protection and

Health Physics 40

Biological Effects

Radiation Survey Instrumentation

Shielding

TOTAL 520

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C-20
6.2 APPLIED FUNDAMENTALS - PLANT SPECIFIC

In addition to the general education require

ments described in Section 6.1, all STAs shall complete

the following training at the college level tailored

to the specific plant at which the STA is assigned or

a plant of similar design. It may be presented

separately from or may be integrated with the education

described in Section 6.1.

Subject/Topics Contact Hours

Plant Specific Reactor Technology

(including core physics data)

Plant Chemistry and Corrosion Control

Reactor Instrumentation and Control

Reactor Plant Materials

Reactor Plant Thermal Cycle

TOTAL 120

6.3 MANAGEMENT/SUPERVISORY SKILLS

Subject

Leadership

Interpersonal Communication

Motivation of Personnel

Problem and Decisional Analysis

Command Responsibilities and Limits

Stress

Human Behavior

TOTAL

Contact Hours

40

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C-21
6.4 PLA.NT SYSTEMS

The training program shall cover the following

systems along with others considered necessary for

a specific plant.

System Contact Hours

Emergency Core Cooling

frEmergency Cooling Water

Emergency Electrical Power, AC and DC

Reactor Protection.

Reactor Coolant

Reactor Coolant Inventory and Chemistry Control

Containment System (including Containment Cooling)

Closed Cooling Water

Nuclear Instrumentation

Non-Nuclear Instrumentation

Reactor Control

Containment Hydrogen Monitoring and Control

Radioactive Waste Disposal (liquid, gas, solid)

Emergency Control Air

Condensate and Main Feedwater

Auxiliary Feedwater

Steam Generator Level Control (PWR)

Reactor Vessel Water Level Control (BWR)

Main Steam

Loose Parts Monitoring (PWR)

Status Monitoring (including Process Computer)

Seismic Monitoring

Residual Heat Removal

Radiation Monitoring

Plant Ventilation

Main Turbine and Generator

TOTAL 200

6.5 ADMINISTRATIVE CONTROLS

Subject Contact Hours

Responsibilities for Safe Operation and Shutdown

Equipment Outages and Clearance Procedures

Use of Procedures

Plant Modifications

Shift Relief Turnover and Manning

Containment Access

Maintaining Cognizance of Plant Status

Unit Interface Controls (multi-unit plants with

one or more units still under construction)

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C-22
Physical Security

Control Room Access

Duties and Responsibilities of the STA

Radiological Emergency Plan

Code of Federal Regulations (appropriate sections)

Plant Technical Specifications (including bases)

Radiological Control Instructions

TOTAL 80

6.6 GENERAL OPERATING PROCEDURES

Subject Contact Hours

Startup

At Power Operations

Shutdown

Xenon Following While on Standby

ECP and S.D. Margin Calculation

TOTAL 30

6.7 TRANSIENT/ACCIDENT ANALYSIS AND EMERGENCY PROCEDURES

Subject Contact Hours

Transient and Accident Analyses

Plant Abnormal and Emergency Procedures

TOTAL 30

6.8 SIMULATOR TRAINING

The plant evolutions, transients and events

listed below shall be conducted along with any

others deemed necessary. The primary objective

should be to demonstrate plant and operator

response to a given condition or event and not

necessarily to develop the control manipulation

expertise of the trainee. The trainee/ instructor

ratio should not exceed 4:1.

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C-23
Simulator exercises should be preceeded by a

period of discussion of the planned exercises addressing

expected response of the plant and applicable plant

procedures to be used. Approximately 100 contact

hours are required with about 50 hours in the classroom

and 50 hours on the simulator.

Following each exercise demonstrating a transient

or emergency event, an incident critique discussion

should be held to enhance the trainees' understanding

of that particular exercise. When the simulator is

not plant-specific, the training shall be tailored to

the specific plant as much as practical.

PWR Simulator Exercises

Reactor and Plant Startup

Load Changes at Power

Shutdown to Cold Condition

Demonstration of Steam Generator Level Manual Control

Load Rejections of Greater than 10%

Failure of Rod Control System

Failure of Automatic Steam Generator Level Controls

Failure of Pressurizer Level and Pressure Automatic

Controls

Turbine Trip from Full Power

Reactor Trip from Full Power

Loss of Normal Feedwater at Full Power

Failure Open of Power Operated Relief Valve

Stuck Open Pressurizer Safety Valve

Loss of Reactor Coolant Pumps at Full Power

and Demonstration of Natural Circulation

Failure Open of One or More Turbine Bypass Valves

while at a) Full Power, b) Hot Standby

Loss of All Feedwater (normal and emergency)

Loss of Reactor Coolant (small and DBA)

Steam Generator Tube Rupture (small and large)

Loss of RHR Shutdown Cooling with the RCS

Temperature 2000 to 3000 F

Inadvertent Safety Injection While at Power

Loss of Offsite Electrical Power

Loss of One Train of Onsite Electrical Power

-18- Rev. 0

C-24
BWR Simulator Exercises

Reactor and Plant Startup

Load Changes at Power (using flow control

when applicable)

Shutdown

Load Rejection of Greater than 10%

Turbine Trip from Full Power.

Turbine Bypass Valve Failure to Open Following Trip

Inadvertent Isolation of MSIV's While at Power

Reactor Scram from Full Power

Reactor Pressure Control Failure

Dropped Control Rod While at Power

Cold Water Transient at Power

Inadvertent Opening of Relief Valve

Loss of Main Feedwater Pumps at Power

Inadvertent Start of Idle Recirculation Pump

Inadvertent Trip of Recirculation Pump(s)

Loss of Reactor Coolant (small break - large break)

Steam Line Break (inside-outside containment)

Loss of Offsite Power

Loss of Shutdown Cooling with RCS Temperature

2000 - 300OF

Demonstration of Natural Circulation Capabilities

Malfunction of Reactor Water Level Automatic Controls

6.9 ANNUAL REQUALIFICATION.TRAINING

Subject Material Hours Required

Review of transient and accident

analyses of FSAR condition III and IV

events emphasizing the individual's

role in accident assessment. Review

selected industry events and LERs

that could have led to more serious

incidents.

Simulator exercises related to

the transients in Section 6.8

conducted so as to emphasize

the-role of the STA.

40 (Lecture)

40 (Simulator)

TOTAL 80

-19- Rev. 0

C-25
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (Assigned by DOC)

(7-77) NRG03

BIBLIOGRAPHIC DATA SHEET NUREG-0737

4. TITLE AND SUBTITLE (Add Volume No., if appropriate) 2. (Leave blank)

Clarification of TMI Action Plan Requirements 3. RECIPIENT'S ACCESSION NO.

7. AUTHOR(S) 5. DATE REPORT COMPLETED

MONTH |YEAR

October 1980

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

Office of Nuclear Reactor Regulation 'Wiber I M

Division of Licensing

U.S. Nuclear Regulatory Commission 6. [Leave blank)

Washington, D.C. 20555

8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (include Zip Code) 10. PROJECT/TASK/WORK UNIT NO.

Office of Nuclear Reactor Regulation

Division of Licensing 11. CONTRACT NO.

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

13. TYPE OF REPORT | PERIOD COVERED (Inclusive dates)

Technical Report I

15. SUPPLEMENTARY NOTES 14. (Leave blank)

16. ABSTRACT (200 words or less)

This document, NUREG-0737, is a letter from D.G. Eisenhut, Director of the Division

of Licensing, NRR, to licensees of operating power reactors and applicants for

operating licenses forwarding post-TMI requirements which have been approved for

implementation. Following the accident at Three Mile Island Unit 2, the NRC staff

developed the Action Plan, NUREG-0660, to provide a comprehensive and integrated plan

to improve safety at power reactors. Specific items from NUREG-0660 have been approved

by the Commission for implementation at reactors. In this NRC report,.these specific.

items comprise a single document which includes additional information about schedules,

applicability, method of implementation review, submittal dates, and clarification of

technical positions. It should be noted that the total set of TMI-related actions

have been collected in NUREG-0660, but only those items that the Commission has

approved for implementation to date are included in this document, NUREG-0737.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17b. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAI LAB I LITY STATEMENT 19ie. -9fie TS~y4D~fhisreport 21. NO. OF PAGES

.20. I SCU nc FaTYSCLJ-tjThis Sf paye 22. PRICE

N RC FORM 335 (7-77)