ML20212A622

From kanterella
Revision as of 18:58, 5 May 2021 by StriderTol (talk | contribs) (StriderTol Bot change)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Summary of 861209 Meeting W/Util in Bethesda,Md Re Mgt Briefing on Util Progress & Accomplishments During 1985. List of Attendees & Viewgraphs Encl
ML20212A622
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 12/17/1986
From: Vissing G
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8612240278
Download: ML20212A622 (96)
v • d • e


Text

g -

s s December 17, 1986 Docket No. 50-313 LICENSEE: ARKANSAS POWER AND LIGHT COMPANY FOR ARKANSAS NUCLEAR ONE, UNITS 1 & 2

SUBJECT:

SUMMARY

OF MEETING OF DECEMBER 9, 1986, WITH AP&L PRESENTING A MANAGEMENT BRIEFING ADDRESSING THE PROGRESS AND ACCOMPLISHMENTS OF AP&L AND AN0-1&2 DURING 1986 I Introduction At the licensee's request, AP&L met with the management of the Division of PWR Licensing - B at the NRC offices in Bethesda, Maryland, for the purpose of reviewing the progress, accomplishments and improvements made during 1986.

The year 1986 was a busy year for the licensee in terms of AP&L general and performance improvements and NRC initiated improvements. Included during this year was a major reorganization, an INP0 evaluation, a NRC safety system functional inspection, a NRC environmental qualification inspection, an emergency feedwater system reliability evaluation and a post INP0 accreditation of training audit. They have completed many modifications to both units, both NPC initiated and AP&L in.itiated. Also several programs have been initiated to support the operation and continued improvement of their activities. The licensee expressed a certain amount of pride in their accomplishments and desired to share this with the NRC management. The attendees of this meetino are identified in Enclosure 1. Enclosure 2 is the copy of the licensee presentation materials. Enclosure 3 is the AP&L policy statement on application of 10 CFR 50.59.

Discussion Enclosure 2 provides a very good documentation of the subjects discussed at the meeting. It was noted that AP&L needs to provide two responses to the results of the Safety Feature Functional Inspection. One is a response for additional information which the licensee indicated would be provided with the response to the civil penalty. The other is an updated LER on the EFW check valve issue, which is under preparation.

/S/

Guy S. Vissing, Project Manager PWR Project Directorate #6 Division of PWR Licensing - B

Enclosures:

As stated cc w/ enclosures:

See next page Ah PWR#

GVi g;eh PWPf6#

GEdison PW (

JSto'

/$ g/86 u/$/86 g/[)'86

$k22gQ $ 3 P

(h

  • ~ .

]/

f

'S ' ' '

MEETING

SUMMARY

DISTRIPlITION PWP PROJECT DIRECT 0PATE c6 Docket File NRC & LPDRs PBD#6 Files JPartlow JStolz -

GVissing RIngram OGC-Bethesda EJordan BGrimes ACRS NRC Participants FSchroeder DCrutchfield CMcCracken JSharkey GKnighton RLee GEdison I \

l l

l t

O I n _

$ l' Mr.~G. Campbell Arkansas Power A Licht Company Arkansas f.'oclear One, Unit 1 pm

'y cc:

Mr. J. Ted:Enos, Manager Nuclear Engineering and Licensing Arkansas Power & Light Company P. O. Box 551 Little Rock, Arkansas 72203 Mr. James M. Levine, Director Site Nuclear Operations Arkansas Nuclear One P. O. Box 608 Russellville, Arkansas 72801 Mr. Nicholas S. Reynolds Bishop Libernan, Cook, Purcell & Reynolds 1200 Seventeenth Street, N.W.

Washington, D.C. 20036 4 Mr. Robert B. Borsum

'N Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Resident Inspector U.S. Nuclear Regulatory Commission

, P. 0.. Box 2090 Russellville, Arkansas 72801 Regional Administrator, Region IV

, U.S. Nuclear Regulatory-Commission l Office of Executive Director for Operations 611 Ryan Plaza Drive, Suite 1000 s

Arlington, Texas 76011 l Mr. Trank Wilson, Director

l.  : Division of Environmental Health Protection Department of Health Arkansas Department of Health s 4815 West Markham Street
  • Little Rock, Arkans s 72201 Honorable William Abernathy County Judge of Pope County Pope County Courthouse Russellville, Arkansas 72801
  1. I Enclosure ATTENDANCE LIST FOR MEETING WITH ARKANSAS POWFR & LIGHT COMPAlY DECEMBER 9, 1986 NAME ORGANIZATION Dale E. James AP&L J. Ted Enos AP&L Pqr. Nuclear Eng & Lic Tom H. Cogburn AP&L Gen Mgr Nuc Sycs Jim M. Levine AP&L AN0 Site Director Frank Schroeder NRR PWR-B Dennis Crutchfield NRR PWR-B Conrad McCracken NRR PWR-B Jeff Sharkey IE/0 RAS / PAS George W. Knighton NRR/PWR-B/PD7 Robert S. Lee NRR/PWR-B/I'D7 Gordon E. Edison NRR/PWR-R

A I 4 8

P O

a

. 1986 NRC MANAGEMENT BRIEFING W .

DECEMBER 9, 1986 WW  % e e

. 4-1 g* \

I k$

B I

t- .

I -

i

'u

' l*e6 4

e

~

k i

L.

.i

?

e f

r -

h 1

i

o s

~

1986 NRC MANAGEMENT BRIEFING DECEMBER 9, 1986 INTRODUCTION................ TOM COGBURN, GENERAL ~ MANAGER

[ .

?- .

NUCLEAR SERVICES

~

t ORGANIZATION................ JIM LEVINE, DIRECTOR 4 ^

SITE NUCLEAR OPERATIONS

, . , 1986 STATION PERFORMANCE.... JIM LEVINE a

1986 SPECIAL INSPECTIONS.... JIM LEVINE

- g- LICENSING IMPROVEMENTS...... DALE JAMES, SUPERVISOR LICENSING N

1986 MAJOR ACCOMPLISHMENTS.. TOM COGBURN

~

f wJ TRIP S TRANSIENT REDUCTION..TED ENOS, MANAGER m

NUCLEAR ENGR. S LICENSING r

1987 OUTL00K................TED ENOS b

l O.

I i

i v .

3 .

1-

+

k-y -

.i6i 13 j;j ARKANSAS POWER S l_IGHT COMPANY ir I NUCLEAR OPERATIONS

~~

u- ORGANIZATION i.

MI WI

-e a

U t

P. -

!l, 1

I; f h

9

k.-. .--

. .o .. . . _ . . - . . . .

.. . e.

I t

SENIOR VICE PRESIDENT . ,

GTE , .'

JOHN GRIFFIN ,

- l l

l VICE PRESIDENT VICE PRESIDENT DIRECTOR l ENGINEERING NUCLEAR *'

FOSSIL

! MARSHALL PENDERGRASS GENE CAMPBELL MICKEY COX 1

i

- GENERATION ENGINEERING

- TRANSMISSION & DISTRIBUTION ENGR.

1

! TECHNICAL SUPPORT l t

- TRANSMISSION & SUBSTATION OONSTRUCTION l

i 1

~

i

. . . - . .. .m._ . , _ _ , ,

) C C C C ( .C -

  • A *

.. 41 s

l I

i SENIOR VICE PRESIDENT -

NUMARC- G T E- -MSU NUgLEAR OVERSIGNT COMMITTEE l

l J. GRIFFIN ,

4.

I, t

VICE PRESIDENT <

i NUCLEAR OPERATIONS -SAFETY, REVIEW COMMITTEE BOARD NUCLEAR COMMITTEE- , G. CAMPBELL

=

1 ,

i GENERAL MANAGER DIRECTOR SITE GENERAL MANAGER NUCLEAR QUALITY NUCLEAR OPERATIONS NUCLEAR SERVICES T. COGBURN L HUMPHREY J. LEVINE I

PLANT SAFETY COMMITTEE

.i l

l

, , ,C ,

] M C -

C C3 '

E. ..

  • =~ .

s

, *D VICE PRESIDENT -

NUCLEAR OPERATIONS ....

G. CAMPBELL r .E NUCLEAR SERVICES GENERAL MANAGER T.COGBURN ,

~

g l MANAGER MANAGER MANAGER -

NUCLEAR ENGINEERING NUCLEAR NUCLEAR -

de UCENSING PROGRAMS SERVICES T.ENOS D. SNELLINGS A. WRAPE NUCLEAR ENGINEERING FIRE PROTECTION - FIVE YEAR PLAN

- LICENSING - ALARA - ENGINEERING SUPPORT FUEL RADWASTE I

EMERGENCY PLANNING

F.

~.

~

'1 F- i G

C3 ' C3

O t.a n6 i r, ,'.'~ . . P. " ~ ' ' '

. . .4 VICE PRESIDENT NUCLEAR OPERATIONS l

T. CAMPBELL ,

(903) ,

DIRECTOR SITE I**

NUCLEAR OPERATIONS ,,

r J. LEVINE ,* .

MANAGER SPECIAL PROJECTS (821) o,sowago l

\ (23) PLANT UCENSING

- eUDoET

-SPECIAL PROJECTS I I GENERAL MANAGER PLANT MODIFICATIONS GENERAL MANAGER PLANT OPERATICNS MANAGER PLANT SUPPORT S. QUENNOZ H. JONES E. EWING i

(34)_ PROJECT ENGINEER

, _ CONSTRUCTION l

i l l l OPERATIONS WORM MAINTENANCE ADMINISTMATIVE ENGINEERING TRAINING TECHNICAL MANAGER CONTROL MANAGER MANAGER MANAGER MANAGER SUPPORT l

- -m (150) (34) (199) (121) (68) (59) (120) l t

_UHff 1 OPS _ Pt.ANNING MECH MATERIALS MOWT PE _ OPS _ CHEMISTRY f _ UNIT 2 OPS SCHEDUUNO ELEC OFFICE SERVICES OAO MAINT _ RAD CHEM i

OPS TECHNtCAL _ l at C ,,,,. COMPUTE't SUPPORT , RE _ T. S. HP SHIFT PERSONNEL _ ES _ ADMIN , RADWASTE

]-

' - - _ SiuutAroR

    • - * * .. cer. e....g

e f 1

l I

t' )

1 1

I am

-Og

=5-o 58 WC L WEEE G8 BC"

=g 5"<

- 5*

e wm g

v 5 55 C"E Y.

005:

W395

- Ew "f0 "r5 8-Q db =-

- O

  • m Z .

O e w $

p e ==E e i .c . -

ogsg E-d

=

- 5e@d .

E

< e 5 LD e M e O e 3 .

D

. e -

. . *  : .E E 5 m

< %E35 SEE amew

.D cr 8 "; "$= =35~

wwwa -

~gr agg --

g ~ W 'aw 5"W"

=E-8 O s

e s=

i-m55-es

=855
c=

25

> _ :5cz5 r5 a 8 i e g5g5 WW I, z e e =8~

~"9M I----------- --.I -

88 m

9 f wb U l 5

. . 9= m e

g Uh U$ 0 s
x --

cc5

WEy g5E o= BE5E- C 55 oggE 8

w - B.

5*

.I .

O'*  %

CM

\

{

f5 3h 28!5 e==

h

. =- - E OW $5E5 W u- -Wa-8 i.

, t e

{

' I e i e I e e i e I e I

1 Jl. .

.a..

e t l

9 e

be 9

6 e

bo

-f .

I u

e- ' e

  • e
    1. -4 a e

. 1986 STATION PERFORMANCE

~

[L .

5 J

l.

d

  • i L

emme i

s U

l i

I&

Le 6 .

._.-,.--,I- ,_,,r_ ______. . _ . _ _ _ _ , , , _

,[ _ , . , _ _ _ _ . _ __ , , ,

4 e PERFORMANCE INDICATORS AD&L INPO E

1. MDC Capacity Factor
2. Monthly Average Availability

. Factor

3. Monthly MWHr Generated
4. Cumulathe MWHr Generated

( - Forced Outage Rate - Forced Outage Rate

5. Forced Outage Rate - Equipment Forced Outages /

1000 Critical Hours

6. Equivalent Availability - Equivalent Availability Factor V r-e r - 7. Daily Average Gross Generation
8. Average Net Efficiency /

Average Cire Water Temp

9. Gross Heat Rate Thermal Performance
10. Reactor Power /0T5G Levels

~

11. Nuclear Fuel Reliability - Fuel Reliability

- 12. Core Burnup vs. Time

13. Acceptable Capacity Factor J - to Refueling
14. Whole Body Radiation Exposure - Collective Radiation Exposure j - Maintenance Radiation Exposure
  • 8 and Goal - Positive Whole Body Counts >1% MPBB *

'T

- Skin / Clothing Contaminations

! J 15. Skin Contamination Events /

  • - Month and Goal t
16. Solid Radwaste Generation

- Volume of Low-Level Solid RaJwaste and, Goal

17. Primary Chemistry Hours out at Spec
18. Secondary Chemistry Hours - Chemistry hours outside of Owners U Group guidelines

- Out of Spec a

- Auxiliary System Chemistry Hours Outside of Station Limits

- Condensate Pump Discharge 0 Level

  • y - Condensate Air In Leakage *2

- SG Blowdown Cation Conductivity * -

- FW Cation Conductivity

  • I 19. Electrical, 1&C, Mechanical, and Shif t Mar. hour Distribution
20. Electrical, I&C, Mechanical, and - PM Items Overdue Shift PM's Scheduled vs. Completed

- Corrective Maintenance Backlog - Maintenance Backlog

21. Corrective Maintenance Backlog
22. Non-Functioning Control Room - Out of Service Control Room Annunciators Instruments

' 23. Audit Finding Report Status

  • AP&L trends these indicators at the Department Level.

i t .

- - - - - - .,s _.____ _.._..-_ -

.-_-_-.r. . - -_.. _- _ . - , - . _ _ - ,

4 e

24. Nonconformance Report Status
25. Number of P'C's Written / Month
26. Number of L Written / Month - Significant Events
27. Number of NRC Violations - Enforcement Action Index
28. Number of NRC Open Items
29. Number of Safety Significant - .

Human Errors L

30. Number of Safety System

. Impairments .,,

31. Unplanned _ Reactor ' Trips'. - Unplanned Auto Scrams While ' - Automatic Scrams While Critical

_',' Critical per 1000 Hours Critical

't - Unplanned Auto Scrams While

'-9( Critical Due to Maintenance L 32. Safeguards Equipment - Safety System Unavailability - Safety System Failures

- Availability

~

33. X of Time Clock Required to Clear LCO Action Statement
34. Number of Times Action Statement

- 3.0.3 was Entered

,, 35. Number of Unplanned Safety - Unplanned Safety System Actuations - Safety system Actuations Challenges

36. Number of Emergency AC Power Actuations
37. Number of Lost-Time - Lost-Time Accident Rate in 3 . Accidents / Month Maintenance

,, 38.' Lost-Time Accident Incidence - Industrial Safety Lost-Time Rate Accident Rate

39. Number of No-Lost-Time Acetdents/ Month
40. Number of Vehicle Accidents / Month

[, 41. D&M Budget and Expenditures

42. Capital Budget and Expenditures
43. X Overtime Worked - Maintenance Overtime

. as

44. Absentee Rate r-b.

6 6

e

m n ~

~

F.

~

. 1 r--) cn m'a E I) t

. 3

~

APal 1986 G0ALS & PERFORMANCE OVERALL APal 1986 APal INDUSTRY PERFORMANCE INDICATORS GOALS PERFORMANCE

  • AVG MIDYR '86" U1 > 65% 59% 63%

EQUIVALENT AVAILABILITY 65% i/

FACTOR (UNIT %) U2 E 71%

U1 < 3 1 4

UNPLANNED AUTOMATIC SCRAMS WHILE CRITICAL U2 2 3

~

5 (SCRAMS PER UNIT)

UNPLANNED SAFETY SYSTEM U1<2 0 1 ACTUATIONS (ACTUATIONS U2 2 2

~

0 PER UNIT)

FORCED OUTAGE RATE Ul < 11.5% 7.8% -

6.8 (UNIT %) U2 s 11.5% 3.5% .

Ul < 10,300 10,020 10,343 THERMAL PERFORMANCE 10,295 (UNIT GROSS HEAT U2 < 10,500 RATE BTU / KWHR)

FUEL RELIABILITY U1 < 0.045 0.056 0.082 (UNIT UCI/G) U2 < 0.070 0.017

-< 370 375 424 COLLECTIVE RADIATION EXPOSURE (MAN-REM PER UNIT)

VOLUME OF LOW-LEVEL SOLID ~

< 297 157 325 RADIOACTIVE WASTE (CUBIC METERS PER UNIT)

EMPLOYEE LOST TIME -

< 0.20 0.00 0.30 ACCIDENT RATE PER 200,000 HR

  • THROUGH OCTOBER 1986 (YEAR-TO-DATE)

" SOURCE INPO

. _ _ _ _-_ _ _ _ _ _ _ _ . _ _ - - _ - . - - -. . - - - - - - - . - - ~ . _ - -

," ,___. g-g-g g g-m F

~'

' n '

]L .3 g

~

~

ARKANSAS NUCLEAR ONE 1987 G0ALS .

OVERALL APal 1987' APal 1990 INDUSTRY 1990 G0ALS PERFORMANCE INDICATORS G0ALS G0ALS MEDIAN UPPER QUARTILE.

EQUIVALENT AVAILABILITY Ul > 81%' U1 > 78% ' 75 80%

FACTOR (UNIT %) (NOTE 1) U2 5583% -

U2 5581%

(NOTE 2) t.:

UNPLANNED AUTOMATIC U1 < 2.0 Ul < 1.0 +2.0 1.0

. SCRAMS WHILE CRITICAL .

U2 2 3.0 -

U2 2 2.0 (SCRAMS PER UNIT) { -

I ,

UNPLANNED SAFETY SYSTEM < 1.0

< 0.0 1.0 0.0 ACTUATIONS (ACTUATIONS PER UNIT)

FORCED OUTAGE RATE < 9.0%

- <i 4.5% 5.0% 3.0%

(UNIT %) (NOTE 1)

Ul < 10,125 U1 < 10,050 10,260 10,100 THERMAL PERFORMANCE (UNIT GROSS HEAT U2 2 10,275 U2 2 10,200 RATE BTU / KWHR) (NOTE 1)

FUEL RELIABILITY Ul c 0.015 Ul c 0.007 .010 0.003 (UNIT UCI/G) U2i[0.025 U2i[0.015 COLLECTIVE RADIATION s 88 5L300 300 250 EXPOSURE (MAN-REM PER UNIT) (NOTE 1) (NOTE 2)

VOLUME OF LOW-LEVEL SOLID 5117 s175 220 175 RADI0 ACTIVE WASTE (CUBIC METERS PER UNIT) (NOTE 1)

(NOTE 2)

EMPLOYEE LOST TIME < 0.20

< 0.20 0.20 0.0 ACCIDENT RATE NOTE 1: 1990 GOALS ARE BASED OH 3 YEAR AVERAGE (1989-1991)

NOTE 2: GOAL VALUES IN 1987 ARE SIGNIFICANTLY AFFECTED BY NO PLANNED REFUELING OUTAGE ON EITHER UNIT.

'b

  • ~

s

p. .

l'i lR I

. . ,H, .6 .

@ AVARABILITY G CMiaC,17

G.. -

ARKANSAS NUCLEA ONE - UNIT 2 7 , , .

l;0 .

~

!E :

, 7

/ 79

- o -,,,,

!; o ,, - -

1-

. , ,, g . g ,. g - g . g :., , . , _

AN0-1 REAC~~OR TR PS 1980 THROUGH 1986 '

30 28 -

t.-

26 - -

m 22 -

20 -

g 18 - ,

16 - ,

14 ' .

g 12 -

  • 10 -

ti

,  % 8- '

6- ,

4-

'~

A A xs A 'f,A ,

A ~~ ,

1980 1981 1982 1983 1984 1986 1986 YEAR  ;

= -

,q ,q ,q _, a a w g . g c. , g +, ,

ANO 2 REACTO R TR PS 1980 THROUCH 1986 ,

28 -  ;.*

+,-

26 - .. .

24 - { '

% 22 -

k 20 y

g g

's -

Vf .

14 - -

d 12 10 -

7 4-/

i 1980 1981 1982 1983 1984 1985 1986 YEAR

e O

~

,' 1986 SPECIAL INSPECTIONS t o INPOPLANT/ CORPORATE

~

g 0-- SAFETY SYSTEM FUNCTIONAL INSPECTION

, Y:

'd o ENVIRONMENTAL QUALIFICATION INSPECTION o EMERGENCY FEEDWATER SYSTEM RELIABILITY EVALUATION L

o CONTROL ROOM HABITABILITY o POST INPO ACCREDITATION OF TRAINING J

w af e m b 4 e

b &

I

INPO EVALUATION PROCESS

!. PLANT EVALUATION 2 WEEKS

' APPROXIMATELY 20 EVALUATIONS, INCLUDING PEERS

.. CORPORATEJ EVALUATION .

i WEEK APPROXIMATELY 10 EVALUATIONS, INCLUDING PEERS

.7 METHODS .

~

CONDUCT INTERVIEW 03SERVATIONS

'g-DISCUSSIONS WITH OTHER EVALUATORS REVIEW OF REPORTS, PROCEDURES, POLICIES

' u FINDINGS .

l I .

RECOMMENDATIONS FOR IMPROVEMENT BASED ON BEST j' - - PRACTICES, EXCELLENCE, RATHER THAN MINIMUM ACCEPTABLE STANDARDS OR REOUIREMENTS

~

-- T _

J NOT NECESSARILY INDICATIVE DE UNSATISFACTORY

' ~~

,, PERFORMANCE GOOD PRACTICES I- - OBSERVATION OF AN UNUSUALLY GOOD PROCESS,

, . PROCEDURE OR METHOD WHICH IS AN EXAMPLE FOR

  • OTHER UTILITIES TO FOLLOW

i

-SAFETY SYSTEM FUNCTIONAL INSPECTION i,]

. o JANUARY 6-31, 1986

! o 12 MAN INSPECTION TEAM

  • NEW INSPECTION APPROACH -

o

. EMERGENCY'.FEEDWATER S'YSTEM PRIMARY FOCUS DF g,. REVIEW COVERING:

l'- -

MODIFICATIONS ANO DESIGN CONTROL MAINTENANCE G SURVEILLANCE i ig ' -

OPERATIONS l !, -

TESTING a

TRAINING GUALITY

' (? -

~

l o ENFORCEMENT MEETING.- JULY di, 1986 l

o VIOLATIONS CITED:

l-CATEGORY III -(ONE VIOLATION)

~- CHECK VALVES, CIVIL PENALTY $50,000 j

ri -

CATEGORY -IV (FOUR VIOLATIONS)

IL -

DESIGN CONTROL l- -

TEST CONTROL MOV MAINTENANCE PROCEDURES DRAWING AND PIPING DESIGN SPECIFICATIONS o RESPONSE TO VIOLATIONS BEING PREPARED j 6. o IN GENERAL THE INSPECTION TEAM FOUND THE DESIGN OF

.. THE AND UNIT i EFW SYSTEM TO BE SOUND s

hp

\*

/

ENVIRONMENTAL QUALIFICATION 2 o JULY 14-18

.0- - 7'M4A INSPECTIDA TEAM

.a71 '

) o DNLY 3 POTENTIAL FINDINGS 1- 0 TEAM LEADER HAD POSITIVE COMMENTS ON OVERALL i- PROGRAM p_.

~>

7 d

.m i

i l.

,o

>4

>1

1 LICENSING IMPROVEMENTS

' o NEL ORGANIZATION

~

f- 0 ' . LICENSING BASIS DOCUMENT BESEARCH SYSTEM 16

.o ADDITIONAL COMPUTER. BASIS TRACKING SYSTEM

! O LICENSEE EVENT REPORT UPGRADE

.,j o SHOLLY PROCEDURE / TRAINING

, 0 --

o SAR UPGRADE i

t -

Y 9

9

~-

l

)

I i j .

I,.

,l ,

NEL ORGANIZATION

[ o COMPOSIJION

!Y -1 MANAGER l~- -3 SUPERVISORS .

!~~ - 10 ENGINEERS

I~

-3 TECHNICIANS

. $. -2 RECORDS CLERKS I

0 GREATER THAN 75 YEARS OF COMMERCIAL NUCLEAR

. m POWER EXPERIENCE o RESPONSIBLE FOR LICENSING, FUEL AND NUCLEAR I id i, ENGINEERING ACTIVITIES 0 .

l-P i

't; i-l . .' -

1 .

{j .

. c 4

t 9

- LICENSING BASIS DOCUMENT RESEARCH SYSTEM L

o LICENSING ~ BASIS DOCUMENTS ON-LINE l

ly. K -

,_: o ON-LINE KEYWORD SEARCHABLE m  ?

o REMOTELY ACCESSIBLE 0

i -

i i m h

{

We s

l l

- ~,

I:

I i

le e

e i

l .

i l

t.

~

TRACKING SYSTEMS / PROCEDURES o 50.59 EVALUATIONS

- MAINTAINS ON-GOING DATA BASE FOR TIMELY 50.59

- REPORTING STANDARDIZE 50.59 ANNUAL REPORT SUBMITTALS

~

- PA'ALLEL'50.59 R AND SAR UPDATE REPORTING

?. o SAFETY ANALYSIS REPORT CHANGES .

[

- IMPROVE EFFICIENCY OF SAR UPDATE PROGRAM

_. - MAINTAINS 0N-GOING DATA BASE FOR SAR CHANGES

- PROCEDUREALIZE SAR UPDATE TASKS B

E i 1 u

t e

a q l

l!

l 4

. e LICENSING EVENT REPORTS 0 10 CFR 50.73 REVISED JANUARY 1, 1984 0

- FEWER REPORTABLE ITEMS GREATER ~ DETAIL REQUIRED

' t;

.j.

. o NUREG 1022, SEPTEMBER 1985

[~

- PREPARATION GUIDELINE

$ - TEXT OUTLINE

- - TEXT OUTLINE CHECKLIST o ANO WRITERS GUIDE, JUNE 1986 o

o DIVISION OF RESPONSIBILITY WITHIN PLANT

'[. LICENSING 6

m rw 6 e i '

Ii

i.

~

"SHOLLY" IMPROVEMENT PROGRAM O G0ALS TO UPGRADING OUALITY OF "SHDLLYS"

. - BE MORE RESPONSIVE IN ADDRESSING THE 3 NSHC

- FACTORS

- - IMPROVE THE ADE00ACY OF THE NSHC ANALYSIS

' ~

~

~ -- STASDARDIZE FORMAT OF SUBMITTALS

  • - CLEARLY EXPLAIN BASIS FOR AMENDMENT TO

~ -

- EXPEDITE PROCESSING

'{'

- PROVIDE SUFFICIENT JUSTIFICATION FOR NRC TO ALSO CONCLUDE NSHC DETERMINATION IS PROPER I - ADDRESS NRC IDENTIFIED PROBLEM (GL 86-03) 0 STEPS IN ACHIEVING GOALS if

- REQUESTED NRC PROJECT MANAGER ASSISTANCE

- DEVELOPED "SHOLLY" GUIDANCE DOCUMENT

] - CONDUCTED STAFF TRAINING t '

d

  • 8 9 O

G O 9 I

~

, , l I:

. MAJOR ACCOMPLISHMENTS IN 1986 h

'r o APSL INITIATED -- BENERAL TMPROVEMENTS L . _ . - .. .

o APSL INITIATED - PERFORMANCE IMPROVEMENTS

['

[ o NRC INITIATED IMPROVEMENTS

.m .

7

(!D .

A V

b .

s =

i t

'e e io O

be O

O re 9

4

[

  • e *

-  :=,

s. '

.W p-u O

APSL INITIATED - GENERAL U

s a .,

9 s.

e

  • OM*-

d 4

O

,. r*

t.

km i,,.

~

MODIFICATION PROCESS IMPROVEMENTS l

o REORGANIZATION CREATED SITE PROJECT ENGINEERING GROUP o PROCEDURES ISSUED TO BETTER CONTROL DESIGN INPUT

, CRITERIA .

l .u AND SPECIFIC ASME III S XI TRAINING: COMPETENCY o

[.

TESTINGif. -

,  ; "*: o TRAINING ON CONDUCT OF PRE-DESIGN AND POST i' INSTALLATION WALKDOWNS ON PIPING SYSTEMS o ENHANCED INSTRUCTIONS ISSUED ON DESIGN INPUT

- CALCULATIONS AND INDEPENDENT VERIFICATION PROCESS

~

i  : J.~o ' - REVISEDPLANTSDESIGNENGINEERINGPROCEDURES

~

FOR. CONSISTENCY AND INCREASED DETAIL

~ ' ' ~

n o PLANT WALKDOWNS TO VERIFY AS-BUI'LD DRAWINGS AND

u DRAWING UPDATES

) '

o

. ISSUED NEW PLANT MODIFICATION MANUAL; DEFINES

' ' ~  ;

iB. STEP MOD. PROCESS i{

i e '- ~ PREPLANNING REGUIREMENTS

-i v. ~ ~ PRIORITY SYSTEM

.'j.

~ - ' RESPONSIBILITIES

"- o INDUSTRY EXPERIENCE DATABASE REFERENCE F0R DESIGN _

i' INPUT L. o BEGAN INTEGRATED (TRAINING) DEVELOPMENT PROGRAM FOR I. DESIGN ENGINEERS 6

I lW - _ - _ -- - -

. j

' ~

50.59 PROGRAM UPGRADE

. l 0 NEW 50.59 POLICY AND GUIDANCE DOCUMENT DEVELOPED WITH OVERSIGHT BY SAFETY REVIEW COMMITTEE u

[x _

CO.NSISTENT WITH.NRC POLICY

'X -

ISSUED FOR TRAINING PURPOSES a NEW POLICY DEFINES TWO LEVELS ~0F 50.59 REVIEW

' ri u -

50.59 DETERMINATION 50.59 EVALUATION

[

[,e

c. .

(i) 50.59 DETERMINATION f.

WOULD THE PROPOSED ACTIVITY CAUSE INFORMATION

!- IN A LICENSING BASIS DOCUMENT TO NO LONGER BE TRUE OR ACCURATE OR WOULD IT VIOLATE A REGIIRE- J

2. NENT STATED IN THE DOCUMENTS?.

(2) .50.59'$hLUATION -

'~Ik;

[. DOES THE ACTIVITY RESULT IN CONTINUED OPERATION i WITHIN THE BASIS ON WHICH'THE OPERATING LICENSE i ~

,, STANDS?~

J NO TECH SPEC CHANGE REQUIRED NO REDUCTION IN MARGIN OF SAFETY.IN TECH lg SPEC BASIS 1

NO NEW POSSIBILITY FOR~ ACCIDENT

{ -

NO INCREASED PROBABILITY 0F ACCIDENT ,

,0 THE 50.59 EVALUATION IS WRITTEN AND PROVIDES BASIS li - FOR CONCLUSION ~THAT THE ACTIVITY. (CHANGE) DOES OR l DOES NOT. INVOLVE AN UNREVIEWED SAFETY. QUESTION r

9 0

t-ll' a_.

~

LICENSING BASIS DOCUMENTS (1) SAFETY ANALYSIS REPORT (S)

~

u INCLUDES: EMERGENCY PLAN *

~

, b _- ~ d~ 1 0A MANUAL"-*

N-CORRESPONDENCE 0~ ~

50.59 EVALUATIONS r

);' (2) OPERATING LICENSE (S) d ~.

INCLUDES: TECHNICAL SPECIFICATIONS

~

ORDERS

[

,  : SAFETY EVALUATION REPORTS

[ 'i~ " (THAT SUPPORT LICENSE G AMENDMENTS) r-1.

l N 10CFR50.54 REQUIRES ADDITIONAL REVIEWS (ALSO AFFECTS

[, .

SECURITY PLAN) 1  ;. .

i o NEW 50.59 POLICY REQUIRES CERTIFIED 50.59 REVIEWERS o GUALIFICATION AND CERTIFICATION CONSISTS OF:

l,u

.NEET ANSI /ANS-3.1-1981 EXPERIENCE AND p

TR4ININGFORDI$CIPLINE i3

(. -

RECEIVE 50.59 REVIEWER TRAINING il*

i L

, j PASS A GUALIFYING EXAM P

ACCEPTANCE BY PSC/SRC l 0 ' *. '

( '

ln REQUALIFY EACH 2 YEARS iU l

n s

[

II -

l .a I '

t _ _ _ _ - . . _ - . _ _ _ . . _ _ . _ _ _ _ - . . . _ _ _ _ . _ _ _ _ _ _ _ - _ ___

o f COMPETENCY BASED TRAINING PROGRAM HAS BEEN DEVELOPED a TRAINING HAS BEEN PROVIDED TO DESIGN ENGINEERS, PSC .

~

AND SRC MEMBERS

~

o TRAINING ~FOR G.O. TECHNICAL STAFF IN PROGRESS

,) O

. TRAINING FOR SITE OPERATIONS, MAINTENANCE, T.ECHNICAL l

.]- SUPPORT PERSONNEL IN iST GUARTER '87 I L

c o DESIGN PROCEDURES HAVE BEEN REVISED 0- . O SITE PROCEDURE REVISIONS IN PROGRESS 0 PROGRAM.TO BE FULLY IMPLEMENTED BY APRIL 1, 1987 t-W

[

r -

le 4

is no l'~

n b~ _ _ _ _ _

L.

. ANO-i G 2 SIMULATORS L.

ANO-1 o -OPERATIONAL FOR TRAINING IN 1986 INITIAL, REACTOR OPERATIONS LICENSED OPERATOR f

EXAMS i' 7 of 8 Candidates passed

..REQUALIFICATION OPERATIONAL EXAMS rg

[ _- .

' l .All- Candidates passed i N o POSITIVE COMMENTS BY REGION IV ON RESULTS OF

!~. SIMULATOR EXAMINATION

(<!.

,o POSITIVE COMMENTS BY INPD ACCREDITATION TEAM s

l]

i o SIMULATOR UPGRADES IN 1986 g.. ~-

ANNUNCIATOR UPGRADE EFIC i

SPDS l in l 'u -

SURFACE ENHANCEMENTS OTHER

~

-o UPDATE THROUGH 1R7 IN 1987

o. .IN-HOUSE CERTIFICATION TO ANSI 3.5 IN 1987 l

'~

b ANO-2 o COMPLETED MAJOR MODIFICATIONS AND TESTING

,f

i. o TRAINING TO BEGIN IN JANUARY 1987 i;
1. o UPDATE THROUGH 2R5 IN 1987
o IN-HOUSE CERIFICATION TO ANSI 3.5 IN 1987

}

E ANO-i RCP VIBRATION MONITORING SYSTEM

-m o NEW COMPUTERIZED VIBRATION ANALYSIS SYSTEM

~

o ONLIkEINDICATIONINCONTROLROOM Ev ~

O CONTROL ROOM ALARM

o COMPLETE ~ON LINE AND HISTORICAL ENGINEERING m ANALYSIS. CAPABILITY

~

0 O CAPABLE OF DIAGNOSING SHAFT CRACKING AND OTHER PROBLEMS PRIOR TO FAILURE g

l ;0 -

l 'D i;

I t.

b.

ie

I

[jr-

,t I,. O REDUNDANT INTERNAL HYDROGEN RECOMBINERS INSTALLED 1R7

~

b o REPLACES EXISTING HYDROGEN PURGE SYSTEM o IMPROVED ~RELIABI;.ITY'-

{'

}j: -

O LICENSE AMENDMENT GRANTED OCTOBER 7, 1986 l

] Q CONDENSATE STORAGE TANK

[0.

! o CONSTRUCTED IN LIEU OF AUTO SWITCH TO SW (NUREG-0737 ITEM II.E.1.1) .

-0 TANK IS SEISMIC AND PARTIALLY TORNADO PROTECTED o TECHNICAL SPECIFICATION SPECIFING~ DEPTH.0F WATER

~

l (( " . REQUIRED IN TANK APPROVED 9-26-86

!'r .

j o TANK TO BE OPERATIONAL AT THE END OF 1R7 I-0 WANT TO USE THE TANK FOR AND-2 TO ELIMINATE AUTO

!~ t.

SWITCH TO SW t _ _ _ _ _ _ . __

ANO-1 S 2 H P DDSE ACCOUNTING COMPUTER o

INSTALLED AND CURRENTLY IN USE (SOME FEATURES o PROVIDES COMPUTERIZED, ON-LINE, REAL TIME PERSONNEL p DOSE ACCOUNTABILITY

~

t .

1 o VERIFIES INDIVIDUAL WILL NOT EXCEED ALLOWABLE DOSES

, PRIOR TO STARTING WORK

[ o~ - . EXTENSION OF ALARA PROGRAM fl 61 LOW LEVEL RADI0 ACTIVE WASTE STORAGE BUILDING

,0

,{ o LOW LEVEL' WASTE.PDLICY..ACT ALLOWED DISPOSAL SITES

~

TO RESTRICT.USE'TO REGION GENERATED WASTE fij o

FACILITY ~.BUILTLFOR STORAGE AS LOCAL DISPOSAL SITE

[ ~ NOT AVAILABLE r

o CURRENTLY USED AS TEMPORARY. STORAGE m

e fe a

wc- ,y,--y .-.y----,.-- 9 ,,w-,-w,--.--- c-my...-- - - . - , .,m, - - - , - - . - - - -. - --, , - - - ., - - -, - . -- --- - - - - -_ __-- - ---- - - -

b.

e INP0 ACCREDITATION 1986 h

0 THREE ADDITIONAL TRAINING PROGRAMS ACCREDITED FEBRUARY 1986 ELECTRICAL MAINTENANCE

}g

~ I '- -

MECHANICAL MAINTENANCE Li I G C MAINTENANCE O- .

o ONE OF EARLIEST UTILITIES TO HAVE THESE PROGRAMS ACCREDITED g

~

o ALL SELF-EVALUATION REPORTS ON REMAINDER OF PROGRAM l

. SUBMITTED IN JUNE 1986 wi l[ -

1r i

>e l

____--___.a. ._ _ _ . , - - _ . - - _ _ _ . , - - _ . _ _ _ , _ _ . _ _ _ _

HIGH SUCCE$S RATE IN OPERATOR TRAINING

]

$ AND-i

[. o 7 0F 8 REACTOR OPERATORS PASSED r - 7 0F 7 BLOCK HIRES PASSED L o 45 0F 51 (88%) RD S SRO PASSED REQUALIFICATION EXAMINATIONS

[- FIRST TIME

... ANO-2

,= .

o 8 0F 8 SENIOR REACTOR OPERATORS PASSED

'O o 29 0F 31 (94%) R0 S SRO PASSED

{ .

REQUALIFICATION EXAMINATIONS

~FIRST TIME l0 I o NRC REGIDN~IV ACKNOWLEDGED HIGH PASS j

RATE AND COMMENDED'APSL*.s TRAINING g- PROGRAM i,.

Le m

STATION INFORMATION MANAGEMENT SYSTEM (SIMS)/

~

WORK CONTROL CENTER (WCC) l l

1 0 SYSTEM FULLY OPERATIONAL AND WCC FULLY STAFFED

, IN EARLY 1986 I- o. INCLUDES PLANT EQUIPMENT DATA BASE WITH DVER ,

y 100,000 COMPONENTS

.g*

o . ISSUES AND TRACKS JOB ORDERS

. ['-

] a WORK CONTROL CENTER PLANERS PREPARE DETAILED WORK PACKAGES n

U ~

o KEEPS DETAILED MAINTENANCE HISTORY g

6

[

[

'r i

t t.

l l.-

? . - . - - -

t. . .

TECHNICAL MANUAL REVIEW o IDENTIFICATION AND UPDATE OF APPROXIMATELY 25,000 TECHNICAL MANUALS o CONTROL AND DISTRIBUTION OF TECHNICAL MANUALS

' o REVIEW OF UPDATE PROCEDURE S DESIGN CHANGES AGAINST MANUALS

f. .: : .- .

o F6RMAL PROCESS FOR MAINTENANCE AND UPDATE OF

~N -

~ MANUALS AND AFFECTED PROCEDURES L.

o REVIEW OF VENDOR BULLETINS, NRC G'INPO NOTICES PREVENTIVE / PREDICTIVE MAINTENANCE 0 REVIEW UPDATED TECH MANUALS FOR CURRENT VENDOR RECOMMENDATIONS o REVIEW MAINTENANCE HISTORIES o REVIEW IMPACT 0F EQUIPMENT SERVICE CONDITIONS UPON PM NEEDS

{

2 Jo ~ VIBRATION'. MONITORING-PROGRAM

~o OIL MONITORING PROGRAMS r

o VALVE STROKE TIME TRENDING e

i 6.

w _ . - - . - - - - - . - . - -

t COMPONENT LEVEL G-LIST l

0 EXTENSION OF SYSTEM LEVEL 0-LIST t.' o DEFINES 0 STATUS ON A COMPONENT BASIS b o CONSONIDATESSPECIALREQUIREMENTSFORCOMPONENTS s; IN ONE DATA BASE (eg. O, EG, FIRE, etc.)

o INPUT FOR GRADED QUALITY PROGRAM BEING DEVELOPED 1[

~

m 6

l *I -

L,

?

m .

l.

9 Le I~

t*

4 .

Le P*

E

![ _

r-

, APSL INITIATED - PERFORMANCE IMPROVEMENTS bel 9

n "
U .

[

r e

% e , - - - - - - - - - , . - - , . - - - - . - - - - - . , ..,-------w- -- . - - , - , - -. - . - + - - - - .---.- - - - - . .-- - ,-- - - - - , - - - - - - - - - - --- - - - - - - - - - - - - - - - - - - -

APGL PARTICIPATION IN THE DDE .

' I EXTENDED-BURNUP PROGRAM UNIT i

,. o FOUR ADVANCED DESIGN (MARK BEB)

- FABRICATED AND EXPOSED FOR THREE CYCLES

~

. o ONE ASSEMBLE 11) BE EXPOSED FOR A FOURTH CYCLE DURING CYCLE 8

}

" ~

UNIT 2

[

m o TWO ADVANCED DESIGN ASSEMBLIES U .

FABRICATED AND EXPOSED FOR THREE CYCLES o ONE CONTINUED EXPOSURE S NOW IN ITS FIFTH

!r IL CYCLE DURING CYCLE 5 l

a T

t.

) b .

I j , .

J . - . . -

1 1

ANO-1 HIGH PRESSURE TRIP SETPOINT o INCREASED SETPOINT FROM 2300 PSIG TO 2355 PSIG l (0RIGINAL DESIGN VALUE)

. o REFERENCED BAW-1890 AND NRC SER

o. T.Si:AMENOMENT REDUEST BY LETTER DATED JULY 18,

,; 1986 a NRC APPROVAL OF T.S. AMENDMENT BY LETTER DATED it.

i NOVEMBER 14, 1986 (AMENDMENT 104) i,

f fl i bA '

I i

lr'

.~

F j_i' l*

, 5. .

G L___ _ . _ . _ _ ~ - _ . _ _ _ _ . _ . . _ . .. ._ __

ANO-1 ANTICIPATORY REACTOR TRIP ARMING THRESH 0LD i o INCREASED ARMING THRESHOLD FROM 20% TO 45% POWER l

a REFERENCED BAW-1893 AND NRC SER 0 T.S. AMENDMENT REQUEST BY LETTER DATED JULY 31,

~

~

~

_ 1986 :-

O NRC APPROVAL OF T.S. AMENDMENT BY LETTER DATED NOVEMBER 14, 1986 L-

[. 0 ESTIMATED REDUCTION IN TRIP FREQUENCY OF 10%

, (ART IN COMBINATION WITH HPTS) l..

~

di 9

.4 7

i .

i 4 -

l

[ _ . . . . . . - .

L.

MAIN FEEDWATER - 1R7 f

a NEW LOVEJ0Y MFWP TURBINE CONTROL SYSTEM

- 0 EXTENSIVE CONTROL HYDRAULIC, LUBE DIL S THROTTLE (STEAM *. INLET) IMPROVEMENTS

]t ICS - iR7 o POWER SUPPLY RELIABILITY IMPROVEMENTS 7

.oJPOSITION INDICATION FOR S/U S.LO-LOAD CVs,

~

s .

TBVs 0 -

0 RAPID FW REDUCTION SYSTEM o'.TBVs FAIL CLOSE ON LDSS OF ICS POWER g

l e l

J .- ._.-.-5. .-,,._,.,---m ~~ ' ~ " ~ ~ ~ ~ ~ ~ ~ ~ ' ' ^ '

4 STEAM GENERATOR MAINTENANCE PROGRAM

~

ANO-i l

o IGA IDENTIFIED AS PRIMARY CONTRIBUTOR TO TUBE DEFECTS AT OR ABOVE 15TH TSP IN VICINITY

- 0F OPEN LANE REGION o StCONDARY SIDE FOULING LEAD TO REDUCED POWER

}:

~. o APSL ESTABLISHED AGGRESSIVE PROGRAM TO ADDRESS PROBLEMS O'

[ o CORRECTIVE MEASURES

b. - EPRI WATER CHEMISTRY GUIDELINES - UPGRADED INSTRUMENTATION i[7

!2 - MECHANICAL CLEANING - WATER SLAP PROCESS l PATENTED BY APSL AND MPR i" ~- SLUDGE LANCING - REMOVE SLUDGE PILE FROM l

} LOWER' TUBE SHEET

--TUBE SLEEVING - COMPREHENSIVE PROGRAM

- TO QUALIFY SLEEVING PROCESS (SPANNING r DEGRADED REGIONS OF TUBES WITH SLEEVE INSERT) T.S. SUBMITTED APRIL 1, 1986 o RESULTS OF PROGRAM (iR7) '

Il'

- APPROXIMATELY 566# OF DEBRIS REMOVED BY

' WATER SLAP' AND SLUDGE LANCING

! - 40 SLEEVES INSTALLED (32 PREVENTATIVE) w_______.._.,___.______.-_____,-,..._____

i, STEAM GENERATOR MAINTENANCE PROGRAM ANO-2 o DENTING PRIMARY CAUSE OF TUBE DEFECTS 0 APSL ESTABLISHED AGGRESSIVE PROGRAM TO

-ADDRESS-PROBLEMS

'v

~

. O CORRECTIVE MEASURES l

,. - EPRI WATER CHEMISTRY. GUIDELINES

~

- BORATION OF CONDENSATE m

7, 1- SLUDGE LANCING U

o RESULTS OF PROGRAM (2R5)

~

-:NO SIGNIFICANT DEGRADATION 1-r w.

J f

D

, _ _ . _ , - - - -- _ _ _ _ ..-__ - _.-,,_,__.,____.--___,,__-,__._,_-._-_m.

L. .~_'E~~

~

RCP SEAL MAINTENANCE IMPROVEMENT O MORE THROUGH RECEIPT INSPECTIONS f

o IMPROVED TRAINING m

o IMPROVE HANDLING EQUIPMENT

(. .: : -- -

w 0 ANO-1 ICW TEMPERATURE REGULATION J-

]- 0 AECL MAINTENANCE PRACTICES ADOPTED

[ o' PARTICIPATION IN BJ N-9000 SEAL DEVELOPMENT

, PROGRAM d

. m 1

we r-g

m. J 1

i.

r-e 4

AND-2 CPC UPGRADES l l'

o CPC IMPROVEMENT PROGRAM (SOFTWARE) l

- CEOG EFFORT BEGAN MID-1983

.. - PRESENTATION TO NRC NOVEMBER 1984

- - ANO-2 IMPLEMENTED 'CEA DROP FIX' MID-CYCLE 5

~

. (TS' AMENDMENT 70 11/85)

%- - CIP SOFTWARE IMPLEMENTED CYCLE 6

(TS AMENDMENT 79 7/86)

,j - BENEFITS: HAS PREVENTED 2 REACTOR TRIPS TO DATE, MINIMIZES NEED FOR' CYCLE-SPECIFIC 1~ SOFTWARE CHANGES, CYCLE SPECIFIC DATE ENTERED.USING ADDRESSABLE CONSTANTS lg a CPC HARDWARE UPGRADE PROGRAM

-i"

- APSL/CE COOPERATIVE EFFORT

- NEW HARDWARE, PRESENT-DAY TECHNOLOGY

,kj -

- RELOCATED IN NON-CA~ AREA, ENVIRONMENT l ]~

SPECIFICALLY FOR COMPUTER USE

- PLANNING FOR CYCLE 6 TESTING

$ - INSTALLATION FOR CYCLE 7 OPERATION ~

- BENEFITS: MORE RELIABLE, SPARE PARTS VENDOR r

" SUPPORT EASIER MAINTENANCE

- UTILIZES EXISTING SOFTWARE ,

- 50.59

o CPC SPARE PARTS

- COOPERATIVE EFFORT BETWEEN APSL, LPSL, SCE j

i i i

I~ . . . - _ _ _ . _ . _ . _ .

F t

L.

O AND-2 STARTUP/ PHYSICS TESTING IMPROVEMENTS TS CHANGE ON STE 3/4.10.1.2 (AMENDM.ENT 76

1. ELIMINATES UNNECESSARY REACTOR TRIP (SAVE r

~8 HRS.)

~'

rp -

CEOG CEA WORTH BY " ROD SWAP" METHOD .

~

ELIMINATES SYMMETRY TEST (SAVE ~1 DAY) o ANO-2 ACTM MODIFICATIONS INSTALLED IN CEA BANK 6 FOR TESTING m -

WORKING WITH CE

. tj .

. ,- WILL REDUCE ROD DROP EVENTS

~~

[

~ FULL INSTALLATION PLANNED FOR 2R6 o ANO-2 PPS ~ PUSH BUTTON MODIFICATION

r -

CAUSE OF SPURIOUS TRIPS DESIGN CHANGE 2R5 t

>e r .

e i

A ..

i w

FUEL INTEGRIT.Y IMPROVEMENTS 4 rl

. o SPECIAL PRECAUTIONS TO AVOID DEBRIS IN RCS 6

o VACUUMING IN RCS

{

0 AND-2 UT EXAMINATIONS j-

j - 12 PINS IDENTIFIED AS LEAKING AND REPLACED; 2 OTHERS REPLACED ALSO n

l g a NEW AND-i FUEL WITH RECONSTITUTABLE ASSEMBLIES W .

ORDERED a

3 o BWOG AND CEOG EFFORTS ON-DESIGN IMPROVEMENTS

.0 .

e l

i 6.

I i

k-- ~ ~ = :Jt' =:. :_ _:_ - _ - ,2_._. - _. - .. :.=r 2.:= . r-~ 2--= .. = .-. -

- a

,-a - "'""&

e = l l

l FUEL HANDLING EQUIPMENT

}

ANO-1

- FUEL TRANSFER MECHANISM REDESIGNED (iR7)

HkS' CAUSED SEVERAL DAYS CRITICAL PATH

! (~D . EXTENSION PREVIOUS TWO REFUELINGS 0'

ANO-2 m

REFUELING EQUIPMENT REPLACEMENT PLANNED

. [.i .

- FOR FUTURE p- _

L we

.w l

l l e t e 4

N

r*

P-1 i

J' rg.

P.j-

. NRC INITIATED IMPROVEMENTS

. 4 .

.h ,

< (y -

be Ih t - e

,L . . .

5 g N' A\c .

g L.

'.!pr-Mi ',

i L ,-

r' -

t .

' I r*, _ .

I

,f

.u I I 1) t I'

l :

t.

t -

?

i .

, . 1 ,-

I l

_.k_. ______m .. ___ . . _ _ _ - , . - a ,,<.v-, ,a-_ .- . . , ..,-, ,, . _, . , _ . _ _ _ . . . _ . .

CONTROL ROOM DESIGN REVIEW (NUREG 0737 SUPPLEMENT i) o SURFACE ENHANCEMENTS IMPLEMENTED IN ANO-1

<. CONTROL ROOM 1R7 N _

)

REGULATORY GUIDE 1.97 (NUREG 0737 SUPPLEMENT i) 7 AND-i d

a o COMMITTED MODIFICATIONS IMPLEMENTED

~

ANO-2

{ ,

[ 0 ALL BUT T COMMITTED UPGRADES COMPLETED e

1 q-6

_ . . . . _ _ _ _ _ _ _ _ _: ... _ -.721______.__ _ __. . _ _ _ _ . _ . _ _ _ _ _ _ _ __.

I.E.B. # 85-03 P MOTOR OPERATED VALVE TESTING PROGRAM o 79 MOVs TESTED IN UNIT i

.. o 73 MOVs TESTED IN UNIT 2 .

o_ DESIGN A P'S CALCULATED Ik

"".,- o VALVE CALCULATIONS OBTAINED FROM VALVE VENDORS 0 ACTUATOR CALCULATIONS OBTAINED FROM VENDORS a TORQUE SWITCH SETTINGS h

t: _

DETERMINE TARGET VALUES FROM VENDOR CALCS MOVATs TESTING TO SET TOROUE SWITCHES AND VERIFY THRUST o A P TESTING OF SELECTED GROUPS OF MOVs

~

(CONFIRM TOROUE SWITCH SETTINGS) 21 MOVs IN UNIT i 5 -

16 MOVs IN UNIT 2 r '

- o TESTING PROGRAM RESULTED IN EXTENDING BOTH OUTAGES I

!p J .. . _.

~

. . ANO-i INADEOUATE CORE COOLING MONITORING SYSTEM

~

o REQUIRED BY NUREG 0737, ITEM II.F.2 a SYSTEM CONSISTS OF (1) SUBC00 LING MARGIN MONITOR, (2) CORE EXIT THERMDC00PLES, (3)

REACTOR VESSEL LEVEL MONITORING SYSTEM ANO l- -

_ (4)JHOT. LEG LEVEL' MONITORING SYSTEM o RVLMS CONSISTS OF TWO RLI CONTAINING AXIALLY j -

DISTRIBUTED ~RGT SENSORS TO DETECT COOLANT INVENTORY ABOVE CORE. (RLIs PLACED IN CENTER

~

CRA LOCATION)

. . o .HLLMS CONSISTS OF ONE WIDE RANGE AND FOUR NARROW RANGE A P TRANSMITTERS PER HOT LEG o 24 CETs TO BE USED FOR ICC~ MONITORING INSTALLATION OF HLLMS, RVLMS COMPLETED DURING 1R7 0

_ o FINAL DESIGN DESCRIPTION SUBMITTED SEPTEMBER 17, 1986 7-i 0 IMPLEMENTATION LETTER DUE FOR NRC SUBMITTAL 6

~

MONTHS AFTER COMPLETION OF 1R7 h

t I-J t . _ . . _ .

. _ 1 ._.. ,

ANO-2 INADEGUATE CORE COOLING MONITORING SYSTEM

~

o REQUIRED BY NUREG 0737, ITEM II.F.2

. O SYSTEM CONSISTS OF (i) SUBC00 LING MARGIN MONITOR, (2) CORE EXIT THERMOCOUPLES AND- (3) e REACTOR VESSEL LEVEL MONITORING SYSTEM

~

p. o RVLMS CONSISTS OF TWO RLIs CONTAINING AXIALLY DISTRIBUTED RGT SENSORS TO DETECT COOLANT

[: INVENTORY ABOVE THE CORE 5 0 42 CETs (21/ CHANNEL) n FINAL DESIGN DESCRIPTION SUBMITTED APRIL 15, 1986 o

5 o NRC SER RECEIVED OCTOBER.24, 1986

0 IMPLEMENTATION LETTER DUE FOR NRC SUBMITTAL ON

~

i

[ JANUARY ~27 i987 lr F

Ls ~

r*

I!

' _ _ _. . ._. _ __ _ . _ - . _ ..rrrrn _ _

i INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA)

SAFEGUARDS PROGRAM

~

o ANO-2 WAS IDENTIFIED FOR PROGRAM PARTICIPATION ,

IN AUGUST, 1983 l 0 PROGRAM OBJECTIVES

~

p

- ASSURE NUCLEAR FUEL IS NOT DIVERTED FOR WEAPONS PRODUCTION

~! - VALIDATE ACCOUNTING PRACTICES

[ o PROGRAM STATUS

- SUBMITTED CETAILED DESIGN INFORMATION IN AUGUST, (0 .

. 1983

- SPEND APPROXIMATELY 90 MAN-DAYS ON IAEA AUDITS

- PROGRAM COMPLETION JANUARY 21, 1986

~ ['

- LICENSE AMENDMENT DELETING LICENSING CONDITION JULY 22, ~1986 (AMENDMENT 78)

]

r

, F 9

=.-ma nn-- - , _-w,,, - - , , ".,-----,-.-,----,------,..-----,----,,,,--rwn.,n--,

- - -,. ,- - - , , - , - -- - - , - - - - - -.,- ,--.- - , ,- - . - - - - , + - -- - - - - - -

L .

ANO-1 RCP SHAFT INSPECTIONS

~

0 ANO-1 RCP SHAFTS INSPECTED DURING 1R7 DUE TO PROBLEMS FOUND AT CRYSTAL RIVER AND DAVIS BESSE

. o ALL FOUR RCP SHAFTS UT

o. ALL;CAPSCREWS AND DRIVE PINS TESTED USING GE

% METHOD J-

}- 0 RCP 'A' DISASSEMBLED AND INSPECTED

[ o JOURNAL SIDE PLATE FILLET WELD CRACK FOUND

- - NOT SAFETY CONCERN

.- ~.

5 o NO OTHER PROBLEMS IDENTIFIED

.J I'l

.I

!L

,[

r 4

..---.,,,-,,,_,_..--..,-,.--a . . _ - - - - - - , _ , , , , , - _ _ , , , , , , , - -

I*

I ,

f u

^

l. .

g t

AND TRIP REDUCTION S AVAILABILITY IMPROVEMENT PROGRAM

[

6I kb mi 6>

]

f ks i

A L 4 e

D I

L k

f

.. AND PLANT TRIP REDUCTION PROGRAM

~

N WHY HAVE SUCH A PROGRAM?

$ 0 REDUCES CHALLENGES TO SAFETY SYSTEMS ,

0 0 ECONOMICS I

R 1

4 MI N

d F

J . _ . . . - .

\

APSL TRIP REDUCTION S TRANSIENT RESPONSE IMPROVEMENT PROGRAM l NRC i 1 TRIP AND l l l EXPERIENCE l l TRANSIENT l l 1 FEEDBACK i  ! EVALUATIONS l CEOG --- - -------

1 BWOG l

[ l i TRIP.

1 I ~. I SPIP 1 l REDUCTION I l INPO l l l

,.g. ___________

g. .___________

l l 1 ANO l l INDUSTRY l I EXPERIENCE l l EXPERIENCE l

- 1 FEEDBACK l l FEEDBACK l 2, I o EVALUATE APPLICABILITY l

~-

1 o SAFETY SIGNIFICANCE l i o VALUE IMPACT 2 1 0 PRIORITIZE l I o SCHEDULE ~ l l

i _

l 1 1 IMPLEMENTATION l

. I PERFORMANCE I F l MONITORING l l e e

L l

1

  • I P

~ .. --

)

- . .e I-e r*

Le f*

i f

, g.

J.

REACTOR TRIP AND EQUIPMENT OUTAGES.

r N

m .

i  ;

E 1

f -~

4 e

i .

j ' -

. 6.

l.

t i ,

JL - --,-._.-. - - .

.- .- , , e o ma a q7c, .

REACTOR TR PS /

1 1980 THROUGH 1986 , k

i. 30 ^

i l 28 - ,.

26 - '

l 24 -

\  % 22 -

20 -

), g 18 - ..

! o

! s 16 -

j u 14 - -

12 -

b 10 -

Ci 8-6-

1 I 4- ~

2-s O i i i i i l ,

1980 1981 1982 1983 1984 1986 1986 YEAR a AND-1 + ANO-2

~

, o a c;- su en ca ,, j r, c -

g A

I

~

~

/

ANO-1 PLANT TRIPS

1981 - 1986 1

! 13 toww (6.1x) '{

1

, 4 RPs/Rx (12.1%)

j i.-

i i

2 OTHERS (6.1%)

\

m 1311FW UPSETS (39.4%)

i 4

l i

l l

4 12 T/G C (36.4%)

l i

l

7, o

)

3 5

4

(

x R

/

S P

R 9

3

' . . )

r I 4 3

j c

2

(

S R

S E

.PI m

o R 2 T

6 T8 9 s

N1 A

, L-P1 8

u. 21 9 n -

r O N

i A

)

r  %

6 8

I

) 1

% (

E 2 8

9 1

C G

( /

S T 1 T 6 E 1 S

, P U )

W F 0 M 4 1

7 (

1 W

F M

O L

2 1

s 4

[

i ,

7 -

g g

y, g r

x g i 2 N O N

~

A 5

' ,9 8 4 1

E C

NSR '. o .

f-r t .

1 j AOT 4 O

N MCFA ,9 8 A r

R Y 1

OITC R+

A F A E Y

Q RAPC E 3 a PAU L - ,9 8

1 u N N

7 OA E

1 N C

.r A A 2 R

E a ' .

- 8 V

,l.

,9 A 1

6 Y

R 2

T S

U D

7 N I

1

" 1 8

- - - - - - - - 9 0 0 o 0 0 0 0 0 0 0 1 D

. 9 8 T 6 5 4 3 2 1 .

t

~

..F _. b , Cb Q F~ a ED C""") s-- .

,g n F. .

7 . .

AN O CAPAC ~~Y FACTO RS -

uuoR rurActs 110 -

180 mwwx iwxwwx gxxx k xxxx g xxxxxxxxx3 90 ~

Vffff f

f fllllllllh

! hhhkk\ b1 i ::

f//f / /

//V

@ CF D REFUEL. @ EQUIP. @ TRIPS

p REACTOR TRIPS

i. 1981 - 85 f-
h" ANO-1 w,

27 AUTOMATIC TRIPS ,

D AVERAGE TIME = 70.58 HOURS O. -

ANO-2 56 AUTOMATIC TRIPS AVERAGE TIME = 32.57 HOURS

)

' F'

,F I-le l.

- , - ww--.--- - - - - - - - - - -,e-----

[ , VAJO R OU-~AG E V3ACT-2.6

" " " * ~ " * '

2.4 -  ;,-

2.2 - l':

b 2- -

i l

a; r., -

f.6 -

1.4 -

n n n , nr CATEGORY

. . , r2 a '

r_3 to ca ca _,

Fj . r i H q A

ANO-2NfdORINPACTSONCAPACITYFACTOR .

j ITEM %CF SOLUTION STATUS l

! RCP Seals 1.3 Maintenance Procedures  ;,.1986 2R7 l Improved .

.; Project

)

Seal Improvement Project

, CPC Trips 1.2 Software Upgrade 2R5

[ Hardware Upgrade 2R6 4 ..

T/G Trips i.i Various Causes Under Study Condenser / Chemistry 1.1 Inspections, Staking, 2R5, Ongoing Cleaning with Acid '

i RTD Response Time 1.0 New RTDs S Thermowells Complete 2R4

Blowdown Lines 0.8 New Valves Installed Complete 2R4 i

lr- .',_

m

, ,a a gj- . 3 m a c ,g m c

\

l l ANO-2 MAJOR IMPACTS ON CAPACITY FACTOR -

(Continued) f

%CF SOLUTION STATUS ITEM

(  ;.-

0.8 Re-Designed

.' .' Complete 2R4

! RCP Sensing Lines 0.8 CPC Software Changes Complete Dropped Control Rods New ACTM Partial 2R5, 2R6 0.8 Valve G Control Pb1ms Under Study

! Main FW Controls CEOG .

SG Denting Control 0.7 Boric: Acid Treatment Continuing to I

Avoid Denting Pressurizer Valves 0.6 Install Isolation Valves 2R6 0.5 Install New Batteries 2R5 -

Station Batteries SG Level Trips 0.5 Tech Spec Change & Complete 2R4 3

i Improved Training, New Instrumentation

a. . , cm en a om o ca F7, tj . r

~

V AJO R OUTAGE \A 3AC-~

5 YEARS - ANO 1 <

3 2.8 -  ;.*

~

2. 6 -  ::

//

g 2.4 -

22-1.8 - '

o 1.6 -

l'

. k 1.4 - f '

l 1.2 - '

/

/

5 4

Y

0. 8 -
0. 6 -

/ -

5 444 8 4 4 42?8 /

MFW RCP S SCLV SC TU T/C RODS EXT STMRCP SNS CATEGORY

' ~ ~

m c5 cm ca cza ca ea ch o

. 1

, cf n ~ ~

s j ANO-1 MAJOR IMPACTS ON CAPACITY FACTOR

~

ITEM %CF SOLUTION STATUS OF CORRECTIVE ACTION

! MFW Pump Controls 2.7 New Control System Installed iR7 j Training / Procedure Complete Upgrades RCP Seals 2.4 Improved Maintenance 1985 8 1986 Procedures / Parts l New Design 1R8 (BSWOG) -

SG Level Limit 2.4 Water Slap 1R7 (BGWOG) l l' SG Tube Leaks 2.2 Preventative Sleeving iR6, iR7 Secondary Improvement Numerous Projs T/G Trips 2.0 Mod. Test Procedure Complete New Monitoring system iR7, iR8 (B&WOG)

Dropped CRDs/ Repair i.3 Under Investigation -

cm ,g r-ca a

_; . . o m a r, . . .

ANO-i MAJOR IFACTS ON CAPACITY FACTOR (Continued) f i,-

ITEM %CF SOLUTION l;' STATUS OF

- C0FEECTIVE ACTION Extraction Steam Piping 0.8 New Desigri 1R7 ll 0.3 Re-Designed ' Complete iR6 l

RCP Sensing Lines -

Condenser Chemistry 0.3 Better Monitoring iR7 Equipment u

t i

l l

l 1

l

c, r r, e , on a a

.- { < , -

i i

! JISTR BUTION OF CAPF A_ FUN JS .

ANO 60

! . ll-

~' -

to -

l ss.s71

a; so - as.ser '

N j f 26.896 22.732 i y '/ '

l ~ 2. -

,,...a /

1. .

3 /m

/, %y p/,

l

, 2%x' 1986 ik 1986

'/A' 1987 Q REGULATORY Q PIANT IMPROVEMENTS l

i

1987 OUILOOK t

o NO REFUELING OUTAGES o TRIP REOUCTION AND PLANT PERFORMANCE

.p IMPROVEMENTS

~

o - SAF8TY' AND PERFORMANCE IMPROVEMENT PROGRAM y X MODIFICATIONS o INPO EVALUATION A. o TURBINE ROTOR REPLACEMENT ANO-i G ANO-2 lJ -

0 STARTUP BOILER REPLACEMENT 7

i' o

4.

h,

< e, a

i .

t

.. . ~ '

/? c A .ao-c .5 ARKANSAS POWER AND LIGHT COMPANY POLICY STATEMENT ON APPLICATION OF 18CFR 59.59 l

l REVIEWED & RECOMMENDED FOR APPROVAL / /fdI / 7[9/es

~ ~PSC' CHAIRMAN DATE APPROVED

/-

H N/D$Oa /7/9ffff

/ .M NO SITE DIRECTOR / D'TE A

REVIEWED & RECOMMENDED FOR APPROVAL h /4,[2.7/8(.

S C DATE

{

l APPROVED FOR USE JM + ///#///

VICEPRESpNT,NtfCLEAROPERATIONS / DATE I

I i

( \

i

e .

POLICY STATEMENT ON 10CFR 50.59 PURPOSE OF STATEMENT To establish an AP&L policy on application and use of 10CFR 50.59 to be used by personnel and to achieve consistency in application.

REGULATION 10CFR 59.59 - Changes, tests and experiments.

(a)(1) The holder of a license authorizing operation of a production or utilization facility may (i) make changes ir the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests or experiments not described in the safety analysis report without prior commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question.

(a)(2) A proposed change, test or experiment shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the i

margin of safety as defined in the basis for any technical specification is reduced.

(b) The licensee shall maintain records of changes in the facility and of changes in procedures reade pursuant to this section, to the extent that such changes constitute changes in the facility as described in the safety analysis report or constitute changes in procedures as described in the safety analysis report. The licensee shall also maintain l records of tests and experiments carried out pursuant to paragraph (a) of this section. These records shall include a written safety evaluation which provides the bases for the determination that the change, test or experiment does not involve an unreviewed safety question. The licensee shall furnish to the appropriate NRC Regional

Office shown in Appendix 0 of Part 20 of this chapter with a copy to i

the Director of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20055, annually or at such shorter intervals as may be specified in the license, a report containing a brief description of such changes, tests, and experiments, including a summary of the safety evaluation of each. Any report submitted by a licensee pursuant to this paragraph will be made a part of the public 06/27/86

record of the licensing proceeding. In addition to a signed original, 39 copies of each report of changes in a facility of the type described in 50,21(b) or 50.22 or a testing facility, and 12 copies of each report of changes in any other facility, shall be filed. This record of change in the facility shall be maintained until the date of termination of the license, and records of changes in procedures and records of tests and experiments shall be maintained for a period of five years.

(c) The holder of a 1.icense authorizing operation of a production or utilitization facility who desires (1) a change in technical specifications or (2) to make a change in the facility or the procedures described in the safety analysis report or to conduct tests or experiments not described in the safety analysis report, which involve an unreviewed safety question or a change in technical specifications, shall submit an application for amendment of his license pursuant to S 50.90.

DEFINITIONS

" ACTIVITY" means the following:

1. A modification (including temporary modifications) to:
a. the plant;
b. procedures;
c. methodologies (methods used to conduct an evolution even if not described in procedures);
d. analyses;
e. organization;
2. A test; or
3. An experiment.

" BASIS" means the section of the Technical Specification entitled " bases."

It also includes information in the Safety Evaluation Report issued by '!RC accompanying changes to the Technical Specifications.

" CHANGE" means the following:

If the LICENSING BASIS DOCUMENTS discuss information pertinent to the ACTIVITY, and the ACTIVITY will result in that information being incorrect or inaccurate, or a requirement in those documents would nut be met, then the ACTIVITY is considered to result in the need to CHANGE the LICENSING BASIS DOCUMENTS 06/27/86

"50.59 DETERMINATION" is the act of determining whether a proposed ACTIVITY will result in a CHANGE to the facility or procedures as described in the

~

LICENSING BASIS DOCUMENTS. It is also to determine if a proposed ACTIVITY, that involves tests or experiments not described in the SAFETY ANALYSIS REPORT, could result in a degradation of the margins of safety during normal operations or anticipated transients, as described in the SAFETY ANALYSIS REPORT, or a degradation of the adequacy of structures, systems or components to prevent accidents or mitigate accident consequences as described in the SAFETY ANALYSIS REPORT.

"58.59 EVALUATION" is the act of preparing a written safety evaluation on the impact of a proposed change, test or experiment for which the 59.59 DETERMINATION concluded that an evaluation was required. The purpose of the 59.59 EVALUATION is to determine if an unreviewed safety question is involved. If so, NRC review and approval will be required prior to proceeding.

"58.59 REVIEW" is the process of performing the 56.59 DETERMINATION, 50.59 EVALUATION (if required); and the applicable reviews by the Plant Safety Committee, and Safety Review Committee, and AP&L Management.

" LICENSING BASIS DOCUMENTS" means:

In general the term Licer. sing Basis Documents means the formal documented records which form the basis used by the NRC Licensing branch (NRR) in order to grant, amend or modify the Operating License.

These may include certain licensing related correspondence from AP&L to NRC or Safety Evaluation Reports from NRC which were used to justify granting of the origirial license, or amending or modifying the Operating License.

Specifically the LICENSING BASIS DOCUMENTS applicable to 50.59 reviews are:

a. the SAFETY ANALYSIS REPORT
b. the OPERATING LICENSE "0PERATING LICENSE" means:

, a. The Operating License issued by NRC (one for each unit)

b. The Technical Specifications (Appendix A to the Operating License)
c. Orders (including Confirmatory Orders) issued by NRC which modify the License.
d. NRC Safety Evaluation Reports issued in support of the original Operating License and subsequent amendments.

06/27/86

. a

" SAFETY ANALYSIS REPORT" means:

a. The multi-volume document entitled Safety Analysis Report (SAR). There is one for each unit.
b. The Emergency Plan (referenced in Section 12.3.1 of the ANO-1 SAR, and Section 13.2 of the ANO-2 SAR). *
c. The QA Manual for Operations (referenced in Section 1.6 of the ANO-1 SAR, and Chapter 17 of the ANO-2 SAR). *
  • 10CFR50.54 specifies additional review requirements for the Emergency Plan, the QA Manual, and the Security Plan. Those review requirements are not covered by this policy.

PURPOSE OF THE REGULATION The purpose of 10CFR 59.59 is to provide criteria to be used by a licensee in determining whether an ACTIVITY may be conducted without prior NRC approval or whether NRC approval must be obtained prior to conducting the ACTIVITY. It also establishes reporting requirements for notification of the NRC of ACTIVITIES (allowed by 59.59) which were conducted without prior NRC approval.

WHEN NRC PRIOR REVIEW IS REQUIRED NRC review and approval _i_s, s required prior to conducting an ACTIVITY which would result in a change to the technical specifications or the existence of an unreviewed safety question.

WHEN NRC PRIOR REVIEW IS NOT REQUIRED ,

NRC review and approval is,not required if the ACTIVITY to be conducted does not result in a change to the technical specifications or the existence of l an unreviewed safety question.

WHEN A 50.59 REVIEW IS REQUIRED A 59.59 DETERMINATION is required prior to conducting an ACTIVITY.

l A 59.59 EVALUATION may be required prior to conducting an ACTIVITY depending on the outcome of the 59.59 DETERMINATION.

METHODOLOGY OF 59.59 REVIEWS Prior to conducting an ACTIVITY, a 59.59 DETERMINATION will be conducted and documented. Documentation of the review will consist of an indication of the portions of the LICENSING BASIS DOCUMENTS reviewed and whether any CHANGES to those documents are required.

1 06/27/86 1

o -  !

If no CHANGES to the LICENSING BASIS DOCUMENTS are indicated and the ACTIVITY conforms to the general requirements of the LICENSING BASIS DOCUMENTS, the 50.59 DETERMINATION is negative, a 50.59 EVALUATION is not required, and the ACTIVITY may proceed without prior NRC notification, review, and/or approval.

If the 58.59 DETERMINATION indicates that the ACTIVITY would make information in a LICENSING BASIS DOCUMENT no longer true or accurate or would violate a requirement stated in that document, a 59.59 EVALUATION must be conducted prior to conducting the ACTIVITY.

The 59.59 EVALUATION will determine whether the result of the ACTIVITY is or is not within the basis on which the OPERATING LICENSE was issued and continued in effect by NRC.

The 59.59 EVALUATION must be written and must provide the basis for determination that the proposed ACTIVITY does or does not involve an unreviewed safety question. A simple statement of conclusion in itself is not sufficient; however, depending upon the significance of the CHANGE, the EVALUATION may be brief.

If the 58.59 EVALUATION indicates that the ACTIVITY is within the basis on whicn the OPERATING LICENSE was issued and continued by the NRC, the ACTIVITY may proceed without prior NRC notification, review, and/cr approval.

If the 59.59 EVALUATION indicates that the ACTIVITY is not within the basis on which the operating license was issued and continued in effect by the NRC (an unreviewed safety question exists), or a CHANGE to the Technical Specifications is required, NRC approval must be obtained prior to conducting the ACTIVITY.

WHO CAN CONDUCT A 50.59 REVIEW Anyone within AP&L (who is qualified and designated as such by the PSC or SRC) may conduct a 59.59 REVIEW. Outside resources (such as B&W, CE, etc.)

may be used as an input to the 50.59 REVIEW. The 58.59 REVIEW and its results must follow the established approval process.

50.59 REVIEWER QUALIFI(fTION An individual may be qualified to conduct 50.59 REVIEWS as follows:

l

1. Attend Training; and,
2. Pass a Certification Exam; and,
3. Meet the minimum experience requirements of ANSI /ANS-3.1-1981 for their function. A minimum of 6 months work experience related to ANO is also required for those individuals for which ANSI /ANS-3.1-1981 does not require plant specific experience.

Requalification is required every two years.

l 06/27/86

~_ .

NOTE: The designated 50.59 REVIEWER should seek assistance from other designated 50.59 REVIEWERS when the scope of the 50.59 REVIEW extends beyond his area of expertise.

NOTE: At the discretion of the Plant Safety Committee or Safety Review Committee, certification may bc suspended or revoked.

Notification of such action, including minimum requirements for recertification, shall be issued to the individual's supervisor for resolution.

APPROVAL OF 59.59 REVIEWS The Plant Safety Committee (PSC) is required to review 50.59 DETERMINATIONS on ACTIVITIES that affect nuclear safety and all 50.59 EVALUATIONS prior to their implementation. The results of the 59.59 REVIEW are provided to the management approval authorities specified in the Technical Specifications together with a recommendation for approval or disapproval. In the case of tem,00rary procedure changes, PSC review and Management approval can be after implementation (up to 14 days), provided the change did not involve intent.

A similar exception i, allowed for temporary modifications such as jumpers and lifted leads.

The Safety Review Committee is responsible for review of the results of 59.59 EVALUATIONS to verify that the ACTIVITIES did not constitute an unreviewed safety question. This review is not required to take place prior to implementation but should be completed within a reasonably short interval (30-60 days) following PSC review.

REPORTING OF 50.59 REVIEWS TO THE NRC The Nuclear Engineering & Licensing Section is responsible for submittal of a report of all ACTIVITIES that require a 50.59 EVALUATION (conducted without prior NRC review and approval) to the NRC on an annual basis. This may be accomplished via a special report or via the annual SAFETY ANALYSIS REPORT update.

IMPLEMENTATION GUIDANCE Attachment 1 " Guidance for Implementation of 10CFR 50.59" is adopted for use by AP&L personnel in conducting 59.59 REVIEWS.

06/27/86

o.
  • 1 ATTACHMENT 1 GUIDANCE FOR IMPLEMENTATION OF 10 CFR 50.59 PURPOSE This document provides guidance on implementation of AP&L's Policy Statement on 10 CFR 50.59. The bulk of this guidance is taken directly from NRC's internal guidance on 10 CFR 50.59 with some changes specific to AP&L.

DISCUSSION 10 CFR 50.59 is composed of three essential parts:

a. Paragraph (a)(1) is permissive in that it allows changes to be made to ANO and its operation as described in the Safety Analysis Report (SAR) without prior NRC approval, provided change in Technical Specifications (TS) is not involved or an "unreviewed safety question" does not exist.
b. Paragraph (b) requires record maintenance of changes made under

! the authority of Paragraph (a)(1). These records must include a written safety evaluation which provides the basis for determining whether an unreviewed safety question exists.

c. Paragraph (c) requires that proposed changes in Technical Specifications be submitted to the NRC as an application for license amendment. Likewise, proposed changes to the facility or procedures and the proposed conduct of tests which involve an unreviewed safety question must be submitted to the NRC as an 4 application for license amendment.

Within the context of this guidance, any proposed cnange to a system or procedure, as described in the SAR, either by text or drawing should be reviewed to determine whether it involves an unreviewed safety question.

Changes may involve an unreviewed safety question even though they are "beyond the second isolation valves," or they do not serve a normal safety-related function, since alteration may introduce an unreviewed safety question.

When performing 10CFR50.59 DETERMINATIONS, one should decide if the LICENSING BASES DOCUMENTS are affected. In reaching this decision, one should ensure that generally applicable requirements (e.g., Fire Protection, Environmental Qualification, Seismic, drainage, etc.) are met. As an example, an ACTIVITY could be proposed to add a non-safety grade room cooler which would not affect LICENSING BASES DOCUMENTS. However, if the design did not conform to applicable seismic or II/I requirements, the ACTIVITY could involve an unreviewed safety question and a 50.59 EVALUATION would be required. ..

l I

A-1

9 s l

Maintenance activities which do not result in a change to a system (permanent or temporary), or which replace components with replacement parts procured to the same (or equivalent) purchase specification, do not normally require a written 50.59 evaluation to meet 10 CFR 50.59 requirements. However, if components described in the SAR are removed, or their function is altered, or if substitute components are utilized, or if changes remain following completion of a maintenance activity, an evaluation is normally required to meet the provisions of 10 CFR 50.59 and the change must be reported to the NRC as required by 10 CFR 50.59(b).

In all cases requiring a written 50.59 evaluation, the evaluation must provide the basis for determination that the proposed change does or does not involve an unreviewed safety question. A simple statement of conclusion in itself is not sufficient; however, depending upon the significance of the change, the 50.59 evaluation may be quite brief.

EXAMPLES Listed Below are examples of various changes to facilities, systems, procedures, and tests which are typical of those requiring a 10 CFR 50.59 evaluation and those which do not require an evaluation under the requirements of 10 CFR 50.59.

a. Changes in the Facility As Described in the Safety Analysis Report. This pertains to any changes in the facility which would, if accomplished, make information in the SAR no longer true or accurate or would violate a requirement stated in the SAR. This would apply to components, systems, and structures described either in the written portion of the SAR or in the drawings contained therein. Contrasting examples of each case are:

(1) Components. Replacement of a thermocouple in the diesel high-bearing temperature automatic shutdown circuitry (if such a component were described in the SAR) with one made by the same manufacturer, but encompassing different response characteristics, would require a 50.59 evaluation if those characteristics were described in the SAR.

On the other hand, replacement of a thermocouple in the diesel high-bearing temperature automatic shutdown circuitry (if such a component were described in the SAR) with one encompassing equivalent response characteristics, but made by a different manufacturer, would not require an evaluation under the requirements of 10-CFR 50.59 unless the manufacturer was named in the SAR.

Replacement of a Rosemount 1153 pressure transmitter with another 1153 pressure transmitter of the same type would not require an evaluation.

k Regla/ cement of a Rosemount 1153 pressure transmitter with a psemount1154pressuretransmitterwouldrequireawritten b0.59 evaluation to determine that the response characteristics of the new transmitter are acceptable.

A-2

a v Replacement of a LPSI motor with an identical motor would not require a written safety evaluation.

Replacement of a LPSI motor with another with different characteristics (e.g., horsepower, rpm, head requirements, etc.) would require a written 50.59 evaluation.

(2) Systems. Modifications of the diesel shutdown circuitry (described in the SAR) to provide an automatic diesel shutdown on high-bearing temperature (shutdown feature not described ir, application) would require an evaluation to meet the requirements of 10 CFR 50.59 if methods of initiating automatic diesel shutdown are described in the SAR. On the other hand, if the methods of initiating automatic diesel shutdown are not described in the SAR, specific automatic shutdown features may be rendered inoperable or new ones added without the conduct of 50.59 evaluation.

(3) Structures. The erection or deletion of any concrete block shield wall within the scope of drawings included in the SAR would require a 50.59 evaluation. If such walls were not within the scope of any drawings included in the SAR, then no evaluation would be required except in the case of deletion of a wall specifically mentioned in SAR text or erection of a wall in a location where the SAR says there are no wsils or where SAR described features would be undermined.

(4) Jumpers / Lifted Leads.

1. If it is determined that use of a jumper / lifted lead results in a change to the facility, as described in the SAR, then a safety evaluation is required. This approach should apply to all types of temporary modifications. Generally, if a plant system is changed by use of jumpers / lifted leads so that it will function differently than described in the SAR, a 50.59 evaluation would be required.

On the other hand, use of jumpers / lifted leads that result in plant conditions already analyzed and approved by NRC would not require a 50.59 evaluation. For example, bypassing protection channels in a manner already described in the SAR would not constitute an unreviewed safety question and would not require an evaluation under the requirements of 10 CFR 50.59. It is expected that only a small percentage of jumpers / lifted leads will require a written 50.59 evaluation.

A-3

a w s>

b. Changes in Procedures As Described.in the SAR. This pertains not only to procedures discussed in the initial operations and organizational chapters of the SAR, but also to other procedural-type commitments, such as the Emergency Plan and modes and sequences of plant operation described in the SAR. If a procedure results in a deviation from the steps listed in the SAR or will result in a change to a system as described in the SAR, then a 50.59 evaluation should be performed. Contrasting examples of the above follow.

(1) If in the description of the radioactive waste system in the SAR, it is stated that the Sh'ift Supervisor will authorize all radioactive liquid releases, an evaluation to meet the requirements of 10 CFR 50.59 would be required before assigning this function to another individual. On the other hand, if the SAR merely states that radioactive liquid releases will be authorized as detailed by plant procedures, the licensee's redesignation of the authorization function would not require a 50.59 evaluation.

(2) If the reactor startup procedure, as described in the SAR, contains eight fundamental sequences, decisi,on to eliminate one of the sequences would require an evaluation to meet the 10 CFR 50.59 requirements. On the other hand, if the eight fundamental sequences were consolidated but did not alter the basic functions performed, it would not be necessary to conduct a 50.59 evaluation.

c. Conduct Tests and Experiments Not Described in the SAR. This pertains to the performance of an operation not described in the SAR which could have an adverse effect on safety-related systems.

Contrasting examples of such tests or experiments are:

(1) Some plants in the startup testing program have performed a deboration to critical with all rods inserted. Since this test is performed without deference to the "one stuck rod criterion," an evaluation to meet the requirements of 10 CFR 50.59 would be required if the test is not delineated in the SAR. Since this test may decrease the margin of safety defined in the TS basis, it should in most instances, be classified as an unreviewed safety question. On the other hand, a test to demonstrate the calibration of the nuclear l instrumentation system by performance of a secondary plant

heat balance would not require a 50.59 safety evaluation, even if such a test was not delineated in the SAR, since the test does not involve an abnormal mode of operation.

l l

l l

l l

A-4

a > q , _

(2) A test to determine if the boric acid evaporator may also be used for concentration of the steam generator blowdown effluent (function not described in the SAR) would require an evaluation to meet the requirements of 10 CFR 50.59, since secondary system chemicals could possibly have a deleterious effect on some components within the reactor coolant pressure boundary. On the other hand, an experiment to determine the decontamination factor of the liquid waste concentrator with influent activities of 10 2 Ci/ml and 10 5 Ci/ml would not require a 50.59 evaluation since such an experiment would not represent departure from normal operational modes.

Tests and experiments not described in the Safety Analysis Report includes only those tests and experiments that would have been described in the Safety Analysis Report had they been anticipated at the time the Safety Analysis Report was written. Such tests and experiments would be restricted to those which could degrade the margins of safety during normal operations or anticipated transients, or degrade the adequacy of structures, systems or components to prevent accidents or 1

mitigate accident consequences.

Test - The process of submitting a statement to such conditions or operations as will lead to its proof or disproof, or to its acceptance or rejection.

Experiment - An operation carried out under controlled conditions in order to discover an unknown effect.

1 A-5

. . - . . . _ . - ._. .

Retrieved from "https://kanterella.com/w/index.php?title=ML20212A622&oldid=3246533"