ML20195J437

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Confirms 881107 Conversation Re Operability of DHR Sys During Both Long Term Cooling & Shut Down Cooling Modes of Operation & Operability of Containment Penetration Ventilation Sys
ML20195J437
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 11/07/1988
From: Martin R
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To: Campbell G
ARKANSAS POWER & LIGHT CO.
Shared Package
ML20195J432 List:
References
CAL-88-19, NUDOCS 8812020176
Download: ML20195J437 (47)
v • d • e


Text

Enclosurel LN 7 704q In Reply Refer To:

Dockets: 50-313/ CAL 88-19 Arkansas Power & Light Company I

ATTN: Mr. Gene Campbell Vice President. Nuclear Operations P.O. Box 551 Little Rock, Arkansas 72203 Gentlemen:

SUBJECT:

CONFIRMATION OF ACTION LETTER This is to confinn the conversation between Mr. J. M. Levine, Arkansas 'wer &

Light Company and Mr. L. J. Callan, USNRC, Region IV, on Noventer 7, Isod On the basis of this conversation, it is my understanding that the Arkansas Power

& Light Company is addressirig the following issues:

1.

Operability of the decay heat removal (DHR) system during both long tenc cooling and shut down cooling modes of operation; 2.

Operability of the containment penetration ventilation system; 3.

Operability of the safety systems affected by the nozzle loading study performed by the Bechtel Company in 1985; and 4.

Internal review of your corrective action systems to (! Jure that safety or operability concerns have been adequately disposition =d.

It is my understanding that the Arkansas Power & Light Company will not restart ANO, Unit 1 until members of your staff meet with the NRC staff to discuss these four issues. The location and date of this meeting will be established by separate correspondence.

If your understanding of the above is different, plesse notify me insnediately.

Sincerely.

S Rober' D. Martin Regional Administrator cc:

(seenextpage)

CERT 1FIED MAIL - RETURN RECE1PT REQUESTED h0$

C:PSA D:DR '

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DDChan6erlain;df LJCallan n

W 11/733 11/1/88 11/7/88 h

lY 8812O20176 881107 PDR ADOCK 05000313 P

PDC

a' N ovember 18, 1988 ARKANSAS 1 STARTUP ISSUES HEETING AGENDA 8:00 a.m.

Introductions and Statement of the purpose of the meeting 8:10 a.m.

Presentation by AP&L on the safety-related pipe nozzle issue 10:00 a.m.

Break for 10 minutes 10:10 a.m.

Pipe nozzle issue ( continued) 11:00 a.m.

Decay Heat Renoval System - Long term cooling and shutdown cooling modes of operation 11:45 a.m.

Lunch 12: 45 p.m.

Penetration room ventilation system operability 1:30 p.m.

Results of licensee review of corrective action systems to ensure that safety or operability concerns have been adequately dispositioned.

2:30 p.m.

End i

1 2-1 i

i l

--g-o AP&L Participants Jim Levine Executive Director, Site Operations George Jones General Manager, Design Engineering Dan Howard Manager, Licensing Bill Eaton Manager, Mechanical / Civil /

Structural Design Rick Lane Manager, AN0 Engineering Dale James Supervisor, Licensing Bill Greeson Supe visor, Design E,ngineering

( tructural)

Cowper Chadbourn Des.gn Engineer (Structur > i Bill Rogers Supervisor, Design Engineering (Mechanical)

Jay Miller Nuclear Engineer Daryl Williams DesignEngineer(Mechanical)

Jim Vandergrift Manager, Operations Bill Garrison Operations Tech. Support, AH0-1 Dan Preschong Supervisor, Design Engineering (Civil)

Larry Humphrey General Manager, Nuclear Quality Lingagoud Memula Bechtel Power Corporation, Chief Engineer, Plant Design David Maxham Babcock and Wilcox Company, Engineer e

0 0

1 2-2

.s' s

NRC Participants NRR Gary Holahan Acting Director, DRSP Les Rubenstein Assistant Director for RIV Reactors and Special Projects, DRSP Jose Calvo Director, Project Directorate IV Craig Harbuck Project Manager for ANO 182 Lawrence Shao Director, Division of Engineering and System Technology (DEST)

James Richardson Assistant Director for Engineering, DEST Tad Marsh Chief, Mechanical Engineering DEST John Craig Chief, Plant Systems Branch, DEST Jai Rajan Tech. Reviewer, Hechanical Engineering Branch, DEST Bob Jones Chief, Section B, Reactor Systems Branch, DEST Bill LeFaye Tech. Reviewer, Reactor Systems Branch, DEST EDO Thomas Martin Chief, Regional Operations Region IV Bill Beach Deputy Director, Division of Reactor Projects Dwight Chamberlain Chief, Project Section A Ian Barnes Chief, Materials and Quality Programs Section g.3 i

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ARKAkSAS NUCLEAR ONE - UNIT 1 N0ZZLE LOAD ISSUE OUTLINE OF PRESENTATION PURPOSE OF MEETING TO PROVIDE STATUS OF OPERABILITY EVALUATIONS ON ANO-1 N0ZZLES AND BASIS OF THE EVALUATIONS 0VERVIEl!

ACCEPTANCE CRITERIA EQUIPMENT CATEGORIES STATUS OF UNIT 1 N0ZZLE EVALUATIONS CONCLUSIONS

  • O e

3-t

OVERVIEH LISTING OF N0ZZLES BASED ON REVIEW 0F SAFETY RELATED EQUIPMENT DATA BASES TOTAL N0ZZLES LISTED 236 84 ARE SMALL BORE AND 5 ARE NON-0 DEVELOPED ACCEPTANCE CRITERIA FOR THE PURPOSE OF DETERMINING OPERABILITY OF ALL N0ZZLES GENERATED INDIVIDUAL CALCULATIONS TO DOCUMENT THE EVALUATIONS OF EACH N0ZZLE, CALCULATIONS REVIEHED IN ACCORDANCE WITH OUR DESIGN PROGRAM,

  • O O

31

ACCEPTAf1CE CRITERIA MEETS VENDOR ALLOWABLES A COMPARISON OF PIPING INDUCED LOADS FROM PIPING STRESS ANALYSIS AGAINST DOCUMENTED VENDOR ALLOWABLES MEETSEMPERICALLYDEVELOPEDCRITERIA(DESIGNGNIDLINES C0hPARIS0N OF PIPING INDUCED LOADS FROM PIPING STRESS ANALYSIS AGAINST DESIGN GUIDELINE VALUES

~

GUIDELINES DEVELOPED BY BECHTEL POWER CORPORATION TO DETERMINE ACCEPTABILITY OF PIPING

~

LOADS ON N0ZZLES VALUES USED AS ALLOWABLES ARE SHOWN TO BE CONSERVATIVE BASED ON B0UNDING ANALYSIS AP8L CALCULATI0flS THAT VERIFY ACCEPTABILITY OF PIPING LOADS ON N0ZZLES NOTE --

SMALL BORE N0ZZLES WERE EXCLUDED FROM FURTHER EVALUATION AFTER VERIFICATION AS SMALL BORE BECAUSE OF THE CONSERVATIVE DESIGN PPACTICES UTILIZED IN ROUTING AND SUPPORTING SMALL BORE PIPING s-3

i

.,oc p 'D EQUIPMENT CATEGORIES OF AND-1 N0ZZLES 69 PUMPS 42 HEAT EXCHANGERS e

60 TANKS 35 HVAC 16 NSSS 14 MISCELLANE0US O

236 TOTAL O

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s, 0

e STATUS OF AND-1 N0ZZLE EVALUATION 236 N0ZZLES IN SCOPE OF EVALUATION 0F TOTAL NtlMBER, 814 ARE SMALL DORE 118 N0ZZLES DETE31NED OPERABLE BY VENDOR ALLOMABLES lli N0ZZLES DETERMINED OPERABLE BY DESIGN GUIDELINES 15 N0ZZLES DETERMINED OPERABLE BY AP8L CALCULATION /EVALVATION 5 ARE NON-0 0

6 0

0 0

3-i

CONCLUSIONS CONCLUSIONS ARE THAT NO OPERABILITY CONCERNS EXIST G

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AND-i Per,etration Room Ventilation e

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Design Aspects of Penetration Room Ventilation System (PRVS)

Design Basis of PRVS:

O Maintain the Penetration Rooms at a negative pressure to contain Reactor Building (RB) leakage following a Loss-of-Coolant Accident or Maximum Hypothetical Accident such that 1/2 RB leakage is filtered by the PRVS o

Accident analysis assumes 0.2% Reactor Building leakage per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident and 0.1% per day thereafter.

PRVS filters 0.1% RB leakage per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident and 0.05% per day-thereafter, which correlates to an evacuation rate of 1.25 scfm.

o 10CFRiOO requires an exclusion area dose limit of

)

300 Rem thyroid I

o Off-site dose limits must remain below 10CFR100 Jimits 41

Oesign Features to Perfarm Design Basis Function o

Air leakage into any of the Penetration Rooms is discharged to a vent by one of two full-size' constant speed fans.

If lead fan fails to start within 20 seconds of an Engineered Safeguards

signal, the standby fan starts.

i o

Filter assemblies consist of a prefilter, high efficiency particulate HEPA filter, and a charcoal l

bed to filter the air removed from the Penentration Rooms.

Filter design rated capacity is 2000 scfm i

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0 Flow rate and room vacuum instrumentation with Control Room indication is provided.

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O System design flow rate is 1800 scfm based on the normal system in-leakage and meeting the following design requirements:

1 I

Acceptable pressure differential on filters i

Negative pressure in all penetration rooms i

Evacuation of 1.25 scfm filtered reactor 1

building leskage Nhich is equivalent to 50%

l of the RB leakage rate of 0.2% per day i

. 2:C

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ROOMS ARE PPG PEN ROOM ELEC PEN ROOM ELEC PEN ROOM INTERCONNECTED E

PRESSURE gut 3 DIFFERENTIAL TRANSMITTER O

O FLOW ELEMENT N

A FILTER A n

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O TO ATuOSeNERE STANDEW FAN V

B i

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PPG PEN ROOM ELEC PEN ROOM ELEC PEN ROOM

' ROOMS ARE g

INTERCONNECTED BY DUCTWORK

+

PEN ETRATION ROOM VENTILATION SYSTEM SCH EMATIC 4

I NOTE:

NORTH ROOMS ARE INTERCONNECTED BY DUCTWORK AND SOUTH ROOMS ARE INTERCONNECTED BY DUCTWORK.

UPPER SOUTH UPPER NORTH i

ELECTRICAL ELECTRICAL PENETRATION ROOM PENETRATION ROOM REACTOR e

BUILDING e

LOWER SOUTH LOWER NORTH ELECTRICAL ELECTRICAL PENETRATION ROOM PENETRATION ROOM SOUTH NORTH PIPING PIPING PENETRATION ROOM PENETRATION ROOM O

RESSU DIFFERDmAL TRANSMITTER AN O-1 PENETRATION ROOM ARRANGEMENT w

I Results of ANO-1 Integrated Engineered Safeguards Test Performed 11-3-88 o

System Performance i

Negative pressure validated at all but one door by observing air flow direction i

Measured flow rate above the system design flow rate of 1800 scfm 0

Problems Identified Air outleakage was experienced at entry door of the Lower South Electrical Penetration Room due to ventilation system in Lower South

Electrical Equipment Room i

i Lower South Piping Penetration Room door would not close completely, but outleakage was not experienced Upper South Piping Penetration Room suction check valve was stuck closed at start of test l

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REACTOR 1

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LOWER SOUTH ELECTRICAL PENETRATION ROOM DOOR 59 l*

LOWER SOUTH ELECTRICAL 1

EQdePMENT ROOM i-3 i

PLAN VIEW OF LOWER SOUTH ELECTRICAL

.c PENETRATION ROOM WHERE DOOR OUTLEAKAGE 4

23 WAS EXPERIENCED 4

Short-Term Actions Completed to Enhance System Performance o

Repaired and sealed penetration room doors.

o Performed preventive maintenance on and adjusted suction check valves i

t o

Modify ductwork in room adjacent to Lower South Electrical Penetration Room j

i o

Performed preventive maintenance on penetration room floor drain backwater valves o

Sealed conduit in Upper North Piping Penetration Room Wall t

0 Repaired gap seal between penetration roomo and Reactor Building l

I 1

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Results from Design Verification Test o

Validated ES initiation and response of system components o

System design flow rate verified at 1800 scfm

+/- 10%

o Negative pressure validated in all penetration ronmG Instrumentation t

Flow direction verification at interface areas a

Verified system interlocks Lead to standby switching o

Establish setpoints Low flow Vacuum t

t 4-1

i Short-Term Surveillance Test Requirements o

Include acceptance criteria in monthly surve.illance test for vacuum and for fan flow rate and bypass filter cooling flow rate o

Include air flow monitoring at doors in 18-month surveillance o

Include confirmation of check valves pening during 1

18-month surveillance tests by visu; inspection l

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Long-Term Actions o

Clarification of SAR and TS o

Upgrade Preventive Maintenance requirements 9

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SUMMARY

AND CONCLUSIONS o

The current operability analysis and testing of the system indic tes that ANO-i will not operate.outside o

the licensing basis for offsite dose limits in the event of a maximum hypothetical accident, 1

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ANO-1 DECAY HEAT REMOVAL O

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DECAY HEAT REMOVAL SYSTEM FUNCTIONAL ASSESSMENT INTRODUCTION SYSTEM DESCRIPTION REGULATORY REQUIREMENTS SYSTEM OPERATIONS o

LOW PRESSURE INJECTION (LPI) MODE o

LPI FAILURE MODES S EFFECTS ANALYSIS l

0 DECAY HEAT REMOVAL (DHR) MODE O

DHR FAILURE MODES S EFFECTS ANALYSIS SYSTEM ENHANCEMENTS l

SUMMARY

S CONCLUSIONS i

l 5-2

INTRODUCTION O

DECAY HEAT REMOVAL (DHR) OR RESIDUAL HEAT REMOVAL (RHR)

IS RECOGNIZED AS A CUBJECT OF HIGH INTEREST TO BOTH THE NRC AND THE INDUSTRY.

o WE WILL ADDRESS HOW THE DECAY HEAT REMOVAL FUNCTION IS ACCOMPLISHED AT ARKANSAS NUCLEAR ONE UNIT i (ANO-1) AND HOW THE REGULATORY AND OPERABILITY REQUIREMENTS ARE SATISFIED.

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THE LPI/DHR SYSTEM IS A DUAL MODE SYSTEM WHICH SATISFIES MULTIPLE FUNCTIONS.

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LPI DESIGN BASIS INJECT B0 RATED WATER INTO THE REACTOR VESSEL FROM THE BWST TO AID IN THE PREVENTION OF FUEL S CLAD DAMAGE IN THE EVENT OF INTERMEDIATE TO LARGE BREAK LOCA (LPI MODE)

RECIRCULATE INJECTION WATER INTO THE REACTOR VESSEL FROM THE REACTOR BUILDING SUMP TO PROVIDE LONG TERM CORE COOLING IN THE EVENT OF INTERMEDIATE TO LARGE BREAK LOCA (LPR MODE)

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SYSTEM DESCRIPTION (Cont'd) o DHR DESIGN BASIS REMOVE DECAY HEAT FROM THE CORE AND SENSIBLE HEAT FROM THE RCS DURING THE LATTER STAGES OF C00LD0WN MAINTAIN RCS TEMPERATURES DURING REFUELING AUXILIARY FUNCTIONS 5-8

DHR MODE

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REGULATORY REQUIREMENTS o

GDC 34 ' RESIDUAL HEAT REMOVAL" REQUIRES:

ABILITY TO REMOVE RESIDUAL HEAT SUCH THAT SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS (SAFDL'S) AND RCS PRESSURE BOUNDARY DESIGN CONDITIONS ARE NOT EXCEEDED SUITABLE REDUNDANCY TO ASSURE THE SYSTEM SAFETY FUNCTION CAN BE ACCOMPLISHED, ASSUMING A SINGLE FAILURE o

FOR ANO-1 THE SYSTEM SAFETY FUNCTION CORRESPONDS TO ' SAFE SHUTDOWN' CONDITIONS WHICH IS HOT SHUTDOWN O

ANO-i 0VERALL RHR FUNCTION IS ACCOMPLISHED UNDER DIFFERING OPERATIONAL MODES AS FOLLOWS:

4 MODE SYSTEM HOT SHUTDOWN STEAM GENERATOR SECONDARY COOLING COLD SHUTDOWN DHR SYSTEM REFUELING O'4', SYCTEM j

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REGULATORY REQUIREMENTS (Cont'd) a GDC 35 ' EMERGENCY CORE COOLING" REQUIRES:

ABILITY TO TRANSFER HEAT FOLLOWING' A LOCA SUCH THAT FUEL AND CLAD DAMAGE DOESN'T INTERFERE WITH CORE COOLING AND CLAD METAL-WATER REACTION IS LIMITED SUITABLE REDUNDANCY TO ASSURE THE SYSTEM SAFETY FUNCTION CAN BE ACCOMPLISHED ASSUMING A SINGLE FAILURE O

FOR ANO-1, EMERGENCY CORE COOLING IS ACCOMPLISHED BY A COMBINATION OF:

HIGH PRESSURE INJECTION (HPI) SYSTEM 4

LOW PRESSURE INJECTION (LPI) SYSTEM CORE FLOOD (CF) SYSTEM o

POST-LOCA LONG TERM DECAY HEAT IS REMOVED THROUGH THE LPI/DHR SYSTEM IN THE LOW PRESSURE l

RECIRCULATION (LPR) MODE l

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REGULATORY REQUIREMENTS (Cont'd) o LPI MODE MEETS LPI DESIGN BASIS GDC 35 REQUIREMENTS OTHER REQUIREMENTS APPLICABLE TO 1

SAFETY-RELATED COMPONENTS (E.G. EG)

TECHNICAL SPECIFICATION REQUIREMENTS o

DHR MODE MEETS DHR DESIGN BASIS 4

TECHNICAL SPECIFICATION REQUIREMENTS ORIGINAL LICENSING REQUIREMENTS (E.G. ACI) i 5-/2 I

REGULATORY REQUIREMENTS (Cont'd) l 0

ANO-1 SAFETY ANALYSIS REPORT (SAR) STATES IN RESPONSE TO GDC 34:

"THE STEAM GENERATORS PROVIDE A LONG TERM CAPABILITY FOR DECAY HEAT' REMOVAL' o

NRC SAFETY EVALUATION REPORT (SER) FOR ANO-i STATES:

i

'WE HAVE DETERMINED THAT REDUNDANCY OF i

COMPONENTS,

VALVES, AND PIPING PROVIDES ADEQUATE PROTECTION FROM THE EFFECTS OF i

A SINGLE ACTIVE OR PASSIVE FAILURE DURING POST-ACCIDENT LONG TERM COOLING '

0 ANO-1 SER CONTINUES BY DESCRIBING THE DHR SYSTEM:

'THE SUCTION LINE TO THE DHR SYSTEM PUMPS CONTAINS THREE ELECTRIC MOTOR-0PERATED GATE VALVES IN SERIES'.

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DHR FAILURE MODES S EFFECTS ANALYSIS o

SINGLE ACTIVE FAILURES EXIST RCS HOT LEG ' DROP LINE' VALVES ACI INDUCED VALVE FAILURES (IN ' DROP LINE')

LPI/DHR COOLER OUTLET G BYPASS VALVES AT

' DRAINED DOWN' RCS CONDITIONS l

o DHR SINGLE ACTIVE FAILURES ARE CONSIDERED RECOVERABLE CAUSING ONLY TEMPORARY INTERRUPTION 0F DHR.

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o NO DHR SINGLE FAILURE WILL CAUSE A FAILURE OF LPI DURING POST-ACCIDENT EMERGENCY OPERATIONS.

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ANO-1 DHR ENHANCEMENTS COMPLETED ACTIONS 0

OPERATING PROCEDURES (0P)

OP 1203.28 ' LOSS OF DECAY HEAT REMOVAL" OP 1104.04 'DHR OPERATING PROCEDURE' OP 1015.02 "DHR S LTOP CONTROL" MODIFIED j

TO ALLOW ACI OVERRIDE WHEN RCS CANNOT.

RE-PRESSURIZE (PRESSURE BOUNDARY OPEN)

PROCEDURE CHANGES TO ALLOW THROTTLING WITH CV 1400/1401 AT ' DRAINED DOWN' CONDITION i

i i

i 5-us

.?

P ANO-i DHR ENHANCEMENTS (Cont'd) o MODIFICATIONS CORRECTED POWER-INTERRUPTION-CLOSURE FAILURE H0DE FOR CV-1050 S CV-1410 IN RCS

' DROP LINE' (1983)

POSITION INDICATION OF CV-1428/1429 IN CONTROL ROOM (1983)

REDUNDANT RCS LEVEL INSTRUMENTATION (1986)

SAFETY PARAMETER DISPLAY SYSTEM (SPDS)

DIAGNOSTICS DISPLAY (INCLUDES DHR PARAMETERS)

(1987 - 1988)

RESTORATION OF CV 1428/1429 TO DESIGN

' FAIL OPEN' CONFIGURATION (1988)

LOW FLOW ALARM SETPOINT CHANGE (1988) a TRAINING SPECIFIC DHR CLASSROOH AND SIMULATOR TRAINING PRIOR TO REFUELING OUTAGES 5 20

)

SHORT TERM ACTIONS o

COMPLETION OF ACI INTERLOCK STUDY o

COMPLETION OF MOV STUDY FOR CV-1428/1429 o

SPDS PARAMETER DISPLAY TO PROVIDE DHR FLOW VERSUS RCS LEVEL VORTEXING LIMIT O

DEVELOP ADDITIONAL SPECIFIC LOSS-OF-DHR i

l GUIDANCE FOR OPERATORS

~

o EVALUATE AND INCLUDE DHR INSTRUMENT ERROR j

IN LEVEL CURVES

)

)

j LONG TERM ACTIONS 1

I o

IMPLEMENT RESULTS OF ACI INTERLOCK STUDY l

i l

l 0

IMPLEMENT RESULTS OF MOV STUDY FOR CV-1428/1429 l

o REVIEW DHR CONFIGURATION FOR ADDITIONAL

)

IMPROVEMENTS IN INSTRUMENTATION AND OPERATOR GUIDANCE TO ADDRESS CONCERNS IDENTIFIED IN j

GENERIC LETTER 88-17 l

3 - z. i I

l n-CONCLUSION r

o LPI FUNCTION FULLY SATISFIES REGULATORY AND DESIGN REQUIREMENTS.

i o

DHR FUNCTION SATISFIES ORIGINAL REGULATORY AND DESIGN REQUIREMENTS; HOWEVER, INCREASED EMPHASIS ON THIS FUNCTION HAS RESULTED IN CONTINUED SYSTEM ENHANCEMENTS.

l l

0 APSL WILL CONTINUE ACTIVELY PURSUING POTENTIAL ENHANCEMENTS IN ACCORDANCE WITH GENERIC LETTER 88-17.

i i

i I

i f-21

g,s o+

STATUS OF AN0 CORRECTIVE ACTION SYSTEM OPEN ITEMS 0

COMPLETE REVIEW OF VARIOUS ACTION TRACKING SYSTEM, OPEN ITEMS HAS BEEN CONDUCTED.

O REVIEW INCLUDED ANO-1 AND ANO-2 0

TRACKING SYSTEM REVIEWED INCLUDED:

- CONDill0N REPORTS

- REPORTS OF ABNORMAL CONDITIONS (RACS)

- NRC COMMITMENT TRACKsNG (COMTRAC)

- PLANT SAFETY COMMITTEE ACTION ITEMS

- SAFETY REVIEW COMMITTEE ACTION ITEMS

- PLANT ENGINEERING ACTION REQUESTS (PEARS)

- ENGINEERING ACTION REQUESTS (BARS)

- QUALITY ASSURANCE / QUALITY CONTROL TRACKING SYSTEM AUDIT / SURVEILLANCE FINDING REPORTS (AFRS/SFRS)

QUALITY CONTROL FINDING REPORTS (QCFRS)

NON-CONFORMANCE REPORTS (NCRS)

QUALITY ACTION REQUESTS (QARS) i

- NRC INSPECTION ITEMS (INTERNAL)

  • POTENTIAL NUCLEAR SAFETY CONCERNS (PNSCS)/10CFR21 ITEMS

- DESIGN CHANGE PACKAGES (DCPS)

- PLANT CHANGES (PCS)

- PLANT IMPACT EVALUATIONS (VLS, PIES, SOERS, OMRS, SERS)

- (E NDTICE1 G-1

ls se

~

P REVIEW CRITERIA 0

COULD FAILURE TO COMPLETE THE ACTIVITY GR OPEN ACTION CAUSE ECU!PMENT SPECIFIED IN TECHNICAL SPECIFICA,T'ON TO BE INOPERABLE OR RESULT IN OPERATION PRCHIBITED iW TS7 0

IS THE ACTIVITY OR OPEN ITEM REl.ATED TO A DETERMINATION OF WHETHER THE PLANT IS OPERATING AS DESCRIBED IN THE SARS OR OTHER LICENSING / DESIGN BASIS DOCUMENTS (IS PLANT OPERABILITY lilDETERMINATE/0UESTIONABLE PENDING EVALUATIONS OF DISCREPANCIES /DEV'1.T! INS RELATED TO DESIGN REQUIREMENTS OR SAFETY ANALYSIS?

O DOES TFE OPEN ACTIGN, ALTHOUGH NOT IN CONFLICT WITH l

THE TS, SARS, OR OTHER LICENSING / DESIGN BASIS DOCUMENTS, I

REQUIRE AN EVALUATION TO DETERMINE IF THE PLANT IS OPERATING IN A SAFE CONFIGURATION, E.G., A VEND 00 NOTICE DESCRIBING A PREVIOUSLY UNRECOGNIZED ACCIDENT l

OR NON-CONSERVATIVE ANALYSIS?

l l

I 4-L l

_ g -=.

n (Y*

,6Y REVIEW RESULTS 0

/.PPROXIMATELY 4750 ITEMS REVIEWED.

O FOR ITEMS WHICH COULD POTENTIALLY MEET REVIEW CRITERIA, SPECIFIC REVIEWS ARE BEING CONDUCTED AND DOCUMENTED.

~

(APPROX. 1200) 4 0

1HESE ITEMS WILL BE RESOLVED BY 1) CLOSURE (E.G.,

INTERNAL AP8L HEATUP RESTRAINT) OR 2) DOCllMENTED BASES'FOR OPERATION PENDING FINAL RESOLUT10N' (APPR0X. 70 REMAINING)

O NO ITEMS HAVE BEEN IDENTIFIED WHICH PRECLUDE RESTAR,T OF ANO-1 CONSISTENT WITH THE ABOVE CRITERIA.

J 6-3

- - - _.

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