ML17156B153
| ML17156B153 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 05/12/1989 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17156B152 | List: |
| References | |
| 50-387-89-01, 50-387-89-1, NUDOCS 8905190417 | |
| Download: ML17156B153 (79) | |
Text
ENCLOSURE 1
~Aendix A
Notice of Violation Pennsylvania Power and Light Company Susquehanna Unit 1
Docket No.
50-387 License No.
NPF-14 During an NRC inspection conducted on January 1 - February 4,
1989, the follow-ing violations of NRC requirements were identified.
Although the excessive cooldown rate cited in Violation. A. 1 is not of technical significance, the violation is significant in that operations supervision, reactor temperature surveillances, and post-trip reviews all should have identified the excessive cooldown rate more promptly.
In 'ccordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions,"
10 CFR Part; 2,
Appendix C
(Enforcement Policy 1987),
the violations are listed below":
Technical Specification 6.8. 1 requires that written procedures important to,safety shall be established and implemented.
- However, in the below described
- cases, procedures were not properly implemented.
1.
Susquehanna Steam Electric Station (SSES)
Surveillance Procedure S0-100-011, Reactor Vessel Temperature and Pressure Recording, step
- 6. 1.4, requires confirmation of compliance with Technical Specifica-tion (T.S.)
4.'4.6. 1. 1 once every 30 minutes by verifying a
maximum cooldown rate of less than 100 degrees F per hour.
Contrary to the above, on January 12,
- 1989, SSES Unit 1 experienced a
cooldown in excess of the required limits in that during the first hour following a
reactor
- scram, the actual cooldown rate at the reactor vessel bottom head drain was about 137 degrees F in the first 45 minutes and was mitigated to 101 degrees F by natural circulation in the first hour.
The fact that the TS limit was exceeded was not discovered until January 16, after the unit had been restarted.
2.
SSES Emergency Operating Procedure EO-100-101 (Attachment A),
"Scram,"
specifies that CRD flow shall be decreased to 20-25 gpm following a reactor scram if a reactor recirculation pump cannot be started.
'ontrary to the
- above, on January 12,
- 1989, following a
reactor
- scram, flow remained't least 60
- gpm, although neither reactor recirculation pump could be started.
OFFICIAL RECORD COPY CIR SUSQUEHANNA 89 0004.0.0 05/06/89
,P~>0e 190417
++0512 PDR P~DCiCl" 0.A000387 0.
Appendix 3.
SSES Procedure, G0-100-003, "Power Operation,"
requires during plant power ascension, the operator to establish automatic feedwater con-trol, verify that reactor vessel level remains at 35 inches and reactor feedwater" pump (RFP) speed decreases accordingly, and open the RFP discharge valves when RFP discharge pressure is within 50-100 psig of reactor vessel pressure.
Contrary to the
- above, on January, 12,
- 1989, during power ascension, the operator did not verify that the feedwater master controller was in automatic and that the vessel level remained at 35 inches with the RFP speed decreasing accordingly.
In addition, the operator opened the RFP discharge valves at 150 psig above reactor vessel versus the 50-100 psig specified in the procedure.
This resulted in a reactor water level transient and plant scram.
This is a Severity Level IV Violation (Supplement I).
B.
Technical Specification 6.8. 1 and Regulatory Guide 1.33 require that writ-ten procedures be established for activities and systems that are import-ant to safety, including plant shutdown and
- cooldown, operation of the Instrument Air and Feedwater
- Systems, and filling and venting of the Shut-down Cooling System.
Contrary to the
- above, in the cases described
- below, adequate procedures were not established.
The fill and vent procedure for the Unit 1
Residual Heat Removal Shutdown Cooling System (SDC) allowed forming a void in the system during fill and vent following a
Residual Heat Removal Pump trip.
This resulted in a
system isolation on high flow during attempts to restore SDC.
SSES Procedures GO-100-'005, "Plant Shutdown From Minimum Power Oper-ation,"
and GO-100-011, "Plant Cooldown Following A Scram,"
allowed an automatic reactor trip to -occur following a manual shutdown when the operator did not bypass the scram discharge volume high level trip prior to resetting the initial manual scram.
The Unit 1
Instrument Air ( IA) System operating procedure did not adequately address a
change in system lineup to provide air to the circulating water pump house common loads from the Unit 2 IA system.
This contributed to the licensee performing the lineup without the use of a procedure, leading to a reactor trip on January 4.
OFFICIAL RECORD COPY CIR SUSQUEHANNA 89-01 0005.0.0 05/06/89
'Appendix A i
4.
SSES Procedure G0-100-003, "Power Operation,"
did not adequately address the transition from the low load valve to the main feedwater discharge valves thereby contributing to a
fee'dwater transient which resulted in the reactor trip on January 12.
This is a Severity Level IV Violation (Supplement I).
Pursuant to the provisions of 10 CFR 2.201, Pennsylvania Power and Light is hereby required to submit to this office within 30 days of the date of the letter transmitting this Notice, a written statement or explanation in reply, including:
( 1) the corrective steps which have been taken and the results achieved; (2) the corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved.
Mhere good cause is
- shown, consideration will be given to extend this response time.
OFFICIAL RECORD COPY CIR SUSQUEHANNA 89 0006.0.0 05/06/89
ENCLOSURE 2 Penns lvania Power and Li ht Com an Enforcement Conference List'of 'Attendees
- Nercn 21 1989 s
Penns lvania Power and Li ht Com an H.
W. Keiser, Senior Vice President - Nuclear R.
G.
Byram, Plant Superintendent
- SSES J.
A. Novak, Project Engineer - Civil & Design Analysis J.
G. Refling, Senior Project Engineer - Systems Engineer A. M. Male, Manager - Nuclear Plant Engineering F.
G. Butler, Manager Nuclear Design D.
F. Roth, Senior. Compliance Engineer F.
S. Gruscavage, Site Supervisor - Nuclear Safety Assessment E. A. Heckman, Licensing Group Supervisor H.
G. Stanley, Assistant Superintendent
- Outages Atlantic Electric L. Fink, Project Supervisor United States Nuclear Re viator Commission S. J. Collins, Deputy Director, Division of Reactor Projects (DRP)
J.
T. Wiggins, Chief, Reactor Projects Branch No. 3, DRP A.
R. Blough, Chief, Reactor Projects
- Section, 3B, DRP J.
R. Strosnider, Chief, Materials 5 Processes
- Section, Division of Reactor Safety (DRS)
R. J.
F. Conicella, Operations
- Engineer, DRS H. J.
- Kaplan, Reactor
- Engineer, DRS C.
D. Sellers, Senior Material's Engineer, Office of Nuclear Reactor Regulation (NRR)
M. C. Thadani, Project Manager, NRR BE Clayton, Regional Coordinator, Office of the Executive Director for Operations F. I. Young, Senior Resident Inspector - SSES J.
R. Stair, Resident Inspector -
SSES OFFICIAL RECORD COPY CIR SUSQUEHANNA 89 0007.0.0 05/06/89
ENCLOSURE 3
'ERSPECTIVES o
SUSQUEHANNA HAS CONPETENT CONSERVATIVE OPERATORS AT THE CONTROLS.
o AGGRESSIVE PRO-ACTIVE IMPROVPIENT PROGRANS.
o RECENT OCCURRENCES VALIDATED OUR INITIATIVES ARE ON TARGET, o
SUSTAINED POWER OPERATION PUTS A
NEW PERSPECTIVE ON CONTINUING TRAINING.
o THOROUGH INVESTIGATION AND ANALYSIS OF ALL OCCURRENCES RENAINS OUR STANDARD.
RESTRUCTURING OPERATOR REOUALIFICATION TO
-NEW FORMAT INDUSTRY INVOLVEMENT
.PROFESS IONALISM
-SCRAM RECOVERY
-MANAGEMENT CONTROLS
-OPERATOR ISSUES IMPLEMENTED INNOVATIVE REACTIVITY CONTROL PROGRAM INPO TRAINING REACCREDITATION CONDUCTED SUPERVISORY TRAINING REQUALIFIED ALL PERSONNEL IMPROVING THE OPERATING ENVIRONMENT TOTAL ORGANIZATIONAL INVOLVEMENT IN VALUE STATEMENTS COMP REHENS IVE MANAGEMENT INVOLVEMENT PILOTED CANDIDATE ASSESSMENT PROGRAM FOR SELECTION OF NEW SUPERVISORS CONTINUAL INDEPTH ASSESSMENT PROMPT
RESPONSE
TO BWR POWER OSCILLATION ISSUE
OPERATOR ASSESSMENTS SOURCES OF OUR INFORMATION:
o CONTINUING TRAINING o,
NEW LICENSING o
NSAG o
NRC o
INPO o
MANAGEMENT,WALKDOWNS o
NQA SURVEILLANCES
WHERE WE ARE VISION INITIATIVES OPERATORS VALUE AND UTILIZE OUR PROCEDURES.
A,CONSISTENT MENTAL APPROACH TO AI L OPERATING EVOLUTIONS.~
ENHANCED MANAGEMENT SYSTEM STATUS CONTROL REACTIVITY CONTROL OPERATORS ARE TOTALLY ATTENTIVE AND INVOLVED PROMULGATE EXPECTATIONS OPERATORS ARE TECHNICALLY COMPETENT A TEAM PLAYER WHO UNDERSTANDS,
- ACCEPTS, AND UTILIZES DEFENSE-IN-DEPTH DAILY.
ENHANCE A SUCCESSFUL TRAINING PROGRAM JPM EXPANSION NEW REQUAL FORMAT NEW SIMULATOR RECOGNITION OF'INFREQUENT STARTUP AND SHUTDOWN EVOLUTIONS.
RECOGNITION THAT TOTAL EMPLOYEE INVOLVEMENT IS REQUIRED TO CHANGE INVOLVE ALL OUR EMPLOYEES AND TUNE A SUCCESSFUL CULTURE VALUE STATEMENT DEVELOPMENT WITH TOTAL EMPLOYEE INVOLVEMENT SUPERVISORY/MANAGEMENT TRNG SHIFT SCHEDULING EMPLOYEE ENHANCEMENT TEAM
GOAL FROM THESE ACTIVITIES TOTAL OPERATOR PERFORMANCE
MELL TRAINED MELL DESIGNED PLANT MELl NAINTAI NED PLANT GOOD PROCEDURES RIGHT ATTITUDE OPERATOR PERFORNNCE SAFE EFFICIENT GENERATION ZERO DEFECT POlICY C(NPLETE NNAGE%NT SYSTEM PROFESS IONALISN 0
LOM.INCIDENT RATE MITH LOM SAFETY S IGNIFICANCE TEAN PLAYER NNAGEKNT SUPPORT SAFETY CONSCIOUS
LOSS OF INSTRUNENT AIR LOSS OF SMUTDOWN COOLING HI WATER LEVEL TURBINE TRIP RPS ACTUATION DRAINLINE COOLDOWN
LOSS OF INSTRUMfNT AIR UNIT 1 COOLING TOWfR OTHI R LOADS t
t I
I I
I I
L
~%\\0 UI SKIP I
1-25-029 I
I l
I I
1-25-030 UNIT I TURBINE BLDG.
~mes U2
> SKID I
I 2-25-029 l
l I
l I
2-25-030 L
UNIT 2 TURBINE BLDG I
I I
I l'1 l
OTHER LOADS UNI I COOL ING TOMER LEVEL BUBBLER r
h 0-25-153 0 25 081 I
UNIT 2 COOl ING TOMER LEVEL BUBBLER C IRC.
MATER PUtFHOUSE
LOSS OF INSTRUMENT AIR - UNIT 1 COOLING TOWER JANUARY lsd o
MAINTENANCE ON UNIT 1 "B" INSTRUMENT AIR DRYER SKID WAS SCHEDULED FOLLOWING THE HOLIDAYS.
o DUE TO REDUCED CAPABILITY OF SYSTEM IN THIS CONFIGURATION TRANSFER OF COMMON LOADS TO UNIT 2 WAS PLANNED, o
UNIT 1 PROCEDURE TO TRANSFER COMMON LOADS TO UNIT 2 IN THE CWPH WAS NOT FOUND, o
SHIFT SUPERVISION DETERMINED CONFIGURATION TO ACCOMPLISH TRANSFER VALVE 025081 OPENED VALVE 125029 CLOSED o
VALVE POSITION LOGGED JANUARY 4TH o
DIFFERENT SHIFT REVIEWS SYSTEM LINEUP.
o DID NOT FIND WHAT THEY CONSIDERED NORMAL LINEUP FOR TRANSFER.
o OPERATOR DIRECTED TO CLOSE VALVE 025153 TO OBTAIN PROPER LINEUP.
STATUS OF VALVE 125029 ASSUMED OPENED o
OPERATOR STANDS BY VALVE FOR 5 MINUTES TO ASSURE
.NO PROBLEM WITH NEW LINEUP.
o HE LEAVES AREA AND 5 MINUTES LATER COOLING TOWER
'ASIN LOW LEVEL ALARM SOUNDS IN CONTROL ROOM.
o SCRAM IMMINENT ACTIONS TAKEN.
POWER REDUCED TO 60X o
FOUR CIRC WATER PUMPS TRIP.
o LO CONDENSER VACUUM o
REACTOR SCRAM
LOSS OF INSTRUMENT AIR UNIT 1 COOLING TOWER FINDINGS JAN 1ST SHIFT RECOGNIZED LOSS OF SYSTEM CAPABILITY AND NEED TO REALIGN SYSTEM FOUND NO PROCEDURE EXISTED UNDERSTOOD AND IMPLEMENTED AUTHORITY AVAILABLE SELECTED A SUCCESS PATH DOCUMENTED/LOGGED ACTION TAKEN DID NOT UPDATE A SYSTEM STATUS CONTROL MECHANISM JAN 0TH SHIFT REVIEWED PLANT STATUS EXHIBITED QUESTIONING ATTITUDE CONCERNING INSTRUMENT/AIR STATUS INVESTIGATED SITUATION CONDUCTED FIELD WALKDOWN TOOK PRUDENT ACTION AFTER VALVE POSITION CHANGE DID NOT ADEQUATELY UTILIZE FORMAl SYSTEM STATUS FILE COMMON SYSTEMS PRESENT CONTINUING CHALLENGES
CAUSAL FACTORS o
REDUCED SYSTEN CAPABILITY o
PERSONNEL
- ERROR, LOSS'F STATUS CONTROL CORRECTIVE ACTIONS o
ADDED AUTHORITY CONCEPT TO STATUS CONTROL PROGRAN o
ADDED SECTION ON SWAPPING I/A LOADS IN CWPH TO UNIT 1 PROCEDURE o
REVIEWED OTHER SYSTPIS WITH CAPABILITY OF BEING CROSS TIED.
2 PROCEDURES CHANGED
LOSS OF SHUTDOMN COOLING QEACTog VE.SS EL EL '738-9 EL 72.3'-1 t
4 ve~k l
~
pooa EL7to-3 1
Keepfi ll eo
LOSS OF SDC o
o TO SUPPORT MAINTENANCE ON "A" PUMP ROOM COOLER OPERATIONS ATTEMPTED TO SWAP SDC TO C
PUMP.
FOLLOWING START OF THE C
PUMPS THE A
PUMP WAS SHUTDOWN.
0 FLOW/PRESSURE TRANSIENT CAUSES ISOLATION OF RHR F008 VALVE AND RESULTANT TRIP OF C" PUMP.
o OPERATORS BEGIN RESTORATION PER THE OPERATING PROCEDURE CLOSE F009 OPEN F008
~ FILL AND VENT OPEN F009 o
WHEN F009 IS OPENED, A BANG IS HEARD>
AND THE F008 ISOLATES.
o WEAKNESS IN DESIGN HAS LED TO SEVERAL ISOLATIONS OF THIS NATURE IN THE PAST YEARS.
o EXACT CAUSE NOT KNOWN.
I I
LOSS OF SDC FINDINGS o
OPERATOR FOLLOWED THE APPROVED PROCEDURE FOR SWAPPING PUMPS.
o OPERATOR FOLLOWED THE APPROVED PROCEDURE FOR FILLING AND VENTING ~
o PRUDENT ACTION TAKEN IN RESTORATION ACTIVITIES.
o PREVIOUS SIMILAR INCIDENTS HAVE OCCURRED.
INVESTIGATIONS HAVE NOT SPECIFICALLY IDENTIFIED THE ROOT CAUSE; CAUSAL FACTORS o
DESIGN WEAKNESS LEADS TO SPURIOUS ACTUATION OF RHR SDC.
ISOLATION INSTRUMENTATION HI FLOW HI PRESSURE o
PLANT CONFIGURATION PREVENTED NORMAL FILL AND VENT TECHNIQUES
CORRECTIVE 'ACTIONS o
AGGRESSIVELY PURSUING CHANGES TO THE INSTRUNENTATION FOR THE REFUELING OUTAGE IN CONDITION 4 AND 5 o
EXPANDED NONITORING INSTRUNENTATION TO GAIN A BETTER UNDERSTANDING OF ROOT CAUSE.
o ADJUSTED PROCEDURE FOR FILLING AND VENTING
Hl MATER LEVEL TURBINE TRIP IIV 1040$ A IIV-10603B IIV I0603C 30" l6" IO RI ALION VESSL l.
SIARIUP FII FLOM CONIROL RIP "A" RFP "8" RFP "C" IRON III HLAIIRS
HI WATER LEVEL TURBINE TRIP o
REACTOR POWER AT 20X.
o TRANSFER OF FEEDifATER FROM STARTUP LEVEL CONTROL TO AUTOMATIC LEVEL CONTROL IN PROCESS.
o UPON OPENING THE FEED PUMP DISCHARGE VALVES FOR THE B" AND "C" PUMPS.
REACTOR LEVEL ROSE RAPIDLY TO LEVEL 8 TRIPPING THE MAIN AND FEED PUMP TURBINES.
o COLD WATER ADDITION CAUSES REACTOR POWER TO INCREASE ENABLING THE RPS TRIP ON CONTROL VALVE FAST CLOSURE.
o REACTOR SCRAM.
HI WATER LEVEL TURBINE TRIP FINDINGS o
THOROUGH TRANSIENT INVESTIGATION UTILIZING GETAR's DATA o
-INDIVIDUALERRED o
RECOGNIZED AN OPPORTUNITY TO REINFORCE OUR VALUES CAUSAL FACTORS o
PERSONNEL ERROR FEEDWATER LEVEL CONTROL NEVER ESTABLISHED.
LACK OF VERIFICATION NOT RECOGNIZED.
o=
PERSONNEL ERROR PLACING TWO FEED PUMPS IN SERVICE AT THE SAME TIME.
DISCHARGE PRESSURE TOO HIGH.
CORRECTIVE ACTION o
POSITIVE BEHAVIOR PROGRAM UTI.LIZED.
OPERATOR INVOLVED DEVELOPED TRAINING ON THE EVENT AND TRAINED EACH SHIFT.
o ADDED PROCEDURAL GUIDANCE
RPS ACTUATION o
SHUTTING DOWN UNIT 1 TO REPAIR VACUUM BREAKER INDICATION ON B2 VACUUM BREAKER o
FOLLOWING INSERTION OF ALL CONTROL RODS MODE SWITCH PLACED IN SHUTDOWN.
SCRAM VALVES REPOSITION WATER FLOWS INTO SCRAM DISCHARGE VOLUME o
OPERATOR RESETS SCRAM WITHOUT FIRST BYPASSING SDV HI LEVEL TRIP GO-100-005 DID NOT CONTAIN SPECIFIC INSTRUCTIONS TO BYPASS SDV HI LEVEL TRIP.
UNIT SUPERVISOR OBSERVED ERROR BUT IS UNABLE TO STOP OPERATOR IN TIME.
SECONDS LATER WATER LEVEL REACHES HI LEVEL TRIP AND RPS ACTUATION OCCURS,
RPS ACTUATION FINDINGS o
SUSQUEHANNA TAKES A CONSERVATIVE APPROACH TO REACTOR SHUTDOWNS BY INSERTING ALL CONTROL RODS PRIOR TO TAKING THE MODE SWITCH TO SHUTDOWN MINIMIZE CHEMISTRY TRANSIENTS MINIMIZE SYSTEM TRANSIENTS o
OPERATOR IS TRAINED FOR CAUSALITY CONDITIONS PLANT SCRAMS HI SDV ALARM LIT BYPASS SDV HI LEVEL TRIP RESET SCRAM o
SDV ALARM NOT LIT o
PROCEDURE DID NOT CONTAIN EXPLICIT DIRECTION CONCERNING BYPASS OF SDV HI LEVEL TRIP o
UNIT SUPERVISOR OBSERVED ERROR AND ATTEMPTED TO STOP PCO
RPS ACTUATION CAUSAL FACTOR o
PROCEDURE 60-100-005 DID NOT CONTAIN EXPLICIT DIRECTION CONCERNING BYPASS OF SDV HI LEVEL TRIP, o
PERSONNEL ERROR EVEN THOUGH PROCEDURE DID NOT CONTAIN EXPLICIT DIRECTION AND SIMULATOR TRAINING WAS NOT CONDUCTED ON THE EVOLUTION, SYSTEM KNOWLEDGE WAS PROVIDED.
CORRECTIVE ACTION.
o DIRECTION CONCERNING BYPASSING SDV HI LEVEL TRIP ADDED TO G0-100-005,
REACTOR COOLANT SYSTBI 3/4. 4. 6 PRESSURE/TBSP ERATVRE LIMITS REACTOR COOLANT SYSTEM LIbIITIHG CONOITION FOR OPERATION 3.4.6.1 The reactor coolant systea tcsperature and pressure shall be llmitad in accordance with the 1feft lines shown on Figures 3.4.6.1-1 and 3.4.6.1-2, as appl fcabl ~, i'or hydrostatic or leak testing,, heatup by non-nuclear
- means, cooldown i'ollowfng a nuclear shutdown and low power PHYSICS TESTS, and operations with a critical core other than low power PHYSICS TESTS, with:
a.
A maxfaua heatup of 1004F fn any L-hour period, b.
A saxiarua cooldown of 100 F fn any L.hour perfod, c.
A Isaxfme tesperature change of less than or equal to 204F in any one'hour period during inservfce hydrostatic and leak tasting operations above the heatup and cooldown lfaft curves, and d.
The reactor vessel flange and head i'lange Ceaperature greater than or equaI to 70'F when reactor vessel head bolting studs are under tension.
APPLICABILITY: At al 1 tfaes.
ACTIGII:
Nth any of the above 1fafts exceeded, restore the tesperature and/or pressure to within the 1 fmfts within 30 minutes; perfore an engineering evaluatfon to determine'he ef'f'ects of the out-of-lfaft condition on the structural intagr ty oi'he reactor coolant systas; determine that the reactor coolant systaa remains acceptable for contfnued operations or -be in at least HOT SHUTOGW within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO SHUTDOWN within the followfng 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SVRVEILIAHCE RE UIREHENTS 4.4.6.1.1 Ourfng systoa
- heatup, cooldown and fnservfce leak and hydrostatic tasting operatfons, the reactor coolant systaa tesperature and pressure 5hall be deterefned to be wfthfn the above required heatup and cooldown limits and o
the right of'he licit lines of Figures 3.4.6.1-1 and 3.4.6.L-2, as applicaola, at least once per 30 Ifnutes.
SUSQUEHANNA - UNIT L 3/4 4 L6 Amendment No. -'~
yo I
eo
~
~
~
~
e I I
~
loo<<
\\ e
~o
~~e e>>
~ O
~ ~ O oe
~ alee ~
~
ee ao
~
~ <<o '
ae e
e I e o
~
a>>e a
I
<<o'" y
~ ~
~ fo ~
~
~
~
~ oo
~ ~
ee
~~
~
O OM~
~ ~
~ <<W e
~o ae p eoo
~ ~
~e
~
~
I Maoa e
~e
~
~
~
o ~ ~
~
ao o o ~
~
~
~
a imam oe <<
~
~ <<ol oooeomm e ee a e O
~
a F
~oae a ggWA
~ <<e
~
leo a<<
e ae Curve A'ppl)cable fa service pe'ho4s uy to 32 EFPY.
O
~O ~ ~
A e CYST%M
@ANT
$IN44CLQJI I4tATlNO Mlle C a NUCLEI lCNt OllflCALlLl&FtAttO QS Ot SWa UCCaam TOflCaL ASXurr lltOOll~
A', I', r Ulna Mre ae Atiuaaco tW coat 00aTLNt TSIVOAATVNf04IFF~ AJI INTlAL QKL4, ILATS Rf'~ Of 1W 4
CO CO CO
~+4M %CACTOID VCSSSL ~Al, T%Wlh*TLJJlt &1 IIINIKN REACTOR VKSSKi. l%TAL TKWERATURE VS.
REACTOR VESSEL. NESSURE Figure 3.4. &.1-2 QkgVEHANLA - UNIT 1 3/4 4-19
DRAIN II% CmumN OPERAlN RESP86E PE. AC,ToP VessEL P-RPV TIE c,ikt'oc P
() of ~)
ALL RNS IN NINfAINLEVEL GNBm.
=
GNK CINRATION ESTNLIQKO STAN RKIRC OR IMU I/S AN) REDUX CRD IH'RESSINI2E/NINIAIN IOO F/HR COOLD(NN If RECIRC NT I/S RAISE LEVEL PIlN3K NAllNALCIRouTIN
DRAINLINE COOLDOWN JANUARY 12, 1989 0415 0
REACTOR SCRAM.
0 RECIRC PUMPS TRIP ON EOC-RPT.
0 S0-100-011 BEING UTILIZED FOR STARTUP IS RESTARTED.
FOR COOLDOWN.
REACTOR VESSEL BOTTOM HEAD DRAIN TEMPERATURE AT 520 DEGREES F, 0430 0
BOTTOM HEAD DRAIN TEMPERATURE AT 440 DEGREES F.
0445 0
BOTTOM HEAD DRAIN TEMPERATURE AT 405 DEGREES F:
OPERATOR FAILS TO OBSERVE THAT THE MAXIMUM COOLDOWN OF 100 DEGREES F IN A ONE HOUR PERIOD IS EXCEEDED FOR THE BOTTOM HEAD DRAIN, FOLLOWING THE SCRAM.
0500 0
BOTTOM HEAD DRAIN TEMPERATURE AT 383 DEGREES F.
0515 0
BOTTOM HEAD DRAIN TEMPERATURE AT 419 DEGREES F
h 0519 0
REACTOR RECIRC PUMP "A" STARTED.
0525 0
REACTOR RECIRC PUMP "B" STARTED.
JANUARY 12, 1989 (CONT'D) 0530 0
BOTTOM HEAD DRAIN TEMPERATURE AT 032 DEGREES F.
1 0
DAY SHIFT SUPERVISOR, UPON INQUIRING, IS INFORMED BY SHIFT SUPERVISION THAT REACTOR VESSEL PARAMETERS FOLLOWING THE SCRAM WERE OBSERVED TO BE WITHIN THE ALLOWABLE RANGES.
0900 0
TECH SECTION SR.
ENGINEER. INQUIRES OF STA WHETHER ANY VESSEL TEMPERATURE/PRESSURE LIMITS WERE EXCEEDED AS A RESULT OF.THE FEEDWATER FLOW TRANSIENT/SCRAM.
THE STA REPLIES THATi BASED ON A
PRELIMINARY REVIEW OF VESSEL DATA, NO PARAMETERS WERE EXCEEDED, JANUARY 12TH TO JANUARY 15TH REACTOR REMAINS IN CONDITION 3:
S0-100-011 CONTINUED TO BE IMPLEMENTED THROUGHOUT THIS PERIOD.
JANUARY 13, 1989 START UP PORC JANUARY 15, 1989 1755 0
PLACED REACTOR NODE SWITCH IN STARTUP PER G0-100-002.
2003 0
REACTOR CRITICAL
I
JANUARY 16, 1989 1930 0
REACTOR AT 27X POWER.
0 SHIFT SUPERVISION, PERFORMING REVIEW OF S0-100-011, DISCOVERS THAT' VESSEL COOLDOWN DEVIATION OCCURRED DURING THE TIME PERIOD FROM 0415 HOURS (SCRAM) TO 0515.HOURS ON 1/12/89 FOR THE BOTTOM HEAD DRAIN TEMPERATURE.
0 THE DUTY MANAGER IS NOTIFIED.
I' THE DUTY MANAGER NOTIFIES NPE THAT AN ENGINEERING EVALUATION OF THE EFFECTS OF THE TEMPERATURE DEVIATION ON REACTOR PRESSURE BOUNDARY INTEGRITY IS REQUIRED TO BE PERFORMED PER T.S. 3.4.6.1, 0
STARTUP PLACED ON HOLD PENDING NPE EVALUATION.
JANUARY 17, 1989 0115 0
THE DUTY MANAGER NOTIFIES SHIFT SUPERVISION THAT NPE's PRELIMINARY EVALUATION CONCLUDED THAT THE REACTOR COOLANT PRESSURE BOUNDARY WAS NOT COMPROMISED.
Qeoa<<a<<y>>
I iot ~ ~ Jae ao
'e' 4 t 'ega re C ~e
~
a va
~ ~
~. <<a
~
~ eel e te )V<<
SCRAM/ Sg>
SH:.T ~CH A"V:SCR Li~i SCCR COD~ Z A A PRCVIDE ~jt WL PQI;uATXCN OE'tas" B 5 CNS=
SH Si.'P
%V SCP SOLZCI. PRE/POS.
SC":AM/
.jFOMATZCN.=RON:.'.tV
. 'W I
P~WSC~
I I
I o <<RV r'lvB 0'lo 45e
~I Pvei o aJha avijS
(-
5 ci"-
-3R -- 'Q) a toto>> e<<J 5A<<
4 RPt'.v SCBA5j/SHE."CfjN:4T..H CFF<<QO.jQ SH~~
~
ASS-B CA"SZ CF SCWu/SHUT CWW rS ZrmTXrZ:-O Q=.j:-BA.Z A:
CP <jCW P=j a o
~ e>> vvN oct.
V
~ vr eg 4
~ e I )
ee<<e <<>>V o Naa<<g QeC<<V<<Ra rCee ~ <<r 4
aoo 5teav>>V I
a<<<<e a
PRIOR TO 4
ttta>>vo>>4'et la Aa>>
~:-~U B~jQ ANY A"D.:CNAL AC:"N BZS ARa
."-CRN AZ-cA"4 =-.=:-CB nC.
N a>>
epS
- ~S BZgU=-:ZD ae e
~ e
%a>>
~ eeet
~ ~
ee
\\
~te
~
~ \\ 0
~
~ epeea mto
~ e C
p>>a oe i
I e
e>>
~
O ~ I g poeot ht
~/
v
~
B
<<<<ego
~ ~ te <<
~ a eta a>>a gtaQ Q
g e>>of o a>>V 4
e<<( tea a aee>>aeeaoeteq o
e 0
~jriad<<>>
ae
~ ee a
~ a<<rA<<4
'5 "T AR
'D aee
~ t ef
<<'a
~
'aeee
~
vve eP<<<<he4oi a Jit!oD
?
B:-5:ARo
~
e>>raise hC o a
~ I SH:=. ~C~j CAL AZV:.SC? I 1
<<t\\ta o a'fg 5 etta a
I o aetAk 4<<>>
Aa1~0 aleii SC.=AN/ Si: '."C AN BZPC.-.
- AotT St.PKRZ DjM'j.
At.HORZZ-RES.AR
.RACcC B:-'!A¹jQ C?=j.."='!5 iQ B:N 9,
XNQ SCGR
DRAI NLINE COOLDOWN
'INDINGS 0
SINCE THE UNIT REMAINED IN CONDITION 3, SO-)00-01)
"REACTOR VESSEL TEMPERATURE AND PRESSURE RECORDING" REMAINED ACTIVE.
'.0 SINCE SO REMAINED ACTIVE'ECOND INDEPENDENT REVIEW NOT TRIGGERED, 0
SO-)00-011 - EVOLUTION DEPENDENT PROCEDURE.
0 STA AND SHIFT SUPERVISOR INDEPENDENTLY VERIFIED AT DIFFERENT TIMES THAT T-SAT DID NOT EXCEED TECH.
- SPEC, COOLDOWN RATE, 0
WHEN SURVEILLANCE WAS CLOSED INDEPENDENT REVIEW BY UNIT SUPERVISOR FOUND THE PROBLEM.
CASUAL FACTORS 0
PERSONNEL ERROR-LOGGED DATA AND DID NOT RECOGNIZE 101 F
CHANGE IN ONE HOUR PERIOD.
0 CLOSURE OF SURVEILLANCE PROCEDURE NOT EXPLICIT.
DRAINLINE COOLDOlO CORRECTIVE ACTIONS 0
INTERIM MEASURE, SPECIFIC REVIEW OF COOLDOWN SURVEILLANCE DATA ADDED TO POST TRANSIENT REVIEW PROCEDURE.
0 PENDING RESULTS OF TECHNICAL ANALYSIS OPERATIONS WILL ADD CLOSURE MECHANISNS TO S0-100-011.
0 FINAL ACTIONS WILL BE INTEGRAL PART OF OUR CONTINUING TRAINING PROGRAM.
PERSPECTIVES.
o SUSQUEHANNA HAS COMPETENT CONSERVATIVE OPERATORS AT THE CONTROLS.
o AGGRESSIVE PRO-ACTIVE IMPROVEMENT PROGRAMS.
o RECENT OCCURRENCES VALIDATED OUR INITIATIVES ARE ON TARGET, SUSTAINED POWER OPERATION PUTS A
NEW PERSPECTIVE ON CONTINUING TRAINING.
o THOROUGH INVESTIGATION AND ANALYSIS OF ALL OCCURRENCES REMAINS OUR STANDARD,
Nuclear Safety Assessment Group of January 1989
Gl1C USIQAS The Susquehanna Station has been operated competently and safely in 1988.
~ Nuclear safety performance has continued to be excellent.
~ Generation performance was strong.
I
uc e8r a ety eI ermgng8 88 XC8 Slit
~ Qnly 5 unplanned SCRAMs and shutdowns (2 SCRAMs, 3 shutdowns).
~ No serious NRC violations or significant violations of the technical specifications.
~ No incidents which degraded safety to a material degree.
e Radiation exposures were below the goal despite the protracted outage.
88 er ormance tran
,341 QWH were generated 96 6 ~ of station record t
operated for 202 continuous da new station record of 120 da of dual-unit operation was set
cIS er ermaoc8 II XC8 8A't
~ Zero SCRAMs, shutdowns, or forced power reductions were caused by operator error.
~ Operator errors. accounted for less than 11 percent of the incidents and events of 1988.
~ Power was generated for 591 of 620 days available to the operators.
ue i i ence as x8rC~8@
Independent surveillances and inspections
~ A vigorous audit program
~ Oversight by Susquehanna Review Committee
~ Incidents and problems investigated; corrective action taken
~ Senior management involved in operation of the station
DUE DILIGENCE (CONT. )
o DAILY REPORT OF OPERATIONS AND EVENTS IS PROVIDED TO DEPARTMENT AND TO CORPORATE MANAGEMENT INCLUDING THE CEO.
o MONTHLY NUCLEAR SAFETY OVERVIEW TO BOARD.
o SSFI WAS CONDUCTED.
o OUTAGE PLANNING AND EXECUTION IS VIGOROUSLY REVIEWED FROM THE VIEWPOINT OF NUCLEAR SAFETY.
o VIGOROUS EFFORTS MADE TO IDENTIFY PROBLEMS AND IMPROVE RELIABILITY.
oo PLANT RELIABILITYREVIEWS oo INDIVIDUALPLANT EVALUATIONS oo SAFETY SYSTEM PERFORMANCE INDICATOR oo
RESPONSE
TO LaSALLE CORE INSTABILITY EVENT o
STRONG EMPLOYEE CONCERN PROGRAM.
JM/Msc001c (31)
nl't 8CQll 8 ll8 ln 0'ta 8 as eman in
~ There were 2 major events.
~ Outage required 112 days, an overrun of 35 days.
Issues were schedule, not nuclear safety.
e ueln uta e
as eman in
~ Unit 1 was shut down for 23 days.
~ Two major incidents occurred:
- Resin discharge of 3i23
- Steam separator event of 4i23
TABLE 1 OUTAGE EVENTS COMPARISON ESF INADVERTENT WATER SPILLS/
REACTIVITY LOSS OUTAGE ACTUATIONS STARTS HAMMER HP EVENTS SDC UI 1RIO Ul 2RIO U2 1RIO U1 =3RIO U2 2RIO 12 14 2
6 7
1 0
1
.2 3
4 5
2 14 6
0 1
1 2
NO CLEAR PATTERN EVIDENT EXCEPT FOR ESF ACTUATIONS, WHERE SUBSTANTIAL IMPROVEMENT HAS BEEN MADE SINCE THE FIRST TWO OUTAGES.
TABLE 2 U-l 1RIO U-1 2RIO o
RHR WATERHAMMER,
'o ESW LOOP 'A'NOPERABLE FOR 17 DAYS, UNIT 2 AT 100X POWER, o
CONTROL STRUCTURE CHILLER 'A'AMAGED.
o UNUSUAL EVENT - MAN FALLS FROM PLATFORM, CONTAMINATED INJURY, o
FIRE IN CONDENSER WATERBOX o
MAIN STEAM LINE PLUG BLOWOUT FLOODING OF CONDENSER BAY U-2 1RIO o
CONTAINMENT HATCHES NOT PROPERLY TORQUED o
TWO RHR WATERHAMMERS U-1 3RIO o
ONE FEEDWATER WATERHAMMER o
MAIN STEAM LINE PLUG BLOWOUT - OUTAGE DELAYED BY 7 DAYS o
FIRE INSIDE OF CONTAINMENT FROM THE VANTAGE POINT OF NUCLEAR SAFETY, THE ESW LOOP INOPERABLE EVENT AND THE CONTAINMENT HATCHES NOT PROPERLY TORQUED EVENT ARE FAR MORE SIGNIFICANT THAN ANY EVENT OF U2-2RIO, I
Table 3 lists the total events in each category for the years
]985 through 1988.
T&BLE 3
SUMMARY
'OP EVENTS BY YEAR'985 1986 1987 1988 Scrams Shutdowns Power Reductions High Voltage Reactivity Control Inadvertent Starts Emergency Plan Fires Waterhammer Containment ECCS Out if Service ESP Actuations Miscellaneous Equipment Damage Tech Spec Compliance Diesel Generator 9
3 8
12 9
1 0
4 6
1 2
ll 5
h 3
8 8
8 3
4 1
11 5
16 2
16 1
4 3
24 27 24 9
3 12 10 0
23 13 NRC Violation TOTALS 16 121 12 120 5
91 8
- 124 NRC violations are counted in the year the infraction occurred.
For
- example, two violations received in 1988 were based upon inspections in 1986.
They are in the 1986 total.
done
I I
I I
SmjrnmeijI'y o4 Event@ by Veer 0 QQQ SCRAMs Shutdowns Power reductions High voltage ESF Actuations NRC Violations 24 27 16 12 8
12 24 23 Total Events 121 120 91 124
CGM888 GY EvsAKs (Percentages)
Operator error Error non-operator Design Malfunction 16 11 44 41 34 33
.l 987 11 25, 10 30 25 20 39 40
88ÃV~CIGAS
- 1. Total number essentially constant.
- 2. Distribution of causes essentially constant.
- 3. Operator errors account for 10/o.
- 4. Human error accounts for 40%
(operator and non-operator).
Al't2-8COA 8 U8 lA Lllt8 8 88 8MBA IAQ
~ Several unexpected technical problems arose:
- Two major valve failures
- Turbine rotor coupling
- Cracks in CRD bolts
- Drywell insulation
- Target Rock containment valves An outage schedule overrun would have occurred in the absence of the two major events.
THOUGHTS ON OUTAGE 1,
INCIDENTS ARE OF ABOUT SANE FREQUENCY AND SEVERITY AS THOSE OF PAST OUTAGES, 2.
SAFE OPERATING ENVELOPE
~HAS OT BEEN VIOLATED.
3.
TMO BIG INCIDENTS, BOTH OF MHICH INPACTED THE SCHEDULE.
o NO INJURIES o
NO EXPOSURE TO RADIOACTIVITY o
NO RELEASE OF RADIO ACTIVITY o
NO DANGER TO THE REACTOR CORE 4.
SIGNIFICANT INCIDENTS HAVE OCCURRED IN OTHER OUTAGES o
NSL PLUG - DELAYED 'CNPLETION o
HATCH. BOLT TORQUING POTENTIAL FOR A NAJOR T.S.
VIOLATION 5.
UNEXPECTED PROBLE%
MERE COiNPETENTLY HANDLED, 6;
NO SERIOUS DEGRADATION OF NUCLEAR SAFETY OCCURRED.
4 THE EVENTS OF THE UNIT 2 OUTAGE WERE REVIEWED IN DETAIL, AFTER A LENGTHY DISCUSSION THE COMMITTEE AGREED ON THE FOLLOWING POINTS 1
THE BAS I C I SSUE IS INDIVIDUALACCOUNTABI l ITY.
EACH PERSON MUST UNDERSTAND THAT HE IS RESPONSIBLE FOR HIS
'ACTIONS AND THAT HE HAS A PERSONAL RESPONSIBILITY TO ENSURE THAT THINGS ARE DONE RIGHT 2
A SPECIFIC INDIVIDUALMUST BE IN CHARGE OF EACH EVOLUTION.
EVERYONE INVOLVED. IN LUDING THE PERSON IN
- CHARGE, MUST KNOW WHO HE IS.
. 3, PROCEDURES ARE MANAGEMENT S METHOD OF CAPTURING THE BEST THINKING THAT THE ORGANIZATION CAN PRODUCE ON HOW TO DO SOMETHING.
THEY ARE DIRECTIVE IN NATURE AND MUST BE
- FOLLOWED, 4.
MANAGEMENT EXPECTS PEOPLE TO ACT INTELLIGENTLY, THE ACCOUNTABLE PERSON IS RESPONSIBLE FOR USING ALL OF THE AVAIIABLE RESOURCES (PROCEDURES.
DRAWINGS.
WORK PLANS.
TECHNICAL'MANUALS. EXPERT ADVICE. ETC.)
TO DO THE JOB
- RIGHT,
PFGKA8FÃs F888
\\
- 1. Human error: Operations, maintenance, surveillance testing
~ Individual accountability a Specific person in charge
~ Adherence to procedures
- 2. Reliability of systems and components
- 3. Upgrading of design
'A8)GF Human performance
~ Management training program Reactivity control procedures o Reorganization of maintenance section
- 2. Equipment reliability
~ Safety system performance indicators
~ Reliability reviews of systems
- 3. Design
~ Safety system functional inspection
+ Uninterruptible power supplies
UMMQgy The Susquehanna station was operated competently and safely in 1988:
o Nuclear safety was excellent.
o Operator performance was excellent.
~ Generation performance was strong.
~ Due diligence was exercised.
D'able 3 lists the total events in each categorv for the years 1985 through 1988.
TABLE 3
SUMMARY
OF EVENTS BY YEAR 1985 1986 1987 Scrams Shutdowns Pover Reductions High Voltage Reactivity Control Inadvertent Star'ts Emergencv Plan Fires Waterhammer Containment ECCS Out f Service ESF Actuations Miscellaneous Equipment Damage Tech Spec Compliance Diesel Generator NRC Violation TOTALS 12 24 16 121 16
'l 7 12 120 2
16 24 5
91 3
3 3
12 p
4
/
10 4
2 3
'I 2.
3 3
13 124 3 g NRC violations are counted in the year the infraction occurred.
For
- example, tvo violations received in 1988 vere based upon inspections done in 1986.
They are in the 1986 total.
OP RATOR R
SCRAPS 85 86 87 89 SHUTDOWNS 0
0 POSER REDUCTImS 0
'SF ACTUATIONS 0
2 (I) 1 a
(i)
NRC YIOLATIONS 3
TOTALS (ALL EYKHTS) 18 (>'7) 1Q 12
(~)
AYERA6E 12 OPERATOR EYENTS PER YEhR 1989 IN FIRST QUARTER JRM/TRC01C(31)
STKAM DRYER STEAM OllTLET WATKR LEVEL STEAM SEPARATORS FKKQWATER INLET ~
(TYPICAL OF 8)
TOP GUIOK JET PUMP NOZZLE FUKL BUNDLES FUKL SUPPORT PIECE RECIRCULATION LOOP
~
SIJCTION CONTROL ROO NIOK TUSE OP JKT PUMP RKClRCULATIGN INLET (TYPICAL OF lO)
~ RECIRCULATION LOOP SUCTION CORK PLATE
-CRD HOUSING COOLINQ WATER~
TO CLEANUP SYSTES SUCTION REACTOR YESSEL FLOW PATHS
COLD WATER INJECTED INTO BOTTOM HEAT FROM CRD WITH NO SWEEP OUT RESULTS FROM LOSS OF RECIRCULATION FLOW (RWCU ALONE IS NOT ENOUGH TO REMOVE COLD WATER FROM BOTTOM HEAT AREA)
o RESTART AT LEAST ONE RECIRCULATION LOOP o 'EDUCE CRD FLOW AND INCREASE REACTOR WATER LEVEL o
RESTART RWCU IF ISOLATED
REACTOR COOLD(M TRANSIENT OF 1/12/89 E%INEERI% PERSPECTIVES o
ENt I%ERINS S fPORl o
SFEIY SI6NIFICANK o
TEN(ICAL BASES o
FATIQK N3NItGRIN6 RNSRN
ENGINEERING SUPPORT 0
(fF-HmJRS (NACT ON-CALL NPE REPRESENTATIVE BY TELEPHONE OR PAGER KLEPHOtK m %E LIfK%NT tEUM)tK OR PAGER TO KGSICAL StfPORT NNAGER OR E%I%ERING SUPPORT LHSER ACCESS m APPROPRIA'K DC%ICAL EXPHlIS DOES NT 8MIRE ON l I%, l%NT 0
RESPONSE TO 1/16/89 DISCmtERY OF HI6H C(mLDM 3, T ON-CALL REP CALLED STRESS E%INEER (J.D. NVAK)
STRESS E%1%ER %NT m %FICE A% CNTAGED STA'S FOR DATA STRESS E%INEER PHFOP%) ANALYSIS OF DATA-CON%RED % SAFEIY SIGNIFICANCE STRESS ENGINEER ADVISED PLANT STAFF OF CONCLUSINS TIK FRN DISCRIERY TO RESPONSE 'm PLANT STAFF WAS LESS MN 6
SAFEtY SI6NIFICANCE OF RE 1/12/89 CGOLD$5 o
M) E%I%ERING CON(XRNS RN BOTT(N IRS COOS(W FATIQK o
CONCLUSION:
NO SAFEIY SINIFIGKE
BRITTLE FRACTURE EVALUATION FOR 1/12/89 COOLDM o
LNEST TPFERAtURE M1L ABt:NE FRACIURE QANSITIN TPIPERA'nSE o
PRE SECTION XI APPE%IX E ANALYSIS CRITERIA KT o
CtmLBM K%'HN'tuRE AM) PRESSU8E RENI%D m RI6HT (F TECH SPEC P-T CURVES CONCLUSION:
BRIT/LE FRACTURE NT A GNCERN
FATIGUE EVALUATION FOR 1/12/89 COOLD55 o
PP&L ON-GOING RPV FATIQK PROGRN BEGl5 EARLY IN PLANT LIFE o
CONCERN FOR FATIQK IN THIS TRANSIENT WAS AT TtK CRD PBETfQTINS o
TRANSIENT OF 1/12/89 CONSERVATIVELY HSAUSTED ~ 0.0%X 0F FATIQK LIFE OF CRD PENETRATINS, (INITIALREVIEM GKLUDED C'.7X)
CONCLUSION:
FATIQK IS NOT A GNCERN
TECHNICAL BA93 o
PLANT DESI6%D TO:
AVOID BRITllE FRACTURE ACGNNXATE FATIQK o
TRANSIENT OF 1/12/89 AM) ALL Olla RECORDED LSD f68 COOLDM INCIDENIS WERE WITHIN DESI6N BASES o
OXjLDM RATE LINIT(F. 1'F IN 1 tOjR SHXLD BE NNIltNED BY MK TPPERATURE (TsaT) o LNER fKAD DRAIN lBVKRATURE SHOlU) BE USED FOR NNITORING Q, T BEIWEEN VESSEL AREAS SUCH AS RECIRC Lmjp TO LlM3}68 o
EN61%ERI% SKULD N3NITOR FATIHKRN ALL SIGNIFICANI TRANSIENTS
SAFEIY SI6NIFICANCE OF TK I/12/89 COOLD55 o
M) E%1%ERIN6 CGNCERNS FOR BOTlXÃ f68 FATIHK o
CNCLUSIN:
NO SAR7( SI6NIFICAKE
FATIGUE NNITORING R EVALUATIN o
DESIGN BASIS ASSlKO DESIGN BASIS HNNSIERS WHH PARA HERS CONSHNATIVELY
,SELECTED TO ASSURE BOtMI% AGUAL PLANI EVENIS o
CURRENT EVALUATIONTECNIQK VERY CONSERVATIVE SELECT A DESIGN BASIS TRANSIENT WI1H PAfQKl'ERS THAT KlM ACTUAL EVENT FATIRE UTILIZATINIS LESS THAN THAT USED IN STRESS 8EPORT o
FNURE EVALUATIN1HMIQK FATIGlKPRD EVALUA'IES FW NOZZLE a CRD PBETINTINS CAN DO GBERS REALISTIC UTILIZES ACTUAL PLANT DATA TO ELIMINATE OVER-CONSENATI35 PERIODICALLY lHNKO HY DATA TRANSf& FRN PuM GK'UTER SYSTEM CAN RN SPKIAL CASES WItH NNAL DATA ENW RELY OPERATINAI A% BASELI%D WON ACC8%LAIED PLANT DATA SINCE INITIALOPERATION BY E% OF 1989
o E%IHEERING REVIBKD 0LDM EON OF 1/12/89 A% mlKR EVENIS FOR BRITTLE FRACTURE & FATIQK USING C0NSHNATIVE %%S WI% GNCLUSION OF NO SIGNIFICANT SAFElY I5'ACT o
FATIQE NNITORING USING FATIGlKPRO WILL PRmtIIK ACQjRAK FATIQK UTILIZATINBASED N ACtuAL PLANT PARIKllH5