ML18046A976

Review of Operating Experience History of Palisades Nuclear Power Plant for NRC Sep.

ML18046A976
Person / Time
Site:	Palisades
Issue date:	10/31/1981
From:	OAK RIDGE NATIONAL LABORATORY
To:
Shared Package
ML18046A975	List: ML18046A976
References
NUDOCS 8110220434
Download: ML18046A976 (126)

Text

..

.* **~** .. .. *.. --*""--*- -** - - *.

. r - . - . -. . *.~

REVIEW *OF 1THE, OPERATING-* EXPERIENCE HISTORY OF THE PALISADES . - -. . - ~ . ~ .,_ . ..

NUCLEAR POWER PLANT FOR THE NUCLEAR REGULATORY COMMISSION'S_-,- --. . - . --- ....... .

. *. * ' SYSTEMATIC -EVALUATION. PROGRAMS-~ .

• . :: . . ...: *.*_*_* ~: -- -~-, : . -: :: ,,... ..

~--- -

.. * *Performed by-the* Staff of the * * * - ....... - ..... . ---*'*

Nuclear. Safety. Information .Center* tNSIC)

Oak Ridge National. Laboratory. (ORNL) .. ' '

• o

• o 0 o 0 I * * * * ~1.-.*'.;

J. -

-.:. ~

- - ~- . "' .. - . - .. - .. - .

October.1981.

( 81 i0220434 811013 "*1 '

PDR ADOCK 05000255*

P PDR

-- -- -*---*-------------------..---------.--~~------ *****------- --*--.-.----.------ -~;'.:--~- .. -.-, ..

~.-*. *~. -~-*~~ *----::~* . -

.~. ~,~---.--.-.----

,_ *. ---* ..-:..~ ,_*:-. - . *- *, :*. ..~--!.>.------=--**.: .

i

• r.. , : :

.*. } .. CONTENTS

.* .ABSTRAC'r : . . , :: ___ ,

• *--

• INTRODUCTION 1-1

1. ._ SCOPE O~ REVIEW 1-2 Ll Availability and Capacity Factors *****************.***** 1-2.

1.2 Review of Forced Shutdowns and Power Reducti.ons *** : **** 1~3

.2: *.:. i-.3* Review of Reportable Events ** ** * .- **** _. ******* ~ ** ~ * ~ ** ~ ** 1-8 1.4 Events of Environmental Importance*and Radioactivity

~:~;:  :;t. $ * *

• Evai"Uatio~

~of "ciP.*~rat~~*g.--E;q,e~i~~ce.*.

~ ....... .

~* ~ ... *..

~. ~-~~~ * . t:..16

~

. :' **2. - SOl;JRCES: OF. 'moRMATION *lrr.ILIZED m THE .REVIEW ********** 2-1 p:::.::2-~1-..  :: *.:* _Av.a:~"l.abil~t.r*and.:C:apac~ty. Factors ** ;:.~* ** ~.-;*.: * .' ******** 2~1 2.* 2 Forced Reactor .* .* -

Shutdowns. . - ':and Power Reductions

. ******* . **** .. : .. 2-1

2. 3 Reportable- E.vents*. ........ ;..... *....... *................. . 2-2 --

2.4 Environmental *Even.ts and Radioactivity Releases ********

• 2-2*

2.5 Use of Computer Files on RECON and Special Pub lie a tions .. *..................................... ~ ... . 2-2

• 3. *CRITERIA AND CATEGORIZATION FOR THE EVALUATIONS OF THE OPERATING HISTORY *** **.*.*** ***..**.**.*******. ****** . 3..;.1 3.1 Signific_ant Shutdowns and Power Reductions **.**..*****.* 3-5 3.1.l Criteria for significant shutdowns and power reductions*

• e e e

• e e e e e e e

• e

• e e e e

• e e e e e e e e e e 0 8 e e G e e e 0 0 e

• 3-5 3.1.2 Use of criteria for determining significant shutdowns

. an.a. power.. r~ductions . .......... () .... 0 " * * * * * *

• o. 0 ** o.* * .* 3-5

• 3.1.3 Non-DBE shutdown and power. *reduction categorization **** 3-6 3.2 Significant Reportable Everits ************************** 3-9 3.2.1 Criteria for significant reportable events ************* 3-9 3.2.2 Use of criteria for determining significant reportable events ....................... ~ ..................... o

• 3-9 3.2.3 Non-signif~cant reportable events ************ ~ ******. ~. 3-11

4. NOT USED

. __,_:_ **-* ;~...

.: ****-~--*.: ***~~

I I

ii

• CONTENT'S. (continued) .. **--:*

'~.:: .-: -. ;

OPERATING EXPERIENCE REvIEW: OF 'PALISADES *************** 5-1 5.1 Summary of Operational Events of Safety Importance ***** 5=1 5.2 General* Plant Description ********* -****.* *...,-~ *** ***.**.** -**** 5-1 5.3, Availability and Capacity Fac~ors .......--.*.* ; **-**. ~- **** ~*** ;.:. 5-2.

5.4 Review of Reactor Shutdows and Power Reductions *******. 5-2

-. s~~.1 -.. DBE in;Ltiating events* .**** o o

• o o

• o e o o * *e

• o " o o e e " o *

• o

• o

• o e 5-4

• ~5~4.t.i DBE Sect. l event ~increase in heat removal by secondaey system e. 0 e 0 e 0 0 0 0 0 0 0 0 0 0 0 * () 0 0 0 0 0 0 0 0 0 0 0 0 0 0 Cl. Cl 0 0

• 5-6

~: S:. 4. 1 ~ ~ :'~:*!)BE Sect. 2 events - decrease in heat removal by secondary system.

• e 0 e e e 0

• e e G 0 e e e 0 Q O e e e e e e D e O e O e e e O e e

• e 5~6 5.4.1.3 DBE Sect. 3 events - single and multiple reactor.

cool.ant .PtlDlP trips ********** o o o

• o

• Cl

• Q * ., o o

• o o ** o *** o ** G *- 5:..9

• s.4.1.4 5.4.2 DBE Sect. 4 events anomalies e e e e e e e e e *

~ reactivity and power distribution e e e e e e e e e e e e e e Trends and safety implications of shutdowns and power reduct ions ....**.***.....**..*......*...

e e e e e e e e e e e e e o o ****** o

• o ***

ID e ** e e e e 5-10 5-12 5.4.2.1 Steam generator tube failures. o.***** o. o**** o* o *** o*** oo 5~18 5.4.2.2 Vibration-induced cracking - charging systemooo*o**o*** 5-19 5.4.2.3 CRDM seal leakage 0 0 0 ** 0 0 0

• 0

• 0 0 0 0 a 0 0 0 0 0 0 0 0 0 ** 0 a Cl 0 D 0 0 0 Cl 0 0 5=20 5.4.3 Power reductions ***.... o ************* o " ** o o *** o o o a o a o o o 5-22 5.4.4 Non-ElB.E shutdowns ******** o *

• o

• o *

• c * * .

• o o a o * * * *

• o o *o o o o o o o 5=22

5. 5 Review of Reportable Events **********.* ***************** 5~26 5-.5~1 Significant Events .*..* ** o ***** o o Cl ********* o ** o o o o

• ID o o o

• 5~26 5.5.1.l Events involving the loss of onsite power sources in coincidence with the. loss of offsite power sources *** 5~28 5.5 .1.2 Events involving the loss of HPSI capabilitY****o****** .;~29 s . s.i.3 Events involving the loss of containment integrity ***** 5-30 5.5.l.4 Events involving the loss of component cooling*

capability. a * * * * * * * * * * * * * * * * * * * * * * * ., o

• o o

• a

• o a o o a o a a O* a ~ 5=32 Events involving the depressurization of the RCS ***** o. 5=33 Trends and safety implications of reportable events in addition to those categorized as significant*v*o*o 5~33

_..,..,.,..~-~*.*';!"'"*.'7*~,.,-~-~- ....... ~- --_.,,...~...,_.,.~ ..*-=

• • • . G iii CONTENTS (continued)

. Page 5o5.2.l Loss of off site power . ............. ~ ...................

• 5-33 5.5.2.2 Control rod drive anomalies............................ 5-3'5 5o5o2o3 Reactor. internals movement .********.****************.*.. 5-3 9

', 5.5.2.4 Containment purge isolation valves ** o * * * * * * * * * * * * * * *

• o.

• 5-39 5.5.2o5 Procedural and human error 0 Cl Cl 0 0 0 0 Cl *

• ID *

• e *** CD * * * * .

• Cl ** Cl

• Cl 5.6 Events of Environmental Importance *********.**.** ~ *.***

506.l Radioactivity release events ******************.******** 5-42 5o6.2 Nonrad~ological events .. ~. e * * * * * * * * * * * * * * * * * * * * * * *.** * * * . 5-42 5.7 Evaluation of Operating Experience *************.******* 5-44 REF'EB.ENCES * .... Q *** o ***** Cl ** Cl * ** ***** ~ ***** * * ** * * * * * * * * * * **

• 6-1

_/

APPENDIX A: Not Used APPENDIX B. PALISADES

1. Shutdown and p ower Re*d u ction Tables**************************** B-1

2. Reportable Event Coding Sheets ************ * ** * * * * * ** * * * * * * * * * * * ~ B-25

.,J

• L--.----~~----.-,-.,-----~--*-*-*-----*

iv No.

lol Cause of Forced Shutdown or Power Reduction and Method of*

Shutdowno o 0 o o coo o o om o o o o o o o o o Tables Title Q o a o o o o o o o.o o o o o o o o 0 0 o o a o o o o o o o o. 1-=5 1

1.2 Systems Involved With the Forced Shutdown or Power Reduction *** 1~6 lo3 Components Involved With the Forced Shutdown or Power Reduct ion a o o o o o o

• o o o o

• o o o ., o o * " o .,

• o o o o o o o o

• o

• o ., o o o o o * *

• o a;)

• o o o C!I 1-==9 lo4 Data Collected for Reportable Events - Plant Status and S}-ostem Involvedci coo o o ca o o Clo o o o . o go

• o

• v o o o o o Clo o . o o o o a o . a o o o o o...

• l-=13 loS Data Collected for Reportable Events - Equipment Involved and* Instrument Involvedo o a *

• o o . o o c o . o o o a o . o Cl. o o * *

• o c . o o Q.Q C!) co e lc:oal4 1.6 Data Collected for Reportable Events - Component Status~

Abnormal* Conditicra~ and Cause ******* o....................... *. 1=15 3.1 Standard Review Plan Chapter 15 Initiating Event for Design Basis (Events (Revision 3) **** o****************o************* 3=3 3o2 NSIC Event Categories. for Non DBE Shutdowns. o*** ** o************ o 3~7 3.3 Reportable Event Criteria - Significant ................. o** o** ."o 3~10 3.4 Reportable Event Criteria - Conditfonally Significant ********** 3-11 5.1 Availability and Capacity Factors for Palisades ********* o***** 5-3

5. 2 DBE Shutdowns for Palisades **.**...*.****.* o**.****** ~ ***.** o * . 5-5 5.3a Forced Shutdown Summary for Palisades ...****...***.*.********* S-13 5.3b Power Reduction Summary for Palisades ********* oo*o*********o** 5=23 5o4 NSIC Primary Category Summary for Non-DBE Shutdowns for Palisades. Cl '° o o o o o o o o o o o o o o o " o o o o o o o o o '°

• o

• o o c o o o o **** c o *** o o " 5a924 5.5 Summary of Palisades Significant and Conditionally

. Sigtiif icant Events Q

• 0 0 0 0

• 0 0 0 0 0 0 0 0 0 0 0 *** 0 0 *** Cl 0 0

• 0

• Cl

• 0 *** 0 II a a 5-2 7 506 Summary of Systems Involved in Palisades Reportable Events;o ** 5~34

5. 7 Summary of Palisades' CRDM Failures *****.****** : **** o....... * *

• 5-37 5.8 Summary of Radioactivity Released from Palisades ******.**.***. 5-43

---....-;---- ----------------*~~- -~ .. **- - ---------- ~ -~--** *---- ------ -- . -----*-***--*-**---------~- ------*-*-- --- ~-- -:--~--------- ~---~-**-.--~-~---- .....---,.-*~----.--- - --~ -

. :* *,. -* *---~-** _ .... ,;t . ,. ~ "__::_____ -**-- --*--*-*-*" .. --------~- - ~ *- - ._,_,_ - -~- --- " . . -~-- - - .. -- . - ____ _; ______ ~~-- .. - ***- -*-**--*-*******-~~ .. ___ :.- -* *-* .

v Figures

. Title

-No* .

5.1 Simplified Palisades HPSI Injection Paths ********************* 5-31

, S. 2 History of Partial 'Off site Power Interruptions at Palisades. *

• 5-36 5.3 Procedure and Human Error in Reportable Events at Palisades *** 5-41

---

*-------------------- ... ~-~----*---*-------------------------*----*

, _ ~ ,; ______ ....

INTRODUCTION The Systemat,ic Evaluation ~rogram Branch (SEPB) of the Nuclear Regulatory Commission (NRC) has the responsibility for condu~t of the Systematic Eval-uation Program (SEP) in an effort to determine the safet~ margins of the design and operation of the eleven oldest operating commercial nuclear power plants in the United States. These eleven plants are being reevaluated in terms of present NRC licensing requirements and regulations. The SEP program as well is-to~

establish docWiientation which shows how these.operating plants compare with current acceptance *criteria and guidelines on significant safety issues, and provide a technical rationale for acceptable.departures from these criteria and guidelines; provide the capability for making i~tegrated and balanced decisions with respect to any. required backfitting; and provide for*the early.identification and resolution of any potential safety deficiency.

The SEP Program is evaluating specific safety topics (called the Topic List) and is based on an integrated review of the overall ability of a plant to. respond to certain design basis events (challenges), including norma~

operation, transients and postulated accidents. The evaluation will result in a reassessment of the overall safety margins for each.facility and documentation of the reassessment on the basis of current criteria.

In this report the operating experience of the Palisades nuclear power plant is reviewed for the purpose of compiling and interpreting data on plant operational occurrences and events for application and input to the SEP Program

• The results of this report will be used by SEPB in performing th~ integrated assessment of overall plant safety for each plant.

'~~-**-,----------*---** ---;---....-*-:-:-;~- . ,. - * - - : - - - - * - - --~----*----*--:,.,-.-**"'l*~-*---::----7.~---*---"*---****---~~,--.,-.--__,__ ~~:*;~--**c~.--*--:;-_*"'; *** .,...~.._.....,.._.,.....,..*.*.-~<o-,,*!:-;~-*-**~.--_-."'"5

l'-2 The review approach with respect to operational events (forced shutdowns and reportable occurrence~) consists primarily of a three~step process:

1) compile information on the events>> 2) screen the events for significance using selected criteria and guidelines>> and 3) evaluate the significance and importance of the events from a safety standpoint. Trends in equipment faiiures and events where systems failed to perform their intended function are identified.

Other types of operating information as noted in the "scope of the review" section is compiled to provide an overall view of the plants' operating histories.

l. SCOPE OF REVIEW The assessment of the operating experience review for Palisades covered the time from initial criticality through and including 1979. The review included the following aspects of operation: availability and capacity factors; review of forced shutdowns' and power reductions; reportable events; events of environmental importance and radi.oactivity releas-es; -an* -evaluation of the operating experience in total.

. 1.1 Availability and Capacity Factors Both reactor availability and unit availability factors were compiled for all years. Starting with 1974, the tmit capacity factors using the design electrical rating (DER -. net MWe) and the maximum dependable capacity (MDC -

net MWe) were compiled as well. Data for the capacity factors was not available from earlier years.

The two availability and two capacity factors are defined *as follows:

1. Reactor availability =

hours reactor critical + reactor reserve shutdown hours x lOO.

period hours

• - - - - - - - - - - *----* ------ * - - ., -- - - -- ........ **4* - . * * - " ' *-- - - - - * * - - * - * * - - - - , - - - - -.-- - -*** *. - -~---*------*-~----**---- . * ~---.-*-.-------:*--* *.-~-- .. *---~*-*--~- .. **~~--- ******-~----:**---,-~-:-... - -

• --****,(-*---*~-****-***-:__.'. .. ,o.-: *..*c. **.

---*--~_ ~

1-3

I' 2.* Unit availability ..

• hours generator

---

..--....... on-line + unit reserve shutdown hours

-'--~~..--~..--..--~-.-~---~..---.-~~..--~~~~~~ x 100 *

. p~riod hours

3. Unit capacity (DER) ..

net electrical energy generated x 100.

period hours x DER

4. Unit capacity (MDC) =

net electrical energy generated x 100.

period. hours* x MDC net 1.2 *Review of Forced Shutdowns and Power Reductions The forced shutdowns and power reductions were reviewed and data .collected on eacli incident. Scheduled shutdowns for refueling and maintenance were not included in the review. However, if a utility had a refueling outage scheduled, the plant experienced a shutdown as a result of an abnormal event prior to the schedul.ed refueling, the utility reported that the refueling was being resched-uled to coincide with the current shutdown, and the utility reported the cause I

of the shutdown as refueling, then this shutdown was considered as forced.

Only that portion of the outage time concerned with the abnormal event, not the refueling time, was included in the compilations.

The power reductions were included to provide information and details that may have been associated with a previous or subsequent shutdown. The power reductions are included in the proper chronological sequence with the shutdowns in the data tables for the forced shutdowns and power reductions (see Appendices).

The following data was compiled annually for the for.ced shutdowns and power reduction:

1. d~te o~ occurrence;

2. duration in hours;

--,-----~--- . -- ------- ----------------- --*-*** - ----~-------. ****--~-~--. --*- --**---.- ...~-----.,---.-__,.....- -------.-*---~-,_,_....,~-~~

. _,. -* -~-- - *- **-----*- ----'-*-*-*-~-.:- ... -*- *'

1-4 3" noting if the shutdowns were also a reportable event 5 e.g., a: licensee 4.

~vent rep*ort: (LER) or abnormal occurrence report (AOR);

a summary description of the events associated with the shutdown or power reduction; So cause of the shutdown (Table lol);

6. *method of shutdown (Table lol);

7" the system directly involved with the shutdown or pc;iwer reduction (Table l 02.) ;

1___ .* - - - --- - - - - - - *-.****. ----~-------- - -.- **.----***** *-:--*- -** ---* . ---~~-.-*-*--*~-- *-----*** -.-..-...;*---**---~~--..***-*-*-* -~-.-..,..--,.- -- -----*-------"'*---...-.--.-....--=--~~~......,--,~-~;"'....~-::""l

. ,. r~**------**.... -:J.. ****-- --~-------...--.:.--*-* -*--*~* *"*-***-: .... .. -'*---~~.......'.---- *---***-* .:,:., '.-*-*---~--**:.:..:..:._~--=-'--~* .. . . "

.:~. _..._._,____,....._,_,_~. __ .. :. -~-- ...* *~- .. .:_ ..."-'---'.;. __ ,_ -* ....

1-5 Table 1.1 Cause of Forced Shutdown or Power Reduction and Method of Shutdown Cause Equipment failure A*

Maintenance or testing B Refueling c Regulatory restriction D Operator training and license exams .E Administrative F Operational error G

• .Other H

• Method of Shutdown Manual 1 Manual Scram 2 Automatic scram 3 Continuation 4 Load Reduction 5 Other 9

---

*----*------~ ----~-,.._**--------*----....;_-~. ---------*---------- -***~--._.,.-*--------------:--**......_.-,-.--.- ....... -:**------.---- ...... **-..-.. ,*"7-~

1-6 Table 1.2 Systems Involved With the Forced Shutdown or Power Reduction System Description Code Reactor RX Reactor Vessel Internals RA Reactivity ,Control Systems RB Reactor Core RC Reactor Coolant System & Connected Systems

• ex Reactor Vessels & Appurtenances CA Coolant Recirculation Systems & Controls CB Main Steam Systems & Controls cc Main Steam Isolation Systems & Controls CD Reactor Core *Isolation Cooling Systems & Controls CE Residual Heat Removal Systems & Controls CF Reactor Coolant Cleanup Systems & Controls CG Feedwater Systems & Controls CH Reactor Coolant Pressure Boundary Leakage DetectiOn Systems CI Other Coolant Subsystems & Their Controls CJ Engineered Safety Features sx Reactor Containment Systems SA Containment Heat Removal Systems & Controls "

SB Containment Air Purification & Cleanup Systems & Controls SC Containment Isolation Systems & Controls SD Containment Combustible Control Systems &* Controls SE Emergency Gore Cooling Systems & Controls

• SF Control Room Habitability Systems & Controls SG Other Engineered Safety Feature Systems & Their Controls SH Instrumentation and Controls IX Reactor Trip Systems IA Engineered Safety Feature Instrument Systems IB Systems Required for Safe Shutdown IC Safety-Related Display Instrumentation ID Other Instrument Systems Required for Safety IE Other Instrument Systems Not Required for Safety IF Electric Power Systems EX Offsite Power Systems & Controls EA AC Onsite Power Systems & Controls EB DC Onsite Power Systems & Controls EC (Composite AC & DC)

Onsite Systems & Controls (Composite ED AC & DC)  ;.

Emergency Lighting Systems & Controls EF Other Electric Power Systems & Controls EG Fuel S~orage and Handling Systems FX New Fuel Storage Facilities FA Spent Fuel Storage Facilities FB Spent Fuel Pool Cooling & Cleanup Systems & Controls FC Fuel Handling Systems FD

---

~----* -- - *****----~--- * *"*-*-* -~ -**- * ~-- -*- ---**F*----*** *--- **--* *** *---*~ -** ---*-***--r-*~-*- -*

• • r** *--~*-*--;. ..~---,J"**-----~- --* _.;,___: ::.....: ...... ,. ***-* -****--*-~---*""* -'"-***** -- *- ___ :_: _______ .;:.. _ _.__,_, * ...!,. ____ ;. ___ ****** . _ ... ~:~ ... --~---

_.: -----'--~ ..*... ~ ...... :.____ .___ . . --*~ .*.: --~-------***'---'---*---'----**- ..,*...... __ ... ~---'-- -*----

System Description (Cont'd.) Code

' . ' Auxiliary Water Systems Station Service Water Systems & Controls wx WA Cooling Systems for Reactor Auxiliaries & Controls WB De~ineralized Water Make-up Systems & Controls

• WC Potable & Sanitary Water Systems & Controls WD Ultimate Heat Sink Facilities WE Condensate Storage Facilities WF Other Auxiliary Water Systems & Their Controls WG Auxiliary Process Systems PX Compressed Air Systems & Controls PA Process Sampling Systems PB Chemical, Vo~ume Control, & Liquid Poison Systems & Controls PC Failed Fuel Detection Systems PD Other Auxiliary Process Systems & Their Controls PE Other Auxiliary Systems AX Air Conditioning,. Heating, Cooling & Ventilation Systems & Controls AA Fire Protection Systems &. Controls AB Communication Systems AC Other Auxiliary Systems & Their Controls AD Steam and Power Conversion Systems HX Turbine-Generators & Controls HA*

Main Steam Supply System &*Controls (Other Than CC). HB Main Condenser Systems & Controls HC Turbine Gland Sealing Systems & Controls HD Turbine Bypass Systems & Controls HE Circulating Water Systems & Controls 1IG Condensate and Feedwater Systems & Controls (Other Than CH) HH Steam Generator Blowdown Systems & Controls HI Other Features of Steam & Power Conversion Systems (Not Included Elsewhere) HJ Radioactive Waste Management Systems MX Liquid Radioactive Waste Management Systems MA Gaseous Radioactive Waste Management Systems MB Process & Effluent Radiological Monitoring Systems MC Solid Radioactive Waste Management Systems MD Radiation Protection Systems BX Area Monitoring Systems BA Airborne Radioactivity Monitoring Systems BB

---

........... -----------*--- .. ..... ------------.---* *-~-

• • - *-*-------** - ~---*-*-* ..: * ..!-._,_ ...

'1-8 8~

I the component directly involved with the shutdown or power reduction (Table 1.3); and

9. categorization of the shutdowu or power reduction. Each shutdown or power reduction was placed into one of .two sets of categories. The shutdowns and power reductions were first evaluated against design bases event~ (DBE) as described in Chap. 15 of the Standard Review Plan. 1 If the shutdown or power reduction could not be categorized as a DBE initiating events then it was placed into one of a series of NSIC categories. For further discussions of these two sets of categoriesp use of the categories>> and a listing of them. 9 see Sect. 3.1 and following.

The listings for the cause~ shutdown method, system involveds and component involved along with their respective codes are those used in NRC's Gray Book 2 series for shutdowns. Note that the information listed under the "system involvedvv column in the data tables in the appendices indicates (1) a general classification of systems (fu,11y written out) and {2) a specific system within the general classification which is coded with two letters.

1.3 Review of Reportable Events The operating events as reported in LERs and LER predecessors, e.g.~ AORs~ unusual event reportss reportable occurrences (ROs), were reviewed.

These types of reportable events were retrieved from the NSIC computer file.

Approximately five years ago~ operating experience information for aper~

ating nuclear power plants in the NSIC file for the time period predating LERs

~,._. --- ... ~* ..** . **- - - - *- - ****** .- -- .. *--------~,- ..... ~* .

... '*-*----.-~ *-*-': ., . . - **-

., .. -, *- - *-'-* _,,_'.~-***';...-: .**,. "'"*- * *.... -*-* -**"****':..___ **-**"- ' ***-- ***--**-...-*'"-"******--'-o-**F* . -;-*

-. .'~*- ::,. . ~* _: .*.

1-9 Table 1.3 Components Involv~d With the Fotced Shutdown or Power Reduction*

Component Type Component Type Includes:

'Acculliulators Scram Accumulators Safety Injection Tanks Surge Tanks Air Dryers Annunciator Modules Alarms Bells Buzzers Claxons Horns Gongs Sirens Ba~teries & Chargers Chargers

_Dry Cells Wet Cells Storage Ce~ls Blowers Compressors Gas Circulators Fans

• ventilators Circuit Closers/Interrupters Circuit Breakers C~mtactors Controllers Starters Switches (other than sensors)

• Switchgear Control Rods

• Poison Curtains Control Rod Drive Mechanisms Demineralizers Ion Exchangers Electrical Conductors Bus Cable*

Wire i*.

Engines, Internal Combustion Butane Engines Diesel Engines Gasoline Engines Natural Gas Engines Propane Engines

1-10 Component Type (Cont'd.) Coniponent Type Includes.

Engines, Internal Combustion Butane Engines Diesel Engines Gasoline Engines Natural Gas Engines Propane Engines Filters Strainers Screens Fuel Elements Generators Inverters Heaters, Electric

• Heat Exchangers Condensers Coolers Evaporators Regenerative Heat Exchangers Steam Generators Fan Coil Units Instrumentation and Controls Mechanical Function Units Mechanical Controllers

• Governors Gear Boxes Varidrives Couplings Motors Electric Motors Hydraulic Motors Pneumatic (Air) Motors Servo Motors Penetrations, Primary Containment Air Locks Pipes, Fittings Pumps Recombine rs Relays Shock Suppressors and Supports Transformers Steam Turbines Gas Turbines Hydro Turbines

G

. _,.. .: ___ . -*- .* '.J._*_.* - ............ ~ __ ,__.:, __ . . .. **- ......_..,.. -- *. ..: .. - :.*.... ~:,..:.:_. __. .:_,_ ....... _..

1-11 Component TyPe (Cont'd.) Component Type Includes Valves Valves Dampers Valve Operators Vessels, Pressure ' Containment Vessels Drywells .

Pressure Suppression' Pressurizers Reactors Vessels i*.

--....-._ ________ -- -~---* ---------- *-*-** -- ---.--------***r**-,--- -* *--~--------.-- *::* --,-~--** *.* 'h~ ->~

G

--~** ...:._.. ....;_ ..:~~--'-** ~--* -~***-

1=12 was reviewed. Any documents that contained LER-type information (equipment failure 9 abnormal event~ etc.) were coded or indexed so that they could be retrieved in the same manner as au LER.. 'Primarily>> this involved various types of operating reports and general correspondence for the late 1960s and early 1970s.

The following information was recorded for each reportable event reviewed:

l. m I.ER report number or other means of identification of report type;

2. NSIC accession number (a unique identification number assigned to each document entered into the NSIC computer file);

3. date of the event;

4. date of the report or letter transmitting the event descript:f;.on;

5. status of the plant at the time of the occurrence (Table 1.4);

6. system involved with the reportable eyent (Table 1.4);

7. type of equipment involved with the reportable event (Table 1.5);

8. type of instrument involved with the reportable event (Table l.~;

9. status of the component (equipment) at the time of the occunence (Table

10. abnormal condition associated with the reportable event~ e.g., corrosion, vibration>> leak>> etc. (Table 1.6);

lL cause of the event (Table 1.6); .and U. significance of the reportable event. Each reportable event was screened using criteria as a step in the evaluation process (See ,Sect. 3.2 and following for further discussion of the criteria 9 the us~ of the criteria, and a listing of the criteria.)

-~.

- ....*7*-----*.-;:--<*-----~

.. ,...,._.~-*---:=:..-~-"'~~-~-'":":":""::::-*-**"I".

.*', "*( -** __ , __ : . .::...-, *-*-*-*---**-::_. - -...:.*. ----"~ '.~ ** **,-* : - ** - .. .:.~-**--*~u. --.:. ..::.:, ._ ******-*---*-*----- .* .. *

• • • ) .1-13, t Table 1.4 Data' Collected for.Reportable Events - Plant Status and System'Involved PLANT STATUS A Conscruction B Operation C Refueling D. Shutdown SYSTEM A Chemical and Volume Control B Component Cooling C Condensate Purification D Condenser Cooling E Containment F Containment Air Cooling G Conta*imnent Filtering H Containment Hydrogen Control I Containment Isolation J Containment Purge.

K Conta1.nment Spray L Core Reflooding M Electric Power N Emergency Cooling/I.PSI 0 Emergency Electric Power p Engineered Safety Features

• Q Fire Protection R Hydraulic s Main Cooling T Pneumatic u Radiation Monitoring v Reactor Control w Reactor Protection x Safety Injection/BPS!

y Secondary Cooling/Aux.

z Secondary Cooling/Feedwater AA Secondary Cooling/Steam BB Service Water

. cc Shutdown Cooling DD Waste Disposal EE Ventilation FF Reactor Internals

--- ***-----------**--**~-~--****

-* ... *-*-- _,__ -- ~- __ _..:.,___~.,,. ....... ;-.... *. *~- -**** ____ __..,;,:.,,____ *-' .. -: .......;. __ ,_._, __ ,_ ____ .; ....- ~

Table 1.5 Data Collected for Re~ortable Events - Equipment Involved and Instrument Involved

• EQUIPMENT INSTRUMENTATION A Accumulator A Alai:m B Air Drier B Amplifier C Battery and Charger C Electronic Function Unit D Bearing D Failed Fuel Detection Instrument E Blower and Dampers E . Flow Sensor F Breaker F Iu=Core Instrument G Cables and Connectors G Indicator H Condenser H Intermediate Range Instrument I Control Rod . I Level Sensor J Control Rod Drive J Meteorological Instrument K Cooling Tower K Position Instrument L Crane L Power-Range Instrument M Demineralizer M Pressure Sensor N Diesel Generator N Radiation Monitor 0 Fastener 0 Recorder P Filter/Screen P Relay Q Flange Q Seismic Instrument R Fuel Element R Solid State Device

• S - Fuse S Start-Up Range Instrument

- _-T *Generator T Switch U Heat Exchanger U Temperature-Sensor V Heater W Internal Combustion Engine X Motor Y Nozzle

• Z Pipe and Pipe Fitting AA Power Supply BB Pressure Vessel CC Pressurizer DD Pump EE Recombiner FF Seal GG Shock Absorber HH Solenoid II Steam Generator JJ Storage Container KK Support Structure LL Transformer MM Tubing i-NN Turbine /

00 Valve PP Valve, Check QQ Valve Operator

. :: *-* ... ~.. :... ___ .:..... -*. ,---** --* .' - ... -**---*- --***- . -- :*.---~ . ... ~ . ~****---..:.:~ .....;. ~-* ~ .... ...--;__.._;~~~-:~..:*~-~-'- ,_.,*;..:.::_:_.... :......____._:_ ,*:...*. -- .. ~-------:........:...::..__~:..-.--~-.*: -~- --

1- .15 .

Table l. 6 Data Collected for Reportabl:e Events - Component Status, Abnormal

• Condition, and Cause COMPONENT STATUS A Maintenance and Repair B Operation C Testing ABNORMAL CONDITION A Age B Airborne Release C Conc::entration D Corrosion E crack F Crud G Environmental Anomaly H Erosion I Exposure J Fatigue' K Fire

• L Instrument Calibration M Instrument Set Point Drift N Leak.

0 Liquid Level * .........

P Lubrication Q Open/Short Circuit R. Operator Commt.m.ication S

• Operator Incorrect Action T Procedures U Records

• V Sampling W Smoke x Stress y* Stress Corrosion z Vibration AA Waterborne Release BB Wear CC Weld CAUSE A Administrative Error B Design Error C Fabrication Error *.

D Inherent Failure E Installation Error F Lightning G Maintenance Error H Operator Error I Weather

"*--~

- .. ~ ,;'__.. *~ ..

1-16 Events of Environmental Importance and Radioactivity Releases Based upon reviewing forced shutdowus 9 power reductions>> reportable events (environmental LEBs) , _and operating reports, any significant or recur=

ring environmental problems were summarized.a Tue routine radioacti:vity releases were tabulated as well, and releases where limits we~e exceeded were reviewed and discussed" loS Evaluation of Ooerating Exoerience Based upon the review involving screening~ categorizing>> and compiling data, the operating history of the plants.was evaluated. Judgments and conclus~ons were made regarding safety ~roblems, operations, trends (recur-ring problems), or potential safety concerns o From the irl.formation provided through the various operating reports and the review process, events were analyzed to determine their safety significance, using the final safety analysis report to provide specific plant and equip-ment details when necessary.

.* **- * ** )*- .. ' : . : _: *-"~--:;,;_:.. .. _,**. , ..: *::.'*-* *.:: *~*-* -* *-* **-- -**'- __ ,1.**.. . :.~:- **: ; * .;.:;* * . . * * . -*** - * - - * - - - . - ----*--**----- .* -*-. ***--****¥-~---*~--- -- _ .. _.  :,,........:; .

' I 2-1

2. SOURCES OF INFORMATION , UTILIZED. IN 1'HE REVIEW Several sources of information and periodic (annual~ quarterly, and

! l.

monthly), NRC publications were used in the review. Some sources contained.

information relative to m~re than one area withui the scope of the review.

2.1 Availability and Capacity Factors The availability and capacity factors 'were either extracted or calcUlated from data given in the Gray Books 2 from 1974 thro.ugh l979 (the first Gray Book was issued in May 1974). Prior to 1974, the annual or semiannual reports were used to compile the availability factors only.

2 .2 Forced Reactor Shutdowns and Power Reductions The review of the forced power reductions involved checking the following sources for comi)leteness of details and accuracy:

1 *. Nuclear Power Plant Opera:f;ing E3:pezoi.ence for l9XX, for the years J973, 1974-75, 1976, 1977, and 1978 (Refs. 3, 4, 5, 6, and 7). The report for 1979 has .not been published. However, since the work on the section of these reports on outages has been performed by NSIC since 1973, the draft copy of this report for 19 79 was available;

2. The Gray Book - NUREG-0020 Series; 2

3. .Annual or semiannual reports from the time of startup through 1977. For 1977 through 1979, monthly operating .

reports were used because . the utilities I*

were no longer required . . to file annuals. The review of poWer reductions invol:ved primarily the annuals, semi.annuals, and monthly reports *

, 2.3 Reportable Events

'!he NSIC computer file of LERs was the primary source of information in reviewing the reportable events.

• The material on the NSIC computer file consists of the appropriate bibliographic materta.l~ title, 100-word abstract, and keywords. When it was necessa.zy to obtain additional *information on the

. event~ the original LER (or equivalent) was consulted by (1) examining those full-size copies on file at NSIC (for the years 1976 through 1979); (2) the microfiche file of docket material at NSIC; or (3) the appropriate operating report (semiannual, annual~ or monthly report).

2.4 Enviromnental.Evenes and Radioactivity Releases Events of environ.mental importance were obtained as a result of con-ducting the overall review of the plant's operating history, -a.nd -the sources of information involve all types of documents listed thus* far.

The data for radioactivity releases were compiled primarily froDPthe report Radioactive Ma:teria.Zs Re'Z6a.sed from N~"l6a::l' PCRJJeza Plants - AnnuaZ Report 1977, NUREG-0521. 'rhis report presents year-by-year comparisons for plants in a number of different categories (solid, gasp liquids noble gas, tritium, etc.). The data for 1978 was taken from the report Radioactive Materials Released from Nuclear Power Plan.ta - Annual Report 1978~ NUREG/CR-1497, which was published in March 1981. The data for 1979 was compiled from the annual environmental reports submitted by the licensees.

2.5 Use of Computer Fiies on RECON and Special Publications Two computer files on RECON (a computer retrieval system containing data bases operated at ORNL) were used extensively for another purpose in

~35

---=------*--;------------... *-:---------*-*****- ---*****-- .,.~--:----*------ . . ----~--:'"""'*~.. --~----*--*--**.. *.-~.~.-*****- ___. ,. . . . ___

. *-,-*---:-* -------...----~-:- --::-:-* . --:.~.~.~

-,-- ---*-*--* :\- .i- *-*--*--*--- *-*-**-*--**-- -- - *- -* __.. :..-----**---*-~---**-* -- -* - . -.0..--*-**-,-4 ..-*,---~ __

. -*~ - ~ -*- - _;_ -- .... ....:;.. .__ , --* ... '"~*-----* --'* *--- _:__ --'_ :--.....:.-~: ...:..--.--~------- ._ .... __ - -* :. __ -

2-3 addition to those indicated 'thus far. Printouts were obtained from the files for Palisades to provide ~overage on other tVt>es of "docket mRt:Pr:i~.::iJn besides reportable events where the licensee may haye been ill correspondence with NRC [or the' Atomic.Energy Commission (AEC)] concerning a particular event. Licensees are often requested to submit additional information or perform further analysis. Before the LERs came into existence in the mid-1970's 9 it was not unusual for licensees to submit on their own or at NRC (AEC's)*

r~quest more than one letter transmitting information.on a particular event.

Thus these printouts provided additional sources of information on reportable events.

Sewral special publications were reviewed to provide details on events of significance. Events described in the fo'ilowing publications often con- .

tained details, evaluations, or assessments other than those provided in the reportable event (or shutdown) a8 a result of further analyses and examination:

l. Reports to Congress on Abnormal, Ocaurrences, NUREG-0020 series; "'

2. "Power Reactor Event Series" (formerly Current Event Series) published by NRC;

3. "Operating Experience Section" of the Nuc'Lea:zt Safet;y journal; and

4. NRC's Office of Inspection and Enforcement's (I&E) publications

a. Operating Experience Bulletins

b. IE Bulletins

c. IE Circulars **.

I*

d. IE Information Notices.

• -*~---------**------~-.-------,.-------**:*-~__....--~-*---------_,.---:-:---*--*-

--- -~- *-------~-*-7**.,....----*.--*-* -~-~--***-A*-~*-***--.-*- ... , ......._._,..._~

• j'"*-**-*-*-'--*--"j,*~._. __________ -...:. ____ ~ _: .* .' . ..::.. _;. ___,,__. ___ .,..,_, ..... : ..".*:....:.._.* *.*. ~:..... - * . ,, ________ .,_' .* _ ..*,, .* *~--*-.<.:...: .. .:...:. *-*-**--*--**-***-*'--** _.._: . __ ... .'.....:.---*--**-'---*- ..: :~ *.*-- c. ___,.,___.:.,._.:..*-*--**-- '"'* *., **_, *

• _:_ * ..._ _ _ . 'c.'

3-1

3. CRITERIA AND cATEGORIZATION FOR THE EVALUATIONS OF THE OPERATING HISTORY, In rev~ewing the operating history of the plant of interest, the two areas focused on were forced shutdowns (and power reductions) and reportable

.events. Given the large number of both shutdowns and reportable events, it was necessary.to develop consistent review procedures that involved screening and categorizing of both occurrences. Following screening and categorization, the study then assessed the safety significance of events and analyzed the categories of-events for various trends and recurring problems.

The shutdowns were evaluated against the design basis events (DBE's) as set forth in Chap. 15 of the .SRP. The DBE's* are those postulated distur-bances in proc,ess variables or postulated malfunctions or failures of equip-ment for which the plants are to be designed to withstand and for which the licensees are expected to analyze and include in safety analysis reports

- '(SAR). In _the. SAR, the effects of anticipated process dist~rbances and post-ulated component failures are to be examined to determine.their consequences and to evaluate the capability built into the plant to control or accomodate such failures and situations (or to identify the limitations of expected

'performance).

The intent is to organize the transients and accidents considered by the licensee and presented in the SAR in a manner that will:

1. Ensure that a sufficiently broad spectrum of initiating events has been considered,

2. Categorize the initiating events by type and expected frequency of

• 1:

occurrence so that only the limiting cases in each group need to be quanti-tatively analyzed, and

---

~-----.-*--------~--"~----~-;-*-~"7----*- .,--------~-----.,..-----------**-**--~---*---*--***--**~-.-.---:--*-:-:-*~:~--~.~--~:--~-;---:---:-.-~. ..

-. -*~:

3. Permit the consistent application of specific acceptance criteria for each postulated initiating event.

Each postulated initiating event is to be assigned to one of the following categories:

1. Increase in heat removal by the secondary system (turbine plant),

2. Decrease in heat removal by the secondary system (turbine plant),

3. Decrease in reactor coolant system flow rate,

4. Reactivity and power distribution anomalies,

5. Increase in reactor coolant inventory~

6. Decrease in reactor coolant inventory,

7. Radioactive release from a subsystem or component, or

8. Anticipated transients without scram.

Typical initiating events that are representative of those that are to be co~sidered by the licensee in the SAR are presented in Table 3.1 Those shutdowns identified as DBE initiating events were categorized as such. If the shutdown was not a DBE, then it was assigned a category

• from a list developed by NSIC to indicate the nature and type of error or failure. The NSIC categories for non-DBE shutdowns were examined as part of a trends analysis.

The reportable events were screened using the criteria presented in Sect. 3.2 (and following) and were categorized according to their significance.

~

The information collected on the reportable events (as outlined in Tables 1.4 through 1. 6) was used to analyze trends for all reportable events - those identified as significant or non-significant.

- -~--- ... --~-- - .. - *-** ___,...,..___ ,,_. ... ---*-*. - - ............. ~~---*

.. ( .. ..:.. - **-*-*-----~ **-** _____ ....., __ .,_~-~--..:... -- ,*.. *'..'.,_ *' _:;::_*,**-*****--*-**-*--*-****** * *. -~'---__:_ ** ,..'....-~*-*** .*. '******* *_____ :._ ____~_,__-~---~ .c*. __ :.: ***. --.-*.:.:...-.......... :**. ,.* ', -*-*****-

l J.

.i 3-3 Table 3.1 Initiating Eyent Descriptions for Design Basis Events' as Listed in Standard Review Plan- Chapter 15 (Revi.sion 3) *

1. Increase in Heat Removal by the Secondary System 1.1 . Feedwater system malfunctions that result in a decrease in feedwater.

temperature 1.2 Feedwater system malfunctions that* result-in an increase in feedwater flow 1.3 Steam pressure regulator malfunction or failure that results in increas-ing steam flow 1.4 Inadvertent opening of a steam generator relief or safety vaive 1.5 Spectrum of steam system piping failures :Ulside and outside of contain-

- ment in a PWR

2. Decrease in Heat Removal by the Secondary System 2.1 Steam pressure regulator malfunction or failure that results*in decreasing s~eam.flow 2.2 Loss of external electric load 2.3* Turbine trip (stop valve closure) 2.4 Inadvertent closure of main steam isolation valves 2.5 Loss of condenser vacuum 2.6 Coincident loss of .onsite and external (offsite) a.c. power to the station 2*. 7 Loss of normal feedwater flow 2.8 Feedwater piping break

3. Decrease in Reactor Coolant System Flow Rate Single and multiple reactor coolant pump trips

• BWR recirculation loop controller malfunctions that result in decreas-

• ing flow rate 3.3 Reactor coolant pump shaft seizure 3.,4 Reactor coolant p1lmp shaft break

4. Reactivity and Power Distribution Anomalies 4.1 Uncontrolled control rod assembly withdrawal from a subcritical or low power startup condition (assuming the most unfavorable reactivity conditions of the core and reactor coolant system), including control rod or temporary control device removal error during refueling 4.2 Uncontrolled.control rod assembly withdrawal at the particular power level (assuming the most unfavorable reactivity conditions of the core and reactor coolant system) that yields the most severe results (low

• power to full power) .

4.3

• Control rod maloperation (system malfunction or operator ~.rror),

including maloperation of part length control rods .

• 4.4 Startup of an inactive reactor coolant loop or recirculating loop at an incorrect temperature 4.5 A malfunction or failure of the flow controller in a BWR. loop that results in an increased reactor coolant flow rate

-~-- -~---* -- ------ - *- - -*------------ --- ---* -~-------- - -- -----.------ - ~~......-----~------~--- -~- .. ,.---**-*-.-~~

3-4 Table 3ol (continued) 4.6 Chemical and volume control system mal.function that results in a decrease in the boron concentration in the reactor coolant of a PWR 4.7 Inadvertent loading and operation of a fuel assembly in an improper position 4.8 Spectrum of rod ejection accidents in a PWR 4~9 Spectrum of rod drop accidents in a B'WR So Increase in Reactor Coolant Inventory 5.1 Inadvertent operation of ECCS during power operation 5.2 'Chemical and volume control system malfunction (or operator error) that increases reactor coolant inventory 5.3 A number of BWR transients, including items 2.1 through 2.6 and item 1.2

6. Decrease in Reactor Coolant Inventory 6.1 Inadvertent opening of a pressurizer safety or relief valve in a PWR or a safety or relief valve in a BWR 6.2 Break in instrument line or other lines from reactor_coolant pressure boundary that penetrate containment 6.3 Steam generator tube failure 6.4 Spectrum of BWR steam system piping failures outside of containment 6.5 Loss-of-coolant accidents resulting from the spectrum of_postulated .--.

piping breaks with;i.n_t;he_reactor coolant pressure boundary, including steam*line.breaksinsid~ of containment in a BWR 6.6 A number of BWR transients, including items 2. 7, 2."8, and 1.3

7. Radioactive Release from a Subsystem or Component 7.1 Radioactive gas waste system leak or failure

• 7.2 Radioactive liquid.waste system leak or failure 7.3 Postulated radioactive releases due to liquid* tank failures 7.4 Design basis fuel handling accidents in the containment and spent fuel storage buildings 7.5 Spent fuel cask drop accidents 8o Anticipated Transients Without: Scram 8.1 Inadvertent control rod withdrawal 8.2 Loss of f eedwater 8.3 Loss of a.c. power 8.4 Loss of electrical load 805 Loss of condenser vacuum 8.6 Turbine trip 8.7 Closure of 'main steam line isolation valves

. -**~ ..-------- ----~-------**-**** ~ ... ~-- . ----*---- --* ___,__.., __ ,_..._,~*--* .,_. __

f __ ..:__*----~*-*1 *-***-*-.:. __ _.. *--~- --***-'~-:._...: __.:_.'._..::... ***- --~--~------ ...*--*-*- - --~

: . --** .. :__ '---~-'"'-~*"--'---.::_. ~ -*- ... *- * -*-** *..:._.._., __...:_...:..._~-*:..*~.-** .. :*.: .....-..:""'-..:....... _~-*- ~ _..:_ __ **** .,..J..._**-*'-"* ~..:.._, __ *.,_::,. **~:..i-.:1.~S .. *.*. .' ....c**** * * *. :*

3-5 The review approach with respect to operational events (forced shut-

• downs. and reportable occurrences) consisted primarily of a three-step process:

1) compile information on the events, 2) screen .the events for significance using selected criteria and guidelines, and 3)" evaluate the-significance and importance of the events from a safety standpoint. The evaluations were to determine those areas where safety problems existed in terms of systems 9 equipment::111 procedures~ and human error.

The reviewers worked semiindependently (brief exchanges of ideas and information) and then were br~ught together periodically for discussion and final resolution as to how events were to be categorized and how the criteria were to be used.consistently~  :

3.1- Significant Shutdowns and Power Reductions For the purpose of compiling information and for evaluation, the power reductions were treated in the same manner as the forced. shutdowns *.

3.i.1 Criteria for significant shutdowns and power reductions

• As indicated previously, the occurrences identified as design basis events were used as criteria to categorize and note significant shutdowns.

These events are listed in Table 3.1 as they are found in SRP Chap. 15.

3.1.2 Use of criteria for determining significant shutdowns and power reductions The generic DBE initiating event types, e.g., "increase in heat removal by the secondary system" or "decrease in reactor coolant system, rv were used i*

as primary flags.for reviewing the forced shutdowns (and power reductions) *

. Once the generic tYPe of event was identified~ the""_partiCular initiating*

event was determined from the details associated with the shutdown. For example., if the reactor shuts down because of an increase in heat removal

• ---*----..... ,.,,____ -* ---~----"- ----* --- -~--

due to a feedwater regulator valve failing open, the shutdown is a DBE generic type l evento Specifically~ based upon the initiating event (valve failed open)p it is a lo2 DBE~ feedwater system malfunction that results in an increase in feedwater flow. Some shutdowns were readily identifiable as specific DBE's, such as tripping of a main coolant pump ~a 3.1 DBE.

Once categorized as a_DBE, the shutdown was considered significant regardless of the resulting effect on the plant (because a design basis event had been initiated)o Loss of flow from one feedwater loop Wa.s considered sufficient to qualify as a 2.7 DBE~ loss of normal feedwater flowo The closure of a main steam isolation valve in one loop was considered sufficient to *qualify as a 2.4 DBE ~ inadvertent closures of main steam isolation valves.

3.lo3 Non-DBE shutdown and power reduction categorization

• Those non-DBE shutdowns were assigned NSIC categories (Table 3.2) to provide more information on the failure or error associated with the shut=

down. With these categories, more specific types of errors and failures could be examined through tabular summaries to focus the reviewer's attention on problem areas (safety-related or not) that were not revealed by the DBE categories

• The causes for non-DBE shutdowns taken from the.Gray Boo~. (listed in Table 1 1) are limited and very general, while NSIC cause categories 0

, .... .'1*-*----*-------.:_**}----:.......-.,-....---~--. -- . ~-;:---~ ---*-* ---'-*~-~ ....... *.... ;__ ., --.... :. .... __:_,___ _:. __ ~r-~~- .' *.***.* *------*----*---- . . ~---* _.*- ___ _.__ , __*__ .. : ** _,._, __ ._

' 3-7 Table 3.2 NSIC Event Categories for Non-DBE Shutdowns N

• 1~0 . Equ:i.pmen t Failure N 1.1 Failure on demand under operating conditions N 1.1.1 Design Error

• N 1.1.2 Fabrication Error N 1.1.3 Installation Error N 1.1.4 End of design life/inherent failure/random failure

• Failure .on demand under test conditions N 1.2.1 Design Error N 1~2.2 Fabrication*Error N 1. 2. 3 Ins*tallation Error N l.2o4

• End of design life/inherent failure/random failure N. 2.0 Instrumentation and Control Anomalies N 2.1 Hardware failure N 2.2 Power supply problem N 2.3 *.Setpoint Drift N 2.4 Spurious signa!

N 2.5 Design inadequacy (system required to. function outside *~

design specificati0ns}

N

• 3.0 Non-DBE Reductions in Coolant Inventory (Leaks)

N 3.1 In prilllary system N 3.2 In secondary system and auxiliaries N 4.0 Fuel/Cladding Failure (densification, swelling, failed fuel elements as indicated by elevated coolant activity)

N 5.0 Maintenance Error N 5.1 Failure to repair component/equipment/system N 5.2 Calibration error N 6.0 Operator Error N 6.1 Incorrect action (based upon correct understanding on the part of the operator and proper procedures, the operator turned the wrong switch or valve + incorrect action)

N 6.2 Action on misunderstanding (based upon proper procedures and improper understanding or misinterpretation on the part of the operator of what is to be . done + incor+ect action)

N 6.3 Inadvertent action (purpose and action notjrelated, e.g.,

bumping against a switch or instrument cab.inet)

.... ~;.

. i*

. l___ --***------------ - *-- ------------..-----*---.. - -- -* *-----.---------* ...

--*.--~ ----------~--------- . *r-*--- ......~~~ ----i'~-.. -****-.*-*.*--~--*c**---.-.---=

.. '-*---*-*-~***- **---*"-~**-**- .. ~**

N 7.0 Procedural/Administrative Error (Incorrect operating or testing procedures Incorrect analysis of an ~vent~f ailure to consider certain conditions in analysis)

N 8.0 Regulatory Restriction N 8.1 . Notice of generic event N 802 Notice of violation N 8.3 Backfit/Reanalysis N 9.0 External Events N 9.1 Human=induced (sabotage 9 plane crashes into transformer)

N 9.2 Environment Induced (tornado>> severe weather, floods, earthquake)

N 10.0 Environmental Operating Constraint as Set Forth in Tech Specs

_. . -*** -- :: .. ~ *--****-- :_ ~: .*.. *:... ... _: . - .. -- - **-*--*-*--- *- _...... :: . ... *... ..:. ...~ . --~ **--~-....,. . _;.,._~_...::, ___ - __..-* -.

3-9 are mere spe'cific.. Thua, 'as an ey.2:nple, the numb,er. of. G;ray Book causes *noted

. as equipment failure should not be expected to equal those identified as equipment faf.lures with the NSIC categories. Other NSIC categories, such as component fcP,lure, could be classified as an equipment failure if the*only available designations for cause were those listed in the Gray Book *

. 3.2 Significant Reportable Events 3.2.l Criteria for significant reportable events Two groups of criteria were used in determining significant reportable events. The first set of criteria (Table 3.3)° indicates those events that are definitely significant in terms of safety and are termed significant

• Th.ose criteria in Table' 3.4 indicate events that may be of potential con-cern.. These events, which might require additional inf~rmation or evaluation*

to determine their full implication, were noted as conditionally*sign:ificant.

3.2.2 Use of criteria for determining significant reportable events The reportable events were all reviewed applying the two sets of criteria for significance rather liberally *. A number of significant events and conditionally significant events were noted. The events initially identified as significant or conditionally significant were analyzed and evaluated further based upon (1) engineering judgment; (2) the systems.,

equipment9 or components involved; or (3) whether the safety of the plant was compromised. The final evaluation for significance considered whether a DBE.was initiated or whether a safety function was compromi~ed such that the system could not mitigate the propagation of events for which it was designed. Thus, the number of events categorized finally as significant was reduced considerably by these steps in the review process.

--*---* -*---~ --~----~--~-- ---- .. ----------,...---. *- *---*-:-**----.~~.~~ -~-*--* -* *~-*--*-*--------------,---~--.-. ........-.------.-..- . .--~

Table 3. 3 REPORTABLE EVENT CRITERIA - SIGNIFICANT SIGNIFICANCE CATEGORY EVENT DESCRIPTION Sl Two or more failures occur in redundant systems during the same evento S2 Two or more failures due to a common cause occur during the same event~

S3 Three or more failures occur during the same event.

S4 Component failures occur that would have easily escaped detection by testing or examination.

SS An event proceeds in a way significantly different from what would be expected.

S6 An event or operating condition o.ccurs that is not enveloped by the plant design* bases.

S7 An event occurs which could have been a greater threat to plant safety with different plant condi-tions~ the advent of another credible occurrence, or a different progression of occurrences.

SS S9 Administrative, procedural or operational errors are committed that resulted from a fundamental misunderstanding of plant performance or safety requirements.

Other . (explain).

~ -~i"-*~*-~-***- ----*{-*.: .. ,_*_..:....__:._ -***

3-11

• • Table 3.4 CATEGORY FOR REPORTABLE EVENT CRITERIA"""'. CONDITIONALLY SIGNIFICANT CONDITIONAL SIGNIFICANCE I EVENT DESCRIPTION Cl A sipgle failure occurs in a non-redtm.dant.system.

C2 Two apparently unrelated failures occur during the same evento C3 A problem results in an off~site radiatiqn release or personnel exposureo

• C4 A design or manufacturing deficiency is identified as the cause. of a failure or potential fail~re.

cs* A problem results in a long outage or major equip-ment damage.'

C6 An ESF actuation occurs during an event.

C7 A particular occurrence is recognized as having a significant recurrence rate. . .

ca .Other *

--*-----------~~~--- -* -- - -*- --*-. -:-**** *-----------*--**----;--*--.

3-12 Those events involving radioactiyity releases were automatically gorized as a conditionally significant 3 event and held for discussion 0

cate~

in the respective ' environmental and release" sections of the report.

3.2.3 Non-significant reportable events Those reportable events not identified as significant or conditionally significant were categorized as non-significant (with an "N" in the signi=

ficance column in the coding sheets in the appendices). These events and the events rejected during the additional review step as noted above were further reviewed by compiling a tabular summary of the systems (Table 1.4) to detect trends and recurring problems. The systems selected yield meaningful information. concerning the system's ability to mitigate. accident sequences or mitigate the effects of such accident sequences.

-- - ~ - - ' --

... .' ... ,... :..:._., ....... ..:_ *:.. - : . * * -- -**-*---*~*- ...... ~ * **-**~---.::.....:.._*__:.': . " .:. *** ------.'..J~.:.~ :.._* *:. *' . _.:_.,..:_:.:...;,,.~;_:,":,_,.;:,._,':.... :;,~ . . . *. *'*" ** .~........_.:.;.. .... ~'--::.:.--.,:.!......:.:....:.._,,.....__c.;:.........>.::.. I

' 5-l

5. OPERATING EXPERIENCE REvIEW OF PALISADES 5.l . Summary of Operational Events of Safety Importance

• • .\ The operat~onal history of Palisades has been reviewed to indicate those areas of plant performance that have compromised plant safety. The review included a d~tailed examination of pl~nt shutdowns, power reductions, report-able events, and special environmental impacts. The criteria used to show degradations in plant safety were (1) events that initiated a DBE and (2) events that compromised safety functions.designed to mitigate the propaga-tion of the initiating events.

Shutdowns and power reductions indicated the n~ber and types of DBEs entered. The reportable events and special environmental impacts indicated the number of times each engineered safety function was compromised. The results of the ~nalyses identifi~d-fifty-two DBEs entered. Additionally, eight events.were* identified wher~ loss of safety sys~em function occurred in:*.engineered safety features.

• 5 .. 2 General Plant Description Operated by Consumers Power Company, Palisades is a Combustion Engineering

. pressurized-water reactor (PWR) located on Lake Michigan at South Haven, Michigan. The nearest city, at a distance of 33 miles, is Holland, Michigan.

The population within 30 miles is 210,000. Within 50 miles, there are 1,000,000 people,_ including the populations of Kalamazoo, Michigan, and 1-South Bend, Indiana.  :.**

j

--1 I

_i _________ .. _ ... - *-- ... ~-.------~------*--- - -~---~--."':**--------*--* -~-*:-**-"------*.~-.--~~*

5-2

-~

The reactor is a two-loop PWR. system with a licensed thermal power of 2530 MWT and DER of 805 MWe, which were increased from 2200 MWT and 668 MWe iii. November 1977. The containment structure is a reinforced concrete cylin-d~r with a steel liner. The architect~engineer is Bechtel Company~ Palisades achieved initial criticality on May 24~ 1971, and began commercial operation on December 31, 1971, under full term. operating license DPR~20.

5.3 Availability and Capacity Factors Table 5.1 presents the Palisades availability and capacity factors (reactor availabilityp* unit availability, unit capacity using the MDCP and unit capacity using the DER).* Values range from lows in 1974, when the unit was shutdown for the majority of the year for the repair of steam generator and condenser tube leakage, to highs during down only i9 days throughout the entire year.

5.4 1977~ when the reactor was shut-Review of Reactor Shutdowns and Power Reductions This segment of the review of Palisades' operating experience involved 129 forced shutdowns and 22 forced power reductions. The compilation of data

- the date, duration, power level, description, cause, shutdown method, system involved, cqmponent involved. and SRP or NSIC category - associated with each forced shutdown or power reduction are presented in Appendix B Part 1. In reviewing the forced shutdowns and power reductions, one is examining events of considerable :importance because the plant was forced to shutdown or- to reduce power as a result of some abnormal condition. However, plants are

,.,_ - -.-.':. - *r* ** o* ; ___ --*

--*N* ,_ ****--!

~ * **:.:.:..** *.:~: .. J..'...._:.,;.._.,_:..**:.*:*.:'_'_. -~~ *-'****" --~--- <.J:;:.* .:.,._. ___ *- < <. ** . : , < -**--***-*~~*-* . . . - ........................... - .......

5-3 Table 5.1 .Availability and Capacity :Factorsfcir P'alisades 197'2 ' 1973 i974 1975 1976 1977 '1978 1979

' I

• .React~r Availability

• 61'.l. 47.6 7.6 66.8 59.0 92.8 54.8 61.3 Unit Availability 57 .*o 43.8 5.5 64.5 55.2 91.4 49.7 59.9 Unit Capacity (MDC)

• NDt NDt 1.3 40.5 44.0 91.4 47.2 61.7 Unit Capaci.ty (DER):j: NDt. NDt 1.0 33.7 34.7 72.1 37.2 48.7

• MDC ~ Maximum Dependabl~ Capacity tND ""No Data tnER = Design Electrical Rating

~*-

• ~-~ ...*------. ~--"**

'*-** :_.__ *-*---* - ,.,,. , .. ~ .... . .. **-** **-*-*-**--- **-*--- *-*-~- .. ,.,,._ ......

5-4 1

de'Signed with sufficient redundancy of backup systems and engineered safety features to mitigate these abnormal conditions and bring the, reactor to an orderly, safe shutdown. Some abnormal conditions that result in a forced shutdown are identifiable as initiating events of DBE accident scenarios that could result in potentially serious consequences in terms of damage to the plant and release of radioactivityg i f successive failures in backup systems and engineered safety features occurred *. These DBE events war.rant special consideration from a safety standpoint because the first step in an accident sequence has been realized.

5.4.l DBE Initiating events Palisades has experienced 52 forced shutdowns identifiable as DBE initi=

ating events. Table 5.2 lists these events b~ SRP category and by y~ar. The

- following di~cussion will. provide in£_ormation on these events according to ~.

their respective DBE category.

Note.that for the forced shutdowns and power reductions often only a ""

couple of sentences regarding the details and description of the shutdown are available. If the shutdown is also a reportable event or serious in nature~ additional details are provided. This is especially true after 1977 when licensees were required to submit monthly reports only instead of annual or semiannual operating reports. The shutdowns as discussed in the annuals

~nd semiannuals generally provided more information_tha~ the monthly reports.

~---

- -*-**--*------*----.. . . ---~...-':'"-"'""-....... ..,~ ..~----.... --:-'--**---***---**--*-,,,-*-----:----**-*--...,......~-----**-**-~-----.,.~*--*"'*'*:*...,-*:*~'?"':""-'"'~'.'.**.-*-*or:-

I

• 1 I

i I

I Table 5o2 DBE Shutdowns for Palisades j

I j

Year ~.

I DBE Category r I! 1972 1973 1974 1975 1976 1977 1978 1979 To-tal  ! .

I

• \ 1.2 FW system malfunctions that result 1 1 I

in increase in FW flow 2.2 Loss of external electric load 1 3 2 6

2. 3. Turbine Trip 1 2 1 5 iI. 2.4 Inadvertent Closure of MSIV valves 1 3_ 4 I

2.7 Loss of normal feedwater flow 2 J (1 or 2 pumps) 9 3l 4 8 4 30  :' ".

3.1 Single and Multiple Reactor Coolant 1 1 Pump Trips 4.3 Control Rod Maloperation (system or 4 1 1 6 operator error) including malopera-tion of part length control rods iI TOTAL 1: 11 2 0 7 8 4 12 8 52 1*

. --1 l.11

• • I l.ll I ,

• --**-**------~*-****- li __c  :----.---~~1 5-6 5o4.lol DBE Sect. 1 event - increase in heat removal by secondary system. Only oue Sect. *l event was identifiedg a 1.2 event where a feedwater system malfunction resulted in an increase in feedwater flow.

The event occurred on January 28~ 19795> when a feedwater regulatin~ valve failed to the open position requiring a manual scram of the plant~ After repairs to the valveS> the unit was placed on line after almost 19 h. There were no apparent significant effects from this event.

5.4~1.2 DBE Sect. 2 events - decrease in heat removal by the secondary system. Palisades experienced four different types of Secto 2 events (Table 5.2)g (1) 2.2 loss of external electric load (6); (2) 2.3 tur-

• bine trip (4); (3) 2.4 inadvertent closure of MSIV valves (4); and (4) 2.7 loss of normal feedwater flow (30).

The first loss of external electric load occurred on March 12, 1972, when the reactor was manually*scrammed and the automatic trip failed on opening of the 345-kV plant output breakers due to a faulty pilot wire circuit. (See Sect. 5.5.1.1 of the reportable events for additional details on this event.)

Three other loss of eXternal electric load incidents occurred in 1976 on November 24, November 25, and December l.Jl respectively. On November 24, the main electric generator and reactor tripped while attempting to manually regu~

late the main electric generator voltage. The November 25 and December 1 events were attributed to generator voltage regulator failure with no addi-tional explanation. The cwo*1979 (April 25 and 30) loss-=ef-load events were due to voltage regulator problems - one specified as a defective component in the voltage regulator and the other unknown.

. i

• \

-~----*------ .. --~

., .. -******- ****l- .. *-* -**-****-***** --**-**- "**--*--*---*- **- *-- ***-*-***--- .. .. ------*~'-* '* -* - ~-* . -***-----'**-*:.__._ ..:.. **-

5-7

• ~our DBE (2.3 events) involving turbine trips occurred on April March 6 and July 8, 1973; and in 1975. (This 1975 event probably 5~ 1972; occurre~

in earlr August; it was not listed in the Gray Book but was listed in the annual report_ for. l~.75 .* ). ,_The April 5, 1972.., turbine trip was caused by a plugged filter in the hydraulic oil system. On March 6,. 1973, the turbine

• was manually tripped (automatic reactor trip followed) when the load limiter malfunctioned. The malfunctioning electrohydraulic control system switch was

. replaced. The July 8, 1973, event occurred during turbine valve testing while reopening the No. 1 stop valve; investigation revealed that the trip was caused* by a leaking diaphram in the turbine vacuum trip mechanism* The .f\

only information obtained on the 1975 event indicated that the turbine tripped due to insufficient loading shortly after synchronizing from the previous outage.

Palisades experienced four instances where main steam isolation valves (MSIV) closed inadvertently and led subsequently to the shutdowns. The first MSIV closure occurred on August 31, 1976, when the reactor tripped due to high primary coolant system/pressurizer pressure. The high primary system pressure was the result of a MSIV closure; it closed on loss of air caused by the malfunction of a 4-way valve off the instrument dryer. The second MSIV closure took place on May 20, 1978, arid was induced by steam flow across the valve disc. Low water level in the steam generator occurred as a result of shrinkage caused by closure of one MSIV, and the reactor subsequently 1:

tripped on the low steam generator water-level signal. To pre~ent recurrence,

• ------- ----------------------** -**-**--------.-.-. -.-------~------*-;*-*-*--*-~-------~*---~----*

l; modifications to the valve operation that would increase the opening force on the valve disc were being considered. Two days later (May 22, 1978)p both MSIVs closed resulting in a plant trip. Pending modificationsp plant power was held at about 90%. No other information regarding the shutdown or modifications was found. The fourth and final MSIV closure on June 11, 1978, was caused by low pressure in the air supply to the valve operator. The monthly report for June ~ndicated that this closure appeared to be unrelated to those that occurred in the preceding month of May.

The largest number of DBEs e~perienced by the plant involved the 2.7 events - loss of normal feedwater flow. These 30 events account for over one-half (58%) of the DBE shutdowns. Twenty-three of these 30 events occurred during either reactor startup./power ascension conditions or at ze~o/low

- power .leve-ls .-

• Although they are to"o numerous to discuss individually the events are described in Appendix B Part 1.

loss of feedwater involved the trip of feedwater pumps.

The most prevalent cause of A low steam generatQr signal from the Reactor Protection System in either of the two loops will shut the reactor down. The causes for the f eedwater pump trips ranged primarily from low suction pressure to i'undefined" pump problems to vibration.

Other causes for loss of normal f eedwater flow included unstable flow con-ditions due* to vibration, feedwater regulator and bypass valve problems, and plugged condensate pump strainers. The most noteworthy shutdown resul-ting ftom loss-of-normal-feedwater flow occurred on August 10, 1979>> ~hen the reactor was manually tripped from 88% power, following the trip of both feedwater pumps during te~ting of turbine valves. Again, details are lacking

- =.::

as to what happened, but this event could point toward a common cause candidate since both feedwater pumps were tripped.

---

*-*--**----*** *-**********- *** -~--**-*-* --*-- ****"'""-"'* *--*-... ~r.-*:..-:. *..... _ *"*-* ** *1~._: __ ., ________ *** **--*-...__,..__ , ** *** >'"---*r_ _ ,_ .,_,_,.,._... ,_. -*---.---*******--*-*---*-*-*-.---...-~--,-'"-:-;-___.,.**-"'"7"*:"-"*"".;**-***.,,--<,"":""'{"

5-9 AsstJ1I1ing eight years of commercial operation (actually slightly more than eight because Palisades was declared comm:e;-cially operational in December 1971), Palisades has experienced three to four shutdowns per year due to loss of normal feedwater flow. Although loss of normal feedwater flow is an expected event of moderate frequency, this appears to be a somewhat high number of demands for operation of the auxiliary f eedWater system where only two auxiliary feed pumps (one motor-driven and one steam-driven) are avail-able. Hpwever, to. date.there is no reason to suspect any problems from the auxiliary feedwat.er system because .it has operated reliably on numerous occasions.

5.4.1'.3 DBE Sect. 3 events - single and multiple reactor coolant pump trips. The only DBE Sect. 3 event that* occurred during Palisades'

• operating history took place on February 1, 1979; it was a single reactor coolant pump trip (DBE 3.1 ev~nt). ,T!ie February monthly report stated that the primary coolant pump was inadvertently stopped due to personnel error resulting in an automatic scram of the reactor. The plant trip was followed by a partial main cooling system (MCS) cooldown and safety injection system initiation. The MCS pressure was recovered and no HPSI pump water was intro-duced into the MCS. An LER (LER.79-012) discussing this event in conjunction --

with inadequate shutdown margins was filed on April 3, 1979; however, there was no mention of personnel error.

-~---* *------------..**----.~*~-**----*~--*;-*----,-~---* *-* *-:** _ _,..___..~***-*--~-......-.-----.-**-***-------*.,---.-* **----~-,.-*--:~--- .--: ,...,-~v-7':;*~-:-.

5-10 anomalies.

5.4.1.4 DBE Sect. 4 events - rea*ctivity and power distribution The 6 Sect. 4 events were all 4.3 events tion (system or operator erro'r) including ma.loperation of part-length control

= control rod ma~opera rods. Four events were identified in 1975 - July 27, August 17, August 30, and September 6; the remaining two occurred on August 27>> 1976>> and on June 6, l978s respectively.

On July 27 9 1975 5 the reactor was manually sbutdoWD. when a second control rod drive mechanism (CRDM) became inoperable. CRDM 27 had been declared in-operable shortly after the previous startup with an ap~arent dragging brake.

During testing on July 27, CRDM 33 was inserted but could not be withdrawn.

CRDM 27 would not insert due to a loosened brake rotating element caused ~y loose set screwss and CRDM 33 had borated contacts. The.brakes were cleaned and adjusted. :__.

The other three DBE 4.3 events for 1975 involved dropped roqs. Cont.rel rod No. 19 dropped into the core on August 17, 1975, resulting in a manual .,

shutdown. Investigation indicated that the clutch coil had apparently shorted internally causing the rod to drop; the clutch was replaced. Similarly, on August 20s 1975s control rod No. 11 dropped into the reactor and its clutch was replaced. Finally, on September 6, 1975, rod No. 16 dropped into the core due to a clutch failure and its_ clutch was replaced as well. The failed clutches were due to high ambient temperatures according to the utility and installation of cooling fans has eliminated this problem.

On August 26 (preceding the August 27 5 1976 shutdown) control rod No. 39 failed to withdraw with Group 4 control rods while attempting to escalate power.

A rod drop test was successfully performed while the attempt to withdraw failed.

As a results the plant continued to operate at 25% power with the dropped ~od *

. *-- __ ,. _______ *--** - * - * - - - .--~* _,.. *-***._~,.,--;-i-;-**r.,"* . ---..-~-~--*,-.~------.-'--*--~;:--*~- ~------.--.~,~.***.--......-~

-- -----..,----- - ---- ~------------ -- ---* . -- .... ~----------- --*

--*-----'-~---------*---- -*-*-** _ , ___ ,.. _____ -- -- -----------~-~-. - -- - ----*-----~*_.___** ....-.!.-~- _** .:. .* -*----***-'-*-*-*--*-***---*-.!.:~--._ ____ ;________ ._

--* -*----l"'~~---*-------~-:--**-----*--*- -*~- -.

5-11

• * :' While operating with a dropped rod, a core tilt of 9%.was expe~ienced, pr~pting the plant shutdown. While maintaining the reactor at low power with maintenance 'about to begin on CRDM 39, CRDM 37 became misaligned.

After shutdown, the inoperable motor and brake package of CRDM 39 was replaced. The mode of failure.was determined to be an open motor winding, although a remote resistance check on the motor windings was previously made and did not indicate this mode of failure. The apparent cause of failure was an intermittent brake failure; continuing efforts to insert and withdraw the rod with the sticking brake caused the winding to open due to excessive motor current.

After repairs were complete on CRDM 39, attempts to withdraw CRDM 37 failed. The motor and brake assembly were replaced with the rebuilt brake of CRDM 39 and a new motor. An investigation indicated that CRDM 37 failed due to the brake. Disassembly of the brake revealed an excessive amount of residue from the brake disc that interfered with the close brake clearance of 0.008 in *. , causing a dragging effect.

During testing of CRDMs on June 6, 1978, control rod No. 3 could not be moved by its operator; it was considered inoper'able and the reactor was shut-down. During the shutdown, the rod was manually scrammed and inserted to **

the bottom of the core. Failure of the rod to move was caused by the backing out of a setscrew in the brake assembly. The motor-gearbox-brake assembly was replaced with a spare.

1-*

. *----*-*---**--* -*- .- **-."~*---~. --.-~---- --*-* **-- ---~---~*-*"*-- -----.--*. -----*-- ------.-~-------.-- - ***~*,.*:--~~-~,

5-12 5.4.2 Trends and safety implications of shutdowns and power reductions This section and its subsections will highlight those events not other=

wise designated as DBE shutdowns or as shutdowns that were also reportable events. After a brief summary of these non-DBE shutdowns for Palisades from 1971 through 1979, three specific problem areas will be discussed.

Beginning commercial operation on December 319 197ls Palisades promptly experienced its first forced shutdown (manual) on the same day due to HPSI valve problems. During operations prior to commercial generation~ Palisades experienced two significant events involving a depressurization event and a loss-of-offsite=power event coincident with a failure of one of the diesel generaeors to load. These two events are discussed in Sect. 5.5.1.l and 5.5.l.5, respectively, in the reportable events *section.

reportable events secti9n.

As evidenced in Table 5.Ja~ which summarizes the forced shutdowns for

.Palisades, Palisades experienced some 29 forced shutdowns in 1972. The year began with a contintiation of the outage from 1971 to repair a HPSI valve with damaged internals. On March 12, the plant experienced a failure to transfer power from the auxiliary transfoDDer to the startup transformer (Sect. 5.5.1.1) *.

On March 15, the first problems with seal leaks on the CRDM surfaced. This problem plague~ Palisades for several years, and it had to shut down frequently to repair the seals. Seven of the 29 shutdowns were concerned with repairing

• j Table 5,3a Forced Shutdown Summary_ for Palieadee i . .'

1971 1972 1973 1974 : 1975 1976 1977  : 1978 1979 Total I. Forced Shutdowns

1. Total NU1Uber 1 29 9 5 p 22 13 27- 10 129

2. Total lloura Dow 8 2991.6 4907' 8275 2826 998 778  : : 1914 760 23449.6 .

3. Cause* .!

. ' ~

A. Equipment Failure 1(8) 16(935.3) 6(3466) 5(2305) 13(621) 19(835) 9(663) . ; 24(1888) 9(272) 102 (10993. 3) i I

B. Maintenance or Testing D. Regulatory Reatric-

- 6(1747) 2(1432) 1(25) 'I

\-

- 9(3204) 0(0) r.

i.

• tion I

E.* Operator Training/ 1(12) 1(9) 2(21)

License Exam F. Administrative 1(15) 1(15) i G. Operational Error 5(36.3) 2(50) 1(18) 1(25) 9(129.3) I II. Other 1(113) 2(72) 3(26) . 6(211) . I

! 4. Shutdown Method

1. Manual 1 13 4 4 12 I

j 2. Manual Scram 2 i 8

1 4

4 11 4-57 12 I

3. Automatic Scram 14 4 .1 4 10 5 16 6 60 i l 4. Continuation 1(246) 1(5970) 1(2205) 1(463) . 4(8884) 'I**

I r

II. Total number of SRP Related 11 2 1 8 4 12 8 52 f i

j Shutdowns (These are in- I" t

J eluded in Totala of Part I) ~.*

.l! *Numbefi of hours asaociated with cauaa of shutdown la in parentheses, :i 1

Il I

• j l . ,,*

• VI

.....I w

i

! ~

'A'able S,3a (conUnued) Forced Shutdown Summary lor ~mll.i~adea 1971 !972 1!1173 1974 U7$ 1976 1977 1978 1979 Total XIX. Syst!!lll Involved

l. Reactivity Control Systems (RB) !II 6 4 1 11 31

2. Coolant Recirculation Systems & Controls (CB) l i 2

3. Other Coolant Systems (CJ) l 1

4. Containment Isolation & Systems Control (SD) 1 1

s. ECCS & Controls (SF) l l 1 3

6. Reactor Trip Systems (IA) 1  ! l l 4 7.

a.

Other Instruments Required for Safety (IE)

Offs!te Power Systems & Controls (EA) l

-* JI.

l

ll l

s

9. AC Onaite Power Systems & Controls (EB) 2 2 4

10. Composite AC & DC Systems (ED) l l

u. Compressed Air Systems & Controls (PA) ll. l

12. CVCS & Llqu!d Poison Systems & Controls (PC) 2 2

13. Air Cond,, Heating. Cooling, & Ventilation (M) l 1

14. Turbine Generators & Controls (HA) 3 4. 3 3 4 2 2 21 I 15.

16.

Hain Steam Supply System & Controle (118) l 2 ll. 7 11 Ii' 17.

18.

Hain Condenser Systems & Controls (llC)

Condensate and FW Systems & Controls (1111)

Gaseous Rod Waste Management Systems (HD) ll

!II l 2 l

3 3 6 4 l

2 s 30 9

l I

I Operator Exams ll 1

I Toi:slilA 130

I i I

• • This total exceeds total number of shutdowns in Part I of this table by l because 2 systems were assigned to 4-30-79 shutdow~.

I j'

i

'l

,\

\

i I

J

'l

._

~***~-~ : .";

.'. - /:._ ___. __:;_..

....:_ ~--***~* -****--*-***- -

_:. 5-15 leaking CRDM seals totaling over 1000 h of outage time. Starting in April, 1973,

• a single 1000 h outage occurred when it was necessary to rework the*

charging-system_ discharge piping to correct the continued problem with vibra-tiqn-induced piping cracks. Also:.in April, the first problems with condenser tube l_eaks appeared.

In mid-January 1973, the plant was forced off-line to repair steam gen-erator .tube leaks. Restarted on March 6, the plant received AEC authoriza~

tion for full power operation on March 23. Moisture separator/reheater unit problems caused a 15-MWe derating. On May 24, an outage-was scheduled to modify the moisture separator/reheaters. The unit was returned to power on June 1 and operated at essentially 100% of rated power until Augus~ 11, when the plant was shut down for steam generator tube repairs. Inspection revealed that core internal vibration had occurred during operation; modifications to the reactor internals were deemed necessary, and the unit -was off-line through the end of the year. The plant was not available for 4907 h for 1973.*

The_plant came on line on October 1, 1974, for the first time since August 11, 1973. The period was devoted to repairing the primary-to-secondary leaks on the steam generator. The plant operated only for the month of October; in November, it was again shut down for repair of condenser tube leakage. Some -

5970-h of the year were devoted to repairs on the steam generator, 1451 h for repair of leaking *condenser tubes, 600 h for repair of the turbine that was damaged while rolling it, and 236 h were required for repair of.. several different

---~-~- --* ----*-- - -*-- --~-~-- ----------------**-*-<

__ ....... -~.-~--.--,..,-~-.----

5-16.

items ".""" replacing CRDM seals and* plugging more condenser tube leakso plant was shut down for a total of 8275 h for the ye~ro The **

The continuation of the November 1974 outage for condenser tub'e repairs extended ~2205 h into 19750 This outage included an extension for repair to the steam generatoro CRDM seal problems and shorts in clutch coils (see Secto 5.4.14) resulted in significant outages.

The unit started up .on* Ma.y 16 .P 1976 >> after a,n extended refueling and maintenance outage (scheduled) that began December 20, 1975. During the outage.I> the first~co:re fuel was completely discharged because of* unaccep-table axial growth of poison rods. The outage included steam generator eddy-current testing with subs~quent tube plugging, replacement of tube bundles in the low-pressure feedwater heaters.I> and other maintenance.

During the summer months, power output was restricted due to a turbine back pressure limitation of *3 in.

continued to be problems.

Rg; Condenser tube leakage and Three instances of generator voltage regulator CRD~ seal leakage -*

failure occurred (see Sect. 5.4.1.2).

Operation throughout 1977 was routine and nominally at 100% of rated power. The operation was uninterrupted for three months. On November l.P authorization was received to increase thermal power from 2200 MWT to 2530 MWT; the DER was increased from 668 MWe to 805 MWe. The plant had a reactor availability of 92.8% for the year - the best performance from an operational standpoint ever experienced by Palisades. However>> during 1977 there were three loss-of-offsite-power events (Sect. 5.5.2.1).

• • -**--*--*--** *-*--*- - *---*----*- ---.-**--*-----:--~----------.---------:..... ____ ----------.-,~--~ ----

. . . . ----'-***~--*--: ~'- ~:..._ --~--'--****-:-

-- r-- -*****------.,'- ---*---- ---*-*****-* -~------ ..........-*- *- *--*- ... ~-------*-'"* --*-**** - *..

5-17 A. refueling outage took place betwee.a January 6 and April 20 in 1978~

• Control rod drive seal leakage and repair accounted for 11 outages. The number of f creed shutdowns and hours associat.ed with these shutdowns increased significantly - 27 forced shutdowns totaling 1914 h - after declining for two successive years (Table 5.3a).

For 1979, the forced outages were reducedwith the longest one ("'500 h) initiated by the second loss-of=load condition in a S=d period due to* voltage regulator malfunction on April 30. This outage was extended to resolve inade-quate piping restraints for two safety injection lines.

Table 5.3a reflects the larg~ number of hours lost due to the variety of equipment related probleln.s - CRDM's seal leaks, steam generator tube repairs, condenser tube failu~es, and feedwater system problems. Some 102 out of 129 forced shutdowns were classified as equipment problems within the Gray Book.

Automatic scrams totaled 60 - a 5:1 ratio to manual scrams and roughly equiva-lent to the number of manual shutdowns (57).

• The following problem areas relative to forced shutdowns experienced by Palisades represent trends in terms of their recurrence over some portion of Palisades' history and/or have potential safety implications. These areas are in addition to the .DBE shutdowns discussed earlier and those significant *-

shutdowns that were also reportable events and hence are discussed later.

These problem areas will be reported: (1) steam generator tube failures, (2) vibration-induced cracks on charging systems, and (3) cont~ol rod seal I'.

~

leakage * .

• . ">=':*

--* --------- - ---*-----...-- - ~ ----. --- .... ~-- -----~--*----..,.....~----- -*--~.-~--..-.---*---* '

'\

5-18.

5.4.2.1 I q, Steam generator tube failures. Primaiy-to~secondary leakage through faulty steam generator tubes poses frequent operating pro=

coolant **

blems fD: PWB.s. Although ~he effects of these steam generator tube. failures j

• generally are adverse only to plant availability, these failures may present

• unnecessary radiation hazards to maintenance personnel when extensive repairs o~ modifications are needed.

Palisades was shut down on January 16, 1973 (a little over a year after it initially began generating electricity)$ because of excessive leakage in one-of the two steam. generators~ steam generator vuAn. This outage lasted for 1161 h. Pressure tests and eddy-current tests indicated that three tubes had £ailed in the U-bend area in the sixth and seventh rows from the channel partition and that all tubes in the first eleven rows had varying degrees of thinning of the wall thickness; a minor eddy-current indication of thinning was found in row 12. The first eleven rows of the other steam g*enerator also showed the same tube wastage, but to a much lesser degree.

• All the failures and indications occurred in the 180° bend area that is enclosed by divider strips on either side. In the immediate vicinity of the firs-t eleven rows of tubes, the three divider strips fem a Y" shapes result~

ing in a tight configuration. It was hypothesized that this tight compart- -... -.

mentation leads to steam blanketing~ which concentrates the impurities in the steam generator water.

One of the leaking tubes and one that had thinned (but had intact wa1ls) from the first steam generator were examined by special radiographic techniques *

---

***~*--*****-~---~--~----* ----~*------****- **-*------ ......,..-:---,--......-------*-***-.--~. ~~"'"~- ... -. --~

.. *-r~"---**-***.;- **1--*-~ .. --

. _ --*-* ... :---~----~...~:. :: : *,~*- .. ,.

5-19

'j***

Results of this examination sho~ed a relatively broad area of* attack of vary-ing depth and irregular shape.

I ' p

• The wall thinning was caus_ed by accelerated localized corrosion of the Inconel 600 and was characterized by general tube wastag~ .in a specific area.*

• -*-*~ -: - .. , . .-* .. ********.-:* .--

No intergranular corrosion was evident. The most likely corrosion mechanism was postulated to be a concentration of free caustic 'at a pseudostagnant steam-water interf aceo . The alternate* wetting and drying caused by the tight con~

figuration resulted in.tube degradation in the localized region.

, About. 600 tubes were plugged in the first eleven rows of each steam generator, and the plant was returned on 1ine* in March 1973 with provisions for continuous monitoring of phosphate concentrations and pH. I On August ll, 1973, the reactor was shui: down for repair of steam genera-.

tor tube leaks. This outage continued through the r"emainder of 1973 (3407 h) and on into 1974 (5970 h) when the reactor was returned t~ operation on October 1. An outage that began in November 1974 (for another reasonl was extended for repairs to the steam generator *. The outage from December 20, 1975, through May 6, 1976, included.steam generator eddy-current t~sting

• followed by the plugging of several tubes.

The utility has stated that as all PWRs are required to do, eddy-current testing of the steam generator tubes are performed every refueling outage.

5.4.2.2 Vibration-induced cracking - charging system. As reported to AEC on March 6, 1972, a leak was found in a socket weld just ~pstream of a charging-pump shut-off valve. After repairs were made, a leal{ in the cap on the check valve in the discharge line was found. The vibrations induced by the positive displacement pumps, and the current piping arrangement caused

... the welds to crack*

• • • • -**--*- *---------**--*--*-* . -*-*-*-**----------***--*. *- ---*---------* --------* ---------~--~-----**--* ., *., **.: :***: *: . ~>;.. ,>

.. ~*.:..-.*--- _,, ~ "'""..-...... ..

5-20 On April 23~ the plant was shut down for a three week scheduled outage to perform a major rework of charging-system discharge piping~* To correct I this continued problem with vibration-induced piping cracksa the discharge piping was physically revise4 to reduce bends, wide swe'ep elbows were installed~

and pulsation-damping in~line accumulators were added.

5.4.2.3 CRDM seal leakage. Palisades has been plagued with CRDM leakage problems throughout its entire history. Although most of the problems were experienced from startup through 1976, leakage of the seals still occurred in 1977 and 1978 (Appendix B Part 1).

Palisades' CRDM seals are composed of three basic parts: collar~ rotating

• seal half~ and stationary seal half. The collar rests on a shoulder of the CRDM shaft and is keyed to the shaft. Additionally, it is held by three 1/4=

in. allen head setscrews that also transmit the motion of the collar to the rotating seal half.

collar.

The rotating half has a skirt that extends over the The protrusion of the collar setscrew is accommodated by three slot~

in the sides of the skirt. A carbon seal surface is mounted at the top.of the rotating seal half that is sealed to the shaft by means of an 0-ring.

The stationary seal half is mounted in the housing above the rotating half s and an 0-ring provides a seal between the stationary half and the seal housing.

The seal halves are held in contact by springs in the rotating seal half that act against the collar. The allowable leakage rate is not to exceed 10 gal/min.

--~*

__*...:...:.:.. - , -  :----~--.:...........:.. ..... *- -***** .: *__ :_* _ **** :..:..:......:.*-**- -- v .*

~-* - .' ... ---- ___._ ____ :.,.. ~. ~-*-- ..... ~---**** ---

5-21 When a leaking CRDM was detected, the general procedure was to exercise the rod in an attempt to reseat ~_he seal. If this proved unsuccessful, then that plant was shutdown and repairs made.

In the latter half of 1972, the original seal design was modified and a new type installed in all CRDMs due to the abnormal leakage problems. The original design for.the rotating pressure seal utilized a nickel binder that was found to be inadequate. The units replacing the original ones were of the same physical design but incorporate a seal face with a cobalt binder.

Thus, the design concept was not altered by this change. Installation and reassembly of CRDMs were thoroughly tested by multiple rod drops and torque testing prior to further plant operation.

Nevertheless, problems continued after the modified seal was installed.

In late 1974 and early 1975~*all CRDM pressure seals were being replaced.w~th seals of, a chroJ!le oxide design. Most repairs of leaking seals that were recorded in operating reports for recent years simply noted "replaced seal housing."

Once a seal fails, primary coolant escapes along the motor shaft to the motor package as well as the normal leak-off path (the drain line from the seal leak-off cavity). If leaking is excessive via motor inspection covers, potential exists for wetting the reactor vessel head insulation and other CRDM motor packages, junction boxes, and electrical connectors. Such problems were encountered with CRDM No. 19 on October 23, 1972. Additionally, it was I*~

~-

5-22 p~stulated that moisture due to this event may have led to the failure of both channels of startup range instrumentation.

5.4.3 Power reductions In Table S.3b~ the 22 forced power reductions are summa.riz.ed in a format similar to that used for the shutdowns in te:rms of causes and systems involved.

Power reductions parallel the forced shutdowns in that most of the causes (19 of 22) are attributed to various types of equipment failure~ and the predominant systems involved include the main condenser system and feedwater system. This is indicative of the problems experienced with condenser tube failures and f eedwater pump trips/low-steam generator level occurrences that accounted for a significant number of the forced shutdowns.

5.4.4 'Non~DBE shutdowns Tabla-5.4 summarizes the NSIC c~egories assigned to non-:ORE shutdowns.

Only the major NSIC categories (and not its subdivisions as presented in Table 3.2) are listed in Table 5.4. By using Table 3.2 in conjunction with the data compiled in Appendix B Part 1 on forced shutdowns~ one can examine the subdivisions of the NSIC primary categories.

As in the cases with forced shutdowns and power reductionss these non-DBE categories illus.trate a variety of equipment failures as being responsible for forced shutdowns. In developing the NSIC non~DBE categories, an attempt was made to separate instrumentation and control problems from the general class of equipment failures. Note that Palisades did not experience ~y

.-- ---* -* - -~ .,- ------- -----__,.....,._------***-,...~--~-~ ~----*-----.* - *o:-*-----*--*-----------......-... -.-.--*---:~.~*-~-~---*--*-*.-----*~*-*:---,..,-. .~-.~v:-:r.

~-

Table 5,Jb Power Reduction Summary for Palisades I:

i i>

j ~.

1971 1972 1973' 1974 1975 !976 1977 1978 1979 Total i I. Power Reductions l i**

r..

1. Total Number s 6 0 0 2 1 0 2 22 l

2. Cause '

A, Equipment Failure s 4 2 6 2 19 '

B. Maintenance or Testing 1 1 '

i*.

D. Regulatory Restriction i:

E. Operator Training/License Exam i

r. Ad111inietrative 1 1 '~**

G. Operational Error H. Other

- l "

1

3. Syste111 Involved '**

1. Reactivity Control System (RB) l 2 3

2. Coolant Recirculation and Control (CB) 1 1 *1:

3. Ultimate lleat Sink Facilities (WE) l l 2 t.* _.'

4. Turbine Generators & Controls (HA) 1 ... 1 'r.

I.~ .

s. Hain Condenser System & Controls (llC) 2 4 1 7 i

6. Circulating Water System & Controls (HG) 2 l 3

7. Condensate and FW Systems & Controls (1111) 1 2 2 s 1

\

IJ

• 1 j

.t Va

• 1 I N

J;

• J

• w l

_.l

'l l

.J ,. ,..

Tabia 5.4 NSIC Primary Category Summary for Mon-DBE lihutdowl!Ul foir ll'eli11ac!lse 1971 1972 1973 !974 197.5 Jl976 nn :8.97S 1979 '!'ottal 1.0 Equipment Failurea l s 2 4 4 e 4 u 4 43 2.0 InstrU111entation and Controll Anomalies 1 2 l!. l s 3.0 Non-SRP Reductions in Coolant Inventory (Leaks) 10 s 1 2 s 6 2 31 4.0 Fuel/Cladding Failure 0 s.o Maintenance Error 2 l 3 6.0 Operator Error 2 ll. 1 4 1.0 Procedural/Administrative Error l 1 ll. 3 8.0 Regulatory Restriction 1 l 2 9.0 External Events 1 1 l 3- ll. 7 10.0 Environmental Operating Constraint - -:: 0 Tech Specs

• 1-I

... - ... . :.. _ .. -- -~.. ). _.:. ~. *. *___ . *, _.., __ . - :. .... **--** ----* ,* -- _, . -~ -.. '*-~~----'-* **' ---~.'.- *-----** :._.......;.._, -*-~

5-25

e serious (5) I&C problems that resulted in, shutdowns.. '!'.he NSIC category 3, non-DBE reductions in coolant inventory (primary and secondary systems),

represents the second'largest number of non-DBE shutdown events. These reflect the large number of shutdowns for repairing steam generator tube leaks and condenser tube leaks.

The NSIC hµman-factor-related c{ltegories (5 - maintenance error, 6 -

I

. i operator error>> ~~ 7 - pro~edural/administrative) were associated with only 10.of the 98 ev~ts for which NSIC categories were assigned. Six of these 10 were reported during 1972 and 1973. This would apparentiy indicate an adjust-ment period on the part of operators, staff, and personnel in general. They might be more error-prone initially, but after having acquired experience with the plant, these errors would be significantly reduced or eliminated.

. . ~*

This seems to be the case at Palisades; *however, this is not borne out in___ the review on reportable events where a*significant number of operator-error, maintenance-error, and procedural-error events are discussed in Sect *. 5.5

• and following.

The seven external events include four lightning strikes that resulted in instances of line faults and loss of offsite power.

"j i

..........--------~--- - - --------------~-- --------*-----------------~-----~--,,...._--,. ... -; -----*---:-.--::----~----. ~-.--""'"~-----.

5-26 5.5 Review of Reportable Events This study reviewed 341 events reported by Cm:1.sumers Power Company

• concerning abnormal occurrences at Palisades. The coding sheets describing each eve~t are in Appendix B Part 2. The review of this data was two-phased.

The first place produced signifieant events which~ in themselves>> represented a degradation of plant safety. These events are discl.issed in Sect *. 5.5.l.

The second phase resulted in occurrences which by them.selves did not demon~

strate a degradation in a plant safety function>> but because of their recur-rence represented trends with safety implications. The second phase occur-rences are in Sect. 5.5.2

• 5.5.l Significant events Reportable events have been' classified as significant if they meet either of two criteria:-

~.

1. events in which the failure or failures initiated a DBE>>

2. events in which the failure or failures incapacitated a safety function designed to mitigate the propagation of a DBE.

Table 5.5 summarizes the categories of significance used to indicate such events and the number of events classified in each category. 'nle safety functions lost or DBEs initiated by the indicated events are:

l. loss of onsite power sources in coincidence with the loss of of fsite sources,

2. loss of RPSI capability>>

3. loss of containment integrity,

--**-**-* -'* ~ ... .. ,., . '*

!J i' I

Sununary of Palisades Sign:i.f II Table 5.5 and Conditionally Significant Events l I.

I l

l Significance Category 1970 1971 1972 1973 1974 1975 1976 1977 1978' 1979 Total f,.

I \

Sl Two or more failures in redundant systems 1 1 1 -1 4 S2 Two or more failures by conunon cause 1 1 2 j SJ Three or more failures - 0 I

S4 Failure would escape detection 1 -1 i

\ SS Event proceeds differently than expected - 0 I:'

S6 Event not enveloped by plant design - 1 1 2 l

~* .

I I

i S7 Greater threat with one more failure 1 1 -2  !

-1 SB Fundamental misunderstanding of plant .... o* ' '.

i* .*

i I S9 Other 1 1 1 J i Total Significant Events 0 4 1 0 0 l 1 2 2 J 14 l1 Cl Single failure in a non-redundant system 0 II C2 'l'Wo failures ~ 0 I

CJ Off site release or personnel exposure 1 5 2 5 1 2 2 2 2_0 r j 13.

I C4 Design or manufacturing deficiency 2 2 6 8 8 2 '. 4 45 t t.:;-

I! cs Long outage or major equipment damage 1 - . _l I

C6 ESF actuation 0 )

I!

I.

C7 Significant recurrence rate 6 12 J 1 10 12 12 4 7 67 I ca Other. 1 2 1 2 l' J 10 I

li Total Conditionally Significant Events 0 8 16 11 9 24 28 22 9: . 16 -14J ln

, N I

-..J

, I' ,*.

5-28 4.

5.

loss of component cooling capability, and depressurizat:ion of the Reactor Coolant System (RCS).

S.S.1.1.* Events involving the loss of onsite power sources in coincidence with the loss of offsite power sources. The loss of onsite and offsite power sources is the initiating event of a DBE (loss of external electric load) and compromises the ability to remove the residual and decay

• heat from the reactor core. Three instances of total or partial losses of both power sources for short periods occurred at Palisades.

The first occurrence was during a cold shutdown on September 2, 1971.

The Palisades-Argenta No. 2 line fa.ulted,.out the air-blast circuit breaker I 27R8 failed to isolate the faulted line. A breaker* failure relay then tripped redundant breakers that isolated Palisades from all off site power sources. The power interruption la-sted 56 min with no partial recovery.

When the off site sources failed 9 the diesel generators started and came up to speed. Generator 1-1 picked up its load properly; however, the output breaker for generator 1-2 failed to close. An operator manipulated the unit synchroscrope in preparation for manually closing the breaker>> and the breaker closed. During the time from the loss of off site power until the operator was able to manually close the diesel generator output breaker 5 there was no power to one-half of the engineered safety-related equipment.

During a. DBE (loss of external load) on March 12 i 19 72, one of the two buses powering engineered safeguards equipment was unavailable until opera.tor action was able to restore power. Both generator output breakers

-*~-...

,--*-'-*--*-*****--..****-'--~*-- -**..:.*.- .. ....

• ---*----*~----- -*~- --*- '-* . ~-------**-:..-*----**-- ... -*** *----...:.---:--**:.-

5-29 opened due to a pilot wire trip.

• The reactor tripped from 30% power~ AS desighed, plant power should transfer from the unit generator to offsite power sources through the startup t~ansformers. However, due to.a failure I

to directly trip the plant protection relays' no such transfer occurred.'

. The backup power source, the diesel generators, started an.d came up to speed.

For some unspecified reason, one of the diesel generators failed to pick up its load~ leaving the bus dead. Power was restored by manual switching of the loads to the startup transformers.

The final power loss incident occurred on June 30, 1975 *. When operating, plant, load~ are powered by the station generator. On reactor scram or ge~~

erator trip., the loads transfer automatically to offsite sources. :.However, follow~g a trip on March 30, the breaker to bus ID failed to close. This

• • failure caused a loss of offsite power on bus ID.

HPSI pumps, l out of 2 LPSI pumps, 2 out of 3 service.water pumps, and other safety-related equipment. As designed, diesel generator 1-2 should Bus ID powers 2 out o.f... 3

~ve started and picked up the loads On. the failure to transfer the load offsite.

The generator started,* but the generator breaker interlock contact was out of adjustment* and the breaker did not close. This condition of interrupted power lasted until the operator was able to tilanually close the breaker to the of fsite power source.

5.5.l.2- Events involving the loss of HPSI capability. HPSI system limits the propagation of effects from a small leak or break ~ the primary coolant system that results in a rapid depressurization. The irPsI system's operation is essential to maintain the reacfaor coolant level when the Main

--*--*-- ..... *;.. ***--....* - - - - . .- . -.-----*.. --- ..- - - - - - - * - __ .... - - *~-- .... , .. , .. ~-*- - .- - .~ ...-,- .. ----.--.-.--c.. - - - - - -.. -~.. *--- -~~~~

5-30 Cooling System pressure is* too high for operation of the I.PSI pumps.

On December 14, 19789 Combustion Engineering (CE) notified Palisades that during a sma.ll break LOCA 9 insufficient HPSI flow would *be delivered to the core in the event of loss of oifsite power and one diesel genera.tor.

'nlis classifies the events described in Secto 5.5.1.1 as a loss of HPSI capability. HPSI is desi~ed to be able to supply borated water through

~5!~gb.:t_ iii:je~tion_p_atl_t.$ __ (_Fig~a_l) *. _-iloweverp the redundant foor paths could not be used since valves 3018 and 3036 are normally locked closed because of HPSI pump net positive suc.tion head (NPSB:) cons~derations. During the loss of offsite power and a diesel generator 9 either valves 3007 and 3009 or valves 30ll and 3013 would be inoperable, leaving only two injection paths.

Prior to this analysis 9 the analysis for CE 2650-MWT series plants was .con=

sidered applicable to Palisades.

5.5.1.3 Events involving the loss of containment integrity. Containment is the structure that completely encloses the primary system to minimize the release of radioactivity in the event of a serious accident. A study revealed that containment isolation would be sacrificed on the single failure of the valve air supply; an incident in 1978 confirmed that this was possible.

Additionally, in 1977 there was a degradation of containment integrity when two isolation valves were left open.

On September 16, 1977, a study concluded that the loss of air supply to the containment purge isolation valves (CV-1803, 1805, 1806, 1807~, 1813, and 1814) would cause loss of containment integrity. The loss would result from deflation of the valve seal rings. This failure mode was not considered

- .*-**-----*-- - --*-*-- --....--------*-* . --~** .. -*------.-------~-*-*--****.**~- ..... ------.

li' I

I 1

11.EAC'TOR.

COOLAllT~

.-----~~_o_o7__ r--.*_

) Loo? .. , Jot.is l I A II

\

i I* jI I

I i

I I

\:

i I I il l*

1*

Ii I

i . i i

l:

l l:

i

• • .~**"'

r I . '*

I 1 I j Figure 5.1 Simplified Palisades HPSI Injection Paths 1

j i

/

l i

.J ,. *: !

5-32 in the FSAR.

A redundant air supply was to be installed as soon as practicalo However, on April'21>> 1978 9 maintenance errors caused this loss of capability to occur. A valve supplying air to these valves was opened to supply breath-hlg air to workmen performing CRDM maintenance. The result was deflation of valve seals and breach of containment integrityo During a test for leaks in containment penetrations on September 14 9 1979~

it was discovered that *the two 3-in. manual isolation valves in the contain=

ment exhaust were left open due to inadequacies iD procedures. The valves vent contaimnent to atmosphere through high efficiency and charcoal filters.

Following an accidents these valves would control long range hydrogen buildup by venting containment after allowing sufficient time for containment pressure and activity decay. The valves were not used during the severe conditions initially following an accident. Since this occurrence, Palisades has

~**

installed recombiners for hydrogen control.

5.5.1.4 Events involving the loss of component cooling capability The component cooling water (CCW) system cools equipment carrying radioactive and potentially radioactive fluids~ In the event of a DBE>> part of the system is isolated. Howeverp two of three pumps and two heat exchangers are necessary to co~l the shutdown heat exchangers, the charging pumps>> the engineered safety feature pumps 9 the waste gas compressor aftercoolers, and the. seal water cooler for the vacuUDl degasifier pump. Only on one occasion was the component cooling water system degrad~d below the level necessary for a DBE~

On Au~st 19, 1976 9 CCW pump P52-C failed off while CCW pump P52-A was down for preventive maintenance. The reactor was operating at 90% power.

. *.... ,_,., - _,:.~:....:*i*.*-*~--:-:.

' "I - *~ --*** _,:_,_:i ,._ ** ---~*~---*~.'.-.I******:.:..-*-*--'- ..* :*** ,.. ---- -*:-:..**w-'*'-'*'" ******' *-** --~**---***---*; ** ~ * ** .:.:.~ ***::.:..:...._., :_~*..__ * -**- * . ..: --~--..-*-***-** *.. ;_( *'

5-33 The cause of the pump fail.ure was failure of *the inboard bearing. Either fail:ure to perfoi:m.a hot-end alignment. to the !>UDlP after repairs or bearing slippage contributed to the bearing failure. Pump P52-A was returned to service later the same day. ,The bearing was replaced on pump P52-C.

5.5.1.5 Events involving the depressurization of the RCS. Depressuriza-tion of the RCS is an initiating event for the DBE - decrease in reactor coolant inventory~ The !>articular occurrence at Palisades involved the inadvertent opening* of the pressurize~ .electl;'omagnetic relief*valve.

In preparing ~or system modifications to the re'actor protection system (RPS) in September 1971, a technician deenergized the RPS breakers. This act also deenergized the solenoids on the pressurizer electremagnet~c relief.

valve and caused it to open. Nonstandard designation of breaker contacts in the control circuit drawing led the t'echnician to believe that the valve*

would remain closed. The primary syst~ pressure *dropped ~o "11280 psia

• over a period of 2 to 3 min before an operator closed the motor-operated

• block valve. The drawings were corrected to show as-built conditions with standard notation.

5.5.2 Trends and safety implications of reportable events in addition to those categorized as significant.  :.

Using the systems involved in reportable events listed in Table 5.6,

&pecific trends and problem areas were identified for safety-related functions: (1) loss of offsite.power, (2) control rod drive anomalies, (3) reactor internals movement, (4) containment purge isolatimtvalves, and (5) procedural and human errors.

5-.5.2.1 Loss of offsite power. Pal.isades has experienced 98 partial

---

*-<**-***-.-~------*---- -:-- *---- ------------*-**-**- *-------.-~----~---..---...-.,--* --*.-------..---------.-~-- --~-- . ~-~:-*~--.-~...~--;:~- ::','

. *_.*~ ..'._ ...... _:~~--.---*--** *. .

..-*::.::.____ ' .--- -.-----*---*~-: ... J:~ .. ._. __ .....:.... _.:.:__ ......:.:.~:....--,*~--:-*. *.* *~: ,_._ ...-_,

• s-Js

-losses of offsite power and 4 total lo~ses.

• Partial losses include failures of up to*five of the six incoming lines as*well as failures of components (such as breakers) that cause an interruption of power to critical buses. Report-able events, by definition, were only those event~ that iiivolved component failures *. Therefore, Table 5.6 summarizes only partial power interruptions caused by component failures. Other partial power interruptions were reported only as they caused unit shutdowns. Fig. 5.2 shows a histogram of the total number_of.partial power interruptions experienced at Palisades.

The four total losses of .offsite power occurred on September 2, 1971, September 24, 1977, November 25, 1977 and December ll, 1977. The first occur-rence. (.1971) is discussed in Sect. 5. 5.1. l. The causes of the three ".power losses in 1977 were never discovered. In each case, the bus common to all incoming lines, theR bus, was isolated and became deenergized. The tripping.

scheme for the R bus was altered after the last occurrence.

• No similar failure has occurred since the trippi:ng scheme alteration. ..

The large number of failures in the emergency diesel electric power system (Table 5.6) cannot be centributed to any one specific component or trend. However, they do indicate a high unreliability of the backup emer-gency electric power system.

5.5.2.2 Control rod drive anomalies. Numerous problems with CRDMs have resulted in functional failures of individual control rods. These failures are summarized in Table 5.7. There have been 24 control function.failures 1._

and 8 shutdown or trip function failures. In addition, four other reported failures were of the type causing internal damage to the CRDMs. Three function

1*

20 19 18 17 16 (IJ

=

0 15 4J

g. 14

.... 13 1-1 Cl) 4-1 r::1 12 H

1-1 11

~

0 10

~

r-i 9

.~

4J 8 1-1 Ill

~ 7

- 0 IM 0 6

....Cl) 5 1

z; 4

3 2

1 1:

1971 1972 1973 1974 1975 1976 1977 1978 1979 Year Figure 5.2 History of Partial Qffsite Power Interruptions at Palisades

' 9 I'

i*:

i

• I

,. I . :

failures and one internal damage failure resulted from: common cause failures or potential common cause failures*.

The first of the potentia1. common cause failures occurred September 15>> 1971.

During lo~power physics testing 11 two control rods failed to drop on individual rod drop tests because of epoxy and other foreign matter on the clutch face.

During installation of new springs* and.pins in the magnetic clutches 11 epoxy was used to hold the spr:Uigs in place and some of it flowed onto the clutch face. Subs~quent to these failures 11 all the epoxy was removed.

Damage to several drive units occurred on November 5, 1971 during drop tests a.t atmospheric pressure. The damage resulted from the lack of water in the drive unit. Water in the units absorbs the impact energy. However, on November*s the rods were dropped with dry units ~ecause drive housing did not occur at atmospheric pressure *. Procedures were revised to provide proper venting prior to test drops.

proper venting in the On February 3 9 1972, a control rod failed to drive following a feedwater -

transient trip. Investigation revealed that boron crystals had formed on the drive mechanism bearings. Borated water had backed up the CRDM seal drain line from the quench tank due to above normal pressure in the tanko The tank was pressurized to about 25 psig as the* result of leaking pressurizer relief valves. Several other CRIM bearings also bad boron deposits. AJ.l CBDMs were cleaned and the CRDM seal drains were rerouted to a floor drain.

Four control rods £ailed to drop in the f ina.l noted comm.on cause f ailurec During tests on August 29 and. 30, 1971, four control rods failed to d~op because of frozen clutches. An investigation showed that th~ clutches froze

~-....--~* ...

.. ---- ..--- --*------*-~--

_: -~-*-* ,,____ .,;._:~~-- *_ ..... ::*. .:. _, ____ ... ,.,_._, -- . -*--** --*-"*--~----

.. , _._._ .. ,.. _. _*;-*--- ----~* .

5-39 I

'because of. baked-on lubricant. As a result, lubricant on all CRDMs was removed and , a new coat was applied."

• 5.5.2.3 Reactor internals movement. As the result of increased noise

• ~

levels on the ex-core neutron flux detectors, Palisades removed the reactor vessel head OU October 6, 1973 to inspect the reactor internals. Arullysis showed that the core support* barrel had been relocated with respect to the aligtiment keys. The movement of 'the reactor vessel internals had increased the amount of water between the core and the flux detectors, resulting in the apparently anomalous flux indicators.

Months of investigation revealed.that the movement of the internals resulted from a relaxation of the preloading of the vessel head studs and the mating surf ace between the reactor vessel and the vessel head.

• Because of the decreased. clamping force, the core barrel was able to move under the influence of noxmal primary* coolant flow. This movement caused observable wear and broken fasteners. As a solution, the expansion/compensating ring was replaced with a thicker ring, and a German-designed spring hold-down system was installed with a minimum hold-down force of 733,000 lbs during full power operation (the original hold-down force was 3400 lbs). Also 5 all worn surfaces were smoothed, new alignment keys installed, and all **

snubber shims replaced.

5.5.2.4 Containment purge isolation valves. During the period from 1975 through 1979, the contaimnent purge isolation valves have fa;i.led leak tests or failed open approximately twice per year. There are three

---

*------* ~-. -. ~-~----.* --------~-~---*-------~-------*--------

., ' *-~ :_______ ."_..__ ... ~.:.. *-* ...:..:_..::,_.__ _:*_*__ , - .. *- **--**-* -'~** .. - ....... ~ .. :;_*_ ...::.~-'.~.c::_..::.::.,~*...:..:,.:_.._ ____-_:.;:.. *.... .* :.... _ . ..., .... ~-**' -~-- - *.... ..:.~. :.:: ~

I ~, 'i

% of Total Reportable Events Attributable to Procedure or Human Error (IQ c:

ID 0

N 0

w 0 ""'0 Q\

0 co 0

.w""' "'......0 1 out of 2 - 50%

~

.... 5 out of 8 - 28%

l"i ....,

\Cl 0 Id n

m Q. .....

.,c:ID "'....,

N 8 out of 28 - 29%

m

I Q.

=

c:

w 9 out of 19 - 47%

a Ill .....

I t"J >< "'......

~

19 out of 35 - 54% I 0 .,

11) m

"'1'

\C 26 out- of 40 - 65%

Vi

~

~ ..... *-

Ill "O 23 out of 37 - 62%

0 l"I' 0\

m ....

• er

....ID ...,

\Q' 23 out of 53 - 44%

t"J Ill

I l"I' Ul

"'......co 23 out of 44 - 52%

m l"I'

--0 "'...... 20 out of 44 - 45%

m

....r.a....

~

0 m l"I' 157 out of 341 - 46%

Q.

Ill Ul

....m 5-41

• ---** -.,.-----*--*----------* ----* ---** ------~. -----* ..

-~ ..............--:.... *:.;. __ .,..:*.:.~- ... -- *~-*.:... ~

5-42 5.6 Events of Environmental Importance 5.6al Radioactivity release* events Twenty events at Palisades have resulted in radi:oact:ivity releases.

Table 5.8 summarizes the amounts released annually. Seventeen of the twenty events resulted from mishandling or failures in the waste disposal system.

The rema~g three events. were releases or overexposur'es during maintenance cm components that normally car:ey radioactive U.quids.

On four occasions ~August 10 5 1973i August 11 5 1973i July 22, 1975 9 and August 28>> 1975.- release rates were exceeded during discharge of radio-active wastes to the environment *. Two overexposures to maintenance personnel occurred during the 1973 maintenance on the steam generators. A Consumers Power employee and a Combustion Engineering employee received quarterly doses of 3.40 and 3.19 rems, respectively.

5.6.2 Nonradiological events The sole reported nonradiological environmental event resulted from the failure of the design for the cooling towers to properly account for transient conditions 5 such as startup and shutdown. Cooling towers were installed in 1975. For an average of eight reported events during each of 1975, 1976, and 1977, Palisades violated the maximum 5°F differential between the ambient temperature of the receiving water and the temperature of the water discharged back to Lake Michigano All of these violations were during transient startup or shutdown conditionso In.19.77~ Palisades amended their Technical Specifi-cations so that the limit could be exceeded during the transient states.

Thus, no events wer'e reported during 1978 and 1979*. """'"

-*------- --*------.------.. ---~. ---*-:..* - .

- ----*--- - ...-----...--.,.--- *~---**---*---~-----~-~~~

i r**_.

j:

i i'i .*

j ,*

• I (*

r \

Table 5.8 SUMHARY OF RADIOACTIVITY RELEASED FROH"PALISADES i*

• i '.

J

~'. *. '*

j EFFLUENT (CURIES) !ill. !ill. ll.ll 1974 !ill 1976 1977 1978 1979 1 AIRBORNE I Total Noble Ga1ea 3.84E-06 5.05E+o2 4.54E+o2 3.36£-02 2.61E+o3. 2.99E+ol 5.99E+ol l.21B+o2 7.62E+o2

) Total I-131 8.70E-03 2.85£-01 9.94E-03 3,75E-0! 2.92E-02 l.52~-02 l.51E-02 9.57£-03

• • j Total llalogen1 (Including I-131) i.40E-02 2.89E-Ol 2,30E-02 4,29E-Ol 3.06E-02 2.35£-02 2.llE-02 l.95E-02 Total Particulates (T1/2>8 day) l.OOE-03 l.15E-Ol 3.48E-03 6.lBE-04 l.43E.;,02 1.lOE-03 5,51£-03 l.19E-05, Total Tritium 2.15E+o2 l.BOE-01 2.21E+o0 4.44E+o0 9.66E+ol LIQUID: "'...

I

~

Total Mixed Fission ' 7.48E-05 6.81E+OO 2.78E+ol 5.87E+o0 3.45E+o0 4.JOE-01 9.29E-02 8. llE-02 8.09E-02 Activation Products Total Tritium . l.45E-03 2.08E+o2 1.85E+02 8.llE+OO 4.16E+Ol 8.95Et00 5.58E+O! l.01E*02 2.40E+o2 Dissolved Noble Gaeea l.84E+o0 2.57E+o0 4.74E-02 5,18E-02 l.94E-03 l.lOE-02 l.49E-Ol

• O.OOE+oO SOLID; Total 3.00E-03 2.SOE+oO 6.29E+02 4.98E+03 2,21E+o2 9.58E+Ol 8.7!E+Ol 3,40E+o3 1. 53E+o2

).*

II I.

.. - ---~ **-*-*-* ... *, :' .

.5-44 5.1 Evaluation of Operating Experience The primary objective of the review of the operating experience at Palisades was to identify the substandard performance of safety systems.

'Ihe two criteria for this evaluation were:

1. events that subjected the plant to DBEs and

.2. events that ca.usea a loss of a safety function designed to mitigate the effects of the DBEs.

Two problems have been identified that have the potential to satisfy both criter:ia. The remaining problems meet only one criterion.

The most serious failures identified are the procedural and human errors.

These failures have the potential of not only starting a DBE (inadvertent tri.pping of a reactor coolant pump on Februa.ry*l, 1979) but also defeating a safety function (leaving containment exhaust valves open on September 14, 1979).

The second most serious events were the electric power interruptions that resulted in a loss of generator load DBE. In itself 9 the loss of offsite power does not cripple the supply of electric power to engineered safety feature.so However, coupled with the large number of failures in the emergency diesel power systemp the loss of of £site power becomes a serious situation with potential for loss of electric power to all safety funct:ionsc The *-

large nU111ber of failures in the diesel generator power source are not attributable to a single failure source; ratherp they simply exemplify the higp. unreliability experienced in diesel generator power sources industrywide.

The only DBE experienced with regular frequency is the loss of* normal feedwate.r flow. Of these occurrences, only one has been reported as the*

'".:.~. 'i*-*~ .....:* *-¥'-***'* __ ...... ~~ **

. . ::_; *~~:.. .. -""*:. _.__._. -**-----*-- .. --*

s.-45 total loss of normal feedwater. In all cases, engineered s~fety features I * ,

performe~ their intended functions to mit.igate the effect *of the fee~water i

loss and bring*the reactor to a safe shutdown.

Other potentially significant system failures include fail:ures in the control rod drive system, the containment isolation system and the HPSI system. The CRDM problems were nonexistent during 1979, indicating that they may have been solvedo Containment isolation difficulties have been solved by both hardware and administrative changes~ The HPSI problems have mainly been those of design and other human errors. Due:* to the importance

'of these systems in mitigating the effects of a DBE, their reliability and performance should be a matter of continuing concern.

j.

.i i

- * - - - ; - - * - * - ****----*--- * - - - .- - - - - -~-!.- - - - .----------.---

.,_,._.._.~..::.-* ...:..._.:~.-*

• • - .~ ... _.. *.' *----**-*-----*-.... - -**** - **-- -

6-1 REFERENCES ,.

1. "Standard Review Plan," Chapter 15, U.S. Nuclear 'Regulatory Commission. *

2. Operating Units Status Report Licensed Operating Reactors, NUREG-0020 May 21, 1974.issue through January 1980 issue, Vol. 4, No. l. *

3. NucZeazt PObJer PZant Operating Experience DuZ"'ing 1973~ OOE-ES-004, U.S.

Atomic Energy Commission, December 1974.

4. Nuclear Power Plant Operating Experience 1974-1975, NUREG-0227, U. s.

N~clear Regulatory Commission, April 1977.

5. NucZeaz- Power. PZan~ Operating Ezperience 1978, NUREG-0366, U.S. Nuclear Regulatory Commission, December 1977.

6.

• NucZeaia Power PZant Operating Ezperi_,ence_ 1977, NUREG-0483,_ U.S. Nuclear Regulatory Commission, February 1979.

7. Nuclear Power PZant Operating Ezperience 1978, NUREG-0618, U.S. Nuclear Regulatory Commission,-December 1979.

- ---- --~---------------=----=--=--=--=-- ;::::::----- ------;- --~----,__,,,_--- ---

B-1 Appendix B: Palisades Part .1. Shutdown and Po~er Reduction Tables i*,

• -- ---------*----*--*----~-:-~ --*------------ ... ........----.-.--,----.,----*--.---.-----:----*-**.**-:* - --- ....

l!'ALl!SAl>ES Teb!e !11.A

!FORCED OUTAGES AHi! roWER IRIEDUCTUONS DBE(D)/

Datu Durat!on !'Olllll!' !hiportab!u Shutdowl!8 Syst111111 Cu111pommt .HSXC(H)

No, Ev1mt DllScr!ptAon Cauau Event (1971) (lln) (%) Hutilodl R11111oived R11voh111d Category

1) 12/31 8 u LTR Repa!r of e high g1reeeure 111ofe~y !Engineered '11.mlves NI.I.~

l/14/72 fojccQ: Aon valve found to havu Safety

• damagelft i11ternalls and correct !on 1Fe111~111l"<Hl b:I of dlscrepnncies noted with tlui (SF) I i Snlttan use nf ~he accm11!11ry N

&ystem.

.1 I

1I I

I l

I~*

~

I*

l'Al. ISAf>E!l 1: Tabla Dl.2

• j; t"ORCEP OUl'AllES Affll l'UHER REPUC'&' IONS

____ ,. ___ f*

DBE(D)/

11 llu111n: t nb I u ffSIC(ff).

Plltu Duration Puwuir ShuLduwn SlfHltUllB Cn111punu11t 1 tin.

( 1912) (Hrs) (2) t:11u11t Puucll'lrtAon Couoli th*lhud Rn11oh11118 lnvolvud Evunt C11t11gory 1/1 246 10 . !.'A'll lllegi111h* uf 111 hlgl! pu1111un. liaruty A 4 !11111!1001!'1111.!I Valves Nl.l.4 1/A4/12 Anjoctjon volhu found tu b11vu d111~- Sni!'illl:y as.,*I lntern11!11 and cu1*roctlm1 of Fu2tur!l!I! bj olbcrupunchu1 11ot.id with s.hu !nltlai UH') I uliu uf tho aecondnq* sy11tu1*. w I) 1/ll 9 20 Tclp fro111 luw aluuia gu11ocatur luv.,J A ) Steaie ~ Fl llun *112. 7 I

due to un11tul>l1! fuc<lwalur cuntrol l'ow.ir uy11tl!ta cau:;ed by a*lua,:i:u*I CCt11du11611lu (Ill!)

p111u11 11t r11 l11ur11.

1

! 2) l/lZ 5 20 Trip from amuirutor girotcct Ive d If- c ) Sa.aie111 !!. Ope1"11tnri.1 N6.!

i fo1*.mt1ul rulay 481Z due I u tocl11~ I- l'uwel!'

c 11111 error. OIA~

I!

i

)) JIU O.l 20 Tr.Ip from low 11tc1111 11u11orator level duu to unstable fuedwatur control aystuaa c1111sed by pluggud COlldl!lllllllU 4 l S!enie &

Powor fl II ors 112.* 7 (1111) i: puwp strulau!a*a. *

4) l/ll 10 20 Trlt* fro11 low 11tc1111 1unuretu( levi!I c ) Stft)l_llM i **auure 112. 7 duu to u1111table fuudwnlur conu*ul r4lHOll' ll)'llll!UI C:llUilcd by t*luggud condo111m lo (1111) puu11* strainers.

S) !Ill 9 20 'l'rlp fro111 guuurator ,..rutuctlvo JU- c l Su111111 6 011er111ur11 115.1 fur1ml Joi rulny 481Z due tu wlrh1g l'owuir foult. (!IA)

I *1

.l fl

ll'ALHSADES Tob!e in.a t'ORCED OUTAGES ANO l'OWER !\EllUCTROt!S Datu Duration Powul!' !h:po1ruble Silutdov111 Syst11!1l Co111pommt DBE(D)/

NSIC(N)

Ho, Dusc1r!pt!on 1!:111111111t (1972) (llrs) oa Evunt He~D1od Rnvollvedl Bmrt!ll.hud Event Cstegoiry Iip '6) 2/l g 20 LTRs 'A'ri11 hom 1111te11m ge11e1rato1r luY levell A J Steam & Va!11e111 D2.7 2/15/72 due to ll!unUunct !on of fceClwnter ll'ower J/6/72 1reg111.mting 11a!ve. Hn11pected C!UlH, um~

bl Repnill'etB co11tsi111111m1t siir coo!er I llllX-2 habe irupltures, llep!llc11d the 34Skv trn1111former bushings. """

7) J/U !9 ]() ft,'l'R Tll" ipg1ed mnnuaHy when automatic tttr !p A. 2 IEAectdc <<:!rcu8t D2.2 l/'l.2/7l [n!!ed on opening of J45kv pBunt ll'o..,1iir CR0S<11r11/

output breakers duu to fnully p!lo~ (EU~ Kntell"l!"UH*teirs 111he c!l!'CIGH, II) 3/15 20! JO Reps h' Ae11k!11g CRIJH eenle. I) Reec~o!" Control fill. [

!.: {RD~ !lolil DirKve

&iedum isms

9) J/211 <24 40-i Power l!'educt!on. Lose of vacuum A 5 Sttesm !. 1'1BirMne RH.B .lo due to lol!ll of Bevel !n !ll litearn  !:'0'1111!11" gene1rotor ~eed p1~np turbine ilroln {!!Ii) tnp.

10) J/21 9 *:-:-:*6a Trip du5 to ~eedwoter pwni> trip A Jl Sll:eum ii> Tr.urbfoe 1!12.7 from 11fbir11t1on in the turbine ll'ower dr!ver. ... (!!II}

11) Jl/]! 67 60 ilep&uce three !euk.\11g CRlltll oeols. Ill llte111ch>ir IC011troi l!U.l (RB} ilolil l!)irhe Hecium Re1us

' 12) 4/5 0,] 60 Trip due to e ~ul!'b!ne g1r!p causud ] Su*n111 !lo IFAH<!!rs D2. 31 I

by a pluggc~ f !!ter An the hydrou- il'<llWCI!'

I 11cal oU syotcm, (Iii\)

j 1

J

I * ....

~ ..

I' A&. l:iAlmS Table Bl.2 '

FOllCED OUl"Al:t:s AHi> rollF.R Rt::IJUC'UONS .

DDE(D)/

Dntu Durntlon Powur K111t0rt.11b I u Siautduwn S!fSt&lll C11a1punu11t HSIC(N)

Hu. 1Ju11c:r t11llun C11uliu (1912) (Hrs) (2:) Evu11t tlulhod l111vuh1ud lnvulvud ~Ydllt Category ll) 4/14 10 68 Rup11lr 111~ In condo1111er tu~e loulrnge. II Steiiiti 8. lle11L .-n.:.i

• Puwur Kxd11u1gen (UC) bj.

I

14) 4/14 10 l'A Tr Jp froia low 11te1111 ge1111rntor level l Ste11111 ll. V11lve11 IJ2. 1_ VI duu to 111111func:tfon of the fuud11111e_r l'owuir by1111uli vulva. (1181)

IS) 4/U l PA 1'r Jp duo to rein~ uper11tl1111 us 11 A ] l!lecu* ,fie t:luctr lcul H9.2

&"Clllllt Of II lightning 11&rJk11 1111 tho l'uwur Conduc~ors l4Skv line. (EA)

16) 4/2l IOOO f!O 8.Tlle Reworked the cluarnlng ll)'Ulc*t die- D l E11gAneuk*ed l'l11ee, Hi, I. I 5/4/12 chorcu piping to currccL Liu* cout 111- Sd!!t!f .-aul111111-_

6/15/72 ued 11roblc111 wllh vll>rnt_l1111-l11duccd **e1Uurc111 1/11/72 gtlplna cucke. tlodlflud CROH 1101118. 4SF) tludlf led the 111lnl-fluw lh1c11 fur hcd puiu1111,

• ltJYl*llt lgatuJ .boll 111l1111lnai from the cu1ul'c11a11L Ina r Ing i;h!na fr111u the r"11ct11r lnto~1111 I e.

11) 6/l 15 *:PA' '!'r A11 frn111 l111t lhNaaol 111arsl11 1*nlloc.:. l" l IUfllll'UIH&llllll- Just r11ue11la- Hl.O I*.

tlv11 clrcul&ry duu lo 11cw uyute111 l Alnn a. l &1111 '

co1111tral11LH. Conuo la. Controlu (IA)

18) 6/22 8l 60 Tr lp tto 1*llduco CKDH :ual le11kugu. A z Reacttor Control Nl. I (Kb)* llod llrlvu H**cl111n 11111111

'\;

j I

,.j I'

I I ll'Ai.lSADES table Bl.2 fORCIED OUTAGES . Mm l'Ollf.R llE!lUCTlONS i.iBJE(D)/

RuporUble NSXC(N)

Do tu Duration Powa!r Sim~down Slfsti!!1111 Co11ponu11t No, Evu11t Duacrlpt!on l!:auau Event (1972) (Hrs) (%) Huft.hod Rn110Avud Kn11olvud ICal!:egory

19) 7/4 65 60 Repa!r of preaauri&ur apr~lf valve 1 Reactor VaB.11e11 1n.1 due l!:o excoasivo leakage on Ines !Coolant of packlns. (CJ) tr.I I

20) 7/6 5 PA Trig) from low steom generator level due to unstable feedwater control A Steam &

ll'owor

'!'urb &11ea D2.7 caused lby low .feedw111*er turbine (Ill!)

speed, 2l) 7/15 S1 60 Repa!r of ateo111 leak 011 rclu!llter o A 1 Stemm S, Va!llll!e 10.2

&Lop valve CV-0548. x*uwer (llU)

22) 7/28 19 60 Repair of CRDH seal Reokoge, R Renctol!' ICoamrol 10.U

~RD) lltocll Dl!'ive Hechrmiaiaa U) 1/'Jl l!'A Trip fro111 Row 111tH111 g1merator J eveR A S~eam a. !n11tn111e11te- D2.7 due to fee~ater pump trip caused ll'ower ~foll a, by a defective 11!bratinn detector. (1111) Contl!'oh

24) 8/6 <24 ".~.i~-50 Power reduction. Dropped CmJH Ho. A Reach11!' Conuoa !Ill. I 9 !n se effort to stop ueal leakage. ... (RO) Rod PrA1111

~lecllumiems

25) 8/U 10 l!i0-4$ Power reduct!on. llgh dbration A Reactor 1i'u111ps H2.4 alarms on B" primary coolant Coo!oBlt pump. (CU)

26) 8/26 .20! 60 Repn!I!' of CRDH Ho, 9 seu! leakage. ll Reactor Conu11.1n 10.D Repair of 11111rBoue slenm lenkn. (RD) Rodi Pri110 H1*cium i 111119 .-

l

• I e . ..

• i

+ ..

'(

l'Al.IS!\llt:S Tabla Ul.2 1-'0ltCEP OUTAt:t:S AHll l'Olll!ll lll!llUC'r&OHS

---

~-

DBE(D)/ i' ll11tu ll11r11 tlo1l l!'owur lh:1mirt nll I u '-- Shutdown illy1o1R111a C11u1pommt HSIC(H) \.

llo. Evoiut  !>ll sc l"l p ti on Cauau (1912) (llru) * (%) Hulhod l!awabud l11volvud Event Category

21) 9/9 <24 60-42 l'ower 1red11ctJon. Tewporary lose of A 5 S1ten111 6 Blm1era 111.1.1, 11 Jra Jn 1111al 011 tho Antercondl!nuer l!'&Dr.llU!'

ataam aJr ejector causing a VIU!Ull .. (llC)

Jro1** ........°'I  :

28) 9/12 268 60 L1'P lnapectJun of cuuoe for high cm)cen- !l l!!hcitl!'ftc 1'ro11a(on1ers tH.1.4 1/29/12 Cration of cowb11ulJblll ga11* Jo n*owell' 1111al11 trone1foniler. Ke l11bod CllllH '"" o:u) '.

c !urclu:u. Srakoid condonsl!r lubllu, I

j HodlfJod 11111111 utu111a relJd v11lvc11.

I

] 29) 9/21 12* 0 Reactor brouaht er ll lcnl and ul111t  ! l 81eaca:or Opernloca HU.l *'.

down twelve t Imus for dl!w1111sr 1*11- (1181>>

t Jon port Ion of At:C 01u:r11t11r 111 l lce11su 111uaui I nut Jun.

JO) 9/24 <12 60-55 Power reduclJ011. J11terco11Jc11sl!r A 5 Stt1ta1e ' *lllowl!r 115. l loop aual on tho steam jot air l'nwcr "

t:Jucto1*11 tco111urar lly lost. (llG) . ..

]I) 10/2l 508 ~o a:rR Repulr of CRPH Hu. 19 11ual 111111 A Reocu1r Control Ill. I 11/2/72 u11te11uJvu water dnu111no to othul" (1111) Rod P1*tva tluch1111 I H-uis L**..

l!RPH pucknsea. ...

12) 12/9 60 60 ltupalr J11011er11Uve CR(Jtl Hu. 26. A Rllac~or Cu11Uul Hi. l._4 II (Kil) Rod Prlvo Hucl11111 is111a

! ))) 12/ll l2 l'A 1'rlp fi:1>111 ~henaul aaare,ln/low 111*eu- G lhmcsor Ol'Cl"lltOl'll 'H6.I I uua*e duo to 111ho111tch of turblnu & <<1110 reactor powur.

'.j 1

]4) 12/21 16 85 Tr i11 Jue lll closure of A" alcaui A ] Steam 8. V11lvu11 ll2. 7 guauirator fecdwatcr control valve *. D'"wno*

<<mu

B'AK.HSAPES Tabl<11 Bl.J 1-'0RCED OUTAGES AMII l'OWER REUUCTXONS DBE(D)/

Dato Duration Powur lll!portable Shu1tdow111 Sye1tu1111 Coai11onl!o1 t NSIC(N)

Ho. Evunt Descr 111Uon Ca11a11 --Event (1973) (llru) (%) Hl!thmB !l'IVOfivll!cli illvoivl!dl Category l) 1/16 1161 7S  !.1'Ra Repo!nd steam generator a1A1a tube B l Stea111 & l!eo_t tn.1 1/29/73 leakage, Powell' Exciu.mgeira 3/6/73 (llB~ bl1 a

OD

2) 3/6 u PA Tr!p due to turb!!'le trip whee tur-. 31 Slt<llBl!I & !net rumentn- D>2. 31 lb~ne load l!.tmiteir 11111HunctAoned. ll'owel!' Ul!llll  !.

0111.) ConUole J) 3/lS <24 1112-7 ?owor reduction, One of two clrcu~ QB Steam g, FU ten a.!9.2 llat!ng waiter p11111ps removccll hom il'oWll!I!'  ;.

ue1r11!ce because of large amounts of (HG) and Beoves were collected on lho travelling screen as a result of l!Sgh winds encl wave acl!:!oo.

4) 3/19 20 82 Trip from apur!oua high pressurizer 31 111sltB'ume11ta- lll!ltlrlllllC!'l~lll- N:l!./a preosure signal, lion & 'Ulll!ll '

ConUoie ICon~ll'oie (KA}

5) 3/25 g ... u AMC operator 2xama

• IE ll. Not 011ermton1 NII.]

l

• i\p11ll !c:a!J!e (ZZ)

• l 6) 3/28 <24 98-76 rower reduct!on, Minor fire in oil- A 5 Steo111 & Turbdnl!!lil NI. i.4 j so1.1ked 1turb1n2 ile1uUlg !nsuiallon. Powell'

'1 (llA) 1 l 7) 3/lO <24 98-66 Pover reductAon, Reheat ere removed A 5 Steam & Valv11u1. ND. i.4

• .:i from aerv!ce ~ue lo C:Olll l"lli VlliVl!O, l'Olll!I"

• 1 0111) j I

l

1 I

ll'AB.KSAi>ES Table Bl.4 roRC!!D OUl'ACES ANO row*:a .REDUC'll' !ONS DBE(D)/ -

!leportable Shut di own $ystl!!1111 IC0111po1u!Blt NS!C(N)

Do tu Durot!on Power No. Evi.mt Descl'!ptlon C1msu Hud1od - Event 0974) (Hrs) (%) K11avo.llvucll Knvn!ved Category NA 1/1 5970 0 Continuation of shutdown began A 4 S!:lllam ii Heat NJ. J 8/11/73 to nipall' prtnmry* to il'OWd!ll' Exll:91n11gera secondary !tube leakage 1n ete1111 (HD) bj generator.

....i 0

l) 9/6 600 0 While rolling turbine, blBde damage Stlllam Iii 1'urblnes Ni.L4 occul'rcd due lo leakage of feed- ll'ower watel' heateira. (Ill\) #'

Z) 10/Z 1l 0 Dul'ing tul'bl11e overepeed test, A Stems II. !le Rays NI .2.4 pilot '!lil:'e relayRng problem required ll'uwel!'

111anual lturbln_e tdp. (!IA)

J) 10/7 5 u P!lot w!re and ant!motor!ng relay A  ! S1tiao111 !!. Rei11ys NI. D.4 ltl'ipped, no deficienclee found. l'oweir (HA)

4) 10/17 216 50 Repiu!ired CRDH 1111111 and repl11ced one A n Steam ii Biell!~ NL 1.4 CRDH clutd1, plugged ~eak Ing -conden-  !'oWUI' Exdurngel!"e ser tubes, andl repaired pipe fJttjng (llC)

Ii 5) U/! US!

J,.,~.*D 60 leok on PCP uual !e11k-0H Hoe.

Condcnilel!" tube !e11knge. A l Steam Iii Powel' lll'll~

1!11ci111nglll!'11 N"J. 2

.I (l!C) l.

.I I

l

.* i

. ~ -

l'ALISADl!S 1'abl~ Bl,. 5 '

f'ORCED OUl'AGl!S AND l'OIJl!ll REllUC'l" IOHS ~.

i*

DllE(D)/

Datu Durnt lun l'u11u1r lh:1mi:t11b!u Shutdown S11stG11111 Co111*onunt HSIC(N)

Ho. Puiicrlptlon C11HHU Evunt (1!115) (llrs) (%) Evunt Hut hod A11vohG1d lnvolvud Catugory 1/1 2205 60 Co11U11uo1t1u11 of ahutdo~n buaan on 1 ..

"" ll/1/1l to repair <<:ondenaer rubo iuukuge, A Sten*

l'owur (llC)

§. Huot 1ixch11118un NJ.2.

b:I I) 4/6 l 'l'u ll'll!lUtr luak h*mtcu* A *. l!L61!1111 &. vu1vu11 Nl.2 I v11lve11, In '*"* l 1 1DWUI!"

(Ill!)

i::

2) 4/22 14 £.II, oil line faUure on tu1*bl11e A ] S~i!!l!W &. tlechun lcn l u.1.1, contl'ul 11yatu11. A'owt*rr l'1111C t I Oii (HA) Uni tu l) 6/22 221 80 Ru1u1h' CIU>tl aeal leak by ro11lucl1111 A l lte11c&:or Control HI. l. I

' seal, aultnr uud 1*hor111ocou11lu on

.i Clllltl tlo

• 10.

(110) llod Drive  !

,, Huclum I 111*11  ;: .

I 4) 6/30 *8 0 75-14 t*uud1111tor 11um1* trl11 011 Juw 1111cl Jon 11ru911uro, Only ono comlum;a t e A 2 Steuue ii l'owu~*

l'llflll'll II:.!. 7 I 1*111*11 wus ru1111l11a "" s.G. levi!I wuo unstable, Con*ectlve action. F.1-1.

conla:ul wll l hu 1'l11cud tu uutn ot .u 1011 1111wer level, (1111) '

i

5) 1.21 39 80 15-5 "6'o a:op11lr cunu:ul l'U&I drive 111otol' A 1 llOIJ!4:loP: Conlrol 04.l I*

' 1S-1 und 11Jjuot bruk.:11. (Rb) !lud Dr Iv.: i 15-8 t111ch1.111lm*11

6) 1 Feedwatur pump trt1* duo 1:0 loss of A" l StOi!!B Iii 1111111110 02.7 euctton preSliUl'P. l'uwol!'

• UHi)

(:

l" I

1) <l 1*ur!i &110 tr 11*1*11d duo to l1111uff Jc.lent l 02.3 londJug (u111tmutor11111 ru I aye),

A SL!!!u..i &

Powe~*

1lol11y11

'i .

t*

l

~

(!IA) f-

8) 11/!2 u 'il'o u1111 h' CllDH oeui le11k-11ff. A l llt:ll!t.':~Ull' Cunrrol HLl.I [;::,. ..

"* i~ .

(!Ill) llud Drive \:

• 1,' Hocl111111a1us

~*.

II

IPAU.XSADES Table Dl.5 1*

fORCED OUTAGES ANil l'OWER llEDUCrRONS I

I DBE(D)/

Datu Duration A'ovur Re1iortab!u Sbutcllown Sys tea -Compommtt NSXC(N)

Ho, DHCl!"iption C&UStl!

0975) (Hrs) (%) Event H111thodl llnvo.i.ved nnvo!ved Event Category

.... *-~---

9) 8/17 135 75-18 To nplll!r CRDH. Rod No, .19 dropped A  ! !Reactor Control 04. l due u1 11horl!: hi clutch coU. ~RB~ Rod ll>l!'hi!!

JO) 8/30 9 15-19 To II'epair CRPH. Clutch fa!led, llleactoll' Hee: hon 11111111 Connon 1>4.l '

=

N dropped l!'Od no *. Bl. <<nn) Rod Drive Hecbnniflme

!l) 9/6 95 60 75-:lO 'B'o ll'epa!r CRUM. Clutch failed, A l!leectoll' Ccmt iru& Dlo.J droppecll ll'Odl 110. 16. (RB) Rod 8>1l"&vu Hcci11111!11mii

12) 10/28 7! 70 To l!'epa!r moan electrical generator ll Steam i. !le6~- .. tll . .B .lo 81ydrogcn cuolcn. l!'owcr E11cimngers

.(llA)

13) 12/20 0 0 llhU.e sbutUng down for ochcduled A 3 Xnsttnuoente- !net rwucn t e- 11!2.4 I.

• I refueling, ~bcrc waa 11 reactor pro- &:ion I!. t lon !Ii i l

tect1on system flow trip. Co11t1ro8e Coauroh!

(KA)

~ ..

9 i.

• • * );.

,*f.

\,

l'Al.lSAllf.S

'!:'11bh Ul.6 t'OllCEP OUl'AGES ANi>> 1'0Uf.R llt:UU<.'TIONS J

I --------- *-- i DBE(D)/ "

Pntu Purntlon l'owur Hu1*ortablu Slmlduw11 $)111i1!<11lll Cmu1*onu11t NSIC(N) ['

tlo. Evunt Pu11crlpt 1011 C11uuu (11176) (Hrs) (I) tluthoii

• Hge11cBvuiJ lnvolvuil Event i'

\I Category I) 5/J 72 0 l.eaklng CRllH 111ccl11111lcal 11.e11l A !IUUll:!dll!" - C1111t rul NI.I.I rcpalreiJ. (Rll~ Reul llr Iva Hue lm11 I 111u11 bl

2) 5/IJ l 10 Holature 11upocotor relief valve A Ste11,. & Valvua u1.1;1, ....wI ugumud. t'1111hy pilot vulvo 6'11WUI!'

rep11lcud *. (HU)

I I l) 5/10 1 10 l.Oa11 of fuedw11tog* 1*ump c11u111:d by , A l Stemm a. 1'11111p11 02. 7'

. l'

.. I

,j.

I

4) 5/12

!~

146 10 low g*uwp 11uctlon g1rc11111uce. Onllf 0110 co11du1111uto 1*urup wo11 An operut 1011 *

!lu11u!r 4 leokl11g CllllH 11uah. A R'owur (Ill!)

iiuacituir (itU)

Control llod Prive NI. I. I

. *~ tl**c h1111 lll1ta j

S) 6/14 90 100 Count !111 roo111 a!r co11dlt lonor A 1 Ot!1ur !Uowun tll. l.4

• 1

• faUed, coualng tlCA/GuLl. detector to fall.

A1.1119lAorlu11 (AA)

• 1 i 6) 1/l 15 t -:*

~l (lcnuntor no!tao 1111oui11ly. llewulill11g A St11ui111* &. (luau:r11tor11 Nl.1.4 donu 1111 a1111 duct lau1ldcr gcncnuor. i'ower (llA)

1

'1) 1/20 Ill 90 A'lont tripped during thu11ders111r1m II l lE!L"cUAc: Electrical tl'J. 2 ci1.1e to 1Ji1u Coult lfJ' outage planned A'uwes* c.anductor11 l for. 1/22 wus 11turled co rl!pBlr <<A*A)

~ ft !!Ilk I llR CllDtl 11u11 Ill.

1

8) 8/6 20 100-28 Power ruductlon. 0110 hmklng cun- A 5 Stmllll 4. ll**nt NJ.2 il1m11ur tube wus fo1111J and 1*l1111uud. Power Kxclm11gurt1 (IU:)

,,l

-I PAUSADES

'l!'abla 111.6 "FO!ICED OUTAGES l\Nll !'OYER RE!IUC'll'BOUS iI DBE(D)/

I No.

Datu (1976)

Duration (Hrs)

Power

(%)

aupnrtable Evunt Desc rip t!o1i Causis Slmll:ciow1111 Hutliod Sy&t1!!111 l!nvnived Ca1mpontmt fovohud NSIC(N)

Event J  !

Category

9) 8/2l 45 95 1.oaa of vacuWll du!'ing condenser G  ! Steam & V111Hvee N6.2 Denk teat due ten improper' ollgnm1mll: l!'cwug-  ;*.

of air ejector. (llC)

- b;j i

10) 8/25 11 0 lnatob H lty of feetlwater regulot !ng A. Steu111 &!. 1!1s1t rrt1R1enu1- D2.7 ~

-~

system ceused SC low level trip. Power Hon ii

~mo Contiroll.s

~**

11) 8/27 18 4.5 76-029 Hotog-/brake fa!lure of 2 CRDH'e lltl!!actorr Co1otl!"oa D4.l .76-030 caused core t!Rt &n excess of TS (RD) lltod-!>dvo UmAt. Rx remninud cr1t1cal. ~!ecllan il!ms I

i

12) 8/31 5 95 HS1V closuru due to loss of air A Amdllory lfahee D2.4 couaedl by llWi hmct ton of 4-way Q'rrnceas I vaAve o~ tho !natrument oh drycg-. O'M ll) 9/12 18 100-U l!'owerr !'eduction. One Rco!<Jna con- Slteom /!, !it!'lllt IMJ,2

l. denser tube was found and plugged. ll'4JWllll' 1Exc911mgeire

! (liC)

14) 9/19 9 95 One leaking co11den11er tube plugged. *- A S1!.em11 !!. niea*t t!]. 2

• • =~":"": .** ll'ower lr:xch11nReirm, 1I (llC)

I . '*

.1 15~ 9/28 !6 95 Condenser tube Rcak. 91 tubc11 ... ] Stellll! ii 8!e1!ll:.  !"1.2 j II rower (liC) lbch1mgere ~  :

1I 16) 10/20 168 ROG B.eak !n regc!"lel!'et!ve heat exchanger  ! Aa1xAU.1111ry Bierntt iB.2 I of tthc che1olcaR and volu111e con~rol ll'B"OCCR8 i>xd1angel!'e

\ system, (re) i I

l J

j

.J

i i* ...

j.

. ~'

ij r T11blu Pl.6

• I' Al. I S1\ll ES (.

c*

~:*,

t'ORCED OUTAGES AHU l'UHER 111.!UUC? ION!! ~....

---*---* -*- f*.

'l DDt:(D)/

Ualu Dural Ion l'owur lti:portub lu Shutdown Slfe1t111* CuM1pununL HSIC(N) r*

Hu. Du11cr 1ptlon Cau11u ~-

(1916) (llra) (Z) i::vunt Hut hod llnvuhud l11volvi:d Event

• catugory ~-  :.*

1,,

• --*------- i

\*1 11) I0/21 2 0 Pnrh1g t rimufor to st11tlo11 power, ee1*vacu water w1111 Bout .iuid cool l11g A IJ!!luct ll'&c l'uwua*

Electr &cul Cu111luctor11 tll. I , t, ~~*

!owur !'UUll' u*.il'A""' * 'l'urbl1111 wa11 . (t:il) bf ~' .

tr lp11ud. Rx rc11111Jut!d cl'h lcal I (

t-4 I..

ID) 11/12 136 100 C!Ultl acul 1 .. ukngu. Seu.la l'eplaceJ. A

' !lu11c~ul!'

(il!I)

Conlroi Rod Dr Ive HI.I.I

{.'-

Hue l11111 I sou1

19) 11/24 22 9S Ccnorotor voltage regulato~ *failuro. A l S1tam111*!!. G1*11eralur11 D2.2 l!'uwcir (1111>>

20) 11/25 s 8 I.ow 11teu111 go11er1u:or luvel. c 1 Steme S.

• P111ops 112 .7 .

A'uwur (IHI)

I

21) s a.* 02 .2 -

+ l 11/25 22 Generator vi>ltuge rugulator A S!ewm Cl!11er11tors

-,, flliluru. ft*uwHr

.-, (HA)

22) 11/26 s 0 Spu.-10011 high g1r.,s11urlzer 1*n:11su1*e A l Biuu n1111"11t11- IHGl I UlllUlllU- Nl .11 111c11111, ~ ft&>IO I. tlun I.

Cont rule Conll'ol11 (I.A) 1'J i

2l) 12/! ]2 !00 Co11urutor vultaae* ir.fuulutol" full111"e CUllliC!d 11>1111 of loml.

A l. Ste11111 I!.

Powor Cuoea*otucu 112. 2

i UIA)

.* .. ~ 24) 12/9 100 16-0leS !IUA 1u111uencH" am! bucku11* full!ld ug1er11te 111 te:it. Clutch dJscu t~ A 1 Aux8AA111r11 l'll'OC4111R Hochanlcul Funcl 1011 tO. 2. a  ;-.

\".'

hUuil ti> iii 111m1:11ge upon l'l!llC!l (&*(:) UnhR

'l 11 I""" l. ..

'j c:.

.*~1 ,. l.

PA8.ISAIJES

'll'abie Bl.7 ll'ORCED OUTAGES 1\1\l!I i'OWF.R HH>llC'ft'ROl\IS l>BE(D) I il111t0! l>uratJ.011 ll'owuir lllaipoirltabiu SB1Mttdlown Syslt!.!Blil Cumpo111.mt NSIC(N)

No, Event lllescr lp t Aon Cm.1011 0977) (llrs) (%) Hetho..S l!nvoivud invoivlllcl Event Category n) A/6 3~ 100-!5 i!'oweir iredluctt!o!ll. Repair .le11kAng A 5 Stenm !. lie.at tn.2 main co11cl1m111111r tubes, Il'ower 1Exdurngcn1

(!IC) . b:I

2)  !/l! 11 100 CooUng tower ?Willi t1dp due to 11 A 2 illlllllltll'."Wllenta- ll11111tru1ne11tlll- l\l2.li

._.I Cl\

false low !eveR Andtcat!on. Hon I!. ldon Iii Co11tiroh Coo1t1roh

<<na*:~

JI) l/H 8 100 Ho!eture separator dra ln tank dump Slt!.!!11111 lli Vaive111 ~H.ft.4 valve failed open, W'oweir

<<mn

4) 1/17 jl' 8 !00 l'eeciwater !Jump trl11, llfi JI S1ten111 1J, ll'nung1s 1>2.7 n*ower (Iii!)

5) I/JO 82 !00-35 l\'ower reciuc It !on. Repoill'."edl lea Ung Steam ti. Bieatt 1tn.2 aiain condlenseir tubes. ll'ower IExcD1na1ge1r111 OICb

6) 2/9 32 .::100-35 l!'oWel!' re1!ucg; &011. Re1>airecl !eok!n*g A Staiam lli D8e111~ l\l].2 ma!n condcnncr tubes, H'owcl!' 1Exd111a1g1:1riil

. '* (liC)

7) Jl/U 36 100 feedwetell'." pump trig due to broken A St1.111111 if ft'tl!R!'iil D2.7 reverse rotation belt on the pwnp i'owell'."

J ~uri>ine. (Bill}

I 8) 'J/U !JI no t'ceclweteir pUDl!D @: ( fp * ] Sitt!!Bllil !!, 11'&unp111 1>2 7 0

I 6'uwur (1111)

I 1*

i

• * ~

I'*,

t.

/

(C'iS) I t

110.alll:eua,1 I

h"J"S 0

• a111n11 llOJJUIOllJ 111Hm(X3

'1' I" IN 1101111111. P".en:n11Wu;t v D8Jncl lllllUlllJU~llD:"i,llUJl(UD( n .11vnihr11 ~co-u i101 '11.Z 91/9 ('II (ll!l)

'Dll(Vll

.<11lcln11 llllttl!I .JDlUal(OJ 01 8.tJDclnJ p1111

• I lliill( 00!) .Jln11.1n11ttJ lh11.11mq .Jo10111 dmml 1uu1nn;i O"lN B.JOlc>ff .11nt:"lun11 li.Juncpd tlJ t1Df1D3Jp11f 1111101 ii n 1tO'B 001 ~z OC/l (ti t~H) 'lllljlltOp 11.Jnlh1m1:1*;t .111111r.r1 tt 1J .Jattn1 91111 no:> lllO.JJ xoqJ111111r1

• • ~ti Ju;iu .C.i111y u*nv R in"" dnt1113f:"l. 01 °0011:"1npa1 Joftnd (K-OOi Cl 91/l (ZI llftl!lJllmpOlf OllJ.J(I poll

'HOJ 1U.JqJ II '"1"1 P:mpnJ n~

pay l111911f 9Hptl!lll JDRlmpllo:"l 111111 'V9-ll l"l'IN (OJ)lln:J v Ja1r.a11 Jrtl""P'"'J *111mm Hllll:J JJ11dn111 001 I'll U/~ (ti II

' *'"1"1 i'"*'

Jn111mp11n;i Hf 11111 9111111101 prrln.1 oil IV ':J:>j 11.Jne n I UUJll..illl:t.J I!

"'"' .1;1111a*1 .Jn111ttpnnJ "l'l.i. '11""1 01(11' OJ """ llOJ 1U:"lf('llf 1011"1 Ji':tnn (lili) 1111111 *.tpa1ln.11I 9111t101J:it1nJ :JI( OJ u.1n9mnpx;i .llOftn,( (>UnoJ 9111UA *.1a1ua11 ..J911llt('D9J o.Jn!I l'UI Jn1111 'l1 ettl!e>llS v -111>.UI 111111 iHO.IJ Oll(Ull llJU.Jp Jft11lo11 001 .,., 91'1 (01 I'

(:HI) ti

' l'lN RJnl111mp11;i 111011

.ll""'°*'

'! lllr.OJS ~ y

""l":t 811Jllt131 pa.IJ1!1fo11 J:.1u1nr110:> 11111111

'UOJ1:111(>0J JOftOd i-001 6 J;/'I (6 iI l.JolOJ11:J 1UIJll3 pnllJOllUJ i"'" u011111 pl>llJnH DllDD:J ~11n113 (Z) (D.Jll) (U60 (N):JISN *llU:tuod*n3 1111"i~e1ts Ul'IOpJOlfS JIOl\Od OJl!Q 'OH

llf(t11.IOdnlf UOJ111.JllQ I (0)311<1 Ii SNOi .r..1n0311 lliltlO.I C!NV S3!lV.tnO Oil!llln1 I

.*l l' 111 O{qDJ. I S:-tOVSl1Vd I

II*

I.

lI

.- e ****--**--,*-*......*-**-*

11'1\8.R SAD ES Ta lb le Ill. 7

• "ORCE!ll OUTAGES lltlil l'OllER llllEO!JC'll'WNS DBE(D)/

D11te Duration Powurr llleportabRe Simtdow111 Sysh!SI Component NSIC(N)

No. Dl!SCll"!pt h>ll ICin1J111cs (1971) (llrs) (%) Event tii:ltilocll K1wohccll Knvuhcd !Event Category U) 9/24 64 100 17-047 Losa of off-s!te power.due Ito Ii ] 1Uccll.ll'1!<1: rr:Aec tit' !cmll IM9.2 J lght11!111g wh&cio cmueed cooling l!'oworr Co1111i11ctorr111 tower pumps &o U!g1. (!!A) b:t

!'ower !loduclt!on i

16) 10/1 14 100-:U i. lllepa!r o!ll lleak in HWFP, s Site!!llll lli 1!'1001!l>lil RH. l.4 I-"

00 i!'IDWf!!'

2. Cleaned main cond.onaer west wateir box.

(Ill!)

!7) l!/25 51 100 17-055 Loss of off-aAta IPOweir, cauee IUectrr!<1: IElleUir!<l:ai IM!P .<<Jl uni<11owi1. l?OWl!!I!' ll:o>01d1J1<1:to1r111 (EA)

18) U/27 18 Losa off ~eedwaler pump dl!lr!ng powerr G ] St~.rnm Iii !Pumps* D2.7

()IOCRiatRon. POW<!!'

0111)

19) U/!R 35 !00 17-058 D..oaa of off-aA~e power. Cause A !l:Recttirftc IEiecuRcall N9.0 a.111known. Puweir Comim:ll::oir11 (EA) ll'owcr Meduc~ton *

20) R2/U 15 *. JOll-48 Rupuirud icak B~ & fecdwater  !!nc. A Si Staum ' ll'!pc11, ltn.2.

• • • • • !'UWlll!" ll'Htdng11 Olli)

l'Al.ISAlll(S Tuble Bl,8 tllllCED OU'fAllt:S Alln l'OWU REDUCTIONS DBE(D)/

lllUu Pnrnt ion Powur Ru1*0U11blu Shutdown s,,11~011a Couipona,:nt HSIC(H)

Hu. f.vunt Duacl'll'tlon C11uuu (1918) (Hrs) (%) thilhod !lm1<lllhudl lnvolvud Event Categor1

-u-**-**---*

I) 4/21 12 10 feedwaler plallp trip. A l $U!BIM  !!> Pumpu 02.7 IJ>owor I

<<mn bJ I 2) S/I 2l!I 100 l.enk Ing CIWH aeal 11. A -1 !lie!IC~l!ll!' Control H1. l .....I

\().

<<110~ Rod Pa*lvou

\ l) S/11 16 0 l.uw ateoia generator water level. i A l S&.!!!1111 l!i I nut nw1cn1*11- 02.7 ll'owua*

UID) "" a.

I Cuuuoau

4) S/20 ,. 9 100 Luw 11toam gencroto.K' water luv11 l ] S~1!!111111 a. l/lllvl'U P2.4 1 I duu a:o 11hri11ko11e cuuaud by clouure f!JWO:a' I

I of tlSlV, (!Ill)

I

'I' 5) S/22 ll 100 2 tl.~lV'u closed. A ] .. Sitomllll i Vu Ives 112 .4.

l l'ower (llB)

6) 5/2] 20 ... !D bus fnlled to transfer. A l Eluuir!c Chcutl tll. I ,1, l'owur Cloaun/

(Ell) . lnhirg-uptera f*

t

1) 6/6 36 78-019 l11011er11blc conuol rod drive. The A Reucto!' Control 114 .J drive a&ut!a1bly was ,c-u11lm:ud. (llU~ llod ilrlvuu I1 ,.

8) "6/7 ft.ow I' 11u10111 11unorator wnlua* lovul d1111 a:o S"ecdwutcr p1111111 11rob)C?1u11.

A l SA.&!IDlll ..

0'0IW4!11' (1111)

I nHt r 1111u1ata-tlon' C1111&roh 112. 7

- ~:

. I:

j  !

L..

j .~

I

• 1 <:

l ~--

ll'AUSADIES Table Bl,8

!FORCED OU'CAGES AND Pll\.1£1111 REDUC'R'KONS DllE(D) I Pat11 Durot!on IPOWllll!' Ruporttalblle Shu~<<Bow111 Sy&lt!.!mi <<:u*pOUlllUllt NSXC(lll)

No. DHClr Ip t 10111 CSl!Bti 0978) (lllt'S) (%) EV&!Rt Huthocll Kll!vollv111d nnvC1!veJI Event Categol!'y

9) 6/8 6 Low stoam generator water !evei A 3l S1te111111 Iii. instrU11el'!1>;111- ll2. 7 due to feedwutel!' pump problems, l!'uwcrr t!o111 a.

<<llB) Cm1t1ro!11 td I

10) 6/11 13 100 HSIV cloeurre. Turbine tripped S1ttea111 & Y11Bves D2.4 N 0 '

1111111ua:Uy, !l'oweir '

(llB)

11) 6/13 8 Low steam generatol!' water level j) Slteam /I. . UReu*e.imenfe- D2.7 due Ito feedwoter pump proble1ns, ll'uwnir tion &,_

(1111) CoB1troh

12) 6/18 :u 95 Lightning stl!'ike, iH 3l IEDau:ttir Ac. 1Uec1tricall- ~9.2 ll'owe~ Conci11c1tol!'e

<<ED>>

!l) 6/28  !.23 Excessive lleakage from i!:Ri>H eea!m, A l. llle:icltol!' Co111tron. Nl.2

~lllB) atoc!l ll>a-llves

14) '8/8 H ....... Repalln to CRDH cooUng pan, A llleactoir Contll'oll. eaa. n.111

~llD) !Rod i>ll' nvee J IS) 7/CJ Fal!11re o! slteam supply valve Ito A Rmll!oaclt !vs Vmivelll Nil.R.4

' SJAE reeu!te~ ~n lloeo of vacuum. 5'1111te Om)

J l

• 1

16) 7/U 140 gm Repe!~s to CRDH seals. j) llleactoir Orn~

<<:011uon Roell Ne.a.a

' llrlvee

]'

j

"*. .... **~* .. *-** -- _,,.

lI I

• 1

--1 l

i l ~ *.

I I rAl.ISADES i..:

, Tab\fi Dl.8  !

FORCED OU'l'ACE:J A D l'Olll!ll llEDllCTIONS DBE(D)/

D11tu Duration Power lluport11blu Shutdown Slf&ll:Clllll!l Cu1111u>nunt NSIC(N)

Ho. Event l>uscrlpttun C1111au (11178) (Hrs) (:0 Hut hod KnvuRvuJ Involved Eva:nt Catego*ry

11) 8/1 20 **eedw11ter pwu11 trl1** So11l repl11cetl. -A ) Stemn1 & '

l'l!lllPll P2.7 l*

ll'owoir (1111) bj I

18) 8/28 JS7 !lep Bocud Hven Clllltl 1101& la and prillMlrlf cool11nt ptuap 11uola.

[WO .A Reactoll' (Ro)*

Control Rud NI. I. I .....

N

_ I f.

(*

Drhua I

19) 9/ll 711 S ht CIWH aeab wore r-epl11cod. a  !\G>llC~OI!' control ua: 1.1 ,.

(llU) Rod Drives

~

20) 11/22 95 100 78-0ll CRDH 11eal replace1111nt. A !ltiOll!ltUll' Control* NI .1. l (Rll~ .Rod L.

! l>r Ives  :.

l I 21) 10/2 12] !00  !.Hkb1g CRIJH HRlll. Two seals wuce A lll<1>actor Conuol ua.*1. I I

I replaced. (RU) Rod l

Drives

22) 10/!0 I 'JI, 100 Lun"h111 CllPH aenl11. Suvunteun u~nl11 *A !le1u:tcr Cui1l rul HI.I.I

l. weru repl11cud. Ulll) llu1i

)

Dr lvm1

.1 2]) 10/17 10 65 t'uedw;itor p11111p u*la* uccun*ud during G ) Ste11111 & l'ump11 1>2 .1. ..

an 1u umpt to clumao Liao 11tea11> 1'1>W&11"  !

ii 111111ply from hla;h 1ir't!e11ur11 lu low (1111) '

i'  !'l"f:s11uru. tilenua, I-

• • l l ['

l i

.l c L;

j '

j  ;: ..

J ,. i

i 1l l!'Al!.HSADIES Table 111.0 rfORCF.D OUTAGES At!ll rmmR l!llEDUCT!Ot!S DBE(D)/.

Datu Duration  !.'uwul!' lluportablu Shutdow111 Syet1:1111 1Cu1i1ft1001un1t NSIC(N)

No. DuscrJpt lon Caue11 (1978) (Hrs) (%} 1Evu11t Hu1tlmcll Hn11oho111l !11v0Avud !Event Category

24) 11/29 278 100 Leakina CRDH seais. ll. l!le!llctor Con~irul tU.l. l (RB) !todl i>ll'Jl.VC>lil b:I I

25) 12/10 lo 0 Ba!aeced 111a!n turbine, S1teo111 ., Toori>BB10111 N6.r N N

!Power

(!IA)

26) 12/ll 89 30 !Excess oU lln lllll!!ll genel!'ll!i!:OI!', ll. Sa:eaDI a, Ge11el!'atoire HLl.4 "i

ll'owair

~iiA)

I

27) 12/16 21 30 Feedwatel!' pump ~rip. A jJ Sltemn a. P11111g>e D2.7 I Power

(!Ill)

I.:***

11_

i*

i'Al.ISAIJt:s Table Bl.9 l

t"OKCt:IJ OUl'ACES ANll l'Olll::R REIJUC1' 10115_

i I,

I

\ DBE(D)/

Datu Puration l'owur ltu&>&*r~iablu Shutdown Slf111h1111 Cu111po"'mt NSIC(N)

Nu. Duucrlpt loll C111111u

\ (1979) (llru) (%) F.vu11t tlutlmd &11vu llv.:J l11110Jv1:d Event i Ciacet:or-y tJj l I) 1/28 19 09 Fuedwater r11s11l1itina vulve fa lied A 2 s~ a1111 & viaav-1111 1>1.2 N I

I open. ll'mrnir w l!:u~1ve II" s j 011 I

<<mu i

• 1

2) 2/l 25 89 Operator Inadvertent l)J tdppeJ 11 G l Kunc gut* I nut n111eut11- Pl.I pr J1111&rlf coolnnt PWlll'* 'Ida> 111111 Cuu8unt t Inn '

(nllowoll by 8'llll"l811l prlmllrlf cuoU11g (Cll) Cuntrolu oyste11 (PCS) coolduwn 1111d safety l11ject:Ju11 oy11t1.1111 !ultJ11t lu11, J'CS pru&liUrG recovored Ulld 110 111'51 Wiiler Willi put In l'CS.

l) l/l 22 89 FaU11re of II conl ro l VD Ive rcault:e!d A ] S1t1H1111 & Va1Ye11 112. 7 -

111 11 lumLur da*oaJn pUlllp trip 'th.1t §'uW<Hll" cia11nu1I the 1111111 feud purup tu trip. Cum1t:IJ'llll 1011 (Ill!) \'

l.*

Ii) 4/1 67 ... 8.1 t'eedwatcr p11111p Ulppud due U* a11ur 1- .A - l S!1!!111a 6. 1 l 11mp11 02.7 ou11 vJbrag lon tr 111 a li;nnl. 1111d caused l'uwer rcmctor t** L*a* Jp 011 low S/G levol. C4lnvun1 I on (1111)

Sll.11111* ti!.

5) 4/25 42 86 B.uoa of gc11c&'11lur load co111I h fo11 due A ] -t:U11u ral or D2.2 l'owur tu vultogu re:gulutor IDll If UllC t Ju11 , (Ha In Cu11vei IJ"a lo1g Ccm:ralur)

{llA).

611) 4/lU 41ljb 1111  !.t111a of !uad condlt Aon due t11 _v11 lta~u A ) S~t!!in!io ii a:e1u~caa*1*c 112. 2 rcguiat11r nmlfmu:t 11111, - l'uwur (tl11I11 Cugwe rs! 011 Uc11uraL01*)

dliA~

I l

j Ill t:rlt hnnl i:*l t

ll'A!.A:SAIJt.:.

Table lllL'J FORCED OU'l'AGllS ANil !'OWER llEllUCT!ONS

' DOF.(D) I Dnte Duration ll'ower !leporublu Shutdown Syelta:m Cnm31onont NSIC(N)

No. Evllnt a>u11cr!pL!un Caueu 0979) (Ura) (%) IH1!1thod Knvolv11d involvlldl Event Catll!gory

, I 6b) 4/30 463* BB 79-020 Outage was extended to re~olve inade- A ~ Engineered Shock Nl.2.1 quate p!ping restraints for two Safety Suppreasors s11fe1ty injection lines. lFeaturee

~SIF)

7) 6/'J 23 u Yhile performing condenser tube leak repairs. trip occ~rred due Ito a reverse power aituat~on.

A l Steam &

ll'ow_er Convers!on lleat Exchangere (Condenser)

Nl.ll..4 r

N

~

UiCb *1 B) 6/16 lS 64 Unit taken off Une for condenser A Steam & Bleat Nl.A.4*

tube leak repeirs, l!'ower Exchangers Conversion (Condenser)

(!IC)

9) 7/11 85-73 Power n:eduction. Cool!ng tower A 5 Auxiliary Pumps Nl.l.4 pump trip, power* reduction required Weter to prevent unlit from tripping on (WE) loss of condenser vacuum.

10) 7/25 88-26 l!'owen: n:educt!on, Reps!~ condenser A 5* Steam a. ileat tH,l!..4 tube leak. Powen: Exchangen Conversioll (Condenser)

(UC)

11) 8/10 21 88 Both feedwater pumps tripped during A 2 Steam &  !!'umps DZ.7 turbine valve !testing. l!'ower Conversion (UH)

12) 8/24 2l 91 Loss of feedwater flow wh!le cuttins A 2 Steam ' P11111p111 D2.7 in the condensate demineral!zera. Power

<<:onvere!on

{HH)

Refueling outage 9-8-79 through end of*yesr.

- ;i*.

s~tmatud.

B-25

• Appendix B: Palisades Part 2. Reportable Event Coding Sheets

---

*---*- - ----------*~- ..---- -*-~---------- ------- ..

---

;~-.-~*-~-~-.---*-~ -.----

T11ble 82.1 Cod lng Sheet for Reportable Event!! for l'el!aodoa - 1910 NSIC

!Even It !lepo1rt l'innlt Co111g>o111ae11t llb1ml!'minfi Sfg11H Acancc N11111beir Acceselon Systcu1 1Eciulp1111mlt Bn11tir11men1t ~IDUSlil Date Pate Statum iSll:AltUl!I C<mdUU0111 Ce~egury Humbeir 47817 4/23- 7/28 A s H Ill N D N 4/'J.S 58468 11/30 A v Joi!. A s.x ia if t>>

I N

i Ij I

1

'I i

Ij

• )

j l

j j

• r i

I I

,,i

• Numbeg-H:i1C Acc.,u11lm1 llnmber E11c11t 8lut1:

IUUIO: P"*'

lh*1uu*t RlP~B l.uu11111 l'lunt StlllUll D l l l ! O : . l ! l l l l r L l l l l l l l 11\ICllLll Sysu..

E11ul111Aen.t IUI" l"HlllHl*lllll -

11111l1"u111e11L Cou151u111.1sa &

S~nau:a Abn111111Rl Condition Couuu SlcnU kance Ca1ei;u1*y

~

61126 l/B-2/U 4/6 A s

" - 1111 ~* D 6120] 5/1 5/16 A s - H D N D N 61&16 5/8 5/18 A s &Ill - II F B C4 641aJ1 5/H 6/1 A D !I - I.I N D H 66501 6/18 6/25 A N 00.1111 - c E c H 61.652 6/25 1/10 A T D - Im  !( .D 65910 8/19 8/21 A *S on.n* - !I J,H "

61565 9/2 9/9 A H,O ** II' I&" l)

E II Si,Sl,CJ 65969 9/16 A s 00 .. II

' N,S,U D S4,S9 66511 9/U 9/24 A 14 J - <<: ~* E C1 66412 9/29 A H 11 I' Ill Q 68606 ll/5 ll/19 A 14 .i - c E D

D CJ 66281 11/5 ll/19 A 14 J - I!: o.x D C1

!2/l 66415 12/16 A v C,JI  ;..

lli Q D Cl 1

i 12/16 A ff ll - DA x fl

" bi 661176 !2/l !2/11 A FF O,DD - IBl I) D I

• l 68000 B2/2l 12/lO B v J - 8 DD D C1 N

. I 60605 !2/28 l/1/72 B AA. 00 - Im H c -

1

• I I*,.

j*

I*

II

!J J

l r

J ,.

Tablo 82.4 Codins Sheet fo~ Reportable Event* for Palisades - 1973 ,

NSlC t:vcnt Report Plant Com11ommt . A!mol!'!lllal '

Syateui SAg11H.i.c111nce Numbor Acceaalon Dnte Dnte Stotua E1gulpmc11t foetrwacnt Ceuse Sttinttus Condh.80111 Cin~cgory N1101ber 77962 l/lS 1/2!11 D M n.tat IS D,F,~ D C5 79379 3/6 8504P. 1/16- 5/!4 D l! A Cl i/28 81865 l/25- J/2] D R £ Cl 2/28 Ao-1-n 81266 5/18 5/29 D v s Ill Q  !!) C7,C8 A0-2-n S/1!11 5/H D AA 00,QQ Ill F,K' Ill! *N AO-J-7j) 8:U9l 7/5 7/17 B v c Ill '!l',11 A N A0-4-73 82958 7/23 8/2 B v s IH N A0-5-13 83206 7/27 8/U II B O,X,IDllli IC ' HID D N A0-6-13 83206 7/27 8/U B 0 IN! c z D N

. A0-7-73 83206 8/l 8/U !I w l!fi c H ill N Ao-8-73 83207 8/6 8/16 Ill v s  !! N A0-9-13 83219 8/8 8/20 B w J B Q D C7 A0-10-73 832l!il 8/8 8/20 II AA 00 s I!

8707] 8/10 12/10 B [JO JJJJ "D Ill A ill Cl.

AD-11-n 83218 8/U 8/21 D v $ lil Q ID C7,C8 llJj i

N 87073 8/!A 12/10 I DD .B.D Ill Ill £ Cl 00 83222 8/U 8/22 D M 00 a 11" ,ft'  !) N 83214 8/15 8/24 D DD lli,S . ii Cl i 83564 8/U 8/29 D x 00,QQ c B. c 00 I:

I I I I,

1; Tt* * ~ (Cunl lnm:J) tl:HC ~.

F.11cnt !lc11urt I' l1111t Cu111po11<<:a1~ Ab1111r111:al SI 1:114i1&:.111ce HU111ber Acce11ulun Syalc1>1 Y.1111 l puu'!n t lnel rU111cnl C11u&u D11te 1>11te Stntue Stu~um Cundlt Ion Cuto?gory Hu111bcr 1540~ 6/29-6/10 9/8 9/26 D

". J c r,ua D C7 75015 9/21 D I 00,l!'f c ti IJ ti 75612 75907 9/'J.l 10/21 9/29 U/2 b

u M

V,11 00,QI)

J,n*

- c ** D N.

16064 10/29 ll/10 D 0 , IJC,S !I c

ff,11 Q

D G

C8,C7 76752 U/26 12/6 u v J c Q .D "C7  ;

Uli: 72-86 ll/29- 12/4- B s 00 18568 12/l 6/17 II

'*" D N 12/9 12/ll D D,V FF,Dlll s Ill N,(l,CC D Cl 11414 12/18 12/21 II llD 00 Ill il,F,S,V II C1

• • ~ .. '

I I

b:I I

N

'° I

I 1

I ~.

j ,'

Toble .82,J Coding Sheet for Reportoble ~vents for ralisadoo - 1972 HSIC Event Re11ort Plnnt Compomm It Al.mo ~*ml - S!gn!r8ro111ct!!

Nwuber Acce11s1on Sy11tem Er1ulpiaentt !niallrW1111B1t Stai.us Condition C11use Dute Date Stotus C111icgu1ry Number 69198 A/5 1/14 D x 00 "." c IE, Ill' ii) N 6047] 1/28 2/7 B v JJ Iii 1119 HJ C7 39569 Ul 2/11 IS 1 00 <<: li A N

19563 2/3 2/14 B w JI ii F IS C4,C7 2/5 2/15 II F U,HH  !! G ll Ril 55271 2/26 'J/6 D A Z,l'I!' Iii IE,N,CC Jll) N 55270 2/29 3/!0 D 0 ill If' c  !,lF D N 69527 l/U J/22 II v JJ I DD D C1 69527 l/12 3/22 B H,O 'if  !' la Q,S G S2,C7 69521 l/U 3/22 Ill w ll c D!I ii) C7 69527 l/ll J/'H. 8 v J Iii Ill!) - D C1 5/4 v P,DB C7 71415 4/23 II JI I! D "l/U 4/23 5/22 D x ll'P Iii 18 c N nu1o 5/7 5/16 )!) s o,n c IE Kl> IM 72193 6/A5 bl I

71696 5/lB. 5/22 Jl) x QQ llA s ta Ril w 0

71691,, 5/U 5/26 I> lll,X I!' c Q Im C4 j* 7!694 5/18 5/26 D 0 IHI K Ii ilJI l!!l R,j 71962 6/2 '1/U D w .ll c S,1 8R c?

72105 6/5 6/14 D JI QQ m II' Jll) INI 1 **- -

I I

_I II 1-l

• ~ ..

• H1ualiur HSIO Acce11&lu11 Hunabc:r

!!11u11l Dille

• 11u11: "'**"

P1111on Dato t..U'l*llB r&ang Sc.nruu

~ll&:U._

Sy11lc111 ltl . 1:pu1 *uo1u r.vuntu E*111 I p1u11 t 1ur n111&i1Qll8 *** A'.11,.

111&1Ull1hClll C11111g11111u11 t _

St.\l1111 Ahnuuuu I Co111lll l11n Cuuail.'

lil1~11l I' l1:1111cc C11tu1:11ry l'

1.

i*.

UE U!llU6 !l/1) l/22 u DD II 8 u G..

3/14 "

89166  !/6 2/20 D v J la Z,1111 D C7 UE 8!1114 D

,.I 2/16 4/l II, S . II H 8Yl20 l/H D c *u

1 I

8966S 2/26 l/16 'J/26 D "I' E

H c H,U H ll B C4.

~

AO 89741 l/26 4/S D FF . l ,KK IC IA L,lt,T A 11*

A0-5-14 90584 3/21 4/21 D K,Y Z,Dll,J.1.00 A T,U A H A0-2-74 91159 4/1 S/HI II s l.,UB II F D H p

l AO-l-74 90640 4/16 4/26 D H c H D H

{ .*

A0-4-74 911640 4/11 4/26 D p H c H AQ,-6-74 91161 4/22 S/10 I.I I' H c Cl D

E "

II A0-10-74 9:l44U 5/S 5/26 D s 11,tlH c *11 D 6/14 " l*

A0-1-14 91160 5/6 D c A0-8-14 5/16 D

" AA

" H z

D H 94119 5/14 5/29 1/llJ 8/IJ

" KK Ii B C4 b:I Ir UIH-74 96068 .5/14 9/21 D p c Q D t:4 w I

. t-'

A0-12-14 9420 6/2 1/15 D c I) t:

I I: A0-9-74 92606 6/4 6/14 D 0

0 H,P" c Ii

• • ll II

' A0-11-14 95118 6/9 6/19 D ti ** ~ Q c H I

. 8/'l8

i*

A0-!11-74 7/U 1/'l'J I.I II AID IC 011 I) II..

A0-15-74 95141 1111 7/29 p F,BU 00 11 11,11 c ~

1 I A0-16-74 95144 1/11 1/'l9 D s z I u A l "

I I

I l1 i":

i'

.j J . "

.f .

' t t  !*:*

).*

~'. -.

. ~* .

Table 82.S (Continued)

NSIC Nu1ubeir Acceas!on EVi!!!llt Mcpon Pinnt Slf&lCDI IE'11dpae111t lnauumcn~

Componenll: Abnoraud

• c211se SlgnH lcance-*

Number Da~e Dnte Status Su1gu111 Co1ul!t 1018 CaR1!!£Ult')'

AO-U-7/o 95142 7/28 9/10 D AA <<>O.QQ c IF D ~

A0-17-7/ii 95ll0 8/28 9/9 D A ~ II IF.,J.N <<: .N 9707/o U/U A0-18-7/o 95l4A 9/1 9/U D s s Ii Ill A0-19-7/o 94198 9/6 9/16 p M c B H D ill A0-2l-71o 96562 9/6 10/25 D s Dill s A0-20-7/ii 95399 9/!0 9/19 D 0 .IF "!'

Ii c Q

!I c

K N

A0-21-7/o 96361 10/8 10/!8 D .o JN. mo II' c &11,Q A0-22-74 96168 10/!0 10/18 D w IE 1111  !!

D II "

C4

' U£...o2-74 97578 10/17 11/15 D H I!.' c ' Q Ill C4 A0-24-74 96479 10/19 10/29 D I> DD c $ v Ii C4 97481 li/!8 I

~ I A0-26-74 9714J 11/4 11/ll D 0 DD c I! . D il'I A0-27-74 98570 12/!2 D A,DO

• 00 Ill B D Cl A0-28-74 98569 12/5 12/26 D DD JJ Ill S,'11' 0 M II Cl UE-l-74 !2/7'6- 2/3/75 D v I!. A  !! A -il'I 1/75 b1

.I w

N

! Table 02.6 c:u1ll11u lihut t1111ua*t11hle t:ve11l11 for l'111l11a1h:u - l!l/5 NSIC t:vent lh*pnrl A' I ant Cn.. gtU!lt!ftl t llb1111r*ol Sl1:11ll lcn11.:u .;,.

Nuauber Acc.i11ulon ii1111tm11 1!*111 t g11111m t lust rauaent C11u11e Dote U11te Stntuu St.mgue Ct1111lh iun C11ll*gor11 Numbur i

A0-15-l 91501 J/20 1/ll u l,J 00 c N,1',S c,u N i (12/6/14)

A0-15-1 91502 B/22 2/4 IJ 0 !J ,tltl I t:,11,AA D CJ 110-n-2 91501 1/26 2/4 D DU FF ,I.ID u 11,tl,T A CJ I

[.

Ul!-15.,.1 101405 2/28 l/ll p AA,I' z c 1' u C4 c x,z A0-15-4 100579

!OUS2 l/1 1/11 l/21 D l' C:C,KK u c4 l'

UIH~-2 101114 l/29 4/10 p v J c 1111. D C:l A0-15-5 101116 4/1 4/U Ba PU J.I I!. S,V,AA II Cl II p 11,K B G,'f A A0-15-6 1011]8 4/4 4/14 II A0-15-1 A0-15-8 101111 1022611 ConceAJed 4/6 4111 4/16 4/H

• II B 8,Z,AA D 11,K 11,tlH II G,'!.'

H A

D ""

.I A0-15-10 102100 4/ll 4/21 B v F I u A H I

c

l I A0-15-9 j\0-75-11 lO:Ul4 101411 4/11 5/29 4/28 6/9 c

B l'

"p M H

L c I., T L,ti A

D 6/) rs Cl, T A A()-15-12 110-15-U 101466 l0l10l 6/19 6/l'J 6/)0 II II p lC.

K Iii) C,T .A,B. "

C4,C1 bl I

l..J c l..J r-=

A0-15-14 104044 6/JQ 1/8 D tt,O t' T  !!* 11 SI Ut:-15-4 104220 1/1 1/8 II v J,X u I) D C1 A0-15-15 104725 1/9 1/21 B H DD,00 ti: R,ll,T,U II C:8 *f I *i UIHS-5 104684 7/ll 1/21 B v J,O c z G H f *.

106911 8/9 (

UB-75-6 l0511t2 7/18 8/22 8 . x 00 1 1JQ T c r D ti

\**

i **.

~*

r ,'*... ,_

I I

'j L

,-j I

• .*.** v

• J , ' t ..

I

'\_-:-

I*

" (.

t f*;,.

J

., J .. **

Tot.le D2,6 (Cont_inued)

NSIC Report Number Access ton Event Plont Cll111111oncnt Aimorm111* SlgnH!cance D11te Dute Stntua System 1Eiaulp111ent !nsll!'U111ent Cmuse Number Sutua Cun1Uttio!'! Category A0-75-16 1048]) 7/22 8/4 B DD E Ii c,n. D Cl,Cl1 UE-75-7 104960 1/21 8/7 D v J c F c C7 UE-75-8 7/27 8/7 B v J c F c C1 A0-75-17 !04991 1/29 8/12 B DD JJ,00 mi N D N Ut:-7S-9 105051 8/4 8/!4 B 0 N,W,KK c IE ft>> N A0-15-18 1055U 8/11 8/28 !I v l,J Iii Q iJ C7 A0-75-2] 106450 8/28 9/22 D DD AA A,19 C]

A0-75-H 106ll7 8/lO 9/9 B v I,J !I q llJ C1 A0-75-20 !06Jl6 9/5 9/!5 B v Il,JJ rm Q l!I C1 A0-75-21 106448 9/!2 9/2'1. B i.,x I[)()) lfA IE ii) N A0-75-22 106449 9/U 9/22 II x.. IA c c 18 ii UE-15-10 107225 10/14 D n* Hl Ii ll ,X ii <<:4 A0-75-24 10751! 10/U 10/24 B !J IK,DD ag £ c c C7 v c. N A0-75-25 107762 108246 10/29 U/5 (6/19/7]) U/21 B

D D,BB ric IC Iii

'll' *WI_

c "

Ill Clo ,C7 A0-15-26 !08212 11/Al :U/U D I ,.J 00 IC II! D ii UE-75-U 11/22 !2/22 B N DD,fF (C IM D C4 ~

I UE-15-12 109188 lll/22 12/22 B N i>D,IFF c N II! C4 w

.s:-

llE:-75-U 109187 ll/22 12/22 D s 00 c H Ill N A0-75-27 10945] 12/21 1/5/76 c !J lllD,00 ~ c Iii N

J.

HSIC

&huabr.p; Acces11lun EVl!lllt !lue*urt I' I 111it Sy11tc11 E11u I 11111:11 t J1111tl'11munt . Cu1111wmm It Ah1111rwal Causo s *1:11 Ir lcance _,

8111t11 Date Stutus Staftt1!111 Cu11Jltlon

• Cai:ccoiry Huwbur r'!

l RO 76-1 110120 1/10 l/21 c p P,JJ !In x P,l II RO 76-2 110950 1/16 2/5 c I' AA H,T c ti II I

II KO 76-l 111642 2/9 2/25 c S,M 11,tlH. ,,. c Y,Pll B C4"

. 'I 110 76-4 ll 1!107 2/21 . l/10 c l,M H,T c ti D H- i.*

RO 76-5 CANCl\l.l.IID RO 76-6 111906 c l/4 l/10 v.1m J,11 IC l' II Cl1,C1 }

110 76-7 112110 l/6 l/2'J c L,T c 110 76-11 111951 l/28 5/7 c "1* 00 c ti H,llB ll

,j RO 76-8 111291 4/8 4/22 c DD JJ,00 B 11,S,M II Cl" 110 76-9 11018 4/19 5/14 c IC l' c ... G H I KO 76-10 RO 76-12 lllil79 114180 4/21 4/10 5/20 5/14' c

c x

x P,llll 00,11 c

c F,P x

G D Cit" llO 76-1:1 l l407'J 5/1 5/21 D II I< ,tlH A C,H 8 Cl1,C7 I

110 76-14 l lli640 5/10 5/'J.5 I x,o 11,110 A T,U A cu

'JI KO 76-l5 no 76-J6 CAHCEl.LED CAHCt:l.l.ED bt KO 76-11 lllU96 5/20 9/U p RO 76-18 CANl!ELl.t:ll

" ill Q D

" w I

ln RO 76-19 CANCELl.Ell RO 76-20 CAHCt:U.l!D 110 76-21 116266 6/22 1/22 p ~ 00 v D 110 76-22 CANCEi.i.ED "

I

,I ...

I

~

j l1

/

f'*

i

• j ~.

[*,'

L*.

t. ~

,J ,,. ~*

L:

Table B2.7 (Contl11ucd) i..

NS!C l'hmt llimu r111ia ft S!gn& I Acance IEV*mlt IRe!'ort COBUftllll!ID&alt NU111be1r Accession Date.

System IE13u lp1ae11 ~ Il111s1tirumen~ Csuse Dete Status SU1tu11 Cond!t!o111 Ca&egory lfu111ber RO 16-23  !.17660 7/20 6/!9 B II DD,lll A,li p G Si RO 76-24 1!6886 7/JO 8/13 8 v.z  !,DD 8 s II H  :*

RO 76-25 117968 8/15 9/10 B II EE,V c Q D if llO 76-26 118197 8/11 8/Jl 8 D K c G II C4,C7 RO 76-27 117661 8/23 9/7 B J,J oo,rrr c !NI l!I C4 RO 76-28 117682 8/26 9/9 D v JI /) DB D Cl RO 76-30 118199 8/26 9/2! D v J,X llll !18 llli C7 KO 76-29 118198 8/21 9/21 Fil v JoX .II BB Ill C7 RO 76-3! 118200 9/!l 9/21 II v Jl. IFIF la !ll D llt RO 76-32 CANCELLED RO 76-33 119722 10/20 11/J D 0 C,IF Iii Q D JM RO 76-34 119759 10/20 11/11 B D K1 ilD IB c 8 C4,C7 RO 76-35 !19760 10/26 11/11 B D IJ{ m c D C4 1 C7 RO 76-36 U9786 !1/5 U/19 9 E r.

no 76-l7 120273 11/15 12/U p w " 11.;r IC IC H;ll',!JJ ts D H

RO 76-38 120475 11/17 12/!7 B D  !IC Ill c D "

C4,C7 bj RO 76-39 120475' 11/25 12/!7 II D  !(. la c D C4,C1 I w

RO 76-40 CJ\

120475 11/30 12/17 II D c C4,C7 RO 76-41 CANCELLED

!IC Ii 8

I RO 76-42 120'i15 12/1 12/U D D IK 1lll c llll Cl1,C7 RO 76-43 12/2 12/JO Ill x z Jm,C !NI. a: D C4 RO 76-44 121026 !2/7 1/4/n Ill w X, l! IC H A,D ~

RO 76-45 120661 !2/9 !2/211. II O,P o.ll" c 11.,H,S G ca RO 76-46 121665 12/25 A/ll/71 Ill ll>D oo.JJ Ill C,R !8 N l1:

j II l

)

I

• Nlwbair H~IC Accc:rm lun

'H11u1lu:r t:vc:nlt 1>11tu

'&'able 02.u ttea*on Onto l'la11t Stni:ia~ lilfBh!UI E*aul111~e11t h1 H rwaen t Cu1111111mm1t llg11~m~

Ab11or*11l Condition Can so Sl1111U lc:11nce _

Ci1t1~1~11.-y

.)\

- f'<

-"f I

t*

u:u-11-15 1229114 6/21/16 2/2] II p It - 11* c B C4,C:1 i l.t:k-11-A 121510 l/l 1/11 u v J,FI' - II H D t:1 I

l U:R-17-2 IZ10'J4 1/6 l/21 B z 11,HH - in C0 N D H

.. I J

l l.t:tt-17-8 1215!6 1/1 2/2 u v I - B s II II .

I U:R-17-U 122901 1/1 2/22 B D K,l>I> - Ill ll u C4,CI l 2/8 I

iI um-11-4 !UOO !Ill l/24 I> D K,Db - ii G D C4,C7 i Lt:ll-71-6 C1111ca lied I LER-11-1 l306!H l/U llll p II K,llD - ~ c 8 C4,1:1  ;

I LER-11-9 122208 1/10 Z/9 B p K,IJD - !i c I C4,C1 f i l.l!ll-71-U 121055 l/lO 4/5 II z 11,tlH - II C,H,Z b H

.I l LER-17-10 12'1411 1/11 2,14 D ti G - ifj I B N 1*

l.t:R-17-IZ 122410 2/5 2/22 B A v.JJ - I D.

,J II um-11-16 122!11>5 2/10 2/2l D D K,llU - $ "

G B H

C4,C1

• l l.l!R-11-11

• 12109) 2/10 l/21 u z II ,tut - 8B C,H D II um-11-1a 122919 2/11 l/11 D o,*r H,W,Z - I!! E,H p H

• I U:R-11-19 124094 2/19 l/21 ti Bii 1*r - c p H b:I r u:a-11-21 125025 4/1 4/ZO D D K,UO - I u B -C4,C1 w I

I LER-11-22 Jl2121 4/0 4/20 D D K,blt - fl u B c4*,c1 I

I I LER-11-2] U5045 4/19 5/4 D llD JJ.oo - IB B,N,X II CJ I LEll-17-24 1251114 4/2l 5/2l B v - c,o c* u D H l LF.ll-11-25 Ll!ll-11-26 125404 C1111cellei!

5/19 6/10 0 A,H r, 11 A Q,i*,u A N II  ;*,-

.i I

i

' t***

• • ,t . ~-

~

I I

J . "

I

• I t i!

Table 112.0 (Cont i1111eJ)

\

NSIC  !*

Event 11\eport ll'!ant Corupo111entt AblJ!llllr!DO! sggnlikrnnce Nwnber Acceee!o111 Syate111 Ef!u!pmcnlt lnllllrll!DCRt Co1uii&1oB'il Canuse Cut.,gory Dote Dote Status Stall:llllil Number LER-71-3!. 125595 S/26 6/24 B A,S 00  !! N (l N LER-77-29 126098 6/8 7/8 B L A ll IC 0,1111 II! N 1.ER-77-27 126097 6/10 1/1 D AA FF,GG,IKK IC N D N LER-77-28 126114 r.1u 1/U B x 00,QQ ll' c Q llll N l.ER-77-l2 127916 1/l Bl! B w ft' c ii D N uR-n-n 127917. 714 8/A B L A c c,v Kl! N LER-77-14 1279!8 7/8 8/l B L A,OO c O,N Bli N LER-77-35 126976 1/9 1/26 B D,BB C,C,S,AA Ill N LER-71-36 127919 7/19 BIA Ill IJ. ll ~ 'll ill 1111 LE!l-17-37 Cance!led LER-77-38 128514 8/U 8/26 !I n,JJ 00 c N,Blll l!ll ti

!.ER-77-Jll l!.28SJ5 8/16 11/lO D IE lFIF !IS 1111 Ill N LE!l-71-48 130088 8/!6 10/H IE c 1111 II>> N

!.ER-71-42 129869 8/17/ 9/l'J 11>

D "'w ir. K tC Q !il l 1l

!.EK-77-40 128536 8/18 8/30 ll!l I!. A l! !J,A. 0 I!) N LER-77-41 128939 8/U 9/2 Jl) 0 IN ,JJJJ c 0 A N LF.R-77-U 129868 8/24 9/H Jl)  !.'  ! c L D N tJj a

LER-71-44 129839 9/8 9/22 a v IC ~ S,'ll' A N w CX>

LER-77-45 130027 9/!6 9/lO II 1;r IFF,00 Iii N II S6,S7 DU c D l!ll 1.ER-71-46

• LEK-77-47 130087

!]0119 9/20 9/24 10/18 10/18 B

D H PP G Ill Q F,R "C7

_j

• ~~-***

I 1.

I I_

Il

)

- H1uabu_r um-11-u HSIC Acccuslo11 Huiubcr 110!H5 t:V-111&

Pate

'>/21 Rc1*1tl"E 111110 11/I r11111&

litutuu sy~tii*

0 l!11ul11111&!11<< 11111 t If lllllUll l CCDlllJIO!!ll!!lft. -

SUU11s Ab1111 na;1 ~

Co11dltlun R,s*

Cuuuo-G SI 1111 I ( lc:i111"e Ciuc11or1

-1 I

l LER-71-50 l.ER-71-52 110922 111102 10/18 Ul/26 11/16 11 /:ll D

D I ,.I 0

00 c

A "s D II II

" '~ .

tt,*r 0  !!

I l

I um-11-n U::R-11-54 131165 ll2956 ll/1 11/21 12/'J 12/21 0

I FF JJ R

H D I c b H

l 1

I

.I um-11-55 132941 ll/25 12/16 I }I G I q D C1

I i u:n-17-56 112959 11/29 12/12 D A,H G,Z I H 0 1),AA D c:l,C7 l LER-11-64 1Jl701 12/1 1/6/10 0 *x DD A s.T A,11

• 1.

i l.ER-11-51 132951 12/5 12/21 u w H c* ti D

\* um-11-62 JJ 1 rr c:

l.ER-11-58 lll100 112958 12/1 12/11 1/6/18 12/21 D

B H K

G -

A I

"Q' D "C1 l.EH-17.,;61 lll621 12/22 1/20/18 B K 00,CIQ c .. - G-('

x 00,QQ Ll!R-71-65 114042 12/21 2/1/78 D T IC Q D j td  ; :.

J, I I:

w I l l.O ' I ' .

i

--1 i

j i

I I

II i

I i

I I ~

I I

I lI l

• • • • • ~----*

'!'able 02.9 Cod Ille Sheet for. lleportn,blc Evenls for ~alleades - 1978 HS1C litepoll'l &'I ant Ct1J111gionen1t Abn11nnmi Evcmlt System 1E11uipn11mtt in.l!UWD~nll:

S.ign ~ff !cancc HU111i>ar Acee salon ICause Date Pate Status SUl!:us Co11dB.1tAou tm~CUCJll'lf Humber Ll!R 78-2 "UJ625i i/6 ll/20 c AA oo.QQ H IC iii D H LER 78*1 U3626 i/1 i/20 c E IFIF Ii\ IM llli Iii LER 78-l 134274 1/8 211 c s,cc 8 0 00,QQI II o.s 18 H.

LER 78-4 1J6344 llll J/l c M llt! II: ~ D t'1 LER 78-5' 136457 212 3/2 c w ll IC H II) H LER 78-6 !36389 2/14 3/2 c I' I!. Iii s a I.EK 78-7 136995 3/ll 3/27 c ff c.rr IC H E "

N LER 78-8 !37506 4/5 4/!9 c FF c c 61 A N l.F.R 78-12 138292 4/14 5/12 c w IE IC L ~ H LER 78-9 138230 4/U 4/28 c v H Ii( IC lij,S ~ H LER 78-11 137863 4/17 5/8 c A,X IDD 0.F Ill Q ID N LER 70-14 138924 4/19 5/19 IC H,CC O,GG,KK It: s I!

LER 78-10 138229 4/20 5/4 B ,, JI Iii 118 D ill Cl LER 78-lJ 138257 4/21 5/!5 D 1' ,1.1 00 Kil s 18 SB LER 7e.,1s 4/28 a rr LER 78-16 ll892S J.]8368 5/9 5/19 5/22 D ,,BB JI I

Ill A 0 11li Ii' Bli D

~

C7 bS LER 70-17 ll8320 5/9 5/23 D l.,AA 00 $ s -11 H

~

i U:ll 78-18 ll9661 5/23 6/22 p H IF T IBI Q I!) ll'1 0 LER 18-21 141035 5/26 6/26 B Q z Ill IE Ob N LER 78-22 ll9842 6/4 1/l B A z,c.K!C. i&.z Ill D i

l!

~

.:~"'; ..**

I

.J j

'*M***

.-~-~.-- - --~. -* *'

~ ..... '"**-***** .. -~--***

~

NUC I' l1111t Coiu11011111u.

• Abnnriaa I Slc11I r l1:a11cc:

• _

Event llu11oct !netruiaent Couse J N11111bur Accu11ol1111 l>>ate Slutun srutu1a E*111 l11*cn t StotMe Condit Ion Cat e11ory P*ila N111~hur 1.a::a 78-19 ll!16l7 6/S 6/20 u v J II z D C1 I I

I.HR 711-20. 141014 6/ll* 6/Z6 D v c .!! u A H I.HR 18-24 ll!.1841 6/21 1/20 B u .H II c c C4 I

LER 18-25 1J!l844 6/21 1/20 II A V,J.I .!! D II um 111-21 ll!1841 111 7/14 II S 0AA 11,HH II s,x II " ..

z c I'I 1.F.R 78-26 I.ER 78-27 11!1912 140279 111 1/!0 8/4 8/11 D

B q

t'F K I T

c A

D H.

H I

I I.ER 18-20 140188 8/4 8/18 p 1,J,'r oo,n* c G I I.ER 78-29 140226 8/1 9/6 p K JJ,00 H

s c "

I LER 18-lO 140249 8/28 9/11 p 1.J 00 R>1 II . N,S c "H ."

.I LKR 78-ll 140687 9/11 10/2 B v J,t'F .!! N,11 p t:7 I

u:a 78-12 141491 10/l!I 11/2 8 A c c,v I I.Ell 711-15 141681 10/20 11/16 c

A lI u:a 10-16 141689 10/21 11/16 8

D "Q z H

c ti s

D c

• ff H

l  ;' .*

l 146517 1/25/19  :-

! LER 18-ll 141491 111/25 11/l .8 am . JJ ,uo . ti I H,H,B D CJ

.-:1 LEK 78-34 141416 10/26 11/9 B D,AA CG,KK c A*,*u D II gs ...

.I l I I.EK 78-17 I.Ell 78-18 141961 1422!18 10/ll 11/6 11/lO 12/6 8

B o,x Q

H,PD Fl' A

A l',H,S 1'

II .

A co H

~-

l II I.ER 78-19 142101 11/JO 12/J 8 z 00 Ill C,H A H-I I.ER 78-40 145198 U/28 1/12/19 8 DP JJ,(10 Ill 11,H II Cl

Tablo h2.9 (f!unt!nuud)

NSIC Event Rc11orl !'!ant Com11omm~ l\b11on10l Sign! r ilo.:on1co Numbcn: Acccsslon System IE*! u gpm1111 t ineltl!"Wlll!nt C1111!lfl!E Dote Dote Statue Status Cond!t 8tt1S11 CaKeguiry l!lumbar u:a 78-42 145261 12/9 1/9/19 p z A,C R,S sn LER 711-41 142709 !2/!4 12/26 D H,N,X al B S6,C4 N

ILER 78-44 146628 12/IS l/18/79 8 IJ K,P llJ G  !>> N LER 78-43 146528 12/Jl 1/16/79 8 v <<: s Ill N

. bl i

.t-N

. 1,'

~--

-~******

\;

I l

I.

I:

muc Evt*nt R"a*orl rlunt C1111gu111a!11t Al11111ru111I Sla:nU Ar1111c11 I:

Huaaboir Accr.11111011 9Jl!ll!lll £1111 i g111u*11 t !1111t l'llUH!ll l: Cnuuo 1* H11111her Pate Pule Slalue Sl11t11H Cm1llhh111 C11lt:Kur1 I.Ell 19-1 1461J1 1/2 2/1 II N l.,P <<: s II CB.

i I I.Ell 19-2 146148  !/l 2/1 D z II !I C,H D H

! LER 19-5 151009 l/l 2/2 u 0 H,HH <<.: u.z *o N i

i' U:R 19-1 141230 l/U 2/11 u v c M u u H I.ER 19-J 146140 1/11 l/ll u z,am I E,H,X

• -1I . I.ER 79-4 lli61J9 1/11 l/Jl B

'I B,BB . 110' A s II CB" *,.

I.ER 19-6 141215 2/8 c Ll-:ll 79-8 141290 1/25 1/26 2/21 B

D E

L,T ff c,00,1111 H

Ii ,,

N,T A I>

H H

I.EK 79-lZ 146B02 2/1 11/l D s,u DD *Ill s II I.ER 19-10 148102 2/1 l/6 B A,DD 00 B, N D "N

LER 19-9 148101 2/20 J/6 B l,J 00 c u B *c1, LER 19-11 14B558 2/21 l/26 B V,FF A,C 1 F1 P !I Q D II I.ER 79-14 140180 l/U 4/12 B v c I u D H LEll 79-U 1401104 l/21 4/9 B 1. A,t'f ,00 A0 B C,H,O u N 149190 5/15 I.ER J!J-19 149221 l/ll . 5/10 B l 00 R G I.EK 79-11 105112 le/4 5/4 B H c c 118 lt 0 T A " ,.

LEK 79-16 149225 4/6 4/10 p 0 I' IC Q I> " us I f LEK 19-U 149226 4/9 4/2J B R CG 0 1CK I[: x II "

C4

.I:-

w i*

UR 1'l-J8 149218 4/9 5/J D A z A s II H I.ER 79-22 1501168 4/26 5/25 p AA l'l',GG,ICK ii: .A,H,O p i.

" i j "'

.l i

I i:

rI I

1 .*,,

.j

. .1 iI J ,, t'-

Table oa.10 (Conlt!nued)

NSlC.  !.

Accosslon A!:vent RCftlOl!'lt r!1m1t System COB!l!JOB!Olllllt Ab11111!11111A Sign! Ucllnce N11111ber IF.~u.l!p111ent !n&ltl!'IMDC!llft C!!luse Pate Dote Statue Stn1tu11 Condl&ltiollll Cale.gorir Number

. -~  !.Ell 79-21 150069 4/27 5/25 B N,CC IF!Y ,CG 0 KK c A,N,O p M LER 79-20 !49884 5/1 5/15 D L GG,KK IC x B C4 LER 79-23 150266 5/18 6/lf! p A z Ill N,CC IE N

• LER 19-26 150170 6/4 7/6 B z i1 0 H,HH B IC,N D C7

!.ER 79-24 150765 6/8 6/25. B L A l!. mi N,O p N LER 79-27 150726 6/14 7/12 I! z 11,HH IB C,N I> IC7 LER 79-25 150166 6/19 7/3 B B U,tlH *c IF D N LER 19-28 152663 6/2l 1/24 I! z 11.m1 13 C,N ID Ci

!..ER 79-29 150897 6/28 7/26 B z ii 0 t1H 1111 IC,N 8) C7 LER 79-31 150!100 6/28 8/20 B A,llD 00 A' 8 0 111 G C)

!..ER 79-30 150902 7/2 7/31 II z 11,tlH I c Kb IC7

! I I.ER 79-lZ 151132 7/2S 8/14 B z Dl,HH Iii C,lll 10 C7

'. LER 79-44 7/29 1/21/80 B I' 00 c 'I!' A ca I

LER 79-31 151247 7/JO 8/16 B R IGG,KK c x IS: N

'f . LER 79-J/1 151414 8/4 8/ll a.

,I LER 79-36 152223 8/21 9/26 II ii "

0 RI!

C,IMl,li B IC ll",ft' G

G N

N I.ER 79-JB 151944 8/29 9/28 B Bii Pl' ... Ill i) i) N ,.

~

~

Ll!R 79-19 151941 8/29 9/28 II\ z D!,H,HM 1111 C,N b 4:7 ~

LER 79-35 151763 9/5 9/U IS DD 111.,S,AA G IC]

LER 79-:n> 152220 IJ/U 9/28 I> i 00 111,C 'll' A H,S2,S8 f :*.

II

ltwaber NSIC Accouulon Nu111ber r.ven&

Put11 Reg111rl:

Pata 10/12 A'I nut lit111*u11 z z Cu*g11111un ~

S~BltUll c

Ah1101*:il Condit Ion E

Caus11 p

Slc11H-l*::111co:

Cuaegory LER 79-40 152652 9/27 IJ c s V,S G "

I.ER 1!.Hl LEll 19-41 152990 152762 10/2 I0/9 11/U

.l0/10 c z Dll,00 c 1' 8 "

10/12 10/26 c s H p l.l!R 79-42 152980 CC,00,Qll lC i.

• l i

j J*

i l

f t'

!\

bS I

-~*

- lJ'I L

i

,,.