ML20059C987

Intervenor Exhibit I-MFP-124,consisting of Technical Review Group Meeting Minutes Distribution, & 920124 DCI-91-TI-N047, Reactor Trip Due to Personnel Error & Safety Injection Due to Leaking Steam Dump Valves

ML20059C987
Person / Time
Site:	Diablo Canyon
Issue date:	08/21/1993
From:	Hug M AFFILIATION NOT ASSIGNED
To:
References
OLA-2-I-MFP-124, NUDOCS 9401060294
Download: ML20059C987 (19)

Text

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'"  %$p F/'2//93 DIABLO CANYON POWER PLANT

'93 00: 28 P5:48 TECHNICAL REVIEW GROUP MEETING MINUTES DISTRIliOYI65T 28 PS :48 NCR NUMBER: DC1-91-TI-N047 TITLE: N-42/U1ReactorTdp'- h THE FINAL MINUTES OF THIS NCR ARE ATTACHED. THE CORRECTIVE ACTIONS HAVE BEEN IDENTIFIED AND ARE BEING SENT TO PSRC FOR REVIEW AND APPROVAL.

GENERAL DISTRIBUTION:

NPG G M Rueger 77 Beale Rm 1451 NOS J M Giscion 77 Beale Rm 1459 OA J A Sexton 77 Beale Rm 1800 NSOC R L Russell 333 Market Rm A1112 ENG QC M J Jacobson 333 Market Rm A1408 NECS R C Anderson 333 Market Rm A1411 NECS M R Tresler 333 Market Rm A1409 NRA K A Hubbard 333 Market Rm A6017 DCPP B W Giffin DCPP P W Bedesen 104/6/24A DCPP T J Martin DCPP R K Rhodes FF3 DCPP D B Miklush DCPP R Bell EE5 DCPP NRC DCPP J Shoulders .AA5 DCPP OSRG DCPP D Blum AA2 DCPP OC DCPP R Exner 104/4/11A DCPP QA DCPP K Oliver 104/3/328

• VOTING MEMBERS:

W Crockett 104/5/521 D Clifton 119/2/247 S Szuch 104/3/33A J Bonner 104/3/1A -

G Voboril AA2 Mut2 MARTIN T. HUG Regulatory Compliance cc: PSRC Secretary LER 1-91-009-00 S:\REGCOM\ FORM \FINALMIN.LST-9401060294 930821 h ADDCK 05000275 PDR O PDR A

• -9AP-15.3 '36-286 (3/89) Itzvis ion: 10/25,/90 ATIAO9tDff A Fa e 1 af 3 MNKDNF09tMANCE REPORT El active Date 12/31/90 -

1. 1. ** "* ##*I 1'

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3. Ita=IActivity N-42/U1 Reactor Trip Reference A0231244

5. Description of konconfessances j Unit 1 Reactor Trip due to an I &.C Technician inadvertently manipulatina a j power fuse on an in service nuclear instrument channel (N 4?). This action 1 is considered a nonconformance by manacement in accordance with OAP 15.B.

T paragraph 2.1.7.

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9. DegC' p k f /s) 10. Organts.ation 7 Organiution 8. Date

{In ggp/s) M6t.Svts. 5/17/91 F. aCross I& C 5/17/91 [/ Cfffffri L A 12. McN Evaluation Attached on t.he111. NCR TextofContindtion Cause Problee - Sheets V. J Corrective Actions -

m 1. Plant conditions !Y. Analysis of Problon T1. Additional Information A II. i=scription of Probles L 16. Estimated 15. Responsible T Y ,

Other Coerple i Organitation

13. Trend o.te v E

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1 Cooe. Till-1 A131 11-11' i1-11I N/A 92

-3/01/E.,% Instrument Contro' s H s

17. Potentially 10CTR21 18. 10CTR50.9 19. Reference other 1 R 16. 10CrR50.73 Reportable Reportable Reportable, if app.

Report.able C

A E

P Yes [)() No l) Yes [ ] No IX) Yes l1 No 1XI N/A L 0 R 20. Easis Refer to Attachment 7 23. Notified By 26 Time 25. Date .

A Initial 21. Time Limit 22. f%ethod N/A N/A N/A N/A

• B Report N/A E I ,28. LER No. 29. Date V L Followp 26. Required P. Time Limit 1 Report YesIXl No i 1 30 Days 1-91-009-00 06/17/91 E T W Y 10. Other Agencies Notified None it. Remarks IR o 60. Other s) 4 Date U

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32. Chairman (pfs)//

WGCrockett ud f'f,,,f/

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35. Date 62. Other(p/s,*f &. Date

36. QA (p/s) /

P R SNSzuch M/flut b .7 / / f N 2 DRClifton (Ab[-l/[) , 3!/f/91

37. Date 44 s) (, 45. Date V 36. Qc (p/s g Othorp1/88/92

[ JEBonner, e i .

3 -/f M OT i . s 38 P NCRs 3. t h 46. Other (p/s) 67. Date

49. GOhfAAC hottfaratson Dates F5RC p. Meeting Dates Reviev .

Corrective 50. Cosplete - TRC Dairman (p/s) 51. Date 52. QA Verification by (p/s) 53. Date Action r til s t ri t.ut ion Materials PSRC Secreta'ry NPC Manager, QA Station / Hydro Construction Initiator Flant Manager, DCFP TES Appropriate QC GONPRAC Secretary Authorized Inspectore Ot.her Engineering if applicable other L-_-_-_-_-_-_________ _ ___ ______ _ _ _ _ _ _

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-5 DC1-91-TI-N047 D4 January 24, 1992 i

REACTOR TRIP DUE TO PERSONNEL ERROR AND SAFETY "

INJECTION DUE TO LEAKING STEAM DUMP VALVES i

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I MANAGEMENT

SUMMARY

i I

A Unit i reactor trip occurred on May 17, 1991'at 0624'PDT, as a result of personnel error. An instrument fuse'for NI channel N42 was removed when STP I-2D, " Nuclear Power Range .

Incore/Excore Calibration," was being performed on: channel l This satisfied the protective logic for a two out of ~

N41.

four high flux at high power (;t109%) reactor trip. Two condenser 40% steam dump valves failed to close after.

actuation, causing a cooldown of the RCS. One minute.and 38 seconds after the reactor trip a two out of four low I pressurizer pressure ($1850 psig) caused a safety injection.

At 0627 PDT, At 0625 PDT, an Unusual Event was declared.

~

operators closed the MSIVs and MSIV bypass-valves-terminating the cooldown. The RCS reached.1730 psig and 465 T ,during the transient. At 0633 PDT a one-hour emergency report was made to the NRC in accordance with 10 CFR 50.72 (a) (1) (1) . At 0800 PDT. Unit 1 was-stable in Mode 3 at NOP, NOT.

91NCRWP\91 TIN 047.JCN Page 1 of 16 [

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DC1-91-TI-N047 D4 .

January 24, 1992 REACTOR TRIP DUE TO PERSONNEL ERROR AND SAFETY INJECTION DUE TO LEAKING STEAM DUMP VALVES  ;

1 I. Plant Conditions Unit 1 was in Mode 1 (Power Operation) at 100% power.

II. Descrintion of Event A. Event:'

On May 17, 1991, at 0620 PDT, two'IEC technicians l were performing surveillance test procedure (STP)

I-2D, " Nuclear-Power Range Incore/Excore Calibration," on nuclear jnstrumentation power 4 range channel N41. The next step (8.11) in the )

procedure required removal of the fuse for reconnection of that channel's signal and high l voltage cables. One technician went to the ,

instrument drawer and inadvertently pulled.the j fuse for NI channel N42 (ref. 1); This resulted '

in the protection logic recognizing a two out of four high flux at high power signal coincidence for a reactor trip (ref. 2).

On May 17, 1991, at 0624 PDT with Unit l'at 100 percent power, a Unit i reactor trip and main  !

turbine trip occurred due to the protective system two out of four high flux at high power signal coincidence. Following the reactor and main turbine trips, the condenser steam dump valves 1 (SDV) automatically opened to prevent reactor coolant system (RCS) pressure and temperature increase. RCS-pressure and temperatur. decreased and, to close the SDVs and mitigate the cooldown, a close signal was initiated. Two SDVs, 1-PCV-1 and 1-PCV-11, did not close following the close ,

signal.

Control room operators entered-emergency procedure E-0, " Reactor Trip or Safety Injection" and ,

verified proper initiation of the reactor trip.

One minute and 38 seconds after the reactor trip, RCS pressure and temperature had decreased sufficiently to result in a safety injection (SI);

91NCRWP\91 TIN 047.JCN Page 2 of 16

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e' DCl-91-TI-N047 D4 i

January 24, 1992 l

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) on two out of four low pressurizer pressure

] (51850 psig) signal coincidence (ref. 2). l l .

l j An Unusual Event was declared.at;0625 PDT on May 4 1

17, 1991.. Due to previous experience with SDVs' , l j failing to close.following actuation, operators

quickly identified the malfunction. At 0627 PDT'  !

i on May 17, 1991, operators manually isolated the1 i leaking SDVs and~ terminated the cooldown by.

1- closing the_ main steam isolation valves . (MSIV)' and j MSIV bypass valves.

i

}

A one-hour emergency report required byL 01 CFR-1 50.72 (a) (1) (i) was made on May 17, 1991 at 0633 ,

i- PDT'(ref. 3). .This report indicated that'RCS pressure dropped to 1640 psig'andMr.,droppedLto.

j j 507 degrees. Subsequent analysis of recorded' data' .

l determined that RCS pressure and temperature had. i reached as low as 1730 psig and 465 degrees T ,.

On May 17, 1991 at 0800 PDT, Operators returned

Unit l'to' normal operating pressure and .

i temperature (NOP, NOT) ~ and stabilized the Unit in Mode 3 (Hot Standby) (ref. 2).

B. Inoperable Structures, Components, or Systems that  ;

Contributed to the Event

~

j None.

i j C. Dates and Approximate Times for Major Occurrences: L  ;

j 1. May 17, 1991; 0624 PDT: Event / discovery date.

j Unit 1 reactor-trips i when a second power 1 range NI.is j inadvertently. -i j .deenergized by an IEC  ;

} technician.

] 2. May 17, 1991;'0625.PDT:'An SI occurs.due'to i RCS pressure:

! decreasing because (';

two'SDVs leaking'

excessively.following i

automatic actuation. 1 s

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I DC1-91-TI-N047 D4 January 24, 1992 l

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3. May 17, 1991; 0627 PDT: Operators manually close the MSIV and MSIV bypass valves to isolate the SDvs.

4. May 17, 1991; 0633 PDT: A one-hour emergency report required by 10 CFR 50.72 (a) (1)_(i) was made.

5. May 17, 1991; 0800 PDT: Unit 1 was stabilized' in Mode 3 at NOP and NOT.

D. Other Systems or Secondary Functions Affected:

All equipment, other than the two SDvs, functioned as intended to stabilize Unit 1 in Mode 3.

I Valves 1-PCV-1~and 1-PCV-11 opened per design but, following closure signal, failed to'close. It was

! determined that the inner plug stem of the valves I separated from the main stem, and resulted in l

' excessive leakage through the valves. The valve l stem separation was the result of microwalding of l the valve seat (ref. 4). l r

E. Method of Discovery: l The event was immediately apparent to plant .

! operators due to alarms and indications received in the control room.

! F. Operators Actions:

Operators closed the MSIV and MSIV bypass valves.

G. Safety System Responses:

1. The reactor trip breakers opened.

2. The control rod drive mechanism allowed the control rods to drop into the reactor.

3. The main turbine tripped.

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- DC1-91-TI-N047. D4 January 24, 1992

4. An SI signal was initiated on low pressurizer pressure.

5. The SI pumps started.

6. .The residual heat removal pumps started.

7. The charging pumps started.

8. The motor-driven auxiliary feedwater;(AFW) pumps started per design.

9. Diesel Generators (DG) 1-1, 1-2 and 1-3.

started, and per design,-did not load. .

III. Cause of the Event A. Immediate Cause:

An IEC technician, while performing surveillance testing on NI power range. channel N41, inadvertently removed a fuse in NI power range channel N42.

B. Determination of Cause:

1. Human Factors:

a. Communications: lack of communications between the technicians was not a contributing y factor in'this event. ,

1

b. Procedures: ' procedures were. reviewed and i found adequate as they ,

contained sufficient I instructions to safely perform the-task.

c. Training: training was' reviewed and

. determined not a factor. All' I&C technicians have received  !

training regarding the. l requirements of independent i and self-verification.

d.  : Human Factors: personnel' error, failure 91NCRWP\91 TIN 047.JCN Page 5 of 16 4

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. DC1-91-TI-N047 D4 .

j January 24, 1992 l

to perform self-verification, was-determined te be the cause 4

of this event. Subsequent to the event a detailed human factors review of I the NI cabinets was performed (ref. 1). '

Several recommendations l made by this review will

! be implemented to preclude recurrence of this event.

j e. Management System: management systems

, were reviewed and

-determined to be

! adequate. Policies

and procedures are in place emphasizing the i importance of self-i verification. These policies and- .

procedures are being .

reinforced through the J

positive discipline, program. 8

2. Equipment / Material:

a. Material Degradation

N/A.

. b. Design: N/A.

1

c. Installation: N/A.

I d. Manufacturing: N/A.

i j e. Preventive Maintenance: N/A.

f. Testing: N/A.

g. End-of-life failure: N/A.

i C. Root Cause:

The root cause was determined to be personnel error (cognitive) in that the I&C technicians did ,

91NCRWP\91 TIN 047.JCN Page 6 of 16

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1 DC1-91-TI-N047 D4 ' l

! January' 24, 1992 not perform-self-verification.- I&C " Policy For  !

Unit / Channel / Component Self-verification," dated' June 30, 1988,. requires 1that an individual! verify >

his own action as correct prior.to performing the-action.

IV. Analysis of the Event ~

A. Safety Analysis:

1. Reactor Trip and' Safety Injection:

Accidental.depressurization~'of the~-RCS is . ,

identified-in the Final Safety Analysis Report:  ;

(FSAR) Update as a Condition.II Event - Faults of. Moderate Frequency'. .The effects of this event are analyzed in Section 15.2.12. . The ,

results of this analysis assure the'ainimum . i departure from nucleate boiling ratio (DNBR) remains'in excess of 1.30 for this event.. The event reported in this NCR was~ bounded lby the-l FSAR analysis.-

2. Overcooling:

A Westinghouse engineering evaluation (ref.'5)-

of the RCS considered the impact of;the .

thermal transient:upon the pressurizer, ~!

reactor vessel, RCS piping, the thick metal of.

the steam generators,1 and 12ut reactor coolant ,

pumps. ' Westinghouse' reviewed < temperature and' pressure data and compared these with i evaluations of'similar transients at other plants. The comparison showed that the DCPP  ;

Unit i rapid cooldown event:is-bounded by~  !

transients previously analyzed-for:other '.

plants. The Westinghouse evaluation concluded .

that the above described-DCPP-Unit 1. transient i did not adversely affect the structural -!

integrity'of the affected components and  !

system, and that the RCS could be returned'to normal operating temperature and pressure'and' .

the unit restarted safely. ,

B. Reportability:

1. Reviewed under QAP-15.B and' determined:to be l 91NCRWP\91 TIN 047.JCN Page 7 of 16 i

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l DC1-91-TI-N047. -D4 3 '1 January 24, 1992 I l non-conforming in accordance with Section 2.1.2.

2. Revic.wed under 10 CFR 50.72 and 10 CFR 50.73 per NUREG 1022.and determined to be reportable- 1 in accordance with 10 CFR 50.73 (a) (2) (iv) . l

3. This problem does not. require a 10'CFR 21 l j report. '

I i

4. This problem does not require reporting via:an ' '

i INPO Nuclear Network entry.- '

5. Reviewed under 10 CFR 50.9.and determined to be not reportable since/this event.does not have a significant implicationifor public health and safety or common defense and security. I

6. Reviewed under-the criteria of AP C-22 requiring the issue and approval of a JCO and determined that no JCO is required.

V. Corrective Actions H A. Immediate Corrective Actions: 1

1. The plant was stabilized in Mode 3.

l

, 2. An immediate, temporary-IEC work stoppage.was directed by management. LIEC personnel ware l

tailboarded on the necessity of self-verification.

1 B. Investigative Actions:

Investigate methods of evaluating self verification and/or concurrent verification during l performance of maintenance activities;and surveillance testing.

RESPONSIBILITY: J. Bonner Return l QC (PQCE)

AR A0231287, AE # 12 '

C. Corrective Actions to~ Prevent Recurrence:

91NCRWP\91 TIN 047.JCN Page 8 of 16-l

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i G DC1-91-TI-N047' D4.

i 1

f January 24, 1992 l.

f- 1. A temporary physical. barrier has been installed-over the NI drawer faceito prevent l

' inadvertent operation of. switches or. fuse l i

l activity. .This corrective action also tracked-as action 33 of Event Response Plan'91-6. MR

'A0231687 describes.the physical barrier.

I i

i RESPONSIBILITY: J. McCann . Complete f

j. .I&C'(PGIT) i AR A0231287,WAE #102 Not outage.related..

4 l

Not JCO related.. l An NRC-commitment.in LER 1-91-009-00.

l Not a CMD commitment..

2. A memorandum has'been sent to~ Shift Control' l Technicians informing them of.the new-1 4

temporary physical. barriers ^and their use.

4

! RESPONSIBILITY: W. .Crockettl Complete-

! IEC (PGIT) l

'AR A0231287, AE # 09 i Not outage related.

i Not JCO related..

i An NRC commitment.in LER 1-91-009-00.-

j Not a CMD commitment.:

i j 3. A design change will be implemented to instal 1~

a permanent,. removable physical barrier design R j for the NI drawers.

4

!. RESPONSIBILITY: F..Lacross: Complete j i IEC Planning _(PGIS) l 3

AR A0231287, AE # 03 Not. outage related.. l

< Not JCO related.

j An NRC commitment in LER- 1-91-009-00.

4 Not a CMD' commitment.

i Self-verification ~will be integrated into IEC' l

4.

laboratory training.

4 5

i RESPONSIBILITY: D. Clifton- Complete I Training (PATT) 1- AR-A0231287, AE-# 07 i Not outage related..

3 Not JCO related.

j 91NCRWP\91 TIN 047.JCN Page 9 of 16 i

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'DC1-91-TI-N047 - D4

January 24~, 1992 i

I.

i An NRC' commitment in LER 1-91-009-00.

) Not a CMD commitment.

I i 5. The technician. involved in the event has been.

l counselled as to the: necessity for self--

i verification.

4 u i RESPONSIBILITY: J.-McCann- Complete '

)

IEC (PGIT)- '

I AR A0231287, AE'# 10-Not outage related.

Not JCO related.

-An NRC commitment-in--LER 1-91-009-00. I Not a CMD. commitment.

6. An INPO video on self-verification practices will'be presented-to.IEC~ personnel during the. ,

quarterly update meeting. l l

RESPONSIBILITY: D. Clifton Complete Training -(PATT)

AR A0231287,JAE # 06 ~j Not' outage related. 1 Not JCO related. 1 An'NRC commitment in LER 1-91-009-00. .l Not a CMD' commitment.. I

7. Stop all IEC work ~until a shop tailboard is' completed on.the importance'of attention.to ,

detail and self-verification.when working-in  ;

the plant.

RESPONSIBILITY: W. Crockett . Complete I&C (PGIT)

AR A0231287, AE # 01 Not outage related.

Not JCO related. i Not an NRC commitment.

Not a CMD commitment.

8. Perform a human factors evaluation'of the~NISz cabinets.

RESPONSIBILITY: K. Oliver Complete ~

Technical: Services (PTMT)

'AR A0231287,~AE-# 04.

Not outage related.

91NCRWP\91 TIN 047.JCN Page 10 of 16" i e--,rei w w w +-m e -e, w:-e- e + + - er =ve-+. mv+-eve-=v= -,r-

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DC1-91-TI-N047 D4 .-

January 24, lan2 l

Not JCO related.

Not'an NRC commitment.

Not a CMD commitment.

9. Present the " Artificial Island" video on self-  :

verification to IEC personnel.

RESPONSIBILITY: F. Lacross .

Complate' IEC (PGIS)

AR A0231287,-Am,/ 05  ;

Not outage related.  ;

Not-JCO related. l Not an NRC commitment. t Not a CMD commitment. i

10. Post an-INPO Nuclear Network question on.self-  !

verification. enhancement that have'been effective in the industry.

RESPONSIBILITY: 'J. NolanL .

Complete "

3 Regulatory Compliance (PTRC)' s AR A0231287,'AE # 08 Not outage related.

Not JCO related.

Not an NRC. commitment. ]

Not'a CMD commitment.- :r j

11. IssueLan AT EWR AR to NECS to evaluate andf '

label both the front and back of the NI cabinets (all instrument' cabinets withLtwo side access ?) including terminations. This-- .

design should include making the: physical I access barriers permanent. ,

RESPONSIBILITY: D. Weatherby. Complete I&C (PGI9)

AR A0231287, AE # 13 [

Not outage related.

Not JCO related. '

Not an NRC commitment.

Not a CMD commitment.

12. NECS provide the DCNs necessary to make the physical access barriers-on the NI cabinets '

permanent.and label the front'and back of the-NI cabinets,' including terminations.

91NCRWP\91 TIN 047.JCN Page 11 of 16-

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l DC1-91-TI-N047 D4- (

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RESPONSIBILITY: G. Voboril Return NECS - I&C (NCFI) 3 AR A0232287,.AE #.14 j Not outage related. '

Not JCO related. i Not an NRC commitment.

Not a CMD commitment. >

13. Issue an AR to Building Services to change the locks on the NI cabinets such-that each cabinet has a' unique key.. Coordinate'with Operations to establish a policy of-NI' cabinet i key control. Ensure.the-IEC. mockup and q simulator have an identical configuration.. j Request completion by 12/31/91.-

. RESPONSIBILITY: D. Weatherby .

Complete I I&C (PGI9) I AR A0231287, AE # 15 I Not outage related.  ;

Not JCO related.' l Not an NRC commitment.

Not a CMD commitment. 1

~

14. Enhance the I-2 series of surveillance tests (power range NI) to'ainimize the. probability.

of the procedure contributing to work being {

performed on more than one channel at a time.

, RESPONSIBILITY: D. Weatherby ' Complete 1 l IEC (PGI9) l AR A0231287, AE # 16-

{ Not outage related.

Not JCO.related.

Not an NRC commitment.

t

15. I&C to review the policy for, concurrent-verification to determine if it is adequate.

RESPONSIBILITY: . W. Crockett Complete I&C (PGIT) 1 AR A0231287, AE #.17 Not' outage-related.

Not JCO related.

Not an NRC commitment. o Not a'CMD commitment. i 91NCRWP\91 TIN 047.JCN Page 12 of 16 i

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DC1-91-TI-N047 D4 .

i' January 24, 1992-I

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VI. Additional Information -l A. Failed Components:  ;

j Copas Vulcan valves 1-PCV-1.and 1-PCV-11.

B. Previous Similar Events:

1. ESF Actuations Due To Personnel Error:

NCR DC1-91-OP-NO38,~" Actuation of: Wrong Test Switch Causesl Unplanned; Diesel Generator Start >

(EST) Actuation due-to' Personnel Error,"

describes-an event wherein a non-licensed operator inadvertently actuated.the. wrong _

~

Solid State Protection System (SSPS) (JG) test- .;

switch resulting in an unplanned Emergency Diesel Generator start,-an ESF' actuation. The  :

root cause of this' event was failure to-follow self-verification policies.. The' corrective action could not have prevented the'most recent event' reported in NCR_DC1-91-TI-N047 since-the counselling was directed-only at .

Operations personnel.

2. Condenser 40 Percent Steam' Dump Valves:

NCR DC1-90-TI-N090.and DCO-90-TI-N091 reported >

a Unit i reactor trip and.SI.due to afstuck.

open pressurizer spray valve. _During this-event SDV 1-PCV-1 leaked excessively following )

actuation. The corrective actions to prevent- '

recurrence for this event. dealt primarily'with the stuck open pressurizer spray valve which .!

l would not have prevented'the contributory:

cause on the SDVs failure reported--in NCR DCl-91-TI-N047.

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c. Operating Experience Review:

1. NPRDS:-

-Not applicable.

2. NRC Information Notices, Bulletins, Generic-Letters: >

s 91NCRWP\91 TIN 047.JCN Page ~ 13 of 16 t

+e-,.- -n .c y..-e -,. *,,,+w#-a.e-r-e e,.e.w- ve-wv,,.+..,,,w,,e.,...ery.

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=  !

'DC1-91-TI-N047' D4 j January 24, 1992' l-None.

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3. INPO SOERs and SERs: ,

None.

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D. Trend Code:.

TI (I&C) - A3 (personnel error, lack of mental attention).-  !

E. Corrective Action. Tracking: j l- '

The tracking action request is A0231287.

~

F. Footnotes and Special Comments:-

None.

G.

References:

1

1. I&C' personnel; statement..

2. Operations logs.

3. Event Notification' Form, May 17, 1991.

4. NCRs DC1-90-TI-N090 and-DCO-90-TI-N091. j

5. Westinghouse Letter.PGE-91-605, dated May ~

18, 1991.

6. ERP 91-6.

7. Initiating Action Request'A0231244.

8. LER 1-91-009-00.-

H. TRG Meeting Minutes:-

1. On'May 24, 1991, AT 10:00 an PDT in room 604' ~l of the administration building the TRG-tet as; >

part- of the .IDU) forta; reactor trip and' safety.

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injection that occurred 1on: May 17,.1991. 'A .

description of the' events leading up;tolthe: 1 trip was-presented. Thel root cause of the i event was a lack of self-verification on the j 91NCRWP\91 TIN 047.JCN Page 14 of 16 l-

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' DC1-91-TI-N047 .D4 January 24, 1992 -

part of an IEC technician during.the-performance of an STP. This has been a' generic problem at DCPP'and in the rest of the  :

industry.- IEC stated that:they'will work with Training to develop an enhanced program for .

self-verification training. Various means~of alerting technicians while they are performing tasks in the plant were discussed. 'It'was- ,

concluded that a physical barrier would.have the best potential for being successful. It .

was determineu that having-independent '

verification was excessive and would not necessarily be 100% successful.:

The TRG will reconvene in about two weeks to i discuss the global ~ concerns' raised in the .l meeting regarding reactor ~ trip hazards and '

training for site-personnel other'than IEC in self-verification. ,

2. On August 1, 1991, at 10:00 am PDT in Roon 533 >

of the. Administration Building, the 15tG reconvened to discuss progress of the '

corrective actions to prevent recurrence. A '

draft of the human factors-study'was presented and it was determined that the study should be reviewed in depth by IEC. Conclusions as to-what modifications to make will be discussed' '

in a TRG reconvene on or about August L 8,.1991.

3. On September 12, 1991, at 10:00 am PDT in Room 533 of the Administration Building, the TRG -

reconvened to discuss tha' human factors l analysis of the NIS cabinets. ..A presentation j was given by-Joe Cucco. :j Additional corrective actions, as noted in V.C.11. through V.C.15. above were identified..

The TRG will' reconvene on or about October m17, 1991. Operations ~will be invited with emphasis to the next TRG reconvene.

I. Remarks: y The action items identified in the ERP that will require long term tracking will be-incorporated 91NCRWP\91 TIN 047.JCN Page 15 of' 16 ,

E E i

'DCl-91-TI-N047- D4 '

January 24, 1992 into this NCR to permit the ERP to be closed in a timely manner.

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