ML20210R939

Summary of 860324-25 Meetings W/Util Re Plans to Recover from 851226 Overcooling Event.Preparation of Comprehensive Recovery Plan,Demonstrating Safe & Reliable Plant Operation, Needed for NRC Restart Approval.Viewgraphs Encl

ML20210R939
Person / Time
Site:	Rancho Seco
Issue date:	04/23/1986
From:	Miner S Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20210R937	List: ML20210R933 ML20210R939
References
TAC-60462, NUDOCS 8605200136
Download: ML20210R939 (175)

Text

3 April 23,1986 D

Oh LICENSEE:

Sacramento Municipal Utility District FACILITY:

Rancho Seco Nuclear Generating Station

SUMMARY

OF MEETING HELD ON MARCH 24-25, 1986 WITH SACRAMENT 0 MUNICIPAL UTILITY DISTFICT (SMUD) TO HAVE SMUD BRIEF THE SENIOR NRC STAFF REGARDING THEIR PROPOSED PLANS FOR RECOVERY FROM THE DECEMBER 26, 1985 OVERC00 LING EVENT There have been a number of failures at the Rancho Seco plant since completion of their refueling outact in June, 1985. Two of these failures resulted in overcooling transients]due to the actions of the ICS. The latest overcooling transient on December "6, 1985 was caused by the ICS power failure.

In addition, there has been a hist (.y of problems with B&W plants NNI/ICS performance. Since the SMUD recovery plan from the December 26, 1985 transient appears to be narrowly focused, a meeting was held at Rancho Seco between NRC staff and SMUD management to provide SMUD with the opportunity for convincing the NRC senior staff that their proposed plan would provide for safe and reliable operation of Rancho Seco. The attendees list and copies of the view graphs presented at the neeting are enclosed.

SMUD initially provided an overview of the utility including its organization structure. They stated the problems encountered during the last few years were due to a number of factors which included lack of board support and commitment, band aid fixes, inadequate manning and over-reliance on contractors.

SMUD stated that they are ccmmitted to improve performance and as a result have made management changes and additions, increased staffing, and started a root cause program. They further noted that a self-initiated audit by the Bechtel Corporation for total plant performance improvement was in progress.

SMUD has established a management team to develop a plan to review plant performance and management improvement. They have assembled a team of four knowledgeable individuals (consultants) outside the SMUD organization to review the plan.

SMUD then provided an overview of the Rancho Seco ICS, NNI, proposed EFIC system, plant maintenance program and their configuration control progrem.

They described plant modifications that are being made as a result of the December 26, 1985 transient.

The NRC staff noted some positive aspects of the licensees actions. For example, the NRC staff was pleased to note that members of SMUD's Board of Directors and the General Manager attended the meetings. However, the present recovery program appears to be narrowly based to address primarily problems resulting from December 26, 1986 transient and to respond to staff concerns.

The program does not appear to be broad based enough to address some of the basic problems at Rancho Seco that have existed for some time. The NRC staff needs to be convinced that the recovery plan is comprehensive enough so that once restarted there is reasonable assurance that another transient would not occur in the near future and the Rancho Seco plant can be safely and reliably operated.

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The NRC staff suggested that SMUD prepare a comprehensive recovery plan and only after their management is convinced that the plan, when implemented, would result in safe and reliable operation of the plant, should it be presented to the NRC staff. The NRC staff stated that they will not agree to start up the Rancho Seco plant until they are convinced by SMUD the plant can be reliably and safely operated.

SMUD agreed to prepare such a plan and noted that they will nave the plan reviewed by the four independent, knowlegeable consultants.

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,. \\ ; ~ w.. ~o e Sydney Miner, Project Manager PWR Project Directorate #6 Division of PWR Licensing-B

Enclosures:

As Stated cc w/ enclosures:

See next page l

PBD#6 SMiner-A 4/ja86 3

r v-MEETING

SUMMARY

DISTRIBUTION Licensee: Sacramento Municipal Utility District

• Copies also sent to those people on service (cc) list for subject plant (s).

Docket File NRC PDR L PDR PBD-6 Rdg JStolz PM OELD EJordan BGrimes ACRS-10 NRC Participants HDenton JTaylor i

JMartin GEdison JCalvo SMiner HBailey JCrews LMiller DCrutchfield DKirsch GYuhas AChaffee i

WAlbert i

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- GPerez DOE DUessing i

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Enclosure ATTENDEES LIST SMUD X 7 aylor C. Wilcox-D. K. Lowe R. Rodriguez-G. Coward R. Dieterich A. Schweiger P. Turner J. Vinquist D. Whitney M. Ehlinger R. Daniels S. Redeker H Beebe J. Jones R. Little.

P. Bosakowski J. Jewett J. Shetler D. Army R. Croley N. Brock 5

J. Williams T. Tucker J. Jackson, MAC J. O'Hanlon, UES J. Taylor, 3&W NRC H. Denton J. Taylor J. Martin G. Edison J. Calvo.

S. Miner

.H. Bailey J.. Crews L. Miller D. Crutchfield.

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WAPA - Federal Agreement 360 MW 18.0 %

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- POOR PLANT PERFORMANCE

- POOR SALP REPORT

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- POOR PL ANT PERFORMANCE

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BECHTEL/ RANCHO SECO AUDIT ORGANIZA TION BECHTEL POWER CORPORA TION SMUD LOREN HINKELMAN - V.P. & G.M.

DEWEY LOWE - G.M.

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8 A D IMPLEMENTA TION CHUCK LACEY - V.P. & DEPT. G.M.

COMMITTEE REVIEW BOARD CHUCK LACEY LOU PEOPLES HOWARD FRIEND 1

AUDIT TEAM DIRECTOR SMUD COORDINATOR FRED HOLLENBACH CARL ANDOGNINI PROJECT MANA GEMENT--- ~ ~ ~ -NORM CURTIS ENGINEERING - - - - - ~ - -KEITH BURROWES, KEN COOKE START-UP AND OPERA TIONS- -- -DICK WEBSTER COST, SCHEDULE, CONTROL- ~ ~ FRANK ADDEMAN LICENSING & OTHER DISCIPLINES-RA Y ASHLEY I6451 1

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MARCH - JUNE 1986 George Cage l6 5 2l

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DUE TO DETAILED PREPARATION 1985 - SUCCESSFUL CONTROL ROOM DESIGN REVIEW BY NRC -

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• PLAN REVIEW BY KNOWLEDGEABLE INDIVIDUALS

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MAIN FEEDWATER SYSTEM PURPOSE:

TO PROVIDE PREHEATED FEEDWATER TO THE OTSG'S DURING ALL PH.ASES OF :

C

. Normal Startup e Power Operations

. Normal Shutdown

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e M= 9 6 h AUXILIARY FEEDWATER SYSTEM P.URPOSE :

PROVIDE AN ADEQUATE SUPPLY OF FEEDWATER TO THE OTSG'S FOR DECAY HEAT REMOVAL DURING:

.O

• Periods when the normal feedwater supply is unavailable

• Period of abnormal operations i

such as natural circulation I

i l

_,..,____,,...._______,,_,.)

AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION

1. LOSS OF ALL FOUR REACTOR COOLANT PUMPS

• Provides higher thermal center for natural circulation

2. SAFETY FEATURES ACTU ATION

()

• Provides auxiliary feedwater flow for decay heat removal

3. BOTH MAIN FEEDWATER PUMP DISCHARGE PRESSURES <850 PSIG

• Provides for auxiliary feedwater flow on the loss of main feedwater 1624]

CONDENSA TE ICS t- --l py.2 052 7 TORAGE f

b TANK py_,ggg7

A OTSG

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4 EFIC EMERGENCY FEEDWATER INITIAiIONAND r-CONTROL Cs l

4 CiANNEL SAFETY GRADE AFW INITIATION STARTS STARTS O

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LOSS OF MFW; (4/MFW FLOW ARTS)

LOSS OF ALL 4 RCP'S i

\\

t i

PRESSURE (600PSIGSG"A" I

PRESSURE (600 PSIG SG "B" l

SFAS (LOW RC PRESSURE, HIGH RB PRESSURE)

EFIC AFW SG "A" LEVEL (13.5"

~

SG "B" LEVEL (13.5" i

I AFW INITIATE AFW CONTROL REPLACE SLB FUNCTION MFW OVERFILL ISOLATE g,

ADV CONTROL e

e

1 EFIC CONTROLS AUTOMATICA__Y AFW PUMPS AFW VALVES C

AFW TEST VALVE MFW ISOL, VALVES ADV's

~

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f CONTROL ROOM OPERATORS CAN MANUALLY CONTROL ANY EFIC CONTROLLED

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APPENDIX "R" CONTROL ROOM SHUTDOWN

, \\

PANEL v

INITIA TE O TSG'S

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)

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CONTROL AFW VALVES (

RPS

)

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W 7:y' SFAS mT2 V3 I

PRESENT AFW VALVE CONTROL

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\\_.

SEQUENCE OF EVENTS

SUMMARY

l RANCHO SECO TRANSIENT 12/26/85 INITIAL DECEMBER 26,1985 l

l.

CONDITIONS:

76% REACTOR POWER l

ICS IN FULL AUTOMATIC l

CONTROL i

04:13 A.M.

t i

i 1

(

i i

i l

l64 9]

1

r SEQUENCE OF EVENTS

SUMMARY

(CONT'D)

ELAPSED TIME (MIN:SEC)

EVENT 15:53 OPERA TOR A TTEMPTS TO CLOSE AFW VAL VE TO "A " O TSG FUR THER. VAL VE FAILS OPEN.

16:43 OPERATORS BEGIN DEPRESSURING RCS TO MOVE OUT OF PTS REGION.

19:13 AFW VALVE TO "B" STEAM GENERATOR IS FULL Y CLOSED. ALL FEED WA TER FLOW TO "B" STEAM GENERA TOR IS STOPPED.

22:13 OPERATOR ATTEMPTS TO CLOSE MANUAL AFW ISOLA TION VAL VE TO "A" OTSG.

VAL VE WILL NOT MO VE.

26:13 ICS DC POWER RESTORED. OPERATORS SHUT AFW VALVES, TBV, AND ADVs FROM CONTROL ROOM.

COOLDOWN CEASES AT 386 F.

42:00 EXTRA SENIOR LICENSED OPERATOR BECOMES LIGHT HEADED

\\

50:00 MAKEUP PUMP IS DAMAGED

[6483 i

k

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

v----,-

(

SEQUENCE QF EVENTS

SUMMARY

(CONT'D)

ELAPSED TIME (MIN:SEC)

EVENT 01:17 OPERA TORS MANUALLY START "B" HIGH PRESSURE INJECTION PUMP (HPI) TO FURTHER INCREASE RCS MAKEUP.

02:15 OPERATORS MANUALLY TRIP BOTH MAIN FEED WA TER PUMPS, A C TUA L MA IN FEED FL O W WAS ZERO SINCE 00:14 03:10 SAFETY FEATURES (SFAS) AUTO INITIATION AT RCS PRESSURE OF 1600 psig.

03:23 OPERATOR SHUTS SFAS ACTUATED AFW VAL VES.

06:13 PRESSllRIZER LEVEL OFFSCALE LOW.

09:13 TBV AND ADV ISOLA TED LOCALLY.

09:23 AFW VALVE TO "B" STEAM GENERA TOR PARTIALLY CLOSED FLOW DECREASED TO "B" STEAM GENERA TOR BY 60%.

I 12:35 AFW. VAL VE TO "A" STEAM GENERA TOR CLOSED, OPERATOR BELIEVES VALVE ONLY 80% SHUT.

13:00 PRESSURIZER LEVEL BACK ON SCALE k

l64 7l

ELAPSED TIME (MIN:SEC)

EVENT 00:00 LOSS OF ICS DC POWER VALVES MOVE TO 50% OPEN POSITION 1

Turbine Bypass Valves (TBV)

]

Atmospheric Dump Valves (ADV)

Main Feedwater Regulating Valves (MFW)

Auxiliary Feedwater Valves (AFW)

MAIN FEED PUMPS GO TO MINIMUM SPEED 00:14 MOTOR DRIVEN AUXILIARY FEEDWA TER PUMP (AFW) STARTS AT 850 psig MAIN FEEDWATER PUMP DISCHARGE PRESSURE 00:15 REACTOR TRIPS ON HIGH RCS PRESSURE 00:19 TURBINE DRIVEN AUXILIARY FEED PUMP (AFW)

STARTS 00:24 AFW FLOW TO BOTH OTSGs t

00:38 OPERATORS INCREASE RCS MAKEUP l

00:43 OPERA TORS SENT TO ISOLA TE TBVs, ADVs, AND AFW VAL VES, OPERA TORS RECOGNIZE EXCESSIVE COOLDOWN RATE.

16461

__.,r._,,

,,,-,..~.,_,,_-,_,_-.-,,,..,_ ~.,e

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

-_,,,_r_-_,,~,,-_.,_,_,_,r..__,_.

i o 1-p w.

To Other OTSG Dual Drive Cont iV Turbine Bypass Valve i

AFW Pump l

ICS L

Manual Main Steam System From Isolation i (

(-,

I \\

Condensate Valve a

s (MSS)

Motor Driven lIC Ta k Main Throttle Or Site AFW Pump y,g y, Sp N

j m - ICS TG Re ul ting SFAS l

valve Oper l

l Conti Viv

/CS 3( 1 -

i O

l Main Feed ICS i

t y

v

1. • "U

}

D

\\

Atmospheric Steam Qump Valves Main Block Viv i

Startup Control Feedwater System (FWS) l RANCHO SECO l

l 4

3 I

i

1 O

M ROOT CAUSE ANALYSIS.

INCIDENT ANALYSIS GROUP / ROOT CAUSE PROGRAM i

• Independent from Responsibility to Resolve Events /Discrepencies

• Identifies Direct and Root Causes

• Management Review Tearn, (MRT) Reviews and Approves Root Causes i

• Cognizant Departments develop Recovery Plans for each incident

• MRT approves Recovery Plans with IAG concurrance l

' QA Dept. tracks Compliance with Corrective Actions l

l l

Iss41

G

"?

i, a

DIRECT CAUSE:

THE IMMEDIATE TECHNICAL OR ENGINEERING EXPLANATION FOR THE FAILURE l

ROOT CAUSE:

i THE PROGRAMMATIC SHORTFALL OR DEFICIENCY i

THAT ALLOWED THE FAILURE TO OCCUR i

i 4

i 16 6 5l 4

i

a n o

u i

IAG 1985 ACTIVITIES

• 42 Root Cause Reports

• 27 Complete

• 85 Recommendations e 40 Recommendations Complete

• 62 Recommendations from 12-26 event l

l6 6 6]

ROOT CAUSE ANALYSIS Complex Event Required Breakdown into Five

" EVENT THEMES"'

1. LOSS OF ICS

2. RAPID COOLDOWN

3. MAKEUP PUMP

4. HEALTH PHYSICS

5. EMERGENCY PLAN IMPLEMENTATION 9

l6 6 7l 1

9 ISSUES REVIEWED FOR_;

• PROCEDURAL' ADEQUACY j

• DESIGN FEATURES l

• HUMAN PERFORMANCE i

• OTHER ISSUES SUCH AS:

-TRAINING l

-PREVENTIVE MAINTENANCE

-CORRECTIVE MAINTENANCE

-K$Y CONTROL

-VENDOR TECHNICAL. ANALYSES l

l688]

~

l l

i

c ROOT CAUSES AND CONTRIBUTORY CAUSES CONTRIBUTORY i

i EVENT THEME ROOT CAUSE CAUSE i

(1) Loss of ICS i Manufacturing Defect I (2) Rapid Cooldown Delay in implementing Inadequate Procedures l

Design Changes to I Inadequate Training l

Mitigate Effects of Loss of ICS l

(3) Makeup Pumps Inadequate Personnel Error I Procedures l

(4) Health Physics Lack of Clearly Procedural 1 Defined Division of I

Non-Compliance Responsibility j

(5) Emergency Plan l

Procedural inadequate Training j

implementation Non-Compliance 18891

g

. AREA $' FOR IMPROVEMENT l

1. PROCEDURAL ADEQUACY

• Guidance for Recovery from SFAS l

• ICS. Casualty Procedure l

• Guidance E.05 Cooldown Procedure l

• Clarification and/or Reorganization of l

Emergency Plan

2. DESIGN FEATURES l

• AUX FW Valves, TBVs, ADVs Position on Loss of ICS i

i l

i l6 7 0]

i AREAS FOR IMPROVEMENT

3. HUMAN FACTORS l

• Understanding of HP Assignment / Involvement

• Visibility of ICS Power Switches j

• Noise from HVAC in Control

4. HUMAN PERFORMANCE i

i

• Emergency Plan Implementation j

• Limiting Cooldown by Securing AFW Pump.

l

• Adherence to RAD-CON Procedures i

.5.OTHERISSUES i

• Training l

• Preventive Maintenance i

i Is 7 il

u

._h-9 PLANT MODIFICATIONS O

I i

\\

16381

PLANT MODIFICATIONS

• PRECLUDE LOSS OF ICS POWER

• MITIGATE TRANSIENTS

• HUMAN FACTOR ENHANCEMENTS y..

i

120 VAC S1 S2 1

o o

24V 24V POWER POWER SUPPLY SUPPLY i

o o

e TRIP SIGNAL 32 57

,o.

%s AUCTIONEER BUS q

i V

a i

POWER SUPPLY

'g- - w, MONITOR i

i A

2 LOAD BUS VOLTAGE SIGNAL y y y HIGH RESISTANCE CONNECTION LOADS ICS POWER DISTRIBUTION I

i

,--------,se m.e-

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_ _ _ +

- 3 1

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l

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I MAuuAL 24 VOC l

coNTacL POWER I

LED 6thu A L SU P Pt.Y I

(4-10 M A)

O i

i Idd 8

l M AWu AL - RED I I

I

!cS l

120 VAC -- 9 I

AUTO-6REEMI l

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e 8

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i

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4 AaTS wand AUTC. MAM'd AUTC.

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HW ICS ICS POWER f

f AUX, EEL AY !

l. _ _ _ _ _ _ _ _ _,, _ _ _ _.!

lr,^uyA'u i S7 Ail 0M j

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i I

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,' AIR SUPPLY V EuT-a r O~AO" r

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(p. ues a).

AFW ICA c.4MTROL Po$

, y,g VA LvF.

' SUPPLY

\\ /

r e

)s i

i AUXILIARY FEEDWATER CONTROL INDEPENDENT OF I.C.S.

FIGURE I 1

_..._,,....--._s_

_.v,_. _,...,. - -, -. ~.....

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

...,_.-..,....y

a<<

p Q,i NEW SCOPE ICS

,CLOSE CtDSE Cl,0SE RkbY

NOilMAL, MTO

,i s, AlqO _

CONTACT SwtTCH ON LDCAL t

OPEN ON Ns"ABLE 1125D SwlTCH LOSSOF ICS POWER r

II>

A i

k h

IDADS i

z GO Atp~u 6

SOLEHO i

~

(KEHElV>t TO CtDS ADVS 2

?

i ADV MANUAL CONTROL

~

i nu 1

1 i

i

aP p

v NEW SCOPE ICS CLOSE CLOSE REbY NOl]M AL,_

AUTO ny au CONTACT SwlTCHON OPEN ON TABLE t1250 i

LOSSOF ICS POWER

_. o A

k

.hs IDADS 4

ALRM MO SOLENDIC 1

DEEHEIElZI TO ClDSE i

TBv5 1

4 i

j 4

l

~

TBV MANUAL CONTFOL FIGUK 3 1

i i

4 t

3

SUMMARY

• IMPLEMENTING MODIFICATIONS TO REDUCE LIKELYHOOD OF LOICS POWER

• IMPLEMENTING MODIFICA.TIONS TO ENHANCE THE EFFECTIVENESS OF THE OPERATOR FROM THE CONTROL ROOM

• CONTROL ROOM CONTROL 6

INDEPENDENT OF ICS FOR:

-AFW FLOW VALVES

-ADV'S l

-TBV'S l

• SIMILAR TO OTHER B & W PLANTS

• HUMAN FACTOR ENHANCEMENTS 16401 l

MAINTENANCE PROGRAM OVERVIEW WORK REQUEST SYSTEM DAILY WORK PLAN MAINTENANCE INFORMATION MANAGEMENT SYSTEM (MIMS)

MAINTENANCE ENGINEERS CORRECTIVE / PREDICTIVE / PREVENTIVE MAINTENANCE g

PROGRAMS e

MAINTENANCE PROGRAM IMPROVEENTS ADDED SINGLE MAINTENANCE MANAGER REPORTING TO PLANT MANAGER INCREASE MANAGEMENT / SUPERVISORY PERSONNEL TO ALLOW MORE SUPERVISORY TIME IN FIELD MANAGEMENT / SUPERVISORY INVOLVEMENT b

IMPLEMENTING GRADUATED QUALITY PROGRAM FOR WORK ON MAJOR SECONDARY PLANT EQUIPMENT PREVENTIVE MAINTENANCE IMPROVEMENTS PM SUPERVISORS FOR MECHANICAL, ELECTRICAL, l&C PM UPGRADE PROGRAM VALVE PM UPGRADE PROGRAM

i l

PLANNING AND SCHEDULING IMP!<0VENNTS NEW SCHEDULING MANAGER.

IN-DEPTH STUDY COMPLETED.

IMPLEMENTATION STARTED.

m INCREASED STAFF IN MAINTENANCE AND SCHEDULING.

DAILY WORK PLAN.

b i

....,___...,.._.~.,..m_.,

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

g n ;.

PLANT MODIFICATION CONTROL d

\\

i 1

8 DEFINITION 4

i G

HISTORY i

O PRE $ENT l

l S

CONCLUSIONS i

i i

4

o i,,

g 9

MANAGEMENT

\\

PLAN MAINTENANCE ENGINEERING MAINTAIN DESIGN i

i l

l PLANT OPERATIONS CONSTRUCTION i

MODIFICATION OPERATE INSTALL CONTROL i

l START UP QUALITY i

i TEST INSPECT l

i

l O

i l

PLANT MODIFICATION CONTROL-ANSI N 45.2.11 i

RESPONSIBILITIES / INTERFACE CONTROL i

i DESIGN INPUT CONTROL i

\\

PREPARATION OF DESIGN DOCUMENTS DESIGN VERIFICATION i

)

AUDITS / CORRECTIVE ACTION t

h

PLANT MODIFICATION CONTROL ANSI N 45.2.11 DOCUMENT CONTROL MAINTENANCE & RETENTION OF DOCUMENTS MANAGEMENT REVIEW OF PROGRAM EXPERIENCE REPORTS

'69

'75 t-l l

DESIGN AND CONSTRUCTION l

CONSTRUCTION RELATED l

A/E AND SMUD PROCEDURES l

l INTEGRATED CONTROLS P*,

02.

p I

'75

'78 l

i i

EARLY YEARS i

i MINOR CHANGES I

l ECP-1 CONTROLS OK FOR MINOR CHANGES

~

1 i

l l

'79

'85 i

l MAJOR MODIFICATIONS TMI, EQ, APPENDIX R i

A/E AND SMUD PROCEDURES CHANGES TAXED CONTROLS 0 ',

O M;

HIERARCHY OF SMUD PROCEDURES v

a ANSI /

ANSI i

1/p////

rEDeut

t n45.a.n

=

QAP

, 7 OCI'S NEP'S C EP'S A P'S S P'S IMPLEMENTING PROCEDURES i

QUALITY,, ENGINEERING _ _ OPERATIONS.

CONSTRUCTION u

o c i.

p' w.

__ NUCLEAR ENGINEERING PROCEDURE ENGINEERING AND CONSTRUCTION DATA BASE l

l

- /b::7 L:

%=.

4 %-

s Gn. - -

a Tax w.

A k n--.

, G -= -

s fi = v =,

7u a k=

i l

NUCLEAR DESIGN DESIGN STD DESIGN ANALYSIS LICENSE CONST.

NUCLEAR

[

ENG.

CRITERIA GUIDES SPECS BASES GUIDES COMMIT ENG.

ENGINEERING fE CE hBR PROC.

LIST PROC.

.i 1000-5100 5200 5300 5400 5500 5600 6000 7000 1

)

4000

}

L

l p

CONTROLS ON PLANT MODIFICATIONS ENSURE THAT CHANGES ARE:

PLANNED DESIGNED INSTALLED INSPECTED TESTED I

OPERATED MAINTAINED IN. A SAFE AND RELIABLE MANNER

C 7

i

SUMMARY

OF PLANT i

MODIFICATION CONTROL NEP'S MEET NRC REQUIREMENTS RECENT REEVALUATION &

CORRECTIVE ACTION APPENDIX R EQUIPMENT QUALIFICATION 79-14 WALKDOWN HVAC SUPPORT WALKDOWN

• \\'

Q.

I l

1 PLANNED IMPROVEMENTS AND CONTROLS

)

i i

UPGRADE DESIGN BASES f

l ISSUE ANALYSIS GUIDES I

I

=

' PRO

- ACTIVE QUALITY 1

PROGRAM o

e i

i l

DED

l r,

'*r ~

l QUALITY-DEPARTMENT

~

i l

Quality i

Department

)

Manager i

l Staff Senior l

Engineer Secretary 1

l Quality Quality Site. Quality Crpy;r gg ty Control Engineering.

Assurance Assurance

. Supervisor Supervisor Supervisor Supervisor (2 7]

[11)

(19).

(12) i l

1 l

QUALITY CONTROL (27)

Supervise QC Inspectors Ensure Inspector Training Ensure inspector Qualification.

e Ensure Inspector A vailability p

Maintenance

~Modifica tions (y

• Communicate Problems Associated With inspection To District Management O

e

QUALITY ENGINEERING (11)

Supervise Quality Engineers Review Purcha~se Documents To Verify inclusion. Of Quality Requirements Review And Monitor Maintenance Activities Provide Technical Support To QC Inspectors u

o Review And Monitor Nuclear Engineering Design. Practices And Procedures

• Provide Quality Department Training.

l l

---

.-----..-,,--,-,-,_,,,---,,,,---.,-,,-,,,~n.--

,,w,,..

.w,w

SITE QUALITY ASSURANCE (19)

\\

• Supervise. Site QA Personnel e Implement The Trend Program

• Maintain The NCR Program

• Maintain QA Interface With NR C,. INP O,. A NI A nd C o de Review Personnel.

,_c

• Perform Surveillances To Close Root Cause Analyses

• Perform QA Surveillances

-Maintenanc e

-Modifications

-Op erations

• Secretary To The MSRC rat 31 l

C@RP@ RATE QUALITY ASSURANgg

• Supervise The QA Audits.

• Ensure That QA Documents Are Current

-QA Manual

-Quality Instructions

-Quality implementing Procedures

\\

• Ensure. Audits Are Conducted On Schedul. l\\,

\\

-MSRC Audits 7

-Other Scheduled Audits

• Maintain Qualified Vendors Listing

• Maintain The Non-Nuclear. QA Program

• Maintain The Quality Portio,n Of The Coordinated Commitment List t

\\

l i

f 16411 l

f

i i

QUALITY ORGANIZATION OVERVIEW 6

O Present Quality Department

.m t

e Quality Department Training i

i

• Quality Department Duties l

e Quality Function Additions l

l I

16291 l

REACTOR COOLANT PUMP SHAFT s

BINGHAM WILLAMETTE C0 (BWC) PUMPS RANCHO SECO BYRON-JACKSON PUMP PROBLEM SHAFT CRACKS - ALL AT SAME LOCATION IN SEAL AREA 48" - 50" FROM UPPER END OF SHAFT OPERATIONAL ASSESSMENT REVIEWED INP0 SERs ON REACTOR COOLANT PUMP PROBLEMS REVIEWED NPRDS FOR BWC PUMP PROBLEMS BWC RQV 28x28x41 PUMPS DIFFERENT FROM BYRON-JACKSON DIFFERENT DIMENSIONS, SEALS, COMPONENT ATTACH-MENT ETHODS, ROTATIONAL MASS O.

ACTION:

FOLLOW ANALYSIS OF CRYSTAL RIVER PROBLEM (1 SHAFT FAILURE, 3 SHAFTS HAVE UT CRACK INDICATIONS) l FOLLOW ANALYSIS OF DAVIS-BESSE PROBLEM (2 0F 2 SHAFTS EXAMINED HAVE UT CRACK INDICATIONS) i DETERMINE WHETHER ADDITIONAL EVALUATION OR INSPECTION OF RANCHO SECO RCPs SHOULD BE DONE.

I' l

i O

i I

ACTION PLAN FOR SYSTEMATIC TROUBLESHOOTING I

FPJe1 0

• SYSTENDTid '

~

TROUBLESHOOTING I

j NUREG 1154 APPENDIX B METHOD e

1 OC TOBER 2, 1985

\\

Loss of Vacuum /Cooldown Event DECEMBER 5, 1985 ICS Oscillation / Stuck Indication l

l DECEMBER 26,1985 Loss of ICS/ Rapid Cooldown 1

l715l l

l

Qp r

t [GE ND: SU. Stcrtup RIquirgd M

LT. Long I:ro SIAIUS DAIE 03-13 86 PE Power Escalation DECEMRER 26, 1985 1RANSl[NI IIME OC NA. Not Applicable Page I of 25 ST. Short Terin

- ACIl0N LIST -

4

$YNI k$ /Np G4 nes Per GAC 85-1001 Apply DESCRIPTION RESPONSIRilliY SCHEDULE STATUS WR No./NCR/

C(594EN15 ETC.

1.

Post Trlp Report 2.

lksman Factors / Root Cause Analysis 3.

Determine Cause/ Corrective Action for ICS Power Failure 4

Makeup Pump (P-236) Failure 5.

Radiation Monitor R-15001 6

RCS Overcooling 7.

Ilealth Physics 8.

Emergency Plans 9.

Training

10. Operational Review

11. System /Camponent Response

12. I)uarantined Equipment List

13. Procedure Changes Tracking 14 Carryover items from October 2 1985 Event Action List

15. Recommendations From 12-26-85 Transient i

16. NRC Region V Restart Items

17. IRRIG 0667 and RAW-1564 Recomunendations D

j

)

0 it c

Rx TRIP DEC.

26, 1985 RCS B-LOOP TEMPERATURE (cold) 600 000LDOWed ClRVE STARIS AT 550*F 550 -

C00LtWAM CtRVE SLDPE IS 100*F PER llots t

E m

500 -

I i

0 450 -

l i

l N

Ld i

0:

400 -

i 1

i n.

350 -

i 300 -

~

1 j

j 250

)

00:00 10 20 30:00 40 50 60:00 70 80 90:00 t

TIME - Minutes from Rx Trip

l

~

O i

REACTOR VESSEL CRACK INITIATION i

j COOLDOWN RA TE EXCEEDED TECH SPEC 100 F/hr 1

l VESSEL ANAL YSIS FOR CRACK INITIA TION i

)

B & W :.

• Acceptable as Pressure was < 2 750 psig Consumed 0.3 "Cooldown Cycles" o

Total Cycles Designed, 240 l

e Total Cycles Consumed, 5 i'

EPRI PER ASME SECTION XI APPENDIX e Acceptable, as Pressure was < 2500psig 1

and/or T -RTuo 7g > 5 5

• F l

c l

e Actual Conditions were < 1700 psig l

T -R T,,7, =169*F c

l IIT/NRC Staff i

l A cceptable, Tc > 170 *F Actual Minimum Temperature 386 *F DEVELOPING REVISED TECHNICAL SPECIFICATION

\\

L!.uLI

l lll1l1l l

N N

N N

O I

I O

O O

l T

T l

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6 A

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C 0

S P

S S

S C

C C

C C

R R

R R

R 0

O O

0 0

1 T

T 1

1 P

O O

e*

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=

=

l V E L K

A A

V M

E 0

2 V

9 i

4 0

1 0

L 0

I A

8 6

0 8

8 3

3 V

3 3

3 P N 2

2 2

2 2

I M

e

- M l

Ol M V

V V

V V

8i i

S S

S S

F t

C F

F S

F J

NI O

3 W

I H

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0>

M

{

S N

O

~

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P P

P M

M M

d8t A

l 6 U 0 1 P

8t 3 P 3

3 P 2I 2P 2l P

l r

P 6

r l

l H

Pf PE 4

P K

6 A

B A

3 I

2 O

M i

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-u FS N

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5 l

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f k k

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I R 6

E E J I I

0 3

RL 0

2 O

5 L

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3 A

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F S

S V

F

)

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=

(

t

(

i l

M l

iSK d

A N gL A g F T d

3\\

4 0

0 0

0

~

5 5

2 2

V V

F F

S S

i

-l l

)

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O MAKEUP PUMP P-236 INBOARD END OUTBOARD END MECHANICAL SHAFT AREAS OF WEAR RADIAL SEAL (TYP)

BREAK DUE TO CONTACT (TYPICAL)

BE ARING THRUST j

BEARING iz 1,

1 1 11 O

l kT

~# "; O r

aanw A c

wp n

i y

I S

/

um ss.

d b

J J

a ist 2nd STAGE 4th DOUBLE Sth OIL DRAIN SUCTION 6th 7th Bth,g INBOARD STAGE BEARING OUTBOARD

^ " "

OIL DRAIN BRACKET BRAC ET l

Ia il i

l

t O

1 4

FLOODED "A" MAIN STEAM LINE l

A & E Analysis for Stress,. Thermal Shock l

1

• Walkdo wn

- Piping,

- NDI of Turbine Bypass Line.

u i

Change in MSL Failure. Logic Setpoint

\\

i t

(

17:el Ii

• ^

j O

i 1

j i

l Health Physics j

Procedures / Policies 1

I

~

I

• Emergency Plan

• Training I

l i

I

}

l l

i l

1 I

t 17181 E

O

~

ACTION PLAN SYSTEMATIC APPROACH l

Loss of ICS Power Rapid Cooldown i

l l

MU Pump Damage j

Health Physics

\\

1 l

Emergency Plan J

I m

Additional Scope o

e I

)

l717l

O M

~

i NON-EVENT RELATED NUREG 1195 ISSUES-Viability Of FSAR Analyses

• Configuration of " Generic" B&W Plant

" Transient Response of B&W Reactors"

• NUREG 0667 Responsiveness To IE Bulletin 79-27,

" Loss of'NNI"

• BAW 1564 "lCS Reliability" Recommendations Crystal River Loss of NNI, Feb 1980 Rancho Seco Loss of NNI/H2 Explosion, March 1984 1

• BAW 1791 "Probabilistic Evaluation Of PTS", Applicability l6 7 2]

[

t

~

O

~

SMUD/ RANCHO SECO NUREG 1195 issues Review Review Of FSAR/USAR Analyses Assumptions 2

j Review Of Plant Configuration Commitments Compliance with NUREG 0667 Recommendations Compliance With BAW 1564 Recommendations Applicability of IE Bulletin 79-27 and I

Crystal River items Applicability of BAW 1971 Analysis Precursor Review

-LER's

-Trip. Reports

~

-Tap Reports

-SOER's

[673]

l

g 4

l FSAR CONCERNS

• Was Cooldown Transient Bounded By i

FSAR Analysis?

• Is Credit Taken For ICS in The Main Steam Line Break Analysis?

i l

i

\\

i l6 7 4]

l

~

O INITIAL FSAR REVIEW MAIN STEAM LINE BREAK EVENT J

. Main Steam Line Break Analysis Assumed No ICS Or Operator Action

. Final Condition Of Transient is Cold Shutdown

. Cooldown Rate Of Transient initiated SFAS in 23 Seconds And Core Flood i

Tanks in 43 Seconds

• Transient Of 12/26/85 Was Bounded By The FSAR Analysis

. Analysis Performed in Response To NRC Question (Q 14A.3)

• Analysis Description Was Poorly Worded In USAR

[675]

i l

I J

ONGOING USAR REVIEW

• Ensure No Undue Credit Taken For ICS Or Other Non-Safety Systems

. Revise USAR Chapter 14 For Clarity, Update Scheduled For July,1986 i

l6 7 6l

g B&W OWNER GROUP NUREG 1195 ISSUES REVIEW

• Preparing Configuration & Compliance Matrix includes

- Status of NUREG 0667 Recommendations j

- Status of BAW 1564 Recommendations

(

- Review Of IE Bulletin 79-27 For Present t

Plant Configurations

- Response To NUREG 1154 " Davis-Besse l

Event" issues i

- Response To NUREG 1195 " Rancho Seco l

Event" Issues

• Finalize Recommendations / Positions, June I

1986

~

18771 I

(

l I.

REVIEW EFIC 4

i II.

CHANGES TO EFIC

- DUE TO ACCELERATED SCHEDULE

- DUE TO EXPERIENCE GAINED o

O III.

INTERACTION OF EFIC/ICS/AFW 1

(

EFIC EMERGENCY FEEDWATER INITIATION AND O

~

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CONTROL 0

e O

I 8

e i

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e I

k w.

4 CiANNEL SAFETY GRADE AFW INITIATION o

a STARTS STARTS O

PUMPS CONTROLS i

I

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5

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A oTSG I

n CONDENSATE

(

M W.

Mi

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wa I

NT RESERVONI LM 3

l 9

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W

=

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l V

AUXILIARY FEEDWATER SYSTEM contAinutur sunneno i

POST-EFIC sounoAny i

l I

O I

i

O LOSS OF MFW; (4/MFW FLOW ARTS)

I I

I i

l LOSS OF ALL 4 RCP'S I

PRESSURE (600PSIGSG"A" l

l PRESSURE (600PSIGSG"B"

)

SFAS (LOW RC PRESSURE, HIGH RB PRESSURE)

EFIC AFW 4

l SG "A" LEVEL (13.5" j

SG "B" LEVEL

.(13.5" i

j i

AFW INITIATE AFW CONTROL REPLACE SLB FUNCTION 7

s O

MFW OVERFILL I'SOLATE ADV CONTR0E

?

' 6=a 4

g 5

?

e i

.. :11..

-.. - -- - 1

2a Aa n.-

e 9

EFIC CONTROLS AUTOMATICALLY AFW PUMPS AFW VALVES O

AFW TEST VALVE MFW ISOL, VALVES ADV's

3 CONTROL ROOM OPERATORS CAN MANUALLY CONTROL O

ANY EFIC CONTROLLED COMPONENT i

\\

5 EFIC SYSTEMS I

ACTUATION INITIATE SYSTEM SYSTEM o

VECTOR (FEED ONLY b

GOOD S. G.)

t

[ AFW FLOW )

CONTROL l

=

kS.G. LEVEL) j ADV CONTROL p

.m..._g r

---

,r.,-

,.,c..

O M

1 APPENDIX "R" CONTROL ROOM SHUTDO WN PANEL 4

L v

i INITIA TE O TSG'S

)

)

TIE

)

1 AFW EFIC pyypg i

l SFAS

)

)

CONTROL AFW VALVES {

\\

RPS A D V's v

i COMPUTER 9

l l

l6931

-- I CS 45 X-/

1:d SFAS r

PRESENT AFW VALVE CONTROL O

- EFIC D f EFIC A W

tt

--EFIC C r--EF I C B s

v, EFIC VALVE CONTROL

..J

1 T- - - -

r -l- - EFI C 'A I

5 ll r - - -EFIC B l

1 l

6 6

6,

=

x i

MAIN l BLOCK l

6 1

s. u.

Ci EFIC MFW ISOLATION COMMANDS i

l713]

1 s

9 o

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M COND.

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4 a

g s

p A

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4 MAIN STEAM SYSTEM I

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EFIC SCOPE CLARIFICATIONS O

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100 psig L #!

! AIR SUPPLY

/

a

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90 psig N

l 2

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BACKUP AIR SUPPLY l

s i

I

EFIC APPENDIX "R" INTERFACE S.G.

CONTROL INDICATION CHANNEL 7

A FV-20527 WIDE RANGE LEVEL A

A PV-20571 S.G. PRESSURE A

g B

FV-20528 WIDE RANGE LEVEL A

B PV-20562 S.G. PRESSURE B

l 9

e

~ - -

INTERIM EFIC CHANGES O

e e

I

^

EFIC A O

W M

=

--EFIC C

--EFIC B wT2 V R C

EFIC INTERIM VALVE C0f1 TROL 1

i J

e 9

e

,w p---

y-EFIC A 1

A I

i i

W D$

yppy 1

=

BLOCK l

A

s. u.

O EFIC INTERIM MFW ISOLATION p---

y-EFIC A d

, - EFIC B g

r b

I MAIN BLOCK l

A

s. u.

EFIC MFW ISOLATION

i(

(

5 I

m

(

(

g W

6 6

m 8 -><

g->c-m o

o V

t2 SOI - '

S O I -- '

E O

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+><

y)

~

V 8 OldB --.V z

V O l d 3 -- J

..s 4

3 o

o

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-,,..,.w-

,-,---,,----------------------o---

PERMANENT EFIC CFANGES O

\\

EFIC SHUTDOWN BYPASS PERMISSIVE (IEEE 279)

S.G. "A" AND S. G. "B"

< 725 P S I G l

G S.G. "A" OR S.G. "B" < 725 PSIG FACILITATES C00LDOWN WITH TUBE LEAK e

f e

s

l l

0 l

l INTERACTION O

EFIC / ICS / MSS O

e S

e 9

e y

,--,---,v-w-

m

.e

-.m.

-h,

m n

w-

MAIN STEAM SETPOINTS (PSIA)

ICS CONTROL TURBINE HEADER TBV's ADV's SAFETIES NORMAL 900 950 975 1065 o

POST TRIP 1015 1 0l40

,1065

0 ICS CONTROL EFIC TURBINE HEADER TBV's ADV's SAFETIES NORMAL 900 950 1035 1065

)

POST TRIP 1015 1035 1065

umE

{

9 fbe l

l y 3 E.

-y H

G I

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P E

T H

l i

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9 C

1 L

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R 0

s I

6 D

6 A

2 E

1 F

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P E

6

)

W 0 S

0 O

0 q

6 C

5 IW 1

FO EL 6

b q

I rH U

1 l

e yg m@mBaox rQmr BM0 i!;

!1 4

l li1 l

';j-

, i i

4

'l STEAM GENERATOR LEVEL SETPOINTS EFIC ICS (EFIC EQUIV.)

LOW S.G.

LEVEL 13" POST TRIP 27" 32" NATURAL CIRC.

317" 245" O

SMALL BREAK LOCA 381" 374" MFW OVERFILL 100% PWR +25" e

a h

l 400" WIDE RANGE (NNI) 250" bP S.U. 7 (L EVEL)

-150" EFIC L O W 7 0

0 50%

100%

SG THROUGHPUT O

,.<v--.--,

r--

_,_3.-.._,

,_.....,y.,__m.....__

-_,_____,-,,_m.,_,_,_,.--,.-....,..,_,___y-.,

\\

200" LEVEL " SPIKING" 1

\\

\\

m

}S.U.

aW>

~

W

~

=

CL.<0

\\

g fEFIC TIME e----+y--

--w,---

,,--w-.

-,,w,-,,,,_-,

,v--,,

y y-g_,-,.,,-,,

,m,

,--,w,

9 i

i AGREEMENT ON EFIC CONCEPT i

i BY AP&L, FPC AND SMUD i

i FOLLOWING TMI i

i

)

• EFIC concept presented to NRC 9/80 i

• Agreement with staff on value of this system I

I

)

rgTal

a r.

g

}

l CONTRACT SPECIFIED EQUIPMENT DELIVERY 4/83 4

l I,

• First of the 3 systems i

i j

• Support 1983 Outage

\\

i e

an I

i

~ ' '

Q1 l

DELIVERY SCHEDULE ADJUSTED TO 5/84

• Design changes l

Outage schedule adjustments

• Informed NRC 10/82 i Interim Safetyi Grade AFW i

n'odifications in 1983 to satisfy NUREG 0 73 7 I

s NRC agreement 9/25/83 i

\\

i l

l N3:1 l

g 4

}

INSTALLATION SCHEDULE i

ADJUSTED TO CYCLE 8 i

i e Control Room Design Review and RG 1.97 Impact on control panel

{

\\

e Schedule consistent with TDI diesel j

cperation i

informed NRC 4/83 l

l l

i l

j (DE

INSTALLATION SPLIT INTO TWO I

PHASES FOR CYCLE 8 AND CYCLE 9 i

e Living Schedule

• Doncerns over level instrumentation

~

)

• lustrumentation scheduled for Cycle 8 j

to allow observation and familiarization i

~

l by operators

\\;

4 j~

l l

• Control scheduled for Cycle 9 j

{

17001 1

1 t

9 1

l i

FOLLOWING 10/85 TRANSIENT DISTRICT COMMITTED TO ACCELERATE MAJORITY l

OF EFIC INSTALLATION TO CYCLE 8 i

l

• CL ASS I INITIA TION OF AFW

\\

Start pumps and position valves No ICS Control of AFW i

j e CLASS I FLOW CONTROL FOR STEAM GENERA TOR LEVEL

~% Level rate control l

High level for natural circulation if RCP's l

not running e CLASS I FOGG l

• CL ASS I A TMOSPHERIC DUMP VAL VE CONTROL m

l

{

PL ANT PERFORMANCE AND MANA GEMENT IMPRO VEMENT OR G A NIZA TION GENER A L M A NA GER' A GM NUCLEAR s

RECOVERY IND EPEND EN T MANAGER REVIEW PANEL

e i' RECOVERY ORGANIZATION RECOVERY MANAGER J. Vinquis t i

I l

l OPER A TIONS TR A INING ENGINEERING LICENSING QA D. Whitney M. Ehlinger P. Bosak owski R. Little J. Je w e t t 122 4

4 9

• 4 e

O M

DOCUMENTED REVIEW OF PRECURSORS Review lessons learned from a broad perspective PAST RANCHO SECO OPERA TING HISTORY e

LER l

i e

TAP Trip Reports l

incident Analysis Group Reports e

i INDUSTRY EVENTS APPLICABLE TO RANCHO SECO BA W Reports Owners Group TAP Study e

• SOER l

l e

lE Bulletins l

DOCUMENTS REFERENCED IN NUREG 1195 j

Will evaluate ICS and NNI l

aro

~

l

p=m.

d l

SYSTEM EVALUATION RESULTS I

RECOMMENDED MODIFICA TIONS e

l ADDITIONAL TESTING i

e CHANGES TO PROCEDURES AND TRAINING i

l l

l 17251 I

I RESULTS OF REVIEW IDENTIFY SYSTEMS TO BE EVALUA TED e

e PROVIDE GUIDANCE FOR EVALUATION 4

l l

1 l7 241 l

i

g 2

e b

ONGOING PROGRAMS INITIATED OR AUTHORIZED IN 1985 WHICH WILL IMPROVE MAINTENANCE MAINTENANCE MANAGER SCHEDULING MANAGER MAINTENANCE PLANNERS MAINTENANct PM SUPERVISORS ADDITIONAL MAINTENANCE ENGINEERS AND SYSTEM ENGINEERS PM UPGRADES VALVE PM UPGRADE 0

4

+ -

,-,,g

--w

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

e----+-,

e---w-r---

SELECTION OF MANUAL VALVES FOR INSPECTION IMPORTANT TO OFF-NORMAL OPERATIONS NEEDED FOR MAINTENANCE OR ISOLATION 0F MAJOR PLANT EQUIPMENT O

4 e

.._-,..---,_.%4.

MANUAL VALVES INSPECTION e

SELECTION OF 143 VALVES TO BE INSPECTED:

IMPORTANT TO OFF-NORMAL OPERATIONS NEEDED FOR MAINTENANCE OR ISOLATION OF MAJOR PLANT EQUIPMENT STATUS:

1 129 INSPECTED 2 FAILURES AFTER VALVES MOVED TO h-DESIRED POSITION MAINTENANCE PERFORMED ON 8 OPERATORS 1 BENT STEM MINOR ADJUSTMENTS AND LUBRICATION DONE ON OTHER VALVES w

-e e---

y----

-3,--.m go.w--9 y,--m+t-- - - +

---

r-

--9, y

v t--v y-v g

MANUAL VALVES ADDITIONAL INSPECTIONS:

50 90* LIMITORQUE MANUAL VALVE OPERATORS l

50 OTHER MANUAL VALVE OPERATORS (NOT LIMITORQUE)

APPARENT PROBLEMS:

NON LIMITORQUE MANUAL OPERATORS LUBRICATION i

MANUAL LIMITORQUE OPERATORS LUBRICATION AND LIMIT STOP NUT ADJUSTMENT SEAL FAILURES WHICH ALLOW WATER LEAKS INTO LIMITORQUE OPERATORS OF VALVES LOCATED OUTDOORS b

ACTION:

ACCELERATE VALVE PM PROGRAM DEVELOPMENT STARTED IN 1985 ANALYZE RESULTS OF ADDITIONAL INSPECTIONS:

DETERMINE WHETHER MORE ACTION REQUIRED 1

l i

_.,._.__._.____.._._.__.___.._____,_.____._,_,___._,_..____-,______.._.._._._m,_.._.-.___...___,__.~_..

l SURVEILLANCE PROGRAM l

SAFETY RELATED AND ASME CODE VALVES COVERED BY INSERVICE INSPECTION PROGRAM SURVEILLANCE PROCEDURES ARE PART OF PREDICTIVE MAINTENANCE PROGRAM RECENT SURVEILLANCE PROGRAM IMPROVEMENTS:

DEVELOPED CROSS REFERENCE FOR TECHNICAL SPECIFICATION AND COPPIITENT LETTERS TO SURVEILLANCE PROCEDURES i

ISSUED SURVEILLANCE PROCEDURE WRITER'S GUIDE PREPARING SURVEILLANCE PROCEDURE REVIEW GUIDE FORMAL TRENDING PROGRAM IN PREPARATION WILL REVIEW MAINTENANCE AND SURVEILLANCE PROGRAMS AFTER PRECURSOR REVIEW 6

j p

x..

i i

\\

OPERATOR PERFORMANCE AND PROCEDURES Steve Redeker Operations Manager i

f Is s 71 E

l LOSS OF ICS/ RAPID COOLDOWN ABNORMAL TRANSIENT OPERATING GUIDELINE (ATOG) METHOD j

IS SOUND FOR LOSS OF ICS AND OVERCOOLING RANCHO SECO PROCEDURES CONFORMED TO ATOG TO TERMINATE OVERCOOLING CRITERIA WHEN TO TRIP FEEDWATER PUMPS TO TERMINATE OVERCOOLING HAVE BEEN DEVELOPED AND IMPLEMENTED CRITERIA FOR OTHER SYMPTOMS WERE IN PLACE ICS RECOVERY PROCEDURE HAS BEEN IMPLEMENTED MODIFICATIONS HAVE BEEN INCLUDED IN PROCEDURES l

LOCAL VALVE OPERATION PROCEDURES ARE BEING DEVELOPED USE OF VALVE WRENCHES ON CONTROL AND MOTOR OPERATED

+

VALVES IS PROHIBITED Is 5 al 9 '

e

~

g, n

MAKEUP PUMP j

i l

= EMERGENCY PROCEDURES REVISED TO l

ASSURE SUCTION IS AVAILABLE I

. RECOVERY FROM SFAS ACTUATION PROCEDUREIMPLEMENTED l

l l

i l6 5 9l

~

C

'l WATCHSTANDING IMPROVED DIVISION OF RESPONSIBILITIES BETWEEN SHIFT SUPERVISOR (nMERGENCY COORDINATOR) AND SENIOR CONTROL ROOM OPERATOR l

i GUIDANCE TO THE STA IS BEING STRENGTHENED STA AWARENESS OF PLANT STATUS l

l Fs51 1

1

j n'

m, t..,

HEALTH PHYSICS j

4

.I

• IMPROVED PROCEDURES AND CONDUCTED TRAINING FOR ENTRY INTO AREAS OF RADIOACTIVITY l

l I

l lesil

O t

PLANT TRIP RESPONSE TEAM i

i

. DOSE ASSESSMENT

. FOLLOWUP NOTIFICATIONS 1

. RESTART EVALUATION

. REPAIR ACTIVITIES l

l l

\\

le s el l

O

,l EMERGENCY PLAN

)

i lMPROVEMENT TO ASSURE USE OF PROPER FORMS l

CORRECTIONS LONG TERM EFFORT TO SIMPLIFY I

l I

I i

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DECEMBER 26, POST EVENT TRAINING e PRECURSORS TO THE EVENT AND CONCERNS ASSOCIA TED WITH

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OPERA TOR A CTIONS TAKEN.

i PROCEDURES TRAINING o EMERGENCY OPERA TING PROCEDURES i

l Q e LOSS /RESTORA TION OF ICS POWER e RECO VER Y FROM SFAS A CTUA TION e ENTRY INTO AREAS OF UNKNOWN RADIOLOGICAL CONDITIONS i

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ADDITIONAL TRAINING

• SIMULA TOR TRAINING

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• VAL VE TRAINING l

• CONDUCT OF SHIFT OPERA TION (TEAM TRAINING) i

• EMERGENCY PLAN TRAINING J

• REACTOR VESSEL HEAD BUBBLE io RECOGNITION AND RESPONSE l

• DIAGNOSTICS TRAINING

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• SHIFT TECHNICAL ADVISOR -

CONTROL ROOM ACTIVITIES i

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NEW TRAINING CENTER RANCHO SECO SIMULA TOR

e INCREASED TRAINING. PROGRAM QUALITY i

• INPO A ccreditation iC e Newly Developed Task-Based Training l

Ma terials j

e Newly Developed System Descriptions l

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• Increase Instructional Staff i

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• High Quality Training Aids And Mockups 4

i e Direct Feedback From Plant Persbnnel 1

l And First Line Supervision l

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CONTROL ROOM DESIGN REVIEW

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(CRDR) 1 PURPOSE:

i Required in Supplement 1 to NUREG 0 73 7 e

Identify improvements to increase plant safety l

l improve plant availability / reliability e

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CRDR S C OPE:

Main Control Room operating area e

SPDS displays and operator interface e

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j AREAS

• OF CONCERN:

I Non-compliance with human factors prin ciple s I

Deficiencies in information presented to

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the operator Adequacy of instrumentation and controls l

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to perform emergency operations 4

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9 CRDR A PPRO A CH:

Establish multidisciplinary re vie w t.eam and a review program Review operating experieryce including function and task analysis Static survey displays and controls l

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Verify Control Room Assess discrepancies I

Select design improvements that correct discrepancies Coordinate improvements with changes l

from other programs l7 0 5l l

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D O CUMEN TA TION:

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i SUBMITTED TO THE NRC ^FOR REVIEW:

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• Program Plan, September 19.8.3 i

i e Summary Report, December 1985 4

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• 516 Human Engineering Observations SAFETY SIGNIFICANCE:

o One-quarter have safety significance e None have such a high safety importance that a non-recoverable error would occur IMPL EMENTA TION:

i e Priority assigned to reflect the safety significance of the change 1

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• In accordance with the District's Living i

Schedule

• Over 15,000 engineering man-hours have been budgeted in 1986 l708]

WITH 12/26/85 TR ANSIENT:

A NAL YSIS:

Assessed the operational usefulness of the l

I Control Room and operational concerns i

This evaluation discovered 26 items OBSER VA TIONS:

11 items had previously been identified l

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8 observations dealt with items outside of I

the Control Room, not within the CRDR scope

-7 were operational or procedural concerns not within the CRDR scope 17 l

INVOLOMENT

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WITH 12/26/85 TRANSIENT l

IMPLEMENTED MODIFICA TIONS:

Relabel breaker switch S US2 Review labelling of the ICS power loss annunciation l

Install simplified ICS schematic plaque on ICS i

power cabinets i

i Revise hand operator position Indication of AFW i

valves FV-2052 7 and FV-20528 i

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Demarcate control valves needed to isolate a steam generator Add additional controls for the T8V, ADV, and AFW control valves i

l Replace Radiation Multipoint Recorder l

Relabel SFAS valves to identify additional

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channels Modify EOP's to improve clarity i

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Y ONGOING HUMAN FACTORS PROGRAM:

FUTURE CHANGES TO PL ANT:

e Review modifications for Human Factors concerns OBSERVED HUMAN FA CTORS PROBLEMS:

• Investigate Human Factors problems reported in the plant OPERA TING EXPERIENCE REVIEW (INCLUDING POST-TRANSIENT ANAL YSIS):

• Perform a post-transient analysis of transients to uncover problems

• Interview operators for plant / operator concerns e

17 : 21 W

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