ML18152A013
| ML18152A013 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna, 05000000 |
| Issue date: | 02/05/1988 |
| From: | Cantrell F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152A014 | List: |
| References | |
| 50-280-86-12-EC, 50-281-86-12, 50-338-87-32, 50-338-87-38, 50-339-87-32, 50-339-87-38, NUDOCS 8802250416 | |
| Download: ML18152A013 (78) | |
See also: IR 05000280/1986012
Text
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ENCLOSURE 1
ENF.0.RC,EMENJ. CONEERENCE .S.UMMARY
- Licensee:
Virginia Electric and Power Company
Docket Nos.:
50-280, 50-281, 50-338, and 50~339
License Nos.:
DPR-32, DPR-37, NPF-4, and NPF-7
Facility Name:
Approved By:
Surry and North Anna
~Chief
SUMMARY
"'/S~'i ~
Date 1gned
Scope:
An Enforcement Conference was held in Region II on January 21, 1988.
Mr. M. L. Ernst opened the meeting by expressing concern with a number of post
outage problems at North Anna as documented in several Inspection Reports in
.1987.
Virginia Electric and Power *Company (VEPCO) . then made a presentation
covering the failure to place the inoperable steam flow instruments in trip
within the required time and-recent problems at North. Anna. 'VEPCO then covered
the equipment qualification problems at Surry and North Anna.
Enclosure 3
contains information from the presentations.
- Results:. The results of the NRC findings in. this area will be forwarded under
separate*cover *
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- Enclosure 1
2
ENFORCEMENT CONFERENCE DETAILS
1.
Attendees - see Enclosure 2
2.
Enforcement Conference
Mr. M. L. Ernst opened the meeting by expressing concern with a number of
of post outage problems at North Anna as documented in several Inspection
Reports in 1987.
Mr. J. Wilson of Virginia Electric and Power Company
(VEPCO) ~ave a brief description of the VEPCO position including concern
with the lack of sensitivity by operations and management on making timely
decisions for non-conforming conditions. Mr. M. L. Bowling then presented
the background, root cause, description of events, safety considerations
and corrective actions associated with the North Anna Unit 2 inoperable
steam flow instrumentation.
Mr. Bowling then discussed the summary of
events, root causes, and management plan associated with the 1987
- -,;refue-li:ng ,:outage and-:startup problems at North-,Anna. *
Equipment Qualification (EQ) problems identified at Surry and documented
in Inspection Report
280/86-12 and 281/86-12 were then discussed by
Mr. E. S. Grecheck.
Mr. Bowling then discussed the EQ problems identified
at North Anna and documented in Inspection Report 338/87-32 and 339/87-32.
With regard to the failure to place the inoperable steam flow instruments
in trip with in the required time, the licensee stated.that management
tolerance of steam flow indication problems at low steam flows contributed
to the violation.
Technical Specifications (TS) require inoperable steam
flow channels be placed in trip within one hour.
However, at
approximately 1200 on November 4, 1987, two steam flow channels were found
to be inoperable and were not placed in trip. Plant shutdown was started
at 10:00 p.m., ten hours later.
The 1 icensee * stated that the safety
sig.nificance of this event was low since redundant channels and signals: *
. wete aVailabl'e.
Co_rrective actions include _specifying channel *check
"requ'irements and *maxiinum *-power levels for *ins:trument *res*ponse.
The
licensee also committed to revising Licensee Event Report CLER)87-015 on.
this event due to numerous deficiencies in the report.
The 1 icensee s~ated that the root causes for the problems encountered
during the 1987 outages* and startups were due in part to
management/supervisor failure to enforce standards, failure to follow
procedures, and inadequate procedures.
Corrective actions include
strengthening accountability, requiring safety committee review prior to
changing procedure initial conditions, and increasing management
involvement.
3.
Conclusions
"Jh~JfiH: .,~b~res __ :,V{PCO .conc.erns .. with respect to management :tolerance of
- non-conforming conditions, failure to follow procedures, and inadequate
procedure.
Future inspections wi 11 be conducted to monitor performance
in these areas. * We are continuing our review of these issues and *our
conclusions will be forwarded under separate cover.
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ENCLOSURE 2
Enforcement Conference Attendees
Virginia Electric and Power Company
J. L. Wilson, Manager, Nuclear Operations Support
M. L. Bowling, Assistant Station Managert North Anna
E. S. Grecheck, Assistant Station Manager, Surry
N. E. Hardwick, Manager, Nuclear Power and Licensing, Corporate
G. l. Pannell, Director, SEC
R. W. Calder, Manager of Nuclear Engineering
R. Mo Kritch, Licensing Engineer
J. E. McDonald, Public Affairs Coordinator
E. T. Shaub, Licensing Engineer
P. T. Knutsen, Supervisor, Nuclear Engineering
Nucl~ar Regulatory Convnission
M. L. Ernst, Deputy Regional Administrator
C. W. Hehl, Deputy Division Director, Division of Reactor Projects (DRP)
G. R. Jenkins, Director, Enforcement and Investigations Coordination Staff
B. A. Wilson, Branch Chief, ORP
F. S. Cantrell,Section Chief, DRP
F. Jape, Section Chief, Test Programs, Division of Reactor Safety (DRS)
R. P. Croteau, Project Engineer, DRP
W. E. Holland, Senior Resident Inspector, Surry
J. L. Caldwell, Senior Resident Inspector, North Anna
L. B. Engle, Project Manager, Nuclear Reactor Regulation (NRR)
L. P. King, Resident Inspector, North Anna
B. Uryc, Enforcement Coordinator
M.A. Scott, Project Engineer, DRP
C. J. **Paulk, Reactor Inspector, ORS
C. F. Smith, Reactor Inspector, DRS
U. Potapovs, Section Chief, NRR
T. E. Conlon, Section Chief, DRS
E. W. Merschoff, Deputy Director, DRS
A. J. Szczepaniec, Reactor Inspector, DRS
W. Levis, Reactor Inspector, DRS
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ENCLOSURE 1
ENFORCEMENT CONFERENCE SUMMARY
Licensee:
Virginia Electric and Power Company
Docket Nos.:
50-280, 50-281, 50-338, and 50-339
License Nos.:
DPR-32, DPR-37, NPF-4, and NPF-7
Facility Name:
Approved By:
Surry and North Anna
~Chief
SUMMARY
~!sf,}/~
- Date 1gned
Scope:
An Enforcement Conference was held in Region II on January 21, 1988.
Mr. M. L. Ernst opened ~he meeting by expressing concern with a number of post
outage problems at North Anna as documented in several Inspection Reports in
1987.
Virginia Electric and Power Company (VEPCO) then made a presentation
covering the failure to place the inoperable steam flow instruments in trip
within the required time and recent problems at North Anna.
VEPCO then covered
the equipment qualification problems at Surry and North Anna.
Enclosure 3
contains information from the presentations.
Results:
The results of the NRC findings in this area will be forwarded under
separate cover.
. >'..
- -;*** . .,. .- '
~ .**
, Enclosure 1
2
ENFORCEMENT CONFERENCE DETAILS
1.
Attendees - see Enclosure 2
2.
Enforcement Conference
Mr. M. L. Ernst opened the meeting by expressing concern with a number of
of post outage problems at North Anna as documented in several Inspection
Reports in 1987.
Mr. J. Wi 1 son * of Vi rgi ni a Electric and Power Company
(VEPCO) gave a brief description of the VEPCO position including concern
with the lack of sensitivity by operations and management on making timely
decisions for non-confonning conditi.ons.
Mr. M. L. Bowling then presented
the background, rocit cause, description of events, safety considerations
and corrective actions associated with the North Anna Unit 2 inoperable
steam flow instrumentation.
Mr. Bowling then discussed the summary of
events, root causes, and management plan associated with the 1987
refueling outage and startup problems at North Anna.
Equipment Qualification (EQ) *problems *identified at Surry and documented
in Inspection Report
280/86-12 and 281/86-12 were then discussed by
Mr. E. S. Grecheck.
Mr. Bowling then discussed the EQ problems identified
at North Anna and documented in Inspection Report 338/87-32 and 339/87-32.
With regard to the failure to place the inoperable steam flow instruments
in trip with in the required time, the licensee stated that management
tolerance of steam flow indication problems at low steam flows contributed
to the violation.
Technical Specificati-ons (TS) require inoperable steam
flow channels be placed in trip within one hour..
However, at
approximately 1200 on November 4, 1987, two steam flow channels were found
to be inoperable and were not placed in trip. Plant shutdown was started
at 10: 00 p.m., ten hours later.
The 1 icensee * stated that the safety
significance of this event was low since redundant channels and signals*
- were available.
Corrective actions include specifyin*g channel check
requirements and maximum power levels for instrument response.
The
licensee also committed to revising Licensee Event Report (LER)87-015 on
this event due to numerous deficiencies in the report.
The licensee stated that the root causes for the problems encountered
during the 1987 outages and startups were .due in part to
management/supervisor failure to enforce standards, failure to follow
procedures, and inadequate procedures.
Corrective actions include
strengthening accountability; requiring safety committee review prior to
changing procedure initial conditions, and increasing management
- involvement.
3.
Conclusions
... * .. -**.
The NRC sha*res VEPCO :Concerns w'lth respect to management tolerance of
non-conforming conditions, failure to follow procedures, and inadequate
procedure.
Future inspections will be conducted to monitor performance
in these areas *. We are continuing our review of these issues and our
conclusions will,be forwarded under separate cover *
.
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ENCLOSURE 2
Enforcement Conference Attendees
Virginia Electric and Power Company
J .* L. Wi.lson, Manager, -Nuclear Operations Support
M. L. Bowling, Assistant Station Manager, North Anna
E. S. Grecheck, Assistant Station Manager, Surry
N. E. Hardwick, Manager, Nuclear Power.and Licensing, Corporate
G. L. Pannell, Director, SEC
R. W. Calder, Manager of Nuclear Engineering
R. M. Kritch, Licensing Engineer
J. E. McDonald, Public Affairs Coordinator
E.T. Shaub, Licensing Engineer
P. T. Knutsen, Supervisor, Nuclear Engineering
Nuclear Regulatory Commission
M. L. Ernst, Deputy Regional Administrator
C. W. Hehl, Deputy Div-ision Director, Division of Reactor Projects (DRP)
G. R. Jenkins, Director, Enforcement and Investigations Coordination Staff
B. A. Wilson, Branch Chief, DRP
F. S. Cantrell, Section Chief, DRP
F. Jape, Section Chief, Test Programs, Division of Reactor Safety (DRS)
R. P. Croteau, Project Engineer, DRP
W. E. Holland, Senior Resident Inspector, Surry
J. L. Caldwell, Senior Resident Inspector, North Anna
L. B *. Engle, Project Manager, Nuclear Reactor Regulation (NRR)
L. P. King, Resident Inspector, North Anna
B. Uryc, Enforcement Coordinator
M.A. Scott~ Project Engineer, DRP
C. J. Paulk, Reactor Inspector, DRS
C. F. Smith, Reactor Inspector, DRS
U. Potapovs, Section Chief, NRR
T. E. Conlon, Section Chief, DRS
E.W. Mer.schoff, Deputy Director, DRS
A. J. Szczepaniec, Reactor Inspector, DRS
W. Levis, Reactor Inspector, DRS
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ENCLOSURE 3*
NRC Enforcement Conference
North Anna 2 Inoperable Steam
Flow Instrumentation
Agenda
- I. Background
U. Root Cause
UL Description of Event
IV. Safety Considerations
V:' Corrective Actions
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Purposes of Steam Flow Instrumentation
-
Control Room Indication
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Steam Generator Level Control System
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Reactor Protection
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Engineered Safety Features
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Technical Spec*ifications Requirements
Reactor Protection
-
Steam/FeedWater Flow" Mismatch Coincident with tow
Safety Injection and'."
Steam Generator Level ih Any One Loop
- * Two Channels per Steam Line, One Channel to Trip
-
Minimum Operable Channels -
One per Steam line
-
Action for Inoperable Channel -
Place in Trip within 1
Hour
-
Channel Check each 12 Hours, Channel Functional Test
Monthly, Channel Calibration each 18 Months
Steam Line isolation -
High Steam Flow in Two Steam Lines Coincident with
Either Low-Low* TAVG or Low Steam Line Pressure in
Any Two Loops
-
- Two Channels per Steam Line, One Channel to Trip
-
Minimum Operable Channels -
One per Steam Line
-
Action for Inoperable Channel -
Place in Trip within 1
Hour
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Channel Check each 12 Hours, Channel Functional Test
Monthly, Channel Calibration each 18 Months
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Technical Specifications Requirem:ents
Specification 3.0.3
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When LCO is Not Met, Except as Provided in Action
- Requirements, within One Hour Initiate Action to Place
Unit in Mode in Which Specification Does Not Apply
Channel Check
-
A Qualitative Assessment of Channel Behavior During
Operation by Observation
Shall Include, Where Possible, Comparisons with
Independent Channels Measuring Same Parameter
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Figure 21-l
HAIN STE.AH SYSTEH (SHEt:r l)
REV 01/ll/86
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VENTURI STEAM FLONC-*
INDICATION
STEAM FLOW VERSUS VENTURI DIFFERENTIAL PRESSURE
100 _________
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FLOW X 10s LBM/HR
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STEAM FLOW INDICATION ERROR VERSUS REACTOR POWER.
2
....
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ERROR ~S CORRESPOND TO
-
E
I INCH TRANSMITTER/VENTURI ERROR
ZERO ERROR
.c
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MAX ERROR*29 INCHES
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MAXIMUM ALLOWABLE ERROR
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10
15
20
25
30
35
40
I
REACTOR POWER, O/o
.
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Root Cause
Personnel Error
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Inadequate Post-Maintenance Testing
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Untimely Operator Action
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Inadequate Use of Alternate Indications
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Management Tolerance * of Steam Flow Indication
Problems at Low Steam Flows
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Reasons for Untimely Operator Action
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Attention Directed at Delta T/TAVG Instrument Failure
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Influenced by Past Experience with Steam Flow Indication at Low Steam Flows
-
Time Required to Make Containment Entry
-
Concern Over Prevention of an Unnecessary Safety Injection When Placing Failed
Channels in Trip
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Steam Flow Indication from ERF Computer
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,. Sequence of Events
8/29/87
9/11/87
9/14/87
9/17/87
9/19/87
9/21/87
Steam Generator A Steam Flow Channel Ill (FT-2474) was
Calibrated within Specified Accuracies *
-Steam Gen*erator "A" Steam Flow Channel Ill (fl-2474) was
Removed from Service for Raychem Repair
Raychem Repair Completed for FT-2474
Tagging Record was Written for Approximately 40 Instruments
The Circuits were Tagged and Deenergized at the Loop Power
Supply Cards
The Tagging Record Indicates That Card at Location C1-221 was
Tagged for FT-2474
This is Actually the Card for LT-2474
The Reverse Error Occurred for LT-2474
Card C3-425 was Tagged for LT-2474, but is Actually the Card for
FT-2474
Raychem Repair Completed for LT-2474
C1-221 Tag: Removed and Card Inserted to Energize FT-2474
LT-2474 w~s Energized Instead*
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Sequence of Events
9/24/87*
9/29/87
9/29/87*
10/08/87
10/25/87
Post-Maintenance Test of FT-2474 Performed by Reading Voltage
Output at Card location C1-221
Test was Satisfactory Because LT-2474 was Actually Tested
C3-425 Tag Removed and Card Inserted to Energize LT-2474
FT-2474 was Energized Instead
Post-Maintenance Test of LT-2474 Believed to have Been Performed
by Verifying Level in Control Room and Found Satisfactory
EWR 87-206 Documentation for FT-2474 Resigned and Redated to
Meet Documentation Requirements
Maintenance Procedure for Verifying RPS Transmitter Operability
Prior tb Startup was Performed with the Unit in Mode 5
Channel 'Check Could Not be Performed Because Steam How was
Zero
Therefore, in Accordance with Procedure, the Valve Lineup was
Verified to be Correct to Establish Operability for the Mode
Change
- Approximate '
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Sequence of Events
. 11/04/87:
1200*
- cRO Log Readings for Steam Generator Steam Flows Indicated that
Steam Generator A Steam Flow Channel m and Steam
Generator "B" Steam Flow Channel IV were Reading Outside of
the Acc'eptance Criteria
CRO Log flagged Problem with Steam Generator Steam Flow
Channels as Being Due to Steam Flow Conditions at Less Than
5010 Power
1719
Entered Mode 1
1754
1816
1838
1845*
- App_roximate
Unit at 24°/o Power
Steam Generator A Steam Flow Channel ill and Steam Generator
B Steam Flow Channel IV Observed Reading Zero flow
B Loop Delta T/TAVG Channel II Observed Reading 10010 Low and
Declared Inoperable
Entered AP-3
. SRO-on-call Departed Plant After Briefing by Shift Supervisor on
Instrumentation Problems Believing that Steam Flow Channels
Would Indicate with Higher Steam Flow
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POWER ANO STEAM FLOWS FROM STRIP CHARTS
POWER
40%
0
SIG C
STEAM FLOW
20%
0
SIG B
STEAM FLOW
20%
0
SIG A
STEAM FLOW
- 20%
0
2
3
4
5
a
7
8
TIME AFTER ENTERING MOOE 1, HOURS
..,;
J2 "A" S/G
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0000
224'3*"
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llOO
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100 ..
---...-----;...------.~-...._..., _______ :s I .. .. i
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Li2 ;.'B" ;S/G
19lS
I.: 80
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Steam/P<<;?ediPlqw
Level
%
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,___, -4
10
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10
% Power
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Sequence of Events
11/04/87:
1900
Power Stabilized at 26-27DAJ
1904
Instrument Techs_ Started Calibration of B: Loop Delta T/TAVG
Channel I
1915
B Loop Delta T/TAVG Channel II Placed in Trip
2000*
CRO Log Readings for Steam Generator Steam Flows Indicated that
2009-
both Steam Generator A Steam Flow Channel Ill and Steam
Generator "B" Channel IV were Indicating Zero and Not
within the Channel Check Acceptance Criteria
CRO Log Flagged Problems as Being Under Investigation
2041
SRO On-call had Several Discussions on Plant Status with Shift
Supervisor
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. Sequence of Events
11/04/87:
2048
Assistant Station Manager {O&M), Who was the Acting Station
Manager, Notified of Instrument Failures by SRO On-call
2054
SRO On-call Advised Shift Supervisor of the Need to Declare Steam
Flow Channels Inoperable, to Enter T/S 3.0.3 and the EPIPs, and
to Enter Containment to Check the Transmitters
2120
SRO dn.-call Confirmed with Shift Supervisor the Need to Enter IS.
3.0.3
2122
Acting Station Manager Notified Corportation Management of Plant
Status
2130*
Assistant Station Manager {NS&L) Notified by STA that Two Steam
- Approximate
Flow Channels were Not Indicating but Could Not be Placed in
Trip per _the AP
AP Required Entry into T.S. 3.0.3
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Sequence of Events
11/04/87:
2139
SRO On-call Notified Acting Station Manager that Steam Flow
Channels will be Declared Inoperable and Unit will be Declared
Inoperable and Unit will be Ramped Off Line IAW T.S. 3.0.3
2150*
Assistant Station Manager (NS&L) Notified by STA that T.S. 3.0.3
was to be Entered
2150
Acting Station Manager Notified Corporate Management of Planned
Shutdown and Entering EPIP
2153
Steam Flow Channels Declared Inoperable
Although Te.ch. Specs. Permit STARTUP and POWER
OPERATION to continue Provided the Inoperable Channel are
Placed in : Trip, AP-3. 7 Requires Action within 1 Hour to Place the
Unit in HOT STANDBY within the Next 6 Hours
2155
Acting Station Manager Discussed Instrument Problems with
Superintendent of Tech. Services
- Approximate
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Sequence of Events
11/04/87:
2200
2200
Commenced Unit Shutdown
Instrument Department Began Analyzing Reactor Protection and
ESF Logic to Determine if "A" or "B" Steam Flow Channels
Could be Placed in Trip with B Loop Delta T/TAVG Channel II
Already in Trip
2201
Operations Entered Containment to Attempt to Check Steam Flow
Transmitters and Valve Lineups
2205*
Assistant Station Manager (NS&L) Verified with Unit 2 SRO that
Requirements of T.S. 3.0.3 are Being Met and the Emergency
Plan has Been Entered
2207*
Assistant Station Manager (NS&L) Advised NRC Resident Inspector,
Who was in Control Room, of Planned Action
2210*
Assistant Station Manager (NS&l) Notified NRC Senior Resident
2215
2216
2240
2345*
- Approximate
Inspector of Plant Status
Steam Generator B Steam Flow Channel IV Returned to Service
State and Local Governments Notified of Unusual Event
NRC Notified of Unusual Event
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Station Nuclear Safety and. Operating Committee Initiated Telephone
Conference to Review Deviation to AP 3.7 : . i
Safety Commitee Included Acting Station Manager
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Sequence of Events
11/05/87:
0012
0015*
0030*.
0220
0539
0800
12/04/87
12/10/87*
12/17/87
12/18/87
- Approximate
Steam Generator A Steam Flow Channel Ill Placed in Trip
A Temporary Change was Made to AP-3. 7 Allowing this Channel to
be in Trip at the Same Time as "B" Loop Delta T/TAVG Channel
II Shutdown Terminated
Assistant Station Manager (NS&L) Briefed NRC Senior Resident
Inspector on Plant Status
Assistant Station Manager (NS&L) Reviewed T.S. Requirements with
Shift Supervisor to Ensure that Technica*1 Specifications were
Being Complied with
Steam Generator A Steam Flow Channel Ill Returned to Service
"B" Loop .Delta T/TAVG Channel II Returned to Service
Assistant Station Manager (NS&L) Met with NRC Senior Resident
Inspector to Review T.S. and AP-3
LER Submitted
Data from Most Recent Startup was Obtained and an Assessment
. of Steam flow Inaccuracies. at Low Power was Initiated .
NRC Resident Inspectors Met with Supervisor, NSE and then *.
Station Manager on the Steam Flow Issue *
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NRC Resident Inspector Exit Meeting and Notification by NRC of
Potential Violation and Enforcement
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Safety Considerations
Reactor Protection and ESF Logics
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Reactor Protection
-
1 of 2 Steam/Feedwater Flow Mismatch (40°/o of Full
Safety Injection .
Steam Flow) Coincident with 1 of 2 Steam Generator
Low Level (250/o Narrow Range Span)
-
No Credit Taken for this Trip in USFAR
-
1 of 2 High Steam Flow in 2 of 3 Steam Lines
Coincident with Either Low-Low TAVG (543°F) in Any
Two Loops or Low Steam Pressure (600 PSIG) in Any
Two Loops
-
2 of 3 High Differential Pressure (100 PSI) Between One
Steam Line and Each of the Other Two Lines
-
2 of 3 High Containment Pressure (17 PSIA)
-
2 of 3 Low-Low Pressurizer Pressure (1765 PSIG)
-
Manual Initiation
Steam Line Isolation -
1 of 2 High Steam Flow in 2 of 3 Steam Lines
Coincident with Ei.ther Low-Low TAVG (543°F) in Any
Two ~oops or. Low Steam Pressure (600 PSIG) in Any
Two Loops
-
2. of, ~ lntermeqiqte High-High Containment Pressure
(17.8 PSIA)
'
. '. .'
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0
., .
)>
0
I\\)
0
% OF FULL STEAM{Flow
O>
0
0)
0
....
....
0
....
0
0
o-.----------l'-------+------~~-------------1-__.
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0
~
0
m
0
0)
0
...
0
0
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Reactor Protection
No Credit Taken in Safety Analysis for Reactor Trip on Lo Steam Generator Level in
Coincidence with Steam Flow Greater than Feedflow
Loss of Heat Sink Protection Provided by Steam Generator Lo Lo Level Trip
- ,
Safety Analysis
Scope and
Assumptions
-
Assumed an Inoperable Steam Flow Indicator on Steam
Generators A and B
-
Con*sidered Complete Spectrum of Steam Une Breaks
Inside and Outside Containment
-
Evaluated Impact On:
Steam Line Isolation and Safety Injection
Redundancy
Emergency Operating Procedures
Appendix R Safe Shutdown Procedures
Post Accident Monitoring Instrumentation
Core Shutdown Margin
- As sump ti on s:
SCENARIO NO.
TABLE 1
STEAM LINE ISLOLATION AND SAFETY INJECTION
REDUNDANCY EVALUATION - HIGH STEAM FLOW CHANNELS
The Fo 11 owing, '.Stearn flow
Channels
Inoper~ble: Loop A/Channel 3 (LAC3)
Loop B/Channel 4 (LBC4)
BREAK .LOCATION
SOURCE OF PROTECTION
ADDITIONAL
FAILURES
TO LOSE
PROTECTION
~------
1.
Loop A, Upstream of
Loop A Nonreturn
Failure of
Flow Transmitter (FT) . Valve (NRV)* OR
NRV ANO
(4.6 Sq. Ft. Max)
LBC3/LCC3 OR LBC3/LCC4
LBCJ AND
OR Hi-2 Containment
Hi-2 CP
- Pressure {Some delay)
2.
Loop A,*Between FT
Loop A NRV
Failure of
and NRV (1.4 Sq. Ft.)
OR LAC4/LBC3
NRV
OR LAC4/LCC3
ANO LAC4
OR LAC4/LCC4
AND LBC3
OR LBC3/LCC3
AND Hi-2CP
OR LBC3/LCC4
OR Hi-2CP (Some
delay)
3.
Loop B, Upstream of
Loop B Nonreturn
Failure of
Flow Transmitter (FT)
Valve (NRV) OR
NRV AND
(4.6 Sq. Ft. Max)
LAC4/LCC3 OR LAC4/LCC4
LAC4 AND
OR Hi-2 Containment
Hi-2 CP
Pressure.(S~me delay)
4.
Loop B, Between FT
Loop B NRV
Failure of
and NRV (1.4 Sq. Ft.)
OR LAC4/LBC3 *
- NRV
OR LAC4/LCC3 *
- AND LAC4
OR LAC4/LCC4
AND _LBC3,
OR LBC3/LCC3
ANO Hi-2CP
OR LBC3/LCC4
OR Hi-2CP (Some
delay)
- If NRV functions as designed, then a Safety Injection Signal is
Generated on High Steam Line Delta-P .
. .. ;:;,
.
. *~ .... ' ..
SCENARIO NO.
TABLE 1 (CONT.)
STEAM LINE ISLOLATION ANO SAFETY INJECTION
REDUNDANCY .. EVALUATION 7" HIGH STEAM. FLOW CHANNELS
BREAK LOCATION
SOURCE OF PROTECTION
ADDITIONAL
FAILURES
TO LOSE
PROTECTION
*--~--------------------~-----------~---~---------
5.
Loop C, Upstream of
Loop C Nonreturn
Failure of
Flow Transmitter (FT)
Valve (NRV) OR
NRV ANO
(4.6 Sq. Ft. Max)
LAC4/LBC3
LAC4 ANO
OR Hi-2 Containment
Hi-2 CP
Pressure (Some delay)
6.
Loop C ,. Between FT
Loop C NRV
Failure of
. and NRV (1.4 Sq. Ft.)
OR LAC4/LBC3
NRV
OR LAC4/LCC3
ANO LAC4
OR LAC4/LCC4
ANO LBC3
OR LBC3/LCC3
AND Hi-2CP
OR LBC3/LCC4
OR Hi-2CP (Some
delay)
7.
Inside Containment,
LAC4/LBC3
Failure of *.
Outside NRV's
OR LAC4/LCC3
LAC4
(1.4 Sq. *Ft./Loop)
OR L.AC4/LCC4
ANO LBC3
OR LBC3/LCC3
AND Hi-2 CP
OR LBC3/LCC4
OR Hi-2 Cont. Press.
8.
Outside Containment
LAC4/LBC3
Failure of
Outside NRV's
OR*LAC4/LCC3
LAC4
(1.4 Sq. Ft./loo~~
OR .LAC4/LCC4
ANO LBC3
OR LBC3/LCC3
OR LBC3/LCC4
- - ...
-
.:*,
"i': .
. ,
To Safety Valves.Atmospheric Steam Dumps,
and Decay Heat Release Valve
--EH
Flow Restrictor
From S/G B
From S/G C
NRV
NRV
NRV
To Auxiliary Steam System
J
To HP Turbine Throttle Valves
To Steam Dumps, Moisture Separator Reheaters,
and Gland Steam Regulator (Sheet 2)
Safety Analysis .
Conclusions
-
No Loss of Safety Function, Even Assuming an
Additional Single Failure
-
Diverse Means (Other than Steam Flow) of Initiating
Safety Injection and Steam line Isolation were Available
-
No Post.:rrip Return to Power Would Have Occurred for
Core Conditions, at Time of Event
-
Adequate Information was Available to the Operator for
Post-Accident Monitoring, Assessment and Management
as Defined in EOPs
-
Event Bounded by UFSAR Safety Analysis
Safety Committee Reviewed Safety Analysis
and Approved Conclusions
. '\\
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ID
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co
0 ,-
X
3: g
u.
C w 5
C z -
ERROR PROPAGATION IN ACTUAL VERSUS INDICATED STEAM FLOW
2
.5
0
.25
.5
8
MAXIMUM INDICATION ERROR 29 INCHES
- VENTURI ERROR 8 INCHES
.75
18
1
30
ACTUAL FLOW X 106 LBM/HR
2
.. *~*-*. -
_ . .,..,., ___ ._,.,
I
I
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Corrective Actions
Issue Operations
Directive
-
Maximum Power Level for Instrument Response
Enhance Training
Identified *
-
Use of Alternate Indication Emphasized
. -
Channel Check Requirements Specified
-
Review in LORP
-
Incorporate Instrument Failure/Indication Problems to
Simulator Startup Scenarios
-
Emphasize Use of Alternative Indications in Simulator
Scenarios
I .
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- "
'
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Corrective Actions
Review and Upgrade Post Maintenance Test Procedure Development
and Documentation Criteria
Inspect Transmitter Wiring at Next Available Outage
Prepare Revised LEA
Evaluate and Upgrade Instrument Calibration Procedures as Required
to Better Insure Transmitter Operability Prior to Entering Mode 3
Present and Disclass Event with Station Supervisors, and Address
Management Concern with Tolerance of Non-Conforming Conditions
Conclusions
Operations Directive Will Provide Assurance That Steam Flow
Instruments are Capable of Performing Their Intended Safety
Functions
- -
Based on 10 CFR 50.59 Evaluation Reactor Protection and Engineered
Safety Feature Actuations Would Have Occurred as Required Even
Considering an Additional Single Active Failure of a Steam Flow
Instrument
Corrective Actions will be Adequate to Prevent Recurrence
\\ .. :, ... , ..
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NRC Questions
1. Why were the steam flow channels not declared inoperable when they failed the
Tech. Spec. channel 'Check?
2. Why did operation continue at about 21010 power for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
with both steam flow channels indicating zero before declaring the channels
3. Why were the post-maintenance test data sheets not properly signed off?
4. What actions are beihg taken to obtain more reliable steam *flow indication at low
powers?
5. Are the steam flow channels capable of performing their intended safety function in
Modes 1, 2 and 3?
6. What corrective action was taken with regards to flow indicator 2485 being
inoperable but not placed in trip?
7. Was management * aV'!'are of the fact that the steam flow channels were not
indicating properly ahd not placed in trip vriolation of Tech. Specs.?
i'~
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ENCLOSURE 3 .
1987 Refueling Outage and Startup Events
- Agenda
I. Summary of Events
11. Root Causes
111. Management Plan
-
-_. :*-* __ *
I
-1
- (_.'
1987 Refueling Outage and Startup Events
6/21/87
Loss of RCS Inventory (Unit 1)
6/29/87
Trip on High Level in #5 Feedwater Heater (Unit 1)
7/11/87
Shutdown Due to Inoperable MSR Stop Valve (Unit 1)
10/22/87
Accumulator Injection into RCS (Unit 2)
10/26/87
Inadvertently Aligned Charging Pump to RCS Causing PORV Lift
(Unit 2)
10/30/87
Inadvertent Cooldown to Below TIS Limit (Unit 2)
i I,
..
..
Root Causes
Personnel
Procedure
-
Failure to Use Multiple or Alternate Indications
Failure to Follow Procedure
Failure to Meet or Maintain Initial Conditions
Failure to Adhere to RWPs
Failure to Use Appropriate Procedure
Personnel Error
-
Management/Supervisor Failure to Enforce Standards
-
Lack of Procedure
-
Inadequate Procedure
'
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Management Plan
-
Procedures
-
Personnel Performance
-* Management Involvement
- ,
'
,.;
Procedures
Require Safety Cd,mmittee Review and Approval Prior to Use for Any
Deviation to a Prcicedure Initial Condition
Continue Assignnient of Licensed RO from Operations Procedure
Group to PT and ISi Groups for Surveillance Testing Development and
Procedure Revisioh:
Review Current Sqrveillance Procedure Development
Enhance Review Process for Determining Applicability of PTs for Use *
in Post-Maintenance and Post-Modification Testing
Oevel'.op and lmpl.ment Mode Change Reactor Operator Checklists to
Verify Required T.S. Equipment Status Prior to Each Mode Change
f ,t;'
. , .
. ,,
Personnel Performance*
Continue HPES Program
Implement "Coaching" Concept
Strengthening Supervisor and Employee Accountability
Ensure Employees U.nderstand Significance of Events and are Briefed
on Root Causes and; Corrective Actions
-,.,,
Management Involvement
Improve Scheduling ofi' Operability Testing Prior to Startup After Major
Outage
- strengthen Management Involvement in Operability Testing Prior to
- startup After a Major Outage by Assignment of Dedicated Operations
Co9rdinator(s)
Continue Managem*ent Walkdowns and Safety Committee Review of
Deficiencies Prior to Startup After a Major Outage. Extend Walkdown
Concept to Maintenance Personnel
Continue Station and Corporate Management Review of Significant
Events with Vice President -
Nuclear Operations
Revise Station Deviation Report Program to Augment Root Cause
Determination
.. r-*lll.
_j
ENCLOSURE 3
Agenda
EQ Enforcement Conference
January 21, 1988
Introduction
EQ Program
Enforcement Items
-
Surry
-
North Anna
EQ Enforcement Summary
.,
f
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Virginia Power
Equipment Qualification Program
, r *
.. '
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Virginia Power -
Equipment Qualification Program
Program Covers the
Principle Areas of
-
Policy, Standards and Procedures
-
Assignment of Responsibilities
-
Development and Maintenance of Equipment List
(EQ Master List -
EQML) and Qualification files
(Qualification Documentation Reviews -
QDRs)
-
Procurement Process
-
Modifications Affecting/Involving EQ
-
EQ Training
-
Involvement of QA/QC
-
Maintenance of Equipment on EQML
-
Review and Incorporation of Regulatory and Industry EQ
Information
"*
f
-
J:
_ Virginia Power :-- Equipment Qua~ification Program
Program Objective Is
to Ensure that
.;._ Equipment Required fiy 10CFR50.49 to be Qualified is
'I
Identified on the EQML
-
Qualification Files (QDRs) Contain the Necessary
Information to Clearly Demonstrate Qualification in
Accordance with 10CFH50.49
-
Equipment is* Procured and Maintained in Accordance
with Procedures which Incorporate EQ Requirements as
Appropriate
-
Modifications and Engineering are Controlled Such that
Equipment Qualification is Properly Maintained
-
Relevant Information Provided by the NRC and Industry
is Evalu~ted and Incorporated, as Appropriate, in a
Timely* Manner
.1* ,{
- ,
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Surry and North Anna
Enforcement Items
.. '1
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- t*
Surry EQ Inspection Results
June 16-20, 1986
-
Concluded that Program has been Implemented
which Meets Requirements of 10CFR50.49
-
No Deficiencies Identified in Implementation of
SER/TER Corrective Action Commitments
-
Six Potential Enforcement/Unresolved Items
' .
Potential Enforcement Items
I
. '
Unqualified Internal
Limitorque Wiring *
-
Corrective Prior to Startup (June 1986)
-
Generic Industry Issue
High Head SI Pump
Motor Lube Oil
-
Change to QDR Not Reflected in Procedure
-
Maintenance Procedure Changed
-
Revision Process Enhanced
LHSI Motor Rewind
-
Misleading Documentation in File
Raychem Splices
Unqualified :.
-
Documentation Strengthened; Fully Qualifiable at Tim*e
of Inspection
-
Bend Radius Inconsistency with Raychem Installation
- ~ Completed Inspections and Repairs
Rockbestos Cable
-
Inadequate Documentation in File ,
-
Documentation Revised; ODA Updated
. ' '
'!
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Potential Enforcement Items
Potentially Unqualified Equipment in MSVH (IEIN 84-90)
Background
-
June 1984 -
Initial Notification from Vendor
-
July 1984 -
Original Evaluation Concluded Ea
Equipment in MSVH Not Affected
-
December 1984 -
Issuance of IEIN 84-90 on Superheat
with MSLB
-
March 1985 -
Engineering Tech. Report Issued Did Not
Adeq*uately Address Ea Concerns
..:... June 1986 -
Revised Tech. Report Issued -
Determined Capability Existed to Detect, Isolate and
Recover From MSLB
Recommended Action:
- Move Steam Pressure Transmitters
- Isolate Openings in MSH
- Reevaluate New Accident Assumptions
-
April 1987 -
Final Engineering Tech. Report Issued -
Evaluated Operability of Equipment on EQML Including *
RG 1.97 Indication . :
.*,-*:
Potential Enforcement Items
Potentially Unqualified Equipment In MSVH (IEIN 84-90)
Root Cause
. -
Inadequate Review of Industry EQ Information
Immediate Corrective
Actions
-
Moved Steam Generator Pressure Transmitters
\\'
'
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Modification Performed in Unit 1 MSVH to Isolate
Elevations
-
In Accordance with GL 85-15/86-15, Virginia Power
Reevaluated Effects of Superheat on Equipment .
Qualifications in MSVH:
- Equipment Located in MSVH Would Have
Performed Safety Function Prior to Experiencing
Harsh Environment
- Functions of *Equipment Not Capable of Qualification
or Relocation are Provided by Alternate Equipment
- Results of Equipment Failure Due to MSLB
(Elevated Temp), was Analyzed and Found to
Remain Bounded by the Safety Analysis
. ' .
. ;:
-;, ..
Potential Enforcement Items
Potentially Unqualified Equipment in MSVH (IEIN 84-90) ....
Current Qualification
I
.
Status
-
Main Steam Pressure transmitters were Moved to
Long Term Corrective
Provide Qµalificaton
-
RG 1.97 Indication (Valve Position) is Being Provided by
Alternate Instrumentation
-
Therefore Equipment on EOML in MSH is Qualified* in
Accordance with 10CFR50.49
Action
-
Enhanced Review/Evaluation Process for EQ
lndustry/NRC Information
-
Strengthened Potential Problem Review Process
-
Lowered Threshold for Initiating Station Deviations
. '
., '
North Anna EQ Inspection Results
October 5-9, 1987
_
_
Concluded that Program has Been Implemented Which
Meets Requirements of 10CFR50.49
No Defeciencles Identified In Implementation of
SER/TER Corrective Action Commitments
Four Violations/
One Unresolved Item -
Unqualified Raychem Splices
-
Unqualified Limitorque Valve Actuators
-
Ea Maintenance Requirements in ODA Not Adequately
Adequately Addressed in Maintenance Procedures
-
Performance Characteristics
-
Victoreen Qualification Test Anomalies Not Adequately
Addressed in QDR
. : .
Unqualified Raychem Splices
(Heat Shrinkable Tubing)
Background
-
Deviant Raychem Splice Categories:
Installation
Acceptance Criteria
Evolution
Undersize Wire
Oversize Bolt
Splice Seal
Overlap
General Workmanship
Criterion
Overlap
Undersize Wire
. Holdout
Raychem Guide
2"
VP Repair
3.A" to 1/2"
0°/o
. Minimum Bend
Radius
QO/o
2 .0 x Use Range
5 x Outer
Resolved Dia.
2.4 to 2.6-. *
to 2.7 x Use
. _. 1 Range
\\'
VP Operability
1/a"
31°/o
- 3. 7 x Use Range
Unqualified Raychem Splices
Sequence of Events
Fall 1985
Identified Problems with Installation of Raychem Splices Being
Worked During Outage; Repairs Made During Outage
12/13/85
A Particular Configuration Problem Identified on 8 Devices Inside
Containment -
Station Deviation Report Initiated
12/17/85
Deficiency Evaluated and Judged Not to 1*mpact Equipment
Qualification nor Indicate a Generic Problem *
04/07/86
Contract Issued to Test Different Configurations of Raychem Splices
. 06/26/86
NRC IN 86-53 Issued
08/11/86
Raychem ;Issues letter to Customers to Aid in Response to IN 86-53
10/27/86*
Decision Made to Delay Inspections Until Testing was Complete
!
... * . . ;
.;.*:
- ~-*
,,*,,
...
Unqualified Raychem Splices
Sequence of Events
11/9/86
Comprehensive Splice Inspection Effort Conducted During Surry
Refueling Outages
11/26/86
Surry Inspection Identified Additional Splice Configuration
Deficiencies; However Only a Small Number Did Not Meet Virginia
Power Qualification Operability Criteria
Results from Testing Contract Received and Raychem Splices
Determined to be Qualified
12/19/86
JCO Provided for Raychem Splices Installed on Narrow-range RTD
Based on Results of Surry Inspections and Belief that North Anna
had Same Problem
12/29/86
New Qualification Acceptance Criteria for Installed Configurations
Developed,, Based on Applicable Industry Testing Results; New
Criteria Showed that Vendor Installation Requirements were Highly
Conservative
01/15/87
EWA Issued to Inspect and Repair Accessible Splices Outside
Containment
02/02/87
Ouside Containment Inspections* Initiated
- ,* .
,. ,**
- .
1.'.,
..
.. i
Unqualified Raychem Splices
Sequence of Events*
02/1.8/87
Conference Calls with NRC on Status of Inspection and Plans for
and
Additional Inspection
02/20/87
. 02/23/87
Detailed Management Plan for Inspections Developed by
Engineering and Provided to Station
02/87 -
Conducted Remaining Inspections and Repairs for Both Units
'10/87
.
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Unqualified Raychem Splices
Management Plan for
Raychem Inspections
-
Reviewed by NRC (Region and Resident
Inspector)
-
Initial Inspections Outside Containment; Planned
to Inspect Inside Containment During Refueling
Unless* Inspection Results Required Otherwise
-
Inspection Priotity Established; (1) IS.
Equipment, (2) Equipment in LOCA/HELB Area
-
Station Management Updates; Control Points
Established to Trigger Suppl*emental Review/Action
(Greater than 50/0 FaUures)
-
Deviation Reports Initiated for Failures; NRC
Guidance in Gls 85-15/86-15 Followed; Repairs
Completed Immediately Following Inspection
I i'
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Unqualified Raychem Splices
Application of Installation Acceptance Criteria
Use of Virginia Power
_Developed Criteria
-
Splices Meeting Repair/Replacement Criteria
Acceptable for Life of Plant
-
Splices Failing Repair/Replace111ent Criteria
Acceptable Until Next Outage
-
Splices Failing Operability Criteria, Gls
85-15/86-15 Actions Followed; Repairs Completed
Immediately following Inspection
,*
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Unqualified Raychem Splices
Inspection Results
(Total Units 1 & 2)
Inside
- Containment .
No. of Devices Inspected
376
No. of Devices Repaired
319
No. of Splices Inspected
2068
No. of Splices Repaired
1301
No. of Devices with
Splices Outside
Operability Criteria
9
No. of Splices Outside
- Operabiltiy Criteria
9
Outside
Containment
Total
382
758
185
504
1475
3543
578
1879
11
20
33
42
- ,*.
Unqualified Raychem Splices
Root Cause
Corrective Action
-
Inadequate Installation Procedures and Training
-
Followed Guidance in NRG Gls 85-15/86-15; Investigated
T.S. Operability Requirements, Evaluated Safety Significance,
Developed Justifications for Continued Operation, and
Scheduled Repairs
-
Determined that Deficient Splices Did Not Cause
Significant Safety Concerns
-
Repaired Deficient Splices* in Accordance with More
Conservative Vendor Guidance
-
Updated Qualification * Documentation for Raychem
Splices to Incorporate Applicable Test Results
. *
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- ..:
1:
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, ..
Unqualified Raychem Splices
Current Qualiflcatlon
Status
-
Raychem Splices Associated with Equipment on EQML
Corrective Action to
are Now Qualified in Accordance with Updated QDR
Industry Test Results Demonstrated that Majority of
NAPS and SPS Splices Installed in Configurations
Outside Vendor Requirements were aualifiable
Prevent Recurrence -
Detailed .Splice Installation Procedures have been Issued
-
Rigorous Splice Installation Training Conducted tor
Appropriate Personnel
-
Plant Modification Documentation Packages Include
Detailed References to Splice Installation Procedures
Where Appropriate
Unqualified Limitorque Valve Actuators
Background
Root Cause
-
Actuators Procured with Qualification Certification
Covering Complete Assembly; Motors Not Specifically
Identified
-
Uri.usual Actuator Motor IDs Identified on Actuators
During QDR Update Effort in February 1987
-
Vendors Unable to Certify that Existing Qualification *
Documentation Applied to Suspect Motors
-
Improper Documentation Provided by Particular Vendor
,.
.
.} .
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1* :*
I
Unqualified Limitorque Valve Actuators
Sequence of Events
Fall 1986
Engineering Review of Limitorque QDRs
10/21/86
02/87 *
Limitorque Correspondence Indicating Potential Qualification
Problem with One Non IS. MOV
Discover. Serial Number Discrepancy on Several Actuator Motors
During Engineering Review of Walkdown* Data
02/24/87
Engineering Contacts limitorque on Incomplete Qualification
Documentation Provided to A-E in 1980 -
Motors lacked
Qualification Documentation
02/25/87
Limitorque* Responds that Stock Motors _are Qualified and North
Anna Special Motors Should be Qualified but Were Not Specifically
Tested with the Actuators
02/27/87
Engineering Notified Station of Potential Qualification Problems with
9 Actuator 'Motors; JCO also Provided
Q3/02/87
Station Deviation Report Submitted
i, .
~------~-
- ,-:*
,*.
- '
IJnqualif ied Limito-,que Valve Actuators
Sequence of Events
03/03/87
Station Safety Committee Reviews the Qualification Issue
03/04/87
First Motor Replacement Initiated (MOV-SW-1038)
03/05/87
One Service Water Actuator Motor Pulled and Destructively
Examined
03/05/87 *
Station Safety Committee Reviewed Results of Inspection and
Approved the JCO Developed on 2/27/87 for the T.S. Actuators
03/07/87
Initiated and Completed Six Motor Replacements
to
03/25/87 *
03/26/87 ._ *
Completed Replacement of 7 of 8 IS. Actuator Motors
10/06/87
The Remaining One IS. and One Non IS. Actuator Motors were
Replaced 'During the 1987 Refueling Outage as a Result of
Replacement Qualified Motor Procurement Lead Time
Last Two Motors Replaced
- I
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Unqualified Limitorque *. Valve Actuators
Current Qualifcation
Status
-
Valve Actuators Now Fully Qualified *
Corrective Action to
Prevent Recurrence ** * -
Procedures Strengthened to Require that Procurement
of any Limitorque Part, Assembly, or Module that is Not
Explicitly Qualified to Test Reports Previously Reviewed
and Approved by VP (as Specified in the Procurement
Document), Must be Identified and Accepted by VP Prior
to Shipment
Conclusions
-
Followed Guidance in NRC Gls 85-15/86-15; Investigated
T.S. Operability Requirements, Evaluated Safety Significance,
Developed Justifications for Continued Operation, and
Scheduled Repairs'
-
Determined that Unqualified Valve Actuators Did Not
Cause Significant Safety Concern
-
Motors with Proper Qualification Certification Installed in
Timely Manner
.
- . Inspections of Questionable Motors and Subsequent
Vendor Information Indicate that Motors are Qualifiable for
- the Envi~onmental * Zones in Which They. are * Located
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EQ Maintenance** Requirements in QDR Not
Adequately Addressed in Maintenance Procedures
Replacement of LHSI Pump Bearings
Safegu*rcl Bldg Fan Grease Changeout/Sampling
Re-Torque Specifications for Conax Connections in SOY Maintenance
Procedures
Overall Root Cause
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Inadequate Reviews of New/Revised QDRs by the
Affected Station Departments to Ensure Requirements are
Incorporated into Station' Procedures
Corrective Actions
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Evaluated Continued Applicability of QDR Requirements
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Change Requests Issued to Delete LHSI Pump Bearing
Replacements and Safeguard Bldg Fan Grease
Changeout/Sampling Requirements from Applicable ODRs
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Revised SOV Maintenance Procedure to Specify Conax
. Connector Re.:rorqueing Requirements
Current Qualification
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Status
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LHSI Pump Motor Bearings were~ and Remain, Qualified
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Safequards Bldg Fans were, and Remain, Qualified
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Conax Connectors to SOVs were, and Remain, Qualified
_ ~inc_e Conax Procedure Which Includes Torque Valves was
Used for Sov Repair
Corrective Action to
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Prevent Recurre.nce -- * A Cross Reference Index Between QDR Requirements
and Implementing* Maintenance Procedures will be
Developed
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Performance Characteristics Not Adequately
Addressed in QDRs
Root Cause
Corrective Action
Current Qualification
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Deficient ODR Preparation, Note that Performance .
Characteristics were Considered in ODR Development,
but were Not Clearly Delineated in a Unique Section of
- the Appropriate QDRs
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Developed a Bounding Analysis for a Worst Case
Instrument loop
Re_sults Show that Total Installed System Accuracies are
within Design Allowable Levels
Status
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Equipment on EQML is Fully Qualified, Including
Individual and Installed Accuracy Effects, for its Application
Corrective Action to
Prevent Recurrence
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The Appropriate ODRs will be Enhanced to Clearly
Delineate Equipment Performance Requirements and
Characteristics
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Victoreen Qualification Test Not Ar)equately
Resolved in QDRs
Root Cause
Corrective Action
Current Qualification
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Inadequate QDR Preparation
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Evaluated Test Anomalies, Including Accuracy Data
Concluded that Test Anomalies Did Not Impact Monitor
Qualification
Issued a Change Request to all QDR Holders on 11-5-87
Status
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Victoreen Radiation Monitors Are Qualified for Their
Intended Purpose
This Deviation is Considered to be an Isolated Occurrence
Corrective Action to
Prevent Recurrence -
A Revision to the Victoreen Radiation Monitor QDR,
Addressing Each Anomaly (Including Accuracy Data), has
been Prepared and is being Processed
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Equipment Not on EQML Specified in EOPs
NRC IR Documents Commitment
to Review* EOPs and Delete
References to Unqualified Equipment
After Further Evaluation,
Corrective Action Modified
as Follows:
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EOPs to be Reviewed and Where Qualified and
Corrective Action
Status
Unqualified Equipment are Specified, the Unqualified
Equipment will be Identified
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Affected EOPs Revised
Operators Trained
- Guidance for Annual EOP Review has been Amended to
New Convention
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EQ Enforcement Summary
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EQ Enforcement Summary
Surry
North Anna
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Identification of a Plant-Specific Problem and Initiation
of Corrective Action for Equipment in the MSVH was
Delayed Due to Inadequacies in the Evaluation of the
Effect of Superheated Steam on Environmental
Temperatures
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Initial Problems Identified with Raychem Installation
were Evaluated to be Due to the Level of Craft
Workmanship
Once Subsequent Problems were Recognized. Inspection
Evaluation, and Repair Proceeded Based on the
Knowledge that '(1) the Raychem Installation Guide was
Highly Conservative, and that (2) Industry Test Results
were Showing that a Much Wider Range of Installation
Configurations were Oualifiable
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EQ Enforcement Summary
North Anna
Conclusions
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Qualification Documentation Obtained. When the Nine
Valve Actuators were Procured Satisfactorily Addressed
the Qualification of the Entire Assembly
Once the Vendors were Questioned Upon Discovery of
Unique Motor IDs, Appropriate Actions were Taken in
Accordance with NRC Guidance
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Evaluation of Problems Resulted in the Conclusion that
There were Not Programmatic Breakdowns in the Virginia
Power Ea Program
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The Identified EQ Problems Did Not Result in any Safety
Concerns
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