ML18059A559
| ML18059A559 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/06/1993 |
| From: | Kobetz T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18059A560 | List: |
| References | |
| NUDOCS 9312130038 | |
| Download: ML18059A559 (56) | |
See also: IR 05000255/1993024
Text
Docket No. 50-255
Consumers Power Company
ATTN:
Gerald B. Slade
General Manager
Palisades Nuclear Generating Plant
27780 Blue Star Memorial Highway
Covert, MI
49043-9530
Dear Mr. Slade:
DEC o s 1993
SUBJECT:
ROUTINE RESIDENT INSPECTION AT PALISADES NUCLEAR PLANT
This refers to the inspection conducted by Messrs. M. E. Parker, D. G.
Passehl, D. J. Hartland, and J. A. Lennartz of this office, from September 30
through November 15, 1993.
The inspection included a review of authorized
activities for your Palisades Nuclear Generating Facility.
At the conclusion
of the inspection, the findings were discussed with those members of your
staff identified in the enclosed report .
Areas examined during the inspection are identified in the report.
Within
these areas, the inspection consisted of a selective examination of procedures
and representative records, interviews with personnel, and observation of
activities in progress.
The topics of the September 30 and October 21, 1993,
management meetings are also summarized.
A copy of your handouts presented at
'these management meetings are attached to the report.
The purpose of the.
inspection was to determine whether activities authorized by the license were
conduct~d safely and in accordance with NRC requirements.
During this inspection, housekeeping and material conditions were noled to
have deteriorated in various areas of the auxiliary building.
Most notable
was the excessive amount of cluttered and contaminated areas, and an increased
use of catchments to direct or contain leaks.
Some areas had work request
tags to improve these conditions that were several months to several years
old. This is of concern since there is an increased potential for personnel
. contamination events and unnecessary exposure.
We noted that you have begun
efforts to improve these areas and we will continue to monitor your progress.
Based on the results of this inspection, certain of your activities, involving
a procedure performance error during a safety injection surveillance test,
appeared to be in violation of NRC requirements.
However, as described in the
enclosed inspection report, you identified this violation. Therefore, the
violation will not be subject to enforcement action because your efforts in
identifying and correcting the violation met the criteria specified in Section
VII.B of the "General Statement of Policy and Procedure for NRC Enforcement
Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C) .
9312130038 931206
ADOCK 05000255
G
-**-
Consumers Power Company
2
OEC o 6 1993
In accordance with 10 CFR 2.790 of the Corrunission's regulations, a copy of
this letter and the enclosed inspection report will'be placed in the NRC
Public Document Room.
We will gladly discuss any questions you have concerning this inspection.
Enclosures:
.1.
Inspection Report
No. 50-255/93024(DRP)
2.
Attachment 1
3.
Attachment 2
cc w/enelosure:
David P. Hoffman, Vice President
Nuclear Operations
David W. Rogers, Safety
and Licensing Director
~
OC/LFDCB
Resident Inspector, RIII
James R. Padgett, Michigan Public
Service Commission
Michigan Department of
Public Health'
Palisades, LPM, NRR
SRI, Big Rock Point
-<bee: . *PUBLIC .ILOl'"
Rl(le
!V
"
M-
RIII
Axe~n
Orsini
Burdi\\k
U-h/&J1
\\"t.~'- "\\>
(cover letter)
(para.6b'~
Sincerely,
Original sign~d by T.
Kobetz
T. Kobetz, Acting Chief
Reactor Projects Settion 2A
ml
RIII
~etz
~
O~\\~'i
l'Y~
, if v/11
~*
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION Ill
799 ROOSEVELT ROAD
GLEN ELLYN. 11.UNQIS 60137-,5927
Docket No. 50-255
Consumers Power Company
ATTN:
Gerald B. Slade
General Manager
Palisades Nuclear Generating Plant
27780 Blue Star Memorial Highway
Covert, MI
49043-9530
Dear Mr. Slade:
o~c o s 1993
SUBJECT:
ROUTINE RESIDENT INSPECTION AT PALISADES NUCLEAR PLANT
This refers to the inspection conducted by Messrs. M. E. Parker, 0. G~
Passehl, D. J. Hartland, and J. A. Lennartz of this office, from September 30
through November 15, 1993.
The inspection included a review of authorized
activities for your Palisades Nuclear Generating Facility. At the conclusion
of the inspection, the findings were discussed with those members of your
staff identified in the enclosed report .
Areas examined during the inspection are identified in the report.
Within
these areas, the inspection consisted of a selective examination *of procedures
and representative records, interviews with personnel, and observation of
activities in progress.
The topics of the September 30 and October 21, 1993,
management meetings are also summarized.
A copy of your handouts presented at
these management meetings are attached to the report.
The purpose of the
inspection was to determine whether activities authorized by the license were
conducted safely and in accordance with NRC requirements.
During this inspection, housekeeping and material conditions were noted to
have deteriorated in various areas of the auxiliary building: Most notable
-was the excessive amount of cluttered and contaminated areas, and an increased
use of catchments to direct or contain leaks.
Some areas had work request
tags to improve these conditions that were several months to several years
old.
This is of concern since there is an increased potential for personnel
contamination events and unnecessary exposure.
We noted that you have begun
effo~ts to improve these areas and we will continue to monitor your progress.
Based on the results of this inspection, certain of your activities, involving
a procedure performance error during a safety injection surveillance test,
appeared to be in violation of NRC requirements.
However, as described in the
enclosed inspection report, you identified this violation.
Therefore, the
violation will not be subject to enforcement action because your efforts in
identifying and correcting the violation met the criteria specified in Section
VII.B of the "General Statement of Policy and Procedure for NRC Enforcement
Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C).
-*
Consumers Power Company
2
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of
this letter and the enclosed inspection report will be placed in the NRC
Public Document Room.
We will gladly discuss any questions you have concerning this inspection.
Enclosures:
l.
Inspection Report
No. S0-255/93024(DRP)
2.
Attachment I
3.
Attachment 2
cc w/enclosure:
David P. Hoffman, Vice President
Nuclear Operations
David W. Rogers, Safety
and Licensing Director
OC/LFDCB
Resident Inspector, Rill
James R. Padgett, Michigan Public
Service Commission
Michigan Department of
Public Health
Palisades, LPM, NRR
SRI, Big Rock Point .
2A
-*
.,
U. S. NUCLEAR REGULATORY COMMISSION
REGION I I I
Report No. 50-255/93024{DRP)
Docket No. 50-255
Licensee: Consumers Power Company
212 West Michigan Avenue
Jackson, MI
49201
Facility Name:
Palisades Nuclear Generating Plant
Inspection At:
Palisades Site, Covert, Michigan
License No. DPR-20
Inspection Conducted:
September 30 through November 15, 1993
Inspectors:
Approved
Date
2A
Inspection Summary
Inspection from September 30, through November 15, 1993
(Report No. 50-255/93024(0RP))
Areas Inspected:
Routine, unannounced inspection by the resident and regional
inspectors of actions on previously identified items, licensee event report
followup, followup of events, operational safety verification, radiological
controls, maintenance, surveillance, engineering and technical support, and
management meetings.
No Safety Issues Management System {SIMS) items were
reviewed.
Results:
No cited violations or deviations were identified in any of the nine
areas inspected.
One noncited violation was identified and is described in.
paragr~ph 8.
'The strengths, weaknesses, and Inspection Followup Items are discussed in
paragraph l, "Management Interview."
9312130044 931206
ADOCK 05000255
G
-*-
1.
DETAILS
Management Interview (71707)
The inspectors met with licensee representatives (denoted in paragraph
11) on November 16, 1993, and informally throughout the inspection
period to summarize the scope and findings of the inspection activities.
The inspectors also discussed the likely informational content of the
inspection report including the attachments, with regard to document.s or
processes reviewed by the inspectors.
The licensee did not identify any
such documents or processes as proprietary.
Highlights of the exit interview are discussed below: *
a.
Strengths noted:
(1)
Performance of startup testing without incident.
This was
noted to be a significant improvement from the last
refueling outage.
(2)
Operator handling of a rapid power reduction that prevented
a reactor trip.
b.
Weaknesses noted:
(1)
Torque values for main turbine hydraulic fluid hose fittings
found not within specified values.
(2)
Procedure performance error during a safety injection system
surveillance test.
(3)
Housekeeping and material condition in various
radiologically controlled areas of the auxiliary buiJding.
2.
Actions on Previously Identified Items (92701, 92702)
a~
(Closed} Inspection Followup Item 50-255/93010-0l(DRS}:
The-
licensee's technical specification required testing of the main
steam isolation valves did not accurately r~flect their ability to'
perform their safety function. The main steam isolation valves
were tested on power ascensions after they had already been
exercised, repaired, or conditioned.
Therefore, they did not
- reflect the as-found condition.
The licensee revised General Operating Procedure 9 (GOP-9), "Plant
Cooldown from Hot Standby/Shutdown," Rev.13, to specify that the
subject testing be performed when the main steam isolation valves
are closed during a plant cooldown, which would reflect as-found
conditions and accurately demonstrate the main steam isolation
valves' ability to perform their safety function. This item is
closed.
2
b.
c.
(Closed) Notice Of Deviation 50-255/93010-02CDRS):
For a steam
line break inside containment concurre~t with a failure of the
main steam isolation valve to close on the unaffected steam
generator event, the licensee failed to meet their commitment
documented in a response dated April 28, 1986, with regard to the
following:
Go to once-through cooling (OTC).
Ensure maximum feed flow to at least one steam generator.
Maximize containment spray flow and place containment
coolers in emergency alignment.
Maximize service water and component cooling water flow.
The licensee identified where all of the above items would be
accomplished in the existing emergency operating procedures and
conducted a review of emergency operating procedure strategy with
respect. to initiation of OTC. The existing strategy allows
automatic initiatibn of feedwater, ensured by manual actions, with
acceptable cooling verified by steam generator level and primary
coolant system parameters. If continued use of the steam
generators for decay heat remo~al is not po~sible, OTC would be
initiated.
The current emergency operating procedure strategy was compared
with an alternative strategy of immediately initiating OTC upon
observing the symptoms of a steam line break concurrent with a
failure of the main steam isolation valve to fully close on the
unaffected steam generator. The licensee concluded that either
method would result in continued core cooling.*
. The licensee maintains that the current emergency operating
procedure strategy of using.OTC only if cooling using steam
generators cannot be verified is preferred to immediate initiation
of OTC. The current strategy reduces* the risk of losing the
abi] ity. to coo.l the core, does not further compound an -a 1 ready -. -
complicated event, and conforms to the approved guidance for
Combustion Engineering (CE) plant emergency operating procedures.
Based on the inspectors review of the licensee's response this
item *is closed.
(Closed) Inspection Followup Item 50-255/93010-03(0RS):
No
_operator training has been provided on a steam line break inside
containment concurrent with failure of the main steam isolation
valve to close on the unaffected steam generator event. This event
would result in simultaneous blowdown of both steam generators*
into containment. Additionally, no training has been provided on
determining operability of instrumentation which may be adversely
affected by this environment .
3
-*
d.
e.
The licensee has committed to complete classroom training prior to
th~ end of 1993 for this event. The training for the subject event
is to include discussions on the following:
How a blowdown of both steam generators could occur.
An explanation of why there are differences between safety
analyses and simulator modeling of some events.
Symptoms, expected plant response, emergency operating
procedure paths involved, and the potential for significant
error or failure of instrumentation located in the
containment.
J
Verification of instrument reading validity and use of
al~ernate instrumentation for this and other events which
degrade the containment environment.
Simulator training will also be provided once the necessary
simulator modeling changes are completed (Open item 50-255/93010-.
05(DRS)) and the emergency operating procedures associated with
this event are validated. Based on the licensee's commitm~nt to
provide. the described training, this item is closed.
(Closed) Inspection Followup Item 50-255/93010-04(0RS):
A-caution
in Emergency Operating Procedure 9.0 (EOP-9), "Functional Recovery
Procedure," Rev.3,
provided no useful information. The caution
statement at step 12 stated .the shift supervisor may deviate from
the procedure via 10 CFR 50.54X. This caution is unnecessary and
inappropriate.
Emergency Operating Procedure 6.0 (EOP-6), "Excess Steam Demand
Event," Rev.4, Attachment 2, contained graphs to account for
- errors in pressurizer and steam generator narrow range level.
However, the attachment was not referenced in the body of the
procedure.
The licensee committed to delete the unnecessary caution from EOP-
9: o curing the current emergency operating procedure revision
effort. Additionally, the licensee identified that EOP 6.0,
Attachment 2, is referred to in several places within the EOPs,
such as step 6.a.l of Attachment 1 to EOP 6.0, and considers no
change necessary regarding this item. Based on a review of the
licensee's response this item is closed.
(Closed) Inspection Followup Item 50-255/93010-05(0RS):
The
simulator modeling of a main steam line break inside containment
concurrent with a failure of the main steam isolation valve to
close on the unaffected steam generator event was not accurate .
4
Simulator modeling of containment temperature and pressure
indicated lower value~ than those expected for the event based on
previous analyses.
The licensee is conducting a comparison between the simulator,
CPMAAP (an engineering analysis computer code), and safety
analysis calculations for containment response to the event to
determine correct simulator modeling. The corrections necessary
for proper simulator modeling are scheduled to be completed prior
to the end of 1993. This item is closed.
No violations, deviations, unresolved, or inspection followup items were
identified in this area.
3.
Licensee Event Report Followup (92700, 92720)
The inspectors reviewed the following Licensee Event Report (LER) by
means of direct observ~tion, discussions with licensee personnel, and
review of records.
The review addressed compliance to reporting
requirements and, as applicable, that immediate corrective action and
appropriate action to prevent recurrence had been accomplished.
a.
(Closed) LER 255/92009:
Inadvertent Actuation of the Control Room,
Heating, Ventilation. and Air Conditioning (HVAC) due to Damaged
Electrical Cable:
On February 13, 1992, the control room
ventilation system inadvertently switched to the emergency mode.
At the time of occurrence, the plant was in cold shutdown and the
primary coolant system was de-pressurized.
The control room
operators immediately verified that a valid containment high
pressure (CHP) or a containment high radiation (CHR) signal did
not exist. The licensee subsequently determined that electrical
maintenance personnel were replacing damaged flex conduit on the
CHR relay 5R-6 circuit~ When the wire on relay 5R-6, point 16,
was disconnected, the "A" train of control room HVAC automatically
switched to the emergency mode.
This event was caused by inadequate job planning and personnel
error. . The job p 1 an was inadequate -in that the work instructions
did not identify the electrical scheme as a "de-energize to
actuate" scheme.
Personnel erred by not referring to the approved
drawings prior to lifting wires.
The drawings clearly identify
the "de-energize to actuate" scheme.
The licensee's completed corrective actions were appropriate.
They include reiterating management's expectations through
training that the electrical and l&C staff review electrical
schematics against job plans prior to initiating work.
Proper job
planning requirements were also reviewed with the maintenance
planners. This LER is closed .
5
4.
b.
(Closed) LER 255/92015:
Noncomplying Movement of Heavy Loads due
to Procedure Error:
On August 24, 1989, the licensee discovered *
that a commitment for both the reactor engineer and the shift
supervisor to approve heavy load path deviations was inadvertently
removed from their heavy load procedures FHS~M-23, "Movement of
Heavy Loads in the Spent Fuel Pool Area," and FHS-M-24, "Movement
of Heavy Loads in the Containment Building Area."
A search commenced to determine if any deviations from safe load
paths, approved by only the shift supervisor, had occurred.
Such
a deviation from approved load paths had occurred on September 30,
1990.
The event was caused by the inappropriate procedure rev1s1on and
weaknesses in the commitment tracking system used in 1989.
Since
then the reactor engineer approval was re-instituted back into the
heavy load procedures and the commitment tracking system was
improved.
This LER is closed.
No violations, deviations, unresolved, or inspection followup items were
identified in this area.
Followup of Events (93702)
During the inspection period, the licensee experienced several events,
some of which required prompt notification of the NRC pursuant to 10 CFR
50.72.
The inspectors pursued the events onsite with licensee and/or
other NRC officials.
In each case, the inspectors *verified that the
notification was correct and timely, if appropriate; that activities
were conducted within regulatory requirements, and that corrective
actions would prevent future recurrence.
The specific events are as
follows:
October 9, 1993 -
October 12, 1993 -
November 5, 1993 -
November 6, 1993 -
November 12, 1993-
Unisolable through wall leak identified on
Pressurizer Temperature Element TE-0101.
Unisolable through wall leak identified on
Pressurizer Temperature.Element TE-0102.
Hydraulic fluid leak found on main turbine
number 2 stop valve (CV-0571)
Erratic operation of main turbine number 1
governor valve (CV-0570)
Hydraulic fluid leak found on main turbine
number 2 intercept valve (CV-0548)
The following are brief summaries of the events.
The inspectors will
evaluate corrective actions for the events when the respective LERs are
reviewed .
6
a.
-*
- '
b.
I
On October 9, 1993, during a reactor coolant system walkdown the
licensee identified water ieaking from a pressurizer head
penetration around the base of temperature element TE-0101.
This
element is used to determine the vapor phase temperature of the
pressurizer.
At the time the leak was identified, the reactor
plant was in cold shutdown with the plant pressurized to 250 psig.
The leak rate was estimated to be about one ounce per minute.
Shortly after this primary coolant system (PCS) leak was
identified, actions commenced to depressurize the system.
This pressurizer resistance temperature detector (RTD) well was
previously scheduled for a visual inspection during plant startup.
The RTD well
h~d not been worked on during the refueling outage;
however, moisture had been observed arouhd this penetration during
the power operated relief valve (PORV) line repairs some three
weeks earlier (see NRC Inspection Report 255/9302l(DRP)).
At the time the PORV leak was found the licensee was uncertain if
the moisture seen at TE-0101 was due to extensive wetting of the
pressurizer head area and surrounding insulation due to the PORV
leak, or due to an actual leak at the RTD nozzle.
Shortly
thereafter, the licensee performed a pressure drop test of the RTD
thermowell.
Although this test did not identify any concerns, the
licensee was unable to perform further evaluation until the PCS
was pressurized .
On October 12, 1993, a similar leak to that described above, was
found during a followup walkdown of the p~essurizer to inspect
additional nozzle penetrations.
The licensee identified moisture
and boric acid corrosion around pressurizer liquid phase
temperature element TE-0102.
This walkdown was performed with the
PCS depressurized and only a static head of water in the
pressurizer.
The leak was estimated at several drops per minute.
Based upon a review of industry experience and NRC Information
Notice 90-10, "Primary Water Stress Corrosion Cracking (PWSCC) of
Inconel 600," the licensee spe(:ulated that the leakage came froin.
between the inconel sleeve.and-the carbon steel shell of the
pressurizer.
On October 12, 1993, daily conference calls were established
between the licensee and the NRC.
In addition, on October 12 and
21, 1993, technical meetings were held with the licensee to
discuss PWSCC on the pressurizer.
The licensee obtained assistance from the Combustion Engineering
(CE) owners group to evaluate this problem.
Region III and NRR
specialists followed the results of additional PCS inspections in
addition to the PORV line repairs.
The licensee's repair efforts
of both TE-0101 and TE-0102 were satisfactory.
(See NRC
inspection report 50-255/93023(DRS) for further details.)
The licensee continued with startup preparations following
7
successful repairs to the leaking pressurizer PORV we)ds ~nd
pressurizer temperature elements.
No additional leaks were
observed during those repair efforts and during containment
closeout tours.
c.
On November 12, 1993, with the plant at 100 percent power~ the
plant experienced a transient when a steam supply valve to the "A"
low pressure turbine was closed due to a hydraulic fluid leak.
Operators performed a rapid down power maneuver on the primary
side to match the secondary side.
The leak occurred on CV-0548, the "A" low pressure turbine reheat
intercept valve.
Upon discovering the leak, the operators
isolated the fluid to the valve causing it to go closed.
The~
resultant pressure spike in moisture separator reheater E-9A
caused lifting of its relief valves.
The operators then isolated
the steam supply to moisture separator reheater E-9C.
This action
placed the main turbine steam system back into a configuration
covered by plant procedures, namely Standard Operating Procedure 8
(SOP-8), "Main Turbine and Generating Systems," Rev.26.
Operators handled the transient well with good coordination
betwee.n the primary side and secondary. side operators.
The quick
reaction by the plant operators and the action of the automatic
controllers in the steam generator water level control system
likely prevented a reactor trip.
The coordination between the
_ operators in the control room is considered a strength.
The hydraulic fluid leak was repaired and the unit was returned to
full power on November 14, 1993.
No violations, deviations, unresolved, or inspection followup items were
identified in this area.
5.
- operational Safety Verification (71707, 71710, 42700)
Routine facility operating activities were observed as conducted in the
-plant and from-the,main control room. -Plant startup,*steady power*
operation, plant shutdown~ and system lineup and operation were observed
as applicable.
The performance of reactor operators and senior reactor operators, shift
engineers, and auxiliary equipment operators was observed and evaluated.
Included in the review were procedure use and adherence, records and
logs, communications, shift/duty turnover, and the degree of
professionalism of control room activities.
Evaluation, corrective action, and response for off normal conditions
were examined.
This included compliance to any reporting requirements.
Observations of the control room monitors, indicators, and recorders
were made to verify the operability of emergency systems,. radiation
8
.,
monitoring systems, and nuclear reactor protection systems.
Reviews of
surveillance, equipment condition, and tagout logs were conducted.
Proper return to service of selected components was verified.
Periodic verification of Engineered Safety Features status was conducted
by the inspectors.
Equipment alignment was verified against plant
procedures and drawings and detailed walkdowns selectively verified:
equipment labeling, the absence of leaks, housekeeping, calibration
dates, operability of support systems, breaker and switch alignment, as
appropriate.
a.
Genera 1
Plant operators brought the plant back on line on November 8,
1993, ending a 156 day refueling outage.
The plant began the
inspection period in cold shutdown with the primary coolant system
partially drained with preparations for leaving cold shutdown in
progress.
Some relevant dates are:
October 27, 1993 -
October 28, 1993 -
November 3, 1993 -
November 4, 1993 -
November 8, 1993 ~
Reactor left cold shutdown.
Reactor in hot shutdown.
Reactor critical at 10-
4 percent power.
Reactor enter~d power operations at
greater than 2 percent power.
Turbine Generator on line following
suq:essful overs peed testing
The inspectors followed various primary coolant system parameters
during and after plant startup. Primary coolant system leakage
was very low, with no observable trend in containment sump levels.
Primary coolant activity values were normal, at or below measured
values seen following previous refueling outages.
Dose equivalent
iodine averaged less than two percent of the technical
specification limit. Secondary side equipment operated well,
except for some hydraulic fluid leaks associated with the turbine
generator steam supply valves, as previously mentioned.
b.
The inspectors provided expanded site coverage beginning
immediately prior to criticality and extending through the start
of power escalation. Assistance from the D. C. Cook resident
inspector office was obtained. Coverage of all three shifts were
provided daily. Major activities observed were:
Approach to criticality
Reactor criticality
9
6.
Turbine Generator Overspeed trip testing
Synchronization of the turbine generator to the grid for
power ascension
c.
Criticality
The unit went critical at 11:09 a.m.(EST) on November 3, 1993.
This started the low power physics testing portion of the startup
program.
The estimated critical rod height and boron
concentration were within the predicted target band.
d.
Plant Tours
Tours of the control room were routinely made.
Staffing
requirements were met, operators were cognizant of changing plant
conditions, the equipment status and the limiting Condition for
Operation status boards were maintained up to date. *Portions of
the following startup activities were observed:
(1)
GOP-3, "Hot Shutdown to Critical in Hot Standby," Rev.12
(2)
T-191, "Startup Physics Test Program," Rev.4 *
(3)
SOP-8, "Mai~ Turbine and Generating Systems," Rev.27
No violations, deviations, unresolved, or inspection followup items were
identified in this area.
Radiological Controls (71707)
During routine tours of radiologically controlled plant facilities or
areas, the inspector observed occupational radiation safety practices by,
the radiation protection staff and other workers.
Effluent releases were routinely checked, including examination of on-
1 ine recorder traces and proper operation of automatic monitoring
equipment.
Independent surveys were performed in various radiologically
controlled areas.
a.
On one tour the inspector observed the entrance to the East
Engineering Safeguards Room was a posted high radiation area; this
area is typically only a radiation area.
Investigation found the.
reason was a 180 R/hr hot spot lodged in place near the suction of
the low pressure safety injection/shutdown cooling pump P-67A.
The hot spot was appropriately shielded. A gamma scan indicated
the hot spot to be a fuel particle, most likely due to the failed
fuel found earlier this outage and carried through the shutdown
cooling system while that system was operating .
10
-*--**
b.
The licensee's attempt to capture and f~rther char~cterize the
particle proved unsuccessful.
The original intention was to jog
P-67A while tracking the hot particle from the suction piping into
the pump casing.
Using a hot spot flush rig they would flush the
particle through a drain on the casing and capture it in the rig
for further analysis.
Instead, the licensee attempted to flush the hot spot with only
the head of water from the hot spot flush rig, without jogging P-
67A, due to other ongoing testing.
The hot spot did not move
during this attempt.
Later during hot shutdown surveillance
testing of this system, the particle became dislodged and was
flushed out of the piping.
The licensee checked all accessible
piping within the system and concluded the particle is probably
residing in the safety injection and refueling water tank.
There has been no indication of abnormal or excessive radiation
doses received by any individuals nor have any anomalous trends
been noted.
The licensee has an acceptable monitoring program in
place to locate and shield the particle should it reappear in the
system:
Housekeeping and material condition was noted to have deteriorated
in various auxiliary building areas as observed during an NRC
management tour just prior to the end of the current refueling
outage.
Of note was the excessive amount of contaminated area in*
the east and west safeguards pump rooms.
Additionally, throughout
the plant an increased use of catchments to direct or contain
leaks was observed.
Some areas had work request tags to repair
valve and pump leaks and improve material condition that were
several months to several years old.
This item is of concern since there is an increased potential for
personne 1 contamination events a_nd unnecessary exposure.
These observations were discussed with the licensee, who has
commenced aggressive action to reclaim and clean up contaminated
areas. *The inspectors-will continue to monitor the licensee's
progress in this area.
No violations, deviations, unresolved, or inspection followup items were
identified in this area.
7.
Maintenance (62703, 42700)
Maintenance activities i~ the plant were routinely inspected, including
.both corrective maintenance (repairs) and preventive maintenance.
Mechanical, electrical, and instrument and control group maintenance
activities were included, as available.
The focus of the inspection was to assure the maintenance activities
reviewed were conducted in accordance with approved procedures,
11
-*--
- 8,
regulatory guides and industry codes or standards, and in conformance
with Technical Specifications.
The following items were considered
during this review: the Limiting Conditions for Operation were met while
components or systems were removed from service; approvals were obtained
prior to initiating the work; activities were accomplished using
approved procedures; and post maintenance testing was performed as
applicable.
The following maintenance activities were observed:
(1)
Repairs to the pressurizer PORV weld and temperature element
nozzles
(2)
Troubleshooting of an inoperable subcooled monitoring channel
(3)
Containment Air Cooler VHX-2 service water leak repair
(4)
Troubleshooting of the hydraulic fluid leak found on main turbine
number 2 stop valve (CV-0571)
(5)
Troubleshooting of the erratic operation of main turbine number 1
governor valve (CV-0570)
(6)
Troubleshooting of the hydraulic fluid leak found on main turbine
number 2 intercept valve (CV-0548)
No violations, deviations; unresolved, or inspe~tion followup items were
identified in this area.
Surveillance (61726, 42700)
The inspector reviewed technical specifications required surveillance
testing as described below, and verified that testing was performed in
accordance with adequate procedures.
Additionally, test instrumentation
was calibrated, limiting conditions for operation were met, removal and
restoration of the affected components were properly accomplished, and
test results conformed with technical specifications and procedure
-requirements.
The results were reviewed by personnel other than the
individual directing the test, and deficiencies identified during the
testing were properly reviewed and resolved by appropriate management
personnel.
The following surveillance tests were observed:
a.
b.
c.
RI-47, "Rod Withdrawal Prohibit Interlock Matrix Check," Rev.7
Ml-SA, "Containment High Pressure Test," Rev.O
Q0-1, "Safety Injection System," Rev.34
The licensee issued a deficiency report after testing the left
channel of safety injection when the operators performing the test
12
missed a procedure step to place the control switch for certain
valves in their required positions. Step 5.2.5.b required the
operator to position the control switch for each of eight valves
to the "as-left" position.
However, the operators only visually
verified the valve positions without manipulating the control
switches.
Step 5.2.4 of the procedure requires the operator to push and hold
the safety injection actuation signal (SIAS) push button until
step 5.2.7.
Releasing the SIAS button in step 5.2.7 removes the
S1AS test signal.
When step 5.2.5.a was performed with the SIAS button pressed, the
operators properly verified that the eight valves, two of which
were CV-0913 and CV-0950, changed to their required test
positions.
CV-0913 and CV-0950 supply seal cooling water to the *
high and low pressure safety injection pumps.
Both valves
properly went to the "open" test position.
The following step 5.2.5.b required the operator to position the
control switches for CV-0913 and CV-0950 to the "as-left (in this
case the "open") position.
However, the operators only visually
verified the valves were open without t~rning the control switch
to the "open" position.
The intent of using the control switch
was to leave a standing open signal to CV-0913 and CV-0950, so
that after the SIAS button was released the valves would not
reclose with the high and low pressure safety injection pumps
still running.
As a result, the left train of high and low pressure safety
injection pumps (P-668 and P-678, respectively) ran for about
fifteen minutes without cooling flow to the seals and bearings,
until an operator discovered the condition.
Technical Specification 6.8.1.a requires, in part, that written
procedures be established, implemented, and maintained covering
the applicable procedures recommended in Appendix A of Regulatory
Guide J.33, Revision 2 (February 1978), -Quality Assurance Program
- Requirements, as endorsed by CPC-2A, Quality Program Description.
The Quality Program Description in CPC-2A endorses Emergency Core
Cooling System Tests in Appendix A of Regulatory Guide 1.33,
section 8.b.(j).
The inspector considers the failure to implement procedure step
5.2.5.b to be a violation of the above requirement.
The cause of
the failure was personnel error, although the instructions at this
section of the procedure could be clearer.
The safety significan~e of running P-668 without seal and bearing
cooling flow for this short time was minimal.
Component cooling
water flows in parallel to cool the pump bearings, stuffing boxes,
and the seal flushing water cooler.
Final Safety Analysis Report
13
{FSAR) section 6.1.2.2.3 states that the seals are designed for
300°F and are provided with cooling to extend seal life. Si nee
the seal cooling was 60°F water from the Safety Injection and
Refueling Water {SIRW) Tank, little if any seal degradation
occurred. Additionally, correspondence from the pump manufacturer
stated that bearing cooling is not needed for the high pressure
safety injection pumps below 250°F.
A similar argument can be made for P-678.
FSAR section 6.1.2.2.2
states that the seals are designed for 325°F and since the seal
cooling was 60°F water from the SIRW tank, little if any seal
degradation occurred. Also, no significant temperature trends
developed during a 1989 test to track performance of the seals,
bearings, and stuffing boxes without cooling water flow during a
45 minute period.
Although the operators made a procedure error, there were several
positive observations.
The prejob briefing went extremely well.
The operator in charge of the test read through the entire
procedure with the rest of the crew and expectations were laid
out.
The ~ystem engineer was
involv~d at the onset of the prejob
brief until the test was completed and provided some good
comments.
There was good discussion on contingency actions should
problems arise during the test, such as an unforeseen loss of
noncritical service water.
Additionally, there was good involvement by the Nuclear Plant
Assessment Department (NPAD) observer present for the test.
Upon
discovering the loss of seal cooling to the pumps the NPAP
observer suggested that someone be sent to the safeguards rooms to
check on the condition of the pumps.
Therefore, the violation will not be cited since the licensee
discussed the problem and quickly performed appropriate recovery
actions and since the criteria specified in Section VII.B.2 of the
"General Statement of Policy and Procedures for NRC Enforcement
Actions," (Enforcement Policy, 10 CFR P~rt 2, Appendix C), were
satisfied.
No violations, deviations, unresolved, or inspection followup items were
identified in this area.
9.
Engineering and Technical Support (37700,92705)
The inspector monitored engineering and technical support activities at
the site and, on occasion, as provided to the site from the corporate
office. The purpose was to assess the adequacy of these functions in
contributing properly to other functions such as operations,
maintenance, testing, training, fire protection, and configuration
management .
14
a.
Various portions of the startup testing program were observed.
Hot shutdown testing, low power physics testing, and power
escalation proceeded relatively smoothly as a result of being
conducted in a well coordinated fashion.
Unlike similar testing
performed at the conclusion of the previous refueling outage, the
startup program was performed without any unplanned engineered
safety feature actuations.
This is a significant improvement from
the last refueling outage and is considered a strength.
b.
The licensee's investigation into the hydraulic fluid leaks on the
main turbine steam supply valves was followed.
The causes of the
two hydraulic fluid leaks were totally separate.
The cause of the
November 5, 1993, leak on main turbine stop valve CV-0571 Was that
one of the hydraulic fluid hose fittings had been insufficiently
torqued.
The cause of the November 15, 1993, hydraulic fluid leak
on main turbine intercept valve CV-0548 was due to a failed a-ring
in a test solenoid valve that supplies the hydraulic fluid.
The licensee has inspected.the other main turbine steam valves and
stated that no other fittings were found to be under-torqued, but
some were found with a higher than specified torque value. All
the fittings found outside the specified torque range were
corrected.
Further analysis has begun to show that. the as-found
values were acceptable from an engineering standpoint, although
they were outside the licensee's internally specified values.
The
licensee's evaluation of this issue is still in progress .
The licensee has inspected other solenoid valves for signs of o-
ring failure with no immediate concerns identified. Their
investigation into this problem is also still in progress and will
continue to be followed.
No violations, deviations, unresolved, or inspection followup items were
identified i.n this area.
10.
Management Meeting (30702)
A ma,nagement meet.ing was held on September -30; 1993, between 0. P
~ *
Hoffman, Vice President Consumers Power Company, and H.J. Miller, Deputy
Regional Administrator, Riii, and their respective staffs. The purpose
of the meeting was to discuss significant events which have occurred
during the current refueling outage and the short and long-term
initiatives Palisades plans to implement for corrective action.
Attachment 1 is a copy of the material presented by the licensee.
A second meeting was held on October 21,* 1993, to discuss events and
corrective actions associated with primary water stress corrosion
cracking (PWSCC) at Palisades. Attachment 2 to this report is a copy of
the licensee's handout from this meeting .
15
11.
Persons Contacted
Consumers Power Company
- D. P. Hoffman, Vice President, Nuclear Operations
- G. B. Slade, Plant General Manager
- R. D. Orosz, Nuclear Engineering & Construction Manager
- R. M. Rice, Director, NPAD
T. J. Palmisano, Plant Operations Manager
- D. W. Rogers, Safety & Licensing Director
- K. M. Haas, Radiological Services Manager
J. L. Hanson, Operations Superintendent
R. B. Kasper, Maintenance Manager
- *K. E. Osborne, System Engineering Manager
- C. R. Ritt, Administrative Manager
- J. C. Griggs, Human Resource Director
- H. A. Heavin, Controller
- D. J. Fitzgibbon, Shift Supervisor
- G. J. Daggett, Material Management Superintendent
.
- G. B. Szczotka, Staff Engineer, Nuclear Training Department
Nuclear Regulatory Commission (NRC}
- J. B. Martin, Regional Administrator
- H. J. Miller, Deputy Regional Administrator
G. E. Grant, Director Designate, Division of Reactor Safety
W. M. Dean, Acting Director, Project Directorate, 111-1, NRR
- A. H. Hsi a, Project Manager, NRR
.
- B. L. Jorgensen, Acting Chief, Reactor Projects Branch 2
- T. J. Kobetz, Acting Chief Reactor Projects Section 2A
- C. N. Orsini, Reactor Engineer, Reactor Projects Section 2A
- *M. E. Parker, Senior Resident Inspector
- D. G. Passehl, Resident Irtspector
- Denotes those present at the management meeting on September 30, 1993.
- Denotes those present at the management meeting on October 21, 1993.
- Denotes those present at the exit meeting on November.16, 1993
Other members of the plant staff, and several members of the contract
security force, were also contacted during the inspection period .
16
I
SI'RATEGIES TO PRODUCE lisISI'ENT, IDGH PERFORMANCE *
PERFORMANCE ISSU~ .
- ACKNOWLEDGE CHALLENGE - BY
EVERYONE (including DPHoffman and
stafO.
'
Driving Force: Our human perfonnance is
1101 allowing us lo achieve the standards and
goals set in our Business Plan.
AN EXCESSIVE COOI...DOWN RATE OF
THE PRIMARY COOLANT SYSTEM
OCCURRED. WE MUST LEARN ALL WE
CAN FROM THIS AND Pl{EVENT THIS
KINO OF PEIU'Ol{MANCE ERl{01{
FROM HAPPENING IN THE FUTURE.
NOD MANAGEMENT HAS A
RESPONSIBILITY TO ASSURE THAT
THE APPLICATION OF DESIGN
AUTHORITY IS ROHUST AND
EFFECTIVE.
STRATEGY
Senior NOD management will conununicate to ~II
NOD employees the significance of inconsistent
pcrfonnance. Gain acceptance by all NOD
employees that perfonnance needs to be
improved.
l~oot cause assessment of this event will he
performed. The implkat ions for senior NOD
management action will he assessed prior to
.. slm1up. The NRC will he brief eel re~ardin~ tlu.*
aclions rclalcd lo lhe cooldow11 t*vcnl.
A review of rccenl design cngi11ccl'ing relalcd
plant evenls will he cond11ctecl, i11cl11cli11g lhc
inoperable safeguards room cooler fans.
Necessary corrcclive aclions will he clcterr11i11ecl.
l111111ediak co1Tcclivc aclio11s will he co111pkll'd
prior lo slnr1iap.
I RESPONSIBILITY I
TPllagan -
Prior to Critical
GBSladc - Prior to
II ea I 11 p
Rl>Orosz - Prior to
Power Operations
u------------------------------*--------------*-*---**** *----*-***-*------
1-ialis.ad('S Nuclear Planl
St*ph*ml>c~r 29, I 99;\\
Page I
--..
I
I
. STRATEGIES TO PRODUCE,NSISTENT, IIIGH l'EIU<'ORMANCE *
PERFORMANCE ISSUFS
RECENTLY, PLANT EQUIPMENT WAS
CHANGED WITHOUT INVOKING THE
'MODIFICATION PROCESS~ I
A PIUMAR\\' OB.IECTIVE OF NOO IS TO
MAINTAIN THE MATEIUAL CONDITION
OF TllE PLANT WITlllN ITS DESIGN
HASIS. RECENTLY, A PRESSUIUZER
NOZZLE TO PIPING WELD CltACJ<
l>EVELOl'EL> TO BECOME A TllROUGll
WALL CRACK. THE IMPLICATIONS OF
TlllS EVl~NT MUST HE ADDRESSED.
l~1li'>.lult~ Nuclcnr lttlrnt
September 29, 1993
- ~*-----------*
STRATEGY
RESPONSIBILITY
Plant administrative processes will be reviewed to
RDOrosz - Prior to
determine if .they allow 'changes' to the plant
Power Operations
outside the approved Design Basis without
adequate review. Procedure and organizational
responsibilities will he changed as necessary to
prevent this.
A plant specific c11gi11ecri11g cval11alio11 of lhe
primnry wntc~r .'ltl'l'SS COl'l'OSiOll Cl'lldd11g ('OIH'C'l'll
\\viii he pcrforuwd 1111<1 will he <
00111plel eel pl'ior Io
plant star1up. A strategy will he clcvclopcd to
dent with Industry co11<'l'1*11s rdall*cl ,,, pri111ar.Y
mater stress col'l'oslon c1*acld11g. Palisades issues
will he 11sscsscd 11 ncl <1<'11 It with 1a.11 11 gc*11c*ric.
concern relevant to the i11cl11stry.
All adio11s
required to assure safe operation in lhc m*xl
operating cycle will he complelcd prior to
startup.
IU>Orosz - f>rio1; lo
I lc*al ll11
STRATEGIBS TO PRODUCE CONSISTENT, IIlGII PERFORMANCE
PERFORMANCE ISSU~
NOD MANAGEMENT RECOGNIZES ITS
NEED TO BECOME MORE EFFECTIVE
IN CREATING THE CONDITIONS AND
CAPABILITIES WHICH WILL PRODUCE
MORE EFFECTIVE PERFORMANCE. IT
WILL SEEK CANDID AND OBJECTIVE
FEEDBACK ABOUT ITS PERFORMANCE
AND WILL CONTINUOUSLY WORK TO
IMPROVE.
MANY VERBAL COMMITMENTS HA VE
UEEN MADE IN PUHi.JC AND NRC
MEETINGS. NOD TAKES THESE
COMMITMENTS SEIUOUSLY.
TllE SAFETY SIGNIFICANCE OF
DISCREPANCIES IDENTIFIED BY CCP
REVIEW NEEDS TO BE QUANTIFIED.
P*.tlisade-; Nuclear Plant
~~ptcmh<~r 29, 1993
STRATEGY
An assessment will be performed to review
management's current approach to the challenges
we face and will provide information to
corporate management about the effectiveness of
our managemerit process and proposed actions.
An independent consultant (Tencra) will be
utilized to support this assess.ment.
Corporate ma11agc111e11t will he briefed 011 IN PO
field notes and the information obtained fro111 the
site debrief.
Public meetings and enforcement conference
meeting 111in11h.*s will he reviewed to c11s111*t* all
commitments have heen met.
The CCP discrcp1111cy list will he reviewed to
ensure the Plant Review Committee has been
made aware of any signific:1nt iss1u.*s. Any new
issues wlll be dispositioned prior to plant star111p.
Page 3
RESPONSIBILITY
RM.Rice - Prior to
Critical
I> 1'11 off man - Prior
to Power
Ope rat ions
CBSladl' - Prior
lo Crit iral
IU>Orosz - Prior to
I feat up
STRATEGlliS TO PRODUCE CONSISTENT IUGH PERJ:i"'ORMANCE
.
.
'
PERFORMANCE ISSUES
PROCEDURAL PROBLEMS RELATEO
TO TEST ANO SURVmLLANCE
ACTIVITIES HA VE BEEN PREVIOUSLY
IDENTIFIED AND MUST RE RESOLVED.
THE INDETERMINATE STATUS OF
NODS HAS BEEN A SIGNIFICANT
CONTRIBUTOR TO AMBIGUITY. SOME
NODS llA VE NOT BEEN CANCELLED AS
PLANNED; AND SHOULD BE.
HUMAN PERFORMANCE l~OOT CAUSE
EVALUATIONS PROVIDE IMPORTANT
INFORMATION FOR NOD
MANAGEMENT.
NOD MANAGEMENT RECOGNIZES ITS
RESPONSIBILITY AS A NUCLEAR
LICENSEE TO KEEP THE NRC
INFORMED.
l~1li..ad~ Nuclear Pta'nt
September 29, 19'J3
STRATEGY
A 111111fi-cli.c;dpli11nry h*n111 n*vil'w will he
conducled of all new or previously prolJlcmalic
special tests and si1rvcillancc procedures llrnl will
he implcmcnlcd prior lo startup. The review
and any procedural revisions will be performed
prior to their use during plant sta1111p.
An SMSC Meeting will he conducted to cancel
NODS which arc ready for cancellation.
RESPONSIBILITY
CBSlaclt* - All
mill's! ones (as
procedure is
rcq 11 ired)
.J.J Fremeau - Prior
to Critical
Pnlisndcs llPES Coordinalor will review 1993
CBSlade - Prior to
Human Performance Event/Deviation Repo11s lo
Critical
determine if there arc any common causal facl ors
which have not been addressed. The results for
Palisades will be reviewed pri01' lo planl startup
at n Plant Review Commiltce mccling which will
include the NOD Senior M:rnagemcnl tea111.
A. co1111nunication slratcgy lo l<c<.*p NRC i11for111t'cl
of significanl (~V('
1
a1ls will IH* dcv(*lopt*cl. l<t*.v
conlacls, frequency of meetings, and
con11111111ication loots will be developecl.
Page 4
l'M Do1111clly - Prior
to ll1*al11p
PROCESS FOR LONG-TERM HUMAN PERFORMANCE IMPROVEMENT
BUSINESS PLANNING MEETING -
OCTOBER 14
Accelerated Human Performance Action Plan
Long-Term Human Performance ~ctio~ ?lans
Status of Short-Term Human Perforwar.ce Action Plans
NRC UPDATE MEETING -
LATE OCTOBER
TENERA DIAGNOSTIC ASSESSMENT -
OCTOBER 29
Root Cause of Human Perf orrnance Problems
Cultural/In.stitutional Issues
Recommendations for Improvement
BUSINESS PLANNING MEETING
-
NOVEMBER 10
TENERA Diagnostic Recommendations
"Performing on the Job" Strategy
Review of Performance Against Business Plan Targets
Status and Update of Business Plan Action Plans
The outcome of this meeting will be an update to the
Business Plan for 1994.
NRC UPDATE MEETING -
LATE NOVEMBER
BUSINESS PLAN MEETING -
EARLY JANUARY
1994 Goals and Objectives (Reflecting the Updated
Business Plan) .
NRC UPDATE MEETING -
LATE JANUARY
1994 BUSINESS PLAN MEETINGS TO BE SCHEDULED
.
.
--*-*
.
"!'.
- .
.
.::
.*Non** NOW
.News of immediate interest to employees of NOD September 27, 1993
PALISADES MONITORING PROGRAM UPDATE
This year a number of issues and events arose involving human performance at
Palisades. These issues and events have impacted our continued success as an
organization and as such, demands all of our attentio_n to improve.
Over the past several weeks, the Palisades Management Team has been
collectively meeting to create a plan of action to help us understand and address these
human performance issues. The first step of.the plan involves systematic monitoring
of individual and group performance through field observation as well as Corrective
Action and Human Performance Enhancement System (HPES) trending. The field
.onitoring of performance will be conducted as an addition to the periodic reviews
organizational performance, equipment performance, and material condition that
have been on-going to determine the need for action in these areas.
The principal concept of the program is to observe the general environment
- .employees work in: the procedures. the processes, and the barriers that employees
cope with to accomplish their work. Then we will work to break down those barriers.
Field notes will be used by the Management observers, similartoan INPO Field Team,
- to permU recall and follow-up. Each manager will conduct two observations monthly.
We will then meet to compare notes and define actions and additional areas to
monitor the next month.
In addition, within the coming week, Vice President, Nuclear Operations, David
Hoffman, will be issuing an all-employee communication detailing an Action Plan for
the NOD Management Team to address Human Performance improvement for both
the short- and long-term future .
- -*
I.
PALISADES MONITORING PROGRAM
- Revis ion 0
PURPOSE
To systematically monitor organizational performance issues thro~gh
field observation, corrective action and HPES trending, NPAO monitoring.
and NOD management monitoring.
II. CONCEPT
Palisades Plant Management will establish a focused monitoring pian
which incorporates direct field observations, reviews of results of
existing management systems, and outside observation activities by NPAD
and others.
A periodic review of the current level of organizational
performance, equipment performance, and material condition will be held
to determine the need for additional action.
II I. PLAN
A.
Palisades Plant Management staff shall meet and identify specific
topics or areas to be monitored for next quarter.
The purpose of
this discussion is to focus monitoring efforts so that consistent
data is obtained for further review.
8.
A monitoring schedule shall be established imonthly identifying
topics.or activities to be monitored and specific individuals
assigned to monitor.
C.
Monitoring shall be performed as assigned.
Written summary of
observations should be submitted within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of monitoring.
D.
Monthly review meeting shall be held with Palisades Management
Staff to review results of following, as apprqpriat_e:
1.
Field monitoring.
2.
HPES reviews.
3.
Corrective Action trending.
4.
Departmental monitoring systems (RDR's, rework monitoring,
etc.}.
5.
NOD management visit program.
6.
NPAD monitoring results .
7.
Internal and External audit report results.
Page Two
Monthly review meeting should have a set format.
Specific topics
which should be covered include:
1.
Human Performance.
2.
Work Process performance.
3.
Equipment performance.
4.
Effectiveness of previous actions.
5.
Topics for future field monitoring.
- D.
Action assignments shall be made based on the results of the
review meeting.
E.
Quarterly, the Plant General Manager should report the results of
this monitoring program with the VP Nuclear Operations.
IV.
RESPONSIBILITIES
A.
B.
The Plant General Manager is responsible for the impler.iEntation of
this program.
He is also responsible for quarterly reporting of
results to the VP Nuclear Operations .
The Operations Manager is responsible for the scheduling of
monthly review meetings, including written documentation of the
results of the meeting.
C.
Various Management personnel are responsible for perfon:iing
monitoring activities as assigned.
V.
RECORD REQUIREHENTS
A,
The only records required to b~ retained ~re the minutes of the
monthly Management Review Meeting.
These records shall be filed
in DCC.
VI.
PROGRAM CHANGES
The program may be changed with the approval of the Plant General
Manager.
Changes to the program shall be docU11Dented as revisions to the
program .
Plant General Manager
- ----
1.
FIELD OBSERVATION GUIDANCE
Observe as many aspects of the activity as is practical including the pre-job
briefing and the follow-up documentation.
2.
- Observe - don't supervise. You are there to observe not consult
(Unless, of course, you need to intervene to prevent personal injury or equipment
damage.) It is okay to ask questions but do not provide direction_
3.
Log TIME on the Field Notes periodically.
(If something unusual occurs you may want to increase the frequency of your
timing notes. For example: if they are interrupted by the page and it would help
your evaluation to know how long it took them to get back on track.)
4.
lndude every factual observation and interaction that occurs.
5.
Try to withhold judgement while taking Field Notes. Record facts not opinions_
6.
Evaluate the Field Note facts using the Field Observation Checklist
7.
Conduct a feedback session with the workers and supervisor. If a problem or
deviation occurred, determine if a corrective action document is warranted.
8.
Forward all documentation to Debbie Beach .
OBSERVATION FIELD NOTES
Workers Involved
Dateffime
Supervisor
Location
Observer
Activity
I TIME I
oss=RVATION FACTS
I EVALUATION
jj
I .
'
!
I
l
ij
-
--*-
- I TIME I
.
"-
OBSERVATION FIELD NOTES
(Continuation Sheet)
OBSERVATION FACTS
I
Page _of_
EVALUATION l
--*
FIELD OBSERVATION JOB AID
Workers Involved
Supervisor
- Observer
Datemme
Location
Activity
PREPARATION
Individuals are trained and qualified to perform the task
Prejob briefing conducted IAW prejob briefing job aid
.Prerequisites were completed and verified
Procedures, precautions, and limitations were reviewed
Tools/materials were available and ready
Job was properly planned and scheduled
PERFORMANCE EVALUATION
Crew successfully completed the task
Written instructions were used and adhered to and were adequate
Crew{s) worked together as a team to complete the task
Industrial safety standards were rr.et
Radiation safety standards were met
Initiative and innovation were demonstrated in completing the job
Decisions made were conservative
Supervision kept appraised of job status and problem5
Communications IAW departmental communications policies
REMEDIAL ACTIVITIES .
Effective proper feedback provided to workers
Appropriate corrective action documents generated
Procedures changed as required
- CLEANUP/DOCUMENTATION
lessons learned captured in historical file
Tools and equipment returned to the originaVproper condition
Documents/records-were- completed correctly
Work area returned to the originaVproper condition
NII
NOTES/COMMENTS (Include comments for all.needs improvement {NII) items and for any exceptional
performance)
N/O
I
To
From
Date
Subject
cc
MA£ngle, Palisades
BAFarnworth, Palisades
OLBeach, Palisades
September 8, 1993
PALISADES PLANT:
MANAGERS MEETINGS.
CSmith, Palisades
CONSUMERS
POWER
COMPANY
Internal
Correspondence
Below is a list of upcoming managers' meetings,:: be held in the Managers Conference
Room from 1100-1300, including dates and tJpic(s) fer discussion.
DATE
TOP!C(S)
September 14
STANDARDS AHO EXPECTATIONS
16
Development
21
Corrcunicate
24
Monitor
28
Accountability
October 6
HONITORIHG
8
Review Program
14
MANAGING CHANGE
Identification
Assessrnient
Pi ann ing
Imp l e-:nenta ti on
Priorities
October 19
COHHUHICATIOH
27
Content
29
Tools
Feedback
November
3
LEADERSHIP ANO MAHAGE11EHT SKILLS
9
Coaching
--
- .
--
17
Counseling
24
Feedback
30
Development Haves
Department Head Alianment
December 9
PERFORMANCE RECOGNITION
.
~
.
OPERATIONS PERSONNEL AVOIDABLE OCCURRENCES
20
..
CJ)
- (AS DETERMINED BY ER'S AND DR'S)
1993 YEAR TO DATE
Both D/Gs Out Of Service.
Misaligned !so-Phase Bus Cooling.
CV-0521 Mispositioning.
Tagging Error On C-903A.
Improper Lineup For Q0-1.
CROM 20 & 31 Excercizcd In Error.
SFHM Unclerload Bypassed .
w
(.)
I
- ' ' ' * t ''' ' ''''' '* '''
SID Cooling Temperature <70° F.
Failure To Uncouple Control Rod
Failure To Close MV-PCI094A During
R0-65
z
w
a: 15
a:
- >
(.)
(.)
0
w
_J
co 10
<(
0
0 >
<(
0
5
a:
w
co
~
- J
z
..................
- .:.:.:.* .. ".:*.:*,:*.:*.:*.:*.
.*::::::*:::::::::::::::::::*:::::.:.::::::::::::.
.:*. **. **.:* ........ **.:*. ",:*.:*.:
0-----...
1989
1990 :
........................
- .
, ... , ................. .
.. .. ** ........... " .
. .. .. ,. , ............. .
.. , ....... , ........... .
......................
.. .. " ................ .
, ... , ............. ..
..................
1991
1992
1993 (YTD)
. 935090 PCS HEA TlJp I COOLDOWN
PORV LINE CRACKED WELD LOCATED IN HOT SHUTDOWN
550 ------------------------*------*---*---*-----.
500
450
400
350
300
250
200
150
100
SDC RETURN_;>-
TEMP
50
,~.--.---.---.---.-..---..----.--..--..---..---.--.-..---.--.---.----.-,-r--1--r--1-r--r--*-1--*1--*1-- -.--r-,-,-
12:00 AM
12:00 PM
12:00 AM
12:00 PM
12:00 AM
12:00 PM
12:00 AM
Sepl5,'93
Sepl6,
193
Sepl7,'93
Sepl8,'93
935090 PCS COOLDOWN
PCS COOLDOWN AFTER PORV LINE CRACKED WELD FOUND IN HOT SHUTDOWN
500
450
400
350
300
op
250
200
SOC RETURN
TEMP
- >
150
100
50 ~--..--.--.---.-~-.---.--r--,.---,i--..--,
-,--,--.--.-.--.--.--,-,---,-r::-i--,--r--r-r-r-,*-r-r-1--y*-1-T-!-T-T-.-...---.---I
6:00 PM
9:00 PM
12:00 AM
3:00 AM
6:00AM *
9:00AM
i2:00 PM
3:00 PM
6:00 PM
Sep 16,
193
Sep 17,
193
2 MINUTECFMS DATA
. ' .
Attachment 2
.-
CONSU1\\.1ERS PO\\VER CO:\\lPAl'\\i'Y l\\1EETING \\'1TH J\\1RC
INCO~L 600 ISSlJES
INDUSTRY AA"'D PALISADES EXPERIENCE
OCTOBER 21, 1993
AGENDA
1. Introduction
RD Orosz
2 . Industry Inconel 600 Experience
PD Fitton
- Failures and Corrective Actions
- Safety Assessments
3. Palisades. Experience
DJVandeWalle
- Failures and Corrective Actions
- Safety Assessments
--
4. Cause of the Relief Valve Nozzle
RBJenkins
Safe-End Crack
5.
Palisades Inconel 600 Program
DABemis
6. Long Term Corrective Actions
RDOrosz
INDUSTRY INCONEL 600 EXPERIENCE
CEOG formed an lnconel 600 Working Group in late 1989. This
group was formed to investigate the Primary Water Stress Corrosion
Cracking (PWSCC) leaks at:
Calvert Cliffs
San Onofre Unit 3 (1986)
St Lucie Unit 2 (1987)
EdF * (7 Units) (1989) (France)
AN0-1 (B&W) (1990)
GOALS:
1. Determine root cause and contributors
2. Evaluate susceptibility to PWSCC at other CE plants
3. Determine safety consequences if a leak were to occur
4.
~ ~~tablish methods and frequency of inspection
5. Develop repair methodologies
--*
INDUSTRYINCONEL600EXPERIENCE
IN Silltfl\\iARY WE LEARNED:
1.
For Primary Water Stress Corrosion Cracking (PWSCC) to occur
we must have:
A.
Susceptible material
B.
High temperature
C.
High tensile stresses
,,
I
I.
I !
I
!
INDUSTRYINCONEL600EXPERIENCE
- OUR PRELIMINARY EVALUATION CONCLUDED:
1. The pressurizer is by for the most susceptible location due to
normal operating temperatu'res.
2.
High strength material was most susceptible because it could
retain higher
levels of residual stress without yielding, and
- hence limiting the stress.
3. . PWSCC in inconel 600 penetrations is an economic issue, not a
safety issue. This conclusion was based on work demonstrating:
A. We expect axial cracks with J weld configurations
I . Field data
2. Analytical modeling
3. Mockup measurement of residual stresses
B. Leakage could occur for several fuel cycles without
compromising the integrity of the PCS pressure boundary.
4. Leakage is most likely to occur in the pressurizer vapor space .
INDUSTRYINCONEL600EXPERIENCE
PRESSURIZER LEVEL TAPS
1.
2. Contracted CE to measure residual stresses in mockup
3. Conclusion: No significant residual tensile stresses are not
produced by this configuration. (Reference 91-ESP-77 A)
CONTROL ROD DRIVE MECHANISM REACTOR HEAD
1.- Goals same as pressurizer
2. Palisades is participating in these activities
3. Work to date has supported conclusion that PWSCC does not
. Jeopardize PCS integrity for these penetrations ..
-.
PALISADES INCONEL 600 EXPERIENCE
Pressurizer relief valve nozzle safe-end crack .
circumferentially oriented crack in heat affected zone of
safe-end-to-pipe welq
repaired weld with like design
Pressurizer temperature instrumentation nozzle leaks .
axially oriented cracks in nozzle near structural "J" weld to
pressurizer vessel
temporary modification involving installation of exterior weld
"pad" addition and severing of Inconel sleeve between "pad"
and "J" welds
3"
Relief
Valve
~zzte
3"
Safety
Valve
Nozzle
l"
l"
Level
~le
TOQ View
Level
Nozzle-.4-"--..,.
Bottom View
Manw~
3"
Safety
Valve
~zzle
Relief
v at\\'t
Noz::zle
I
L
... ----121f co :Ret----
Su~rt
Skirt
I
llO ii' 0 IA II nsido Cladding1
Heater
Su~e
Noz:z e
, ..
!
T!~O
~.;ozzie
I
130f' 00
Elevation
r/6UR£ 1
PRESSURIZER SHED
. -
.
GENERALEQUIP:MENTLAYOUT
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PflESSUR I Z ER ____.
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C.S.NCZZL~
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. .
. - -
.-
. --
... LONGITUD*INAL *SECTION *SHOWING WEL.D ,-- CRACK I
WELD RE.PAIR .
TE-0101 MODIFICATION
,... ,.... SA.-- -
.......
~.::::>.
rt :.N:_*
INCONC:L
NOZZL~
PRJ 0 n~~n ~xT~~roo
,..._, __ ..__,
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WE:.....O "PAD" ADDITION
C.S. PRESSURIZE=:
INTERNAL
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c = c* o ~
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-'*-*
LP..: 0!.NG
T- 72 PRESSURIZER
-*
PALISADES INCONEL 600 EXPERIENCE
Safety Assessment
Pressurizer relief valve nozzle safe-end crack
Evaluated other pressurizer and PCS nozzles with Inconel 600
for susceptibility *
Examined susceptible nozzles on the pressurizer
Evaluated lifetime of repaired pressurizer relief valve nozzle.
Repair is considered temporary with a lifetime of one refueling
cycle. A permanent repair will be implemented during the 1995
refueling shutdown .
Evaluated margin-to-failure of cracked pressurizer relief .valve
nozzle and concluded that nozzle had sufficient strength to
withstand normal and faulted conditions
Evaluated primary coolant system leakage detection system and
concluded that similar cracks would be detectable. This .leak
detection capability was demonstrated during the pressurizer
. relief valve nozzle leak .
- --
PALISADES PRESSURIZER SAFE-END CRACK
Cause of the Safe-End Crack
For primar*y water stress corrosion cracking (PWSCC), there
are three required factors. These are:
A metallurgical condition
An aggressive steam/water environment
A threshold stress value
The safe-end was made of a material with a high yield stress
which made it vulnerable to PWSCC. These vulnerabilities are
refl~cted by:
A high hardness HRC-22
A high yield stress of 77KSI
An assumed low (1600-1700°F) post"'.forging heat treatment
The safe-end was exposed to stagnant steam at approximately
640°F. The material would be highly sensitive to cracking at
this temp~rature.
- *
Although externally applied piping system loads induced by
pressure, weight and thermal expansion meet design
.
requirements and are relatively low, significant local stresses
. appear to exist at the Inconel/stainless steel safe-end weld.,
These stresses are due to:
The field welding process
The mismatch in pipe/safe-end sizes
The materials mismat.ch thermal effects
To add insight into the load assessment, analysis of dissimilar .
metal effects and temperature stratification were conducted.
These evaluations led to the conclusion that these loadings
along with the external piping loads, as induced by the piping
support system, were not the driving forces for the crack. The
local pipe/safe-end configuration and fabrication process are
judged to be as the primary romributors.
,:i
PALISADES PRESSURIZER SAFE-END CRACK
Repair of the Pressurizer Safe-End
The heat affected zone of the inconel material was removed and
the weld prep machined to assure that the new heat affected
zone does not have intergranular penetration or discontinuities
that could act as stress risers.
The fit-up between the safe-end and stainless steel pipe was
improved.
The inner surface of the weld was ground s~ooth .
There were no repair weld deposits made, to the inner surface _
of the weld .
All of these improvements and changes will extend potential
crack initiation time significantly, resulting in increased lifetime
for the repaired weld as compared with the original weld.
Very conservative crack propagation calculations indicate more
than 20 months at temperature and pressure (640°F, 2060 psia)
would be required for a crack to grow through wall. For the
. 15 month fuel cycle, an initial 0.039in semi-elliptical crack of a
6: 1 aspect ratio would just grow through-wall.
Time to crack initiation in the heat affected zone of the new
safe-end should be comparable with that of the original safe-end
heat affected zone. No credit is taken for this in the calculation
of time to through wall cracking.
Therefore, the lifetime of the new weld will exceed one
operating cycle.
- '
...
PALISADES INCONEL 600 EXPERIENCE
Safety Assessment
Pressurizer temperature instrumentation nozzle leaks
Confirmed that cracking is similar (i.e. axial orientation) to that
experienced in other industry instrument nozzles
Evaluated other pressurizer and PCS nozzles for susceptibility
Examined susceptible nozzles on pressurizer and PCS loop
Evaluated lifetime of temporary modification~ Modification is
considered temporary with a lifetime of one refueling cycle .
Combustion Engineering Owners Group generic safety
evaluation applies
, .
PALISADES INSPECTIONS/LEAKAGE MONITORING
Palisades outage/startup inspections
l. ASME Section XI ISI inspections
2. PCS boric acid walkdowns
3*.
System Engineering inspection of pressurizer heater sleeves
4. PCS walkdown at 2150 psia after each refueling outage.
On line leakage monitoring
1. Daily PCS leakage calculation
2..
Monitoring containment parameters
Sump level
Containment temperature
Containment humidity
Radiological conditions
Biweekly containment. tour of lower elevation
PALISADES INCONEL 600 PROGRAl\\1
I.
BEFORE 1995 REFOUT
A.
Plan for replacement or justification of continued operation
of temporary Inconei 600 penetration repairs.
B.
Implement comprehensive PCS penetration inspection
program, based upon industry and Palisades experience:
1 .
Maintain list of PCS penetrations prioritized for
inspection and maintenance, including the following
considerations:
a.
Material types.
b.
Postulated mode(s) of failure.
c.
Potential safety and economic impacts from
failure.
d.
. Expected reliable life (Reliable life ends when
continued use or repairs are not justifiable for
safety, reliability, exposure, or economic
reasons).
e.
Conditions required for inspection, repair, and
replacement.
2.
Establish enhanced inspection program for PCS
penetrations approaching end of expected reliable life.
3.
Develop plan to improve or replace penetrations prior
to expected end of reliable life.
-
4. . Ensure that inspecJion program effectively bounds
conditions under which cracking is found.
5.
Evaluate and qualify non-destructive examination
techniques for detection of PWSCC .
C.
Develop remediation plan for cracked and leaking Inconel
600 penetrations.
PALISADES INCONEL 600 PROGRAM
.. II.
1995 REFOUT
A.
Replace:
l.
Temporarily repaired penetrations that are not
justified for continued operation past one cycle.
2.
Other penetrations identified for replacement with
improved materials or designs.
B.
Inspect old, removed penetration components for:
1.
Adequacy of previously performed repairs.
2.
Evidence of aging as compared to expected rate.
3.
Verification of conformance to expected failure
modes.
,.
c.
Inspect penetrations in accordance with PCS penetration
inspection program.
D.
Based upon inspection results, expand inspection scope as
needed to bound cracked penetrations.
E.
Remedy found cracked and leaking penetrations in
accordance with remediation plan.
III.
SUBSEQUENT TO 1995 REFOUT
A.
Adapt penetration inspection program and remediation plan
after each Refout, based upon updated industry and
Palisades experience.
B.
Execute penetration inspection program and remediation
plan during each Refout .
- .
PALISADES INCONEL 600 EXPERIENCE
Long Term Corrective Actions
Evaluate design of pressurizer relief valve nozzle and PORV line
and perform modifications necessary to assure a suitable nozzle
lifetime.
Perform permanent modification to pressurizer temperature
instrumentation nozzles.
Develop comprehensive program to deal with Inconel 600 issues at
Palisades. Program to include:
Evaluation and qualification of non-destructive examination*
techniques for detection of PWSCC .
Development of an augmented inspection program for
Inconel 600, including temperature nozzles, safe-ends and
control* rod drive nozzles.
Planning for replacement of Inconel 600.
Contingency planning for inspections/repairs of any future leaks .