ML20134J439
| ML20134J439 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/08/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20134J413 | List: |
| References | |
| 50-348-96-09, 50-348-96-9, 50-364-96-09, 50-364-96-9, NUDOCS 9611150212 | |
| Download: ML20134J439 (44) | |
See also: IR 05000348/1996009
Text
. _ . . _ . . . _ _ - - - _ .- _ -
.
,
.
~'
. t
!
'
.
! 1
l
U.S. NUCLEAR REGULATORY COMMISSION (NRC)
F
REGION II l
l
!
'
!
Docket Nos: 50-348 and 50-364-
Report No: 50-348/96-09 and 50-364/96-09 ,
,
Licensee: Southern Nuclear Operating Company (SNC), Inc. f
Facility: Farley Nuclear Plant (FNP), Units 1 and 2 i
Location: 7388 North State Highway 95 !
Columbia, AL 36319 i
!
Dates: September 1 - October-12, 1996
i
1
Inspectors: T. Ross, Senior Resident Inspector
- J. Bartley, Resident Inspector
J. Blake, Reactor _ Inspector (Sections M2, M3.
'
,
t
M6, and M8) l
R. Caldwell, Resident Inspector (In training) i
R. Chou.. Reactor Inspector (Section E1.1 thru 3)
M. Ernstes Operator Licensing Examiner (Section )
05.1) !
W. Kleinsorge, Reactor Inspector (Sections R1.2, 1
R2. R3, R7) !
l J. Lenahan, Reactor Inspector (Section E7.1 and ;
2) !
l
l Approved by: P. Skinner, Chief. Projects Branch 2
l Division of Reactor Projects
f
f
$
l
i Enclosure 2
,
9611150212 961108
PDR ADOCK 05000348
G PDR ,, l
)
. --. - - -.- - - - - . - ~ ~ - . ~ - - - - -
.
i
-
l .
. I
Farley Nuclear rP Units 1 And 2
- NRC Inspection Report 50-348/96-09. 50-364/96-09 i
This integrated inspection included aspects of licensee operations. !
engineering, maintenance, and plant support. The report covers a 6-week I
period of resident and regional inspections.
.
Ooerations
e Overall, both units operated well at steady state full power. The
l
conduct of operations by Operations personnel and management was l
consistently in compliance with procedures and regulatory requirements.
l and reflected conservative operation (Section 01).
e Shiit operators remained very attentive to plant conditions, and were
quite knowledgeable of plant status and ongoing activities (Section
01.1).
e The transfer of Unit 2 new fuel to the spent fuel pool was very well
controlled (Section 01.2). ]
e Operator performance in response to the Unit 2 main feedwater transient
and shutdown of Unit 2 for the upcoming refueling outage was~ exemplary
l '(Sections 01.3 and 01.4).
I e A violation was identified for two instances of mispositioned v71ves due ,
to personnel errors by non-licensed operators that caused two letdown
system transients on Unit 2 and inoperability of the 18 emergency diesel
'
generator (Section 01.5).
e .Overall. housekeeping and-physical conditions of the plants remained
adequate (Sections 02.1 - 02.3).
e A noncited violation was identified for failing to incorporate a newly
approved license amendment into plant procedures (Section 03.1).
,
e -The requalification program complied with the requirements and standards
of plant procedures as well as the requirements of 10 CFR 55.59 for the
, areas inspected. There continued to be some problems in ensuring the
operators are properly retrained on identified weaknesses (Section
05.1).
e Operator overtime was well controlled (Section 06.1).
Maintenance
e Maintenance and surveillance testing activities were routinely conducted
,
in a thorough and competent manner by well qualified individuals in
, accordance with plant procedures and work instructions (Section M1.1).
Enclosure 2
l
_ _ _. _ _ _ - . ~ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ . _ _
.
. i
'
!
l 2 l
e. Major work activities (i.e., epoxy coating of component cooling water
heat.exchangers and control room air conditioning modification) were
well coordinated, exhibited good craftsmanship, and accomplished
.according to approved work instructions and procedures (Sections M1.2 -
and M1.3).
! e Current status and history of the FNP steam generators indicate that the
licensee has been doing a conscientious job of managing steam generator 3
degradation (Section M2.1).
'
e Licensee and contractor procedures 'for the evaluation and management of
steam generator eddy current data appear to be adequate for the
circumstances (Section M3).
l e . Steam generator degradation has been properly managed even though the i
!
integrated program for management has not been documented (Section M6). !
'
Enaineerino i
e. Design change control process procedures comply with regulatory l
l requirements (Section E1.1).
- Design change packages were generally of good quality. However, several l
'
minor discrepancies with the associated installation drawings were
identified (Section E1.2).
l e Additional examples of recent and pre-existing as-built design '
! discrepancies associated with cable tray and pipe supports were
identi fied. The number of welding discrepancies identified indicates
that the licensee's welding inspection program may be deficient (Section
'
E1.2). ,
e Program for control and handling of heavy loads was adequate (Section .
E1.3).
e Engineering and technical support and oversight of major maintenance,
modification and inspection activities were evident and effective
(Section E1.4).
e Quality assurance audits of engineering activities provided useful i
insights (Section E7.1).
l
e Safety system self-assessment program has been a very effective tool in
, identifying and correcting deficiencies in design and operation of
l critical safety systems (Section E7.2).
d
i Enclosure 2
i
.. - -. . -- --- . - - _ - - .. .
- - - . - ..
,
'
!
'
.
!
.
l
l 3 ,
'
Plant Sucoort
!;
e' Implementation of radiological controls in the radiologically controlled
areas were evident and generally effective. Overall. the radiologically
controlled areas were adequately maintained and posted (Section R1.1). i
- o Primary and secondary water chemistry controls were appropriate, and in
compliance with regulatory requirements and industry guidelines, with !
, some exceptions (Section R1.2 and R3.1). .
o Procedures, equipment, and practices for monitoring primary-to-secondary
leakage were appropriate (Section R2.1). :
o Primary and secondary water chemistry program was subjected to
independent audits, with appropriate action taken for identified
weaknesses (Section R7.1).
e Emergency drills were well coordinated by Emergency Planning personnel.
- and demonstrated the readiness of emergency response personnel and !
facilities (Section P4.1).
L e Routine security _ activities continued to be performed in a conscientious
and capable manner, assuring the physical protection of protected and
, vital areas (Section S1.1). However, in one instance, miscommunication
between Security and Health Physics resulted in a -trailer entering the
protected area without being properly searched. This was identified as.
an unresolved item (Section S1.2).
'
e Two apparent violations were identified regarding inadequate
installation and inspection of Kaowool one-hour fire barriers required '
by the Fire Protection Program. An enforcement conference to consider
escalated enforcement has been scheduled (Section F2.1);
i e One unresolved item was identified concerning the adequacy of
qualification testing on Kaowool fire barriers (Section F2.1).
L e Fire brigade drill demonstrated proficiency of fire fighting capability
l
(Section F5.1).
I-
Enclosure 2
_ _ _
_
-
.
' !
.
.
Reoort Details
Summary of Plant Status
l
1
Unit 1-_ operated steadily at 100% power for the entire inspection period. l
,
. Unit 2 operated steadily at 100% power for the entire ins)ection period.. )
except for two brief power reductions. On September 8. t1e unit was ramped
down to 70% power to search for main condenser tube leaks. Then again on
September 23. Unit 2 was ramped down to 95% power in response to transient l
conditions caused by the 2A steam generator feed pum) (SGFP). After 317 days i
of continuous operation. Unit 2 was shutdown on Octo)er 12 to begin its !
eleventh refueling outage (U2RF11).. !
I. Operations
l
01 Conduct of Operations
01.1 Routine Observations of Control Room Ooerations (71707)
Using Inspection Procedure (IP) 71707, the resident inspectors conducted l
frequent inspections of ongoing plant operations including routine tours
of the main control room (MCR) to verify proper staffing, operator
attentiveness, and adherence to approved operating procedures. The
inspectors also regularly reviewed operator logs and Technical
S)ecifications (TS) Limiting Condition of Operation (LCO) tracking ,
sleets, walked down the main control boards (MCB), and interviewed )
members of the operating shift crew to verify operational safety and
compliance with TS. The inspectors attended daily plant status meetings
to maintain awareness of overall facility operations, maintenance
activities, and recent incidents. Morning _ reports and.Farley Nuclear
Plant Incident Reports (FNPIR) were reviewed on a routine basis to
assure that potential safety concerns were properly reported and
resolved.
-Overall control and awareness of plant conditions during the inspection
period were excellent. During tours of the NCR, the inspectors
regularl
balance ofy observed thatannunciators
plant panel' very few MCB, emergency.
were in alarm atpower board
any one (EPB),
time. One
and
or two persistent annunciator alarms 3revented Unit 2 from achieving
" blackboard". However. Unit 1 and tie EPB were frecuently in a
blackboard condition. Operator attentiveness to, anc knowledge of..
plant conditions and status of ongoing activities continued at a high
level. The combined number of MCB deficiencies was reduced to less than
20. demonstrating management resolve for keeping MCB deficiencies as few
as possible. Many of the remaining deficiencies were on Unit 2 MCB.
awaiting tha upcoming refueling outage.
Enclosure 2
.
'
.
.
2
01.2 Unit 2 Transfer of New Fuel to the Soent Fuel Pool (60705 And 71707)
The resident inspectors observed the transfer of several new fuel
assemblies from the new fuel storage racks to the Unit 2 spent fuel pool
(SFP) in accordance with (IAW) FNP-0-FHP-3.0 " Receipt and Storage of
New Fuel." Revision 28. Licensee personnel were knowledgeable and very
methodical. The handling and transfer of new fuel assemblies was well
controlled and consistent with procedural and TS requirements.
01.3 Unit 2 Main Feedwater Flow Control Transient (93702)
On September 23, 1996. Unit 2 operators observed that the 2A and 2B
SGFPs were experiencing sudden speed oscillations. These oscillations
had a destabilizing effect on main feedwater flow to the steam
generators (SGs). Operators promptly initiated a >ower reduction and
sent a System Operator (50) to physically examine )oth SGFPs in the
turbine building. Operations subsequently concluded that the low
pressure turbine governor valve for the 2A SGFP was malfunctioning. The
steam supply to the low pressure governor valve was isolated and control
of the 2A SGFP was placed on the high In about
45 minutes after the transient began,plantpressure governor
conditions hadvalve.
stabilized
with the 2A SGFP back in automatic control. Unit 2 power reduction was
arrested at 95% power and then returned to full power a few hours later.
During the ramp up to full power, a resident inspector interviewed l
responsible operators, monitored 2A and 2B SGFP operation and walked -
down the Unit 2 MCBs. The inspector concluded that o)erator response to i
the transient was excellent and plant conditions had Jeen restored to )
normal, ,
l
01.4 Unit 2 Shutdown For U2RF11 (71707) ,
1
On October 11, 1966 Operations commenced a rampdown of Unit 2 IAW FNP- l
Unit '
2-UOP-3.1 " Power Operation."and was the removed
main turbine / generator
from the grid, which(MTG) power w
marked the commencement of U2RF11 at one minute after midnight on '
October 12. 1996. Immediately following removal from the grid,
operators conducted a MTG overspeed test IAW UDP-3.1. Appendix 4 and
FNP-2-STP-151.5 " Main Turbine Overspeed Test." The MTG tri) ped at 1934
rpm, well within the required acceptance criteria. Unit 2 slutdown
continued IAW FNP-2-UDP-2.1. " Shutdown of Unit from Minimum Load To Hot
Standby." After entering Mode 2. operators manually tripped the reactor
aursuant to UDP-2.1 and FNP-0-ETP-3661 " Control Rod Test and Evaluation
3rogram." Operators responded to the manual trip IAW FNP-2-EEP-0.
" Reactor Trip or Safety Injection." and then FNP-2-ESP-0.1 " Reactor
Trip Response." A resident inspector was in the MCR observing all of
the aforementioned activities. Overall, operator performance was
excellent and plant equipment response was per design. However, the i
inspector did note a few minor problems, which included: a) Manual i
reactor trip first out annunciator did not alarm (this problem had been
previously identified): b) Nuclear instrumentation system (NIS)
Enclosure 2
1
__
.
.
e
3
intermediate range channel NI-36 appeared undercompensated, and
prevented the automatic energization of NIS source range channels: and
c) Operators had difficulty maintaining Tavg within plus/minus 1.5
degrees of Tref as prescribed by UDP-2.1 (step 5.5.1). Each of these
problems were discussed with site management for resolution.
01.5 System Misalianments By System Ooerators (71707)
Within about a 6 week period, the plant experienced two serious system
misalignment events caused by S0 personnel errors that resulted in two
letdown system transients and the inoperability of an emergency diesel
generator (EDG).
On August 27, 1996, the licensee discovered that two valves (Transfer
Header To Storage Tank valve OSY52V524 and Manual Pump Discharge To
Transfer Header valve OSY52V529A) in the IB EDG fuel oil transfer system
were inadvertently left open. This misalignment of the fuel oil
transfer system effectively 3revented fuel oil makeup from the IB fuel
oil storage tank (FOST) to tie IB EDG day tank. Operations had
discovered the misaligned valves during a refill of the IC EDG FOST from
the auxiliary FOST when o)erators observed a high level alarm on the 1B
FOST and sent a S0 to checc out the problem. Subsequent investigation
by Operations, determined that the two valves in cuestion had not been
closed by the responsible 50 as was required by FhP-0-50P-42.0, " Diesel
Fuel Oil Storace and Transfer System " following a refill of the IB EDG
FOST on August 23, 1996. For about five days, fuel oil makeup to the IB !
EDG day tank was not possible. With just the day tank as a fuel supply,
the 1B EDG would be expected to run at full load for ap3roximately four
hours. The licensee reported this event to the NRC by _icensee Event
Report (LER) 50-348/96-05 dated September 20, 1996.
On October 8,1996, a S0 mispositioned two valves during two separate .
'
incidents while attempting to fluff the 2B mixed bed demineralizer IAW
FNP-2-SOP-50.4 "Demineralizer Resin Removal and Addition." During the
first incident, while performing Step 4.5.1.3 of SOP-50.4 which required
closing demineralizer uutlet valve V167B. the S0 inadvertently selected )
valve V167A and began closing the valve. Upon hearing an unexpected 1
change in flow noise, the S0 immediately reopened the valve and notified
the Shift Supervisor (SS). Prompt action by the S0 prevented a serious
overpressurization of the letdown line which would have occurred had he j
succeeded in fully closing V167A. thereby dead-heading letdown flow. !
During the partial closing of V167A, the Unit 2 MCR did receive a j
" Letdown Pressure High" alarm. After notifying the SS, the 50 proceeded
on with SOP-50.4, even though the SS was under the impression he had
secured from the evolution and was returning to the MCR to discuss the
incident. Unbeknownst to the Unit 2 operating crew, the 50 proceeded to
Step 4.5.1.5 which directed him to open demineralizer Johnson screen
outlet valve V161B. However, the S0 misread the )rocedure and 1
inadvertently selected V161A. which he opened. T1is caused letdown flow ;
to be diverted from the inservice 2A mixed bed demineralizer to the !
Enclosure 2 l
'
,
l
!
___ _ . _ . _ _ _ _ . ___.___ _ __ _ ___ ._
'
.
.
'
4
1
primary spent resin storage tank (SRST). In the MCR operators noticed
that auto makeup to the volume control tank (VCT) had initiated and VCT
level was dropping rapidly from the normal 40% level. Operators
promatly started manual makeup and were able to stabilize VCT level at
21% aut not restore it. The SS managed to contact the S0 and direct him
to close V161A. Operators were then able to restore VCT level.
Subsequent review of SRST level changes concluded that approximately 800
gallons had diverted from the VCT in about six minutes.
In both of the aforementioned events, a responsible 50 failed to perform
applicable system operating procedures as written resulting in
mispositioned valves that adversely affected operability of the IB EDG '
and initiated two transients involving the Unit 2 letdown system. These >
personnel errors are considered a violation of TS 6.8.1 that requires
compliance with plant procedures SOP-42.0 and 50.4. This is identified i
as violation (VIO) 50-348, 364/96-09-01. Multiple Valve Misalignments By
System Operators. ;
02- Operational Status of Facilities and Equipment ,
'
02.1 General Tours of Soecific Safety-related Areas (71707) .
General tours of FNP specific safety-related areas were performed by the i
resident inspectors to examine the physical conditions of plant
equipment and structures, and to verify that-safety systems appeared
properly aligned. Limited walkdowns of a more detailed nature of the i
accessible portions of safety-related structures, systems and components
l were also performed in the following specific areas:
o Control room air conditioning systems (CRACS) and emergency j
ventilation systems, trains A and B
e. Unit 1 and 2 residual heat removal heat exchanger (HX) rooms-
e Unit 1 and 2 turbine building
l e Unit 1 and 2 SFP and SFP HXs
j e Unit 1 and 2 component cooling water (CCW) pump and HX rooms
,
e Unit 1 and 2 low voltage switchyard
e Unit 1 and 2 service water intake structure (SWIS), including
service water system (SWS) pumas, switchgear, and strainers
e EDG Building - 1/2A.18.1C. 23. and 2C EDGs
i
e Unit 1 and 2 vital 4160 volt alternating current switchgear
e Unit 1 and 2 iping penetration room 100 foot elevation
e Unit 1 and 2 13ing penetration room on 121 foot elevation
o Unit 1 and 2 F) ventilation equipment rooms
e Unit 1 and 2 hot shutdown panels
- e Unit 1 and 2 turbine-driven auxiliary feedwater (TDAFW) pump rooms
e Unit 1 and 2 main steam valve rooms
,
e Unit 1 and'2 new fuel storage areas
!
Enclosure 2 l
!
I
l
'
.
1
5
Overall material conditions and housekeeping for both units were
generally adequate. Minor equipment condition and housekeeping problems
identified by the inspectors were reported to the responsible SS and/or
maintenance department for resolution.
02.2 Biweekly Insoections of Safety Systems (71707)
The resident inspectors used IP 71707 to verify the operability of the i
following selected safety systems:
e Unit 2 Containment Integrity
The inspector verified Unit 2 containment integrity using FNP-2-STP-
14.0 " Containment Integrity Verification Test." Revision 14. The
inspector verified accessible valves, breakers, hatches, and capped
pipes were properly aligned. No discrepancies were noted.
02.3 TS LCO Trackina and EDG Test Data Loa (71707)
Resident ins)ectors routinely reviewed the TS LCO tracking sheets filled ,
out by the s11ft foremen whenever a TS LCO action statement was entered. ;
All tracking sheets for Unit 1 and 2 reviewed by the inspectors were
consistent with plant conditions and TS requirements. An inspector also
reviewed the EDG Test Data Log book, and did not identify any
discrepancies.
02.4 Tao Orders (71707)
During the course of routine inspections, the following tag order (TO) I
and associated equipment clearance tags were examined by the inspectors:
e TO# 96-1012-2 2B CCW Pump
All tags and the T0 examined by the inspector were properly executed and ;'
implemented.
03 Operations Procedures and Documentation
03.1 Untimely InCorDoration of TS Amendment Into Plant Procedures (71707) i
By letter dated December 19, 1995. the licensee proposed to amend its TS I
requirements for CRACS and the control room emergency filtration system.
In this letter. SNC requested that the effective date for the pro)osed
amendment be applicable when LCOs were entered to implement the CMCS
design change. However. Amendment number 119 (Unit 1) and 111 (Unit 2)
,
for Facility Operating Licenses (FOL) NPF-2 and 8 were issued on May 21.
l 1996 effective upon the date of issuance. SNC recognized the problem
and called the Office of Nuclear Reactor Regulation (NRR) for
clarification. The NRR project manager (PM) stated that it had been
their intent to issue the TS amendment as requested by SNC. Rather than
Enclosure 2
l
. _ . -, .~ . . - - - . . . --. . . - .-
.
t
'
.
!
t
'
!
6
i
submit a letter to correct the TS-amendment effective date. the licensee <
assumed their discussion with the NRR PM was sufficient. Some time .
later, as implementation of.the CRACS design change became delayed, a !
resident inspector questioned the licensee's basis for not incorporating
these amendments into applicable plant procedures, and implementing the ,
associated CRACS surveillance requirements. Subsequent review by the'
'
NRC's Office of the General Counsel confirmed that both license i
amendments were in effect and could not be deferred without written NRC l
approval. Failure-to implement newly amended TS requirements in a
timely manner is usually a serious NRC concern. However, in this case,
the extenuating circumstances for this violation meets the guidelines of.
Section VII.B.1 of the NRC Enforcement Policy for a non-cited violation !
(NCV) identified as NCV 50-348. 364/96-09-02. Untimely Incorporation of
TS Amendment Into Plant Procedures.
05 Operator Training and Qualification !
05.1 Licensed 00erator Reaualification Proaram !
l
a. Insoection Scone (71001) l
During the period September 17 - 20, 1996, the inspector reviewed the
licensee's-licensed o)erator requalification program to determine '
compliance with 10.CFR 55.59. Requalification. Specific areas of review ,
included job performance measures (JPM) administration.. operating :
examination quality, documentation of results, and remediation, i
b. Observations and Findinas
1
The inspector observed the administration of -JPMs to staff Senior -l
Reactor Operators (SR0s) on the simulator and in the plant. The. '
licensee evaluators' grading was consistent with the inspector's
observations.. The evaluators effectively asked follow up questions
based on operator performance. This enabled them to determine the cause I
for an observed error and determine individual or generic program areas
in need of improvement.
JPM 3erformance'by the staff SR0s was marginal. Three JPMs were failed
by t1e operators. One of the SR0s failed two JPMs resulting in an
overall failure for that operator. Two operators missed the same JPM
when they misdiagnosed a problem with the rod control system and
unnecessarily tripped the reactor. 1
The documented remediation for the operator who failed the JPM
walkthrough adequately addressed the weaknesses observed by the
inspector. The inspector reviewed the documentation of remediation for
other operators who failed their annual examination this year. Farley
- 3rocedure FNP-0-TCP-17.3. " License Retraining Program Administration."
i Revision 14 dated 12/18/95. detailed the licensee's guidelines for
- remedial training. Section 3.15.2 of this procedure stated that "At a
4
Enclosure 2
l
.- - .-- - -, .-
.
-
.
7
minimum, the remedial training will consist of a written study guide and
assignment to an instructor who will monitor the progress of the
trainee. This training will be documented using a memo similar to
Figure 2". In two cases, the guidance was not fully utilized. There
were not clear objectives of the training nor was there any
documentation of an instructor being assigned to monitor the training
and initial for its completion. The operators had been documented as
having weaknesses in several competency areas. However, the remediation
documented in a training department memorandum, listed only two discrete
3rocedure issues and one general communication issue to review.
Remedial training was not consistent with areas in need of improvement
documented in simulator evaluations. Although the remediation was not
properly documented, and its completion was not initialed for by the
assigned instructor, the inspector was able to determine that adequate
remediation had been conducted for the operators' performance
weaknesses. This issue will be tracked as Inspector Followup Item (IFI)
50-348,364/96-09-03. Inconsistent Application of Remedial Training
Documentation Guidance.
The inspector reviewed the licensee's system for providing feedback of
generic weaknesses observed during regualification examinations. Peer
evaluators reviewed the simulator evaluations and listed areas where
additional training could improve performance. These items were
documented in a memorandum to the Operations Training Supervisor. The
last review, dated June 10, 1996 contained fourteen such items. This
system adequately supported the feedback portion of.a systems approach '
to training. However, in one instance the training was not effective.
A crew of operators incorrectly interpreted the EEP-0, " Reactor Trip or :
Safety Injection." foldout page guidance and implemented FRP-P.1. ]
" Response to Imminent Pressurized Thermal Shock Condition." >
unnecessarily during a lar e break loss of coolant accident. During' a 3
emergency preparedness dri 1 cbserved by the inspector on September 18, l
1996, the simulator crew made the same error despite remedial training. l
Training department evaluators had recommended a procedure change in i
March 1996, to help reduce the' susceptibility of operators making this
incorrect interpretation but they had not received any feedback on their
recommendation.
During review of the annual requalification training curriculum outline,
the inspector identified that more simulator training was provided for
shift operators than for staff operators. 10 CFR 55.59(a).
RequaHfication Requirements, does not distinguish between shift and
staff operators regarding participation in the program. Staff o3erators
received one week of simulator training per year. Included in tlat week
was the simulator portion of their annual operating examination. With ;
the annual examination typically being administered on Thursday and )
Friday, there were actually only three days of simulator training for '
staff operators. Control room operators received additional training .
during other weeks based on plant needs such as startups or shutdowns. l
Updated Final Safety Analysis Report (UFSAR) section 13.2.1.2 states the
Enclosure 2
!
i
.
.
.
8
retraining will last approximately one week requiring the student to
perform each of a list of 26 items. The training department tracked
completion of these items using license simulator retraining form OPS-
564, " Task Tracking Sheet." The inspector verified that all licensed
operators were documented as receiving at least this minimum training.
c. Conclusions
The inspector concluded that the licensee's requalification program
complied with the requirements and standards of plant procedures as well
as the recuirements of 10 CFR 55.59 for the areas inspected. The
licensee ceveloped and administered examinations that effectively
identified areas in need of improvement. Documentation of remediation
was in need of improvement to ensure adequate completion. Although
staff operators did not receive as much simulator training time as the
shift operators, all operators were found to have completed the minimum
requirements.
06 Operations Organization and Administration
1
06.1 Administrative Control of Ooerator Overtime (71707)
The requirements for control of operator overtime are contained in TS 6.2.2 and FNP-0-AP-64, " Work Schedules for Personnel." Revision 3. The
inspector interviewed two shift clerks and reviewed the shift manning
and biweekly payroll time records for Crews 1 and 6 for the period July
1 - August 1. 1996. The inspector did not identify any discrepancies
with the records cr control of operator overtime. There were no
instances of excessive overtime. The inspector also reviewed the
proposed schedule for the Fall 1996 Unit 2 Outage. The inspector found
that the schedule maintained overtime within the requirements of TS 6.2.2. The inspector concluded that o)erator overtime was adequately
controlled and that the records were t1orough.
08 Hiscellaneous Operations Issues (92901)
08.1 (Closed) IFI 50-348. 364/95-12-01. Lack of Measurable Performance
Indicators Due to Use of Westinohouse Owners' Grouo Emeroency Response
Guideline
This IFI identified that critical tasks were not an objective measure by
which the facility evaluators could determine if an individual or crew's
performance was satisfactory. Since that inspection, the licensee has
revised the critical tasks to provide the evaluator with measurable
objectives for determining satisfactory performance. This IFI is
considered closed.
l
l
l
l
Enclosure 2
1
.
.
.
9
II. Maintenanqq
M1 Conduct of Maintenance
M1.1 General Comments
Inspectors observed and reviewed portions of various licensee corrective
and preventative maintenance activities, and witnessed routine
surveillance testing, to determine conformance with plant procedures,
work instructions, industry codes and standards. TS and regulatory
requirements.
a. Jnsoection Scooe (61726 and 62707)
The resident inspectors observed all or portions of the following
maintenance and surveillance activities, as identified by their
associated work order (WO), work authorization (WA) or surveillance test
procedure (STP):
e FNP-2-STP-80.2 1C EDG Operability Test
e WA 121388 1A CCW HX Epoxy Coating Application
e WA 123723 2A CCW HX Epoxy Coating Application
e WO 596001466 1A CCW HX Tube Ins)ection Cleaning, Repair.
Stabilization and )1ugging
e WO S96001476 2A CCW HX Tube Ins)ection, Cleaning, Repair,
Stabilization and ) lugging
e FNP-2-STP-608.1 Loop A Main Steam Safety Valve Operability Test
e FNP-2-STP-151.5 MTG Overspeed Test
e FNP-0-ETP-3661 Control Rod Test And Evaluation Program
o WO S00079525 Train A CRACS modification
e WO 0463607 U2 TDAFW Pump (Turbine) lubrication
b. Observations. Findinos and Conclusions
All of the aforementioned maintenance work and surveillance testing
observed by the inspectors were performed IAW WO instructions,
procedures, and applicable clearance controls. No adverse findings were
identi fied. Safety-related maintenance and surveillance testing
evolutions were well planned and executed. Responsible personnel
demonstrated familiarity with administrative and radiological controls.
Surveillance tests of safety-related equipment were consistently
performed in a deliberate step-by-step manner by personnel in close
communication with the control room. Overall, the craftsmen and
technicians appeared knowledgeable, experienced, and well trained for
,
the tasks they performed.
In addition, see the discussion below regarding certain major
maintenance activities observed by the resident inspectors (Section M1.2
and 3).
Enclosure 2
i
>
_- .
,
'
,
!
.
,
,
.
10 ,
'
M1,2 WO 596001466 and S96001476. and WA 121388 and 123723. 1A and 2A CCW HX
Preoaration. Reoair. and Eooxy Coatina '
a. Insoection Scoce (62707)
Resident inspectors observed various aspects of the modification - i
activities associated with e) oxy coating of the service water inlet and !
outlet channel heads, tube sleet, and down tube'(inlet side only) for '
the 1A and 2A CCW HX. The work observed included hydroblasting, grit
blasting, tube sheet weld repairs, CCW tube stabilization, and the epoxy -
application. Inspectors also reviewed and verified imalementation of
applicable WO instructions: WAs: FNP-0-ETP-4418, CCW leat Exchanger i
Epoxy Coating Application; and Plastocor Application Specification for ;
tubesheet cladding, inlet end coating, and channel head coating.
l
b. Observations and Findinas
'
This was a major modification that involved a number of plant. employees )
and contractors for over three weeks. The work on the 1A and 2A CCW HXs '
was performed in parallel. In general, all observed work was
accomplished IAW work instructions and applicable procedures (i.e'., ETP-
4418 and Plastocor Application Specification). However, several process I
control and equipment problems were identified.
e Eleven broken / severed tubes in the 2A CCW HX, some with missing l
sections of tube;
e During application of 2A CCW HX primer inlet coat of epoxy,
ambient air and surface temperatures slightly exceeded the
established-limits: ,
e During the mixing of one batch of epoxy for_ the 2A CCW HX, epoxy
material temperature slightly exceeded the established limit;
e While being stored at a complex 3 warehouse, ambient air
temperature for epoxy materials slightly exceeded the established
limit for about four hours; and
e Inadequate control of combustible materials in the CCW HX rooms
Broken, severed, and fragmented tubes in the 2A CCW HX were subsequently ;
stabilized and fenced in as well as possible. Not all tube fragments
could be located or secured. A Nonconformance Disposition Report (NDR),
including a 10 CFR 50.59 safety evaluation, was issued to address this
condition. In this report the licensee concluded that the " failed tubes
and missing tube fragments within CCW heat exchanger 2A will not impact
_the CCW system capability to Jerform its intended design function or the
continued safe o)eration of t1e CCW system." [ Note, licensee also
discovered a num)er of damaged tubes in the 1C and 2C CCW HXs .sometime
later.] With regard to the temperature limits that were. exceeded during
Enclosure 2
. -. .- -- - - - -. - . - . .. --
.
i
.
I
11
the application and storage of epoxy material, the licensee requested
and received a letter from the e) oxy manufacturer confirming that the
slightly increased temperatures lad insignificant consequences.
Furthermore, the licensee confirmed that appropriate adjustments to the
curing times, and special application instructions, were implemented to ;
,
. accommodate the increased temperatures. ,
During the conduct of CCW HX epoxy coating modification activities, the !
,
licensee failed to exhibit positive control over the quantities of ,
to minimize the storage and placement of intervening combustiales -
between redundant service water inlet valves to the CCW HXs during epoxy ,
work on the CCW HXs. Controls were put in place for the 1C and 2C CCW
HX work. Subsequent evaluations by the Engineering Support (ES) group
confirmed that general fire loadings did not exceed limits established i
by the fire protection program. However, the program was explicit about
maintaining no intervening combustibles between redundant trains. The
inspector did note that hourly fire watches were in effect during the
duration of work on the A CCW HX. although for different reasons (i.e..
Kaowool concerns, see section F2.1). Also, continuous fire watches were
in place during those periods when activities required an Open Flame :
Permit. !
c. Conclusions
i
In general, all observed work was accomplished IAW work instructions and i
applicable procedures by qualified individuals. The overall process was I
'
l well controlled in a conscientious and deliberate manner. Application
of lessons learned was evident as the licensee progressed from the B. to
the A. and then to the C CCW HXs. Problems were appropriately addressed
and resolved. The review of CCW HX epoxy process, and NDRs for the
fragmented CCW HX tubes, was identified as IFI 50-348. 364/96-09-04. CCW I
HX Epoxy Coating And Broken Tubes, pending additional review by the
inspector.
M1.3 WO S00079525. Tie-In of New CRACS. A Train
a. Insoection Scoce (62707)
Resident inspectors observed selected portions of the CRACS modification ;
'
including: 1) installation of duct blankoffs: 2) removal of old
- compressor, receiver, and cooling coils
- and 3) installation and hookup
of new cooling coils. Previous observations of work on the CRACS
modification were documented in Inspection Report 50-348, 364/96-03. .
The CRACSWork
S95-0-8816. modification was implemented
was performed under WO- by#S00079525. Design Change Packa
I
Enclosure 2
!
<
.
.
.
12
b. Observations and Findinas
The licensee planned to take the A train CRACS out of' service (00S) on
Se)tember 5. -1996, to complete modification work. The work was
scleduled to take 25 days. The licensee planned to implement a revised
TS 3/4.7-.7.2 per FOL Amendment No. 119 coincident with taking the A-
train 00S. Amendment No. 119 waived the provisions of 3.0.4 (as applied
to TS 3.7.7.2) during the initial 30 days of implementation of control
room cooling design changes. Also, the revised TS adopted the Improved
TS surveillance for verification that the CRACS units can remove the
design basis accident (DBA) heat loads.
The licensee identified, on September 2.1996. that the test procedure.
FNP-0-ETP-1052 "B Train Control Room Air Conditioning System Test."
Revision 0. would not be a valid test as written for the current .
conditions. The IB control room pressurization filter unit heaters !
would not remain energized in test and the outside air temperature was
not greater than 85 degrees fahrenheit as required by the test ,
procedure. The licensee submitted Request for Engineering Assistance :
(REA) 96-1291 to Southern Company Services (SCS) to determine if l
additional heat was required for the MCR due to reduced outside
temperatures to assure the DBA heat load during test conditions. SCS
determined that the pro)osed test conditions (with outside temperatures
70 to 90 degrees fahrenleit) provided heat loads greater than the DBA' ,
heat loads. SCS also recommended that the direct current emergency
~
lighting be energized during the test to add conservatism. The licensee i
subsequently revised Engineering Test Procedure (ETP) 1052 to '
incor) orate SCS's recommendations. The inspectors reviewed the revision
to ET)-1052 and the SCS response to REA 96-1291 and determined they were
adequate. The licensee performed FNP-0-ETP-1052. Revision 1. i
successfully on September 7, 1996 and placed the A train CRACS 00S on
September 12. 1996.
The modification was completed and A train CRACS was returned to service
on September 27, 1996
On September 27. 1996, during-an internal review. licensee personnel
identified that locking tabs were tack welded to the cooling coil
mounting bolt heads without documented approval from engineering. The
licensee found that craft personnel had experienced difficulties with
tightening the cooling coil mounting bolts because the bolt head was not
accessible with the coils in place. The engineer and craft personnel ;
determined the solution was to tack weld sheet metal locking tabs onto !
the bolt heads to prevent them from rotating. The engineer discussed ;
this solution with SCS. SCS stated that they did not see a 3roblem with '
the locking tabs so the engineer incorporated the locking ta)
installation into the WO. The licensee initiated FNPIR 1-96-263 to
document the problem. Mock-up testing was aerformed to assure that the
attachment process had not degraded the meclanical properties of the
bolts. The licensee produced a sample set of tab attached bolts using !
l
Enclosure 2 '
.- _. - - - - - .- - - . . - - - - -- - .
.
\
'
.
2
.
13
the same material, welder etc.. as those installed. Testing involved
mechanical testing at FNP and metallurgical / nondestructive testing in i
Birmingham. Test results indicated that there was no adverse effect >
upon the life of the bolts with the attached tabs. The inspectors
reviewed the test results and concurred with the licensee's evaluation.
c. Conclusions
'
The inspector determined that overall the work was well coordinated and
craftsmanshi) was good. The licensee's responses to the deficiencies
were thoroug1 and well thought out. I
H2 Maintenance and Material Condition of Facilities and Equipment
M2.1 Steam Generator Intearity Maintenance and Testina
a. Insoection Scoce (73753)
Through licensee procedures, programs and associated records the
inspector reviewed the history of the Farley Units 1 and 2 SGs.
b. Observations and Findinas
Farley 1 and 2 are Westinghouse 3-loop units with series 51 SGs. Each !
SG contains 3388 U-bend tubes made of Inconel 600. The nominal tube
outside diameter is 0.875 inch with a nominal wall thickness of 0.050
inch. In Unit 1. the tubes were ex)anded full length of the tube-sheet
using the Westinghouse Explosive Tu]e Expansion (WEXTEX) process: in
Unit 2. the tubes were expanded using a mechanical hardroll process. l
Both units have replaced their original anti-vibration bars (AVB) with ;
adjustable 405 stainless bars.
'
Unit I reached initial criticality in August 1977, and to date has
completed thirteen (13) refueling outages. Tube degradation modes of
.
the Unit 1 SGs were as follows: l
l
e Outside Diameter Stress Corrosion Cracking (0DSCC) at Hot Leg l
Support Plates: (Confirmed by tube pull in 10/89 - R20C26
'
l
l from SG C.)
e Thinning at Lower Cold Leg Support Plates: (No significant change
l
'
in percent value since 1986.)
e ODSCC above Hot Leg Tubesheet: (Indications in sludge pile and in
free span area to first tube support plate. Confirmed by
tube pull in 10/89 - R20C26 from SG C.)
e Pitting above Cold Leg Tubesheet: (Confirmed by tube pull in
10/89 --R21C48 in SG C.)
e Primary Wate.' Stress Corrosion Cracking (PWSCC) at the WEXTEX Hot
. Leg Tubesheet Transition: (4/91 inspection was first !
rotating pancake coil at the top of the hotleg WEXTEX
. transitions. Circumferential and axial indications have
Enclosure 2
i
__ . _ - _ _ . _ _ _ - _ . _ _ _ _ . . _ _ _ _ _ . _ _ _ _ . _ _ _ _
.
l
'
i
14 i
i
been reported since 4/91. Ultrasonic testing (UTEC) ;
confirmed that all circumferential indications were !
initiated from the inside diameter-of the tube.) i
e PWSCC in Low Row U-bends: (Row 1 tube preventatively plugged in !
12/81. - Row 1 tubes unplugged and Rows 1 & 2 U-bends were ,
heat treated in 4/91.) +
e AVB Wear: (Indications represent old wear spots corresponding to i
locations of original AVB's, which were replaced with l
adjustable 405 stainless steel AVB's in the 1985 & 1986 !
outages.) '
e ODSCC at Free-Span Locations: (First identified during the 9/92 l
tube pull- (R18C40 - SG B.) No clear eddy current signals at
that time. In the 10/95 inspections, eight indications t
confirmed above the hot leg tubesheet and seven other indi- !
cations confirmed in other hot leg free-span areas.) ;
Status of Unit 1 SGs after the 13th refueling outage was as follows: ,
.SG_A SG_B SS_C i
Total Inservice Tubes: 3116 3223 3131 l
Total Inservice Sleeved Tubes: 56 37 115
Total Inservice Sleeves: 67 42 149 >
Total Plugged Tubes: 272 165 257 j
Cumulative Plugging Equiv. for 2.41% 1. 564% 5.212% !
'
Sleeved Tubes:
Cumulative P1ugging Equivalent: 8.10% 4.92% 7.74% ['
Average Plugging Equivalent (All i
SGs) 6.92%
l
Unit 2 reached initial criticality in May 1981, and to date, has !
completed ten (10) refueling outages. Tube degradation modes of the
'
Unit 2 SGs were as follows: ,
e ODSCC at Hot Leg Support Plates: (Confirmed by tube pulls' 4/86 - I
'
R31C46 and R38C46 from SG C.)
e ODSCC above Hot Leg Tubesheet: (Have not been confirmed by tube t
pull.to date. Minor intergranular attack observed on the '
outside diameter surface above the tubesheet in 10/90 - l
R19C48 from SG C.) !
e _PWSCC at the Mechanical Roll Tubesheet: (Confirmed by tube pull' !
in 4/89 - R16C50 from'SG C. Shotpeening was performed in .
10/87 and F* criteria implemented.) !
e PWSCC in Low Row U-bends: (Row 1 tubes 3reventatively plugged in :
10/82. Row 1 tubes unplugged and leat treated in 10/90. No ;
indications to date.) l
Enclosure 2 i
!
, _
. . .
- . -
>
,
- .
.
'
- 15
i e AVB Wear: (Indications represent old wear spots corresponding to 1
'
locations of original AVB's, which were replaced with
, adjustable 405 stainless steel AVB's in the 4/86 & 11/87
.
outages.) i
j- e Free-Span !.ocations: (Responsible for plugging of six tubes. Not
- - characterized by a tube pull, but data suggests that the ;
mechanism is most likely ODSCC.)
,
Status of Unit 2 SGs after the tenth refueling outage was as follows:
1
] SLA S_G_H SLC )
<
Total Inservice Tubes: 3074 3189 3190 i
i
Total Inservice Sleeved Tubes: 77 56 140 l
l Total Inservice Sleeves: 112 58 170
- -Total, Plugged Tubes: 314 199 198 l
l
l Cumulative Plugging Equiv. for 4.04% 2.15% 5.99%
Sleeved Tubes: !
i
j Cumulative Plugging Equivalent: 9.39% 5.94% 6.02% j
Average Plugging Equivalent (All l
-
SGs) 7.12%
i
3 c. Conclusions !
'
The reviews of the current status, and the history of FNP SGs provided )
an indication that the licensee has been doing a conscientious job of
j
managing the SG degradation modes as they are identified.
1
i M3 Maintenance Procedures and Documentation :
i
-
a. Insoection Scooe (73753)
,
The inspector reviewed licensee and contractor procedures for the eddy !
current inspection of FNP SGs.
'
,
b. Observations and Findinas
'
Procedures reviewed included the following:
'
e Farley Nuclear Plant Surveillance . Test Procedure. FNP-2-STP-159.0.
" Steam Generator Inspection. Plugging and Repair." Revision 13. l
5 e Farley Nuclear Plant Systems Performance Procedure. FNP-2-SYP-3.0,
"
"SG Data Management." Revision 3.
'
i
!
i i
Enclosure 2 j
! !
2
, _ _ - , , . ,,
.
.
.
16
e Westinghouse Nuclear Services Division Procedure No. MRS 2 4.2
APC-37. " Steam Generator Eddy Current Data Analysis Techniques."
Revision 1.
e Westinghouse Nuclear Services Division Procedure No. MRS 2.4.2 ;
APC-38, " Steam Generator Data Management." Revision 1 !
I
e Rockridge Technologies Procedure 42-EC-266. " Data Analysis Proce- :
dure For Farley Nuclear Plant Units 1 & 2." Revision 0.
>
e Rockridge Technologies Procedure 42-EC-267. " Data Management
Procedure Farley Nuclear Plant Units 1 & 2." Revision 0.
'
'
The procedures reviewed were either new, or very recently revised.
procedures. These procedures and/or revisions were arepared as a result i
of the Unit 2 data management problem identified in _ER 50-364/96-002-00
and VIO 50-364/96-03-03. The completion of the procedures and revisions
were required to support SG inspections during U2RF11. ;
c. Conclusions I
The licensee and contractor procedures for the evaluation and management !
of SG eddy current data appear to be adequate for the circumstances. ;
M6 Maintenance Organization and Administration
a. Insoection Scooe (73753)
Through review of documentation and discussions with licensee personnel,
the inspector reviewed the management involvement with the SG program.
b. Observations and Findinas
SNC intracom)any correspondence, dated-May 8, 1996 from: D. N. Morey,
forwarded "rarley Steam Generator'- Roles and Responsibilities." This
memorandum provided documentation of the assigned responsibilities for
the various facets of SG management.
The inspector met-with SG management personnel from the licensee's-
corporate office for a- presentation of the licensee's long range plans
and goals for the Farley SGs. The presentation.provided insights into
.the licensee's plans to systematically recover previously plugged SG
tubes by installing sleeves at the locations of the' degradation. The
presentation also provided various computer models of the projected
recovery process, which showed how the recovery and sleeving of plugged
tubes could be managed to maintain the plugging levels at or below 7%
per unit.
Er. closure 2
1
., _ _ ,_ -
- - . . _ -
.
.
.
17
'
c. Conclusions
The review of the history of the Farley SGs. as discussed in section M2
above, and the discussions with chemistry personnel, as discussed in
Section IV Plant Support of this ins)ection report. provided indication
that the licensee has been managing t1e degradation of the SGs even
though the integrated program for management has not been documented.
M8 Miscellaneous Maintenance Issues (92902)
M8.1 (Closed) VIO 50-364/96-03-03. Steam Generator Tube Flaws Within F*
Distance
The licensee's response dated June 27. 1996, provided a commitment to
revise and formalize instructions used by SG inspection data analysts
and data managers to ensure compliance with TS requirements. The
response also stated that a broadness review was in progress to review
FNP's safety related vendor services used to support safety-related
activities.
The inspector reviewed the close-out Jackage for Corrective Action
Report (CAR) No. 2193. Revision 0. T1e package contained copies of the
signature pages for the SG surveillance and data management procedures
which will be used in the up-coming Unit 2 outage. (The inspector also
reviewed the individual procedures as described in Section M3 above.) l
The CAR package also contained a ten question checklist for use in
conducting the broadness review to ensure that other safety related l
vendor activities are properly controlled. The CAR close-out contained l
tables showing seventy-nine (79) safety-related vendor-provided. 1
services and procedures that were reviewed during the broadness review !
for this violation. j
M8.2 (Closed) LER 50-364/96-002. Misacolication of Technical Soecification
4.4.6 Reauirement Reaardina F' i
l
'
Root cause analysis and corrective actions for VIO 50-364/96-03-03. as
reported to NRC and described above, are sufficient to close this LER.
M8.3 (Closed) LER 50-364/95-001 and LER 50-364/95-001-01. Steam Generator
Tube Dearadation and Tube Status
The original LER (95-001) was provided to satisfy TS 4.4.6.5.c which l
requires that SG tube inspection results which fall in the Category of !
C-3 shall be considered to be a reportable event and reported pursuant l
l to 10 CFR 50.73 prior to resumption of plant operation. The LER also
served to satisfy TS 4.4.6.5.a which requires that following each
'
inservice inspection (ISI) of SG tubes. the number of tubes plugged,
repaired or designated F*/L*, in each SG shall be reported to the
Commission within fifteen days of the completion of the inspection.
Enclosure 2
l
!
1
__ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ . _ _ . . _ _ . . _ _ . . .
-
'
.
,
18
plugging, or repair effort. The revised LER (95-001-01) was issued to I
show the change in status of six tubes from F* to L* These LERs are ;
considered closed. ,
!
M8.4 (Closed) LER 50-348/95-009. Steam Generator Tube Dearadation and Tube !
Status i
This LER was provided to satisfy TS 4.4.6.5.c which requires that SG
tube inspection results which fall in the Category of C-3 shall be ,
considered to be a reportable event and reported pursuant to 10 CFR '
50.73 prior to resumption of plant operation. The LER also served to i
satisfy TS 4.4.6.5.a which requires that following each ISI of SG tubes. l
the number of tubes plugged or repaired, in each SG shall be reported to i
the Commission within fifteen days of the completion of the inspection. ;
plugging. or repair effort. This LER is considered closed. '
(
III. Enaineerina
El Conduct of Engineering l
t
El.1 pm ian Chance Control Processes
l
'
n. insoection Scoce (37550)
The inspectors reviewed the licensee's procedures which control
the design change program. ]
>
b. Observations and Findinas :
!
The inspectors reviewed the procedures listed below which control l
design changes and verified that the design control measures were i
consistent with 10 CFR 50. Appendix B. Criterion III and 10 CFR j
50.59. The following procedures were reviewed
FNP-0-AP-8. Design Modification Control. Revision 22. dated 3/1/96
FNP-0-AP-88. Nuclear Safety Evaluations. Revision 0. dated 2/1/96
'
From review of the above procedures the inspectors concluded that
'the following attributes were adequately addressed: design
processes, design inputs interface controls, design verification,
document control )ost-modification testing, control of field
changes, and 10 CFR 50.59 safety evaluations. The inspectors
concluded that adecuate controls were in piace to ensure effective
implementation of cesign changes.
.
1
Enclosure 2
- -. --
_ . _ - . . _ - _ _ _ _ _ _ _ _ . . _.__ _ _
_.__.____.____m __
.
'
.
.
'
19
c. Conclusions
l
The inspectors concluded that the licensee's design change control :
procedures complied with the requirements of 10 CFR 50, Appendix ;
B, Criterion III, and 10 CFR 50.59. ;
E1.2 Review and Walkdown of Desian Chanae and Modification Packaaes !
a. Insoection Scooe (37550)
The ins)ectors reviewed the design change and modification packages
listed ]elow to: 1) determine the adequacy of the safety evaluation
screening and the 10 CFR 50.59 safety evaluations: 2) verify that the ;
modifications were reviewed and approved in accordance with TS and
'
.'
applicable administrative controls: 3) verify that applicable design 2
bases were included: 4) verify that UFSAR requirtwents were met.
'
b. Observations and Findinas
The inspectors reviewed the following design change and
modification packages: ;
Desian Chanae Descriotion
S95-1-8857 Modification on Cable Tray Support No. 427
.
S95-1-8976 Repair of Main Steam Pipe Support MS-R205
l l
584-2-2915 Service Water 2" Diameter and Under Piping Replacement
S94-2-8752 Replacement of William Powell Service Water Valves
i
S95-2-8982 EDG 2B Exhaust Piping Hanger Support Modification
S96-1-9021 Repair of the Unit 1 Auxiliary i
Building / Containment Cork Interface in Room 186 l
Design Change numbers S95-1-8857..S95-1-8976 and S96-1-9021 were
i- completed. Design Change numbers S84-2-2915. S94-2-8752, and S95-2-8982
are scheduled to be implemented during U2RF11.
The inspectors found that the DCPs had been reviewed and approved
in accordance with the licensee's design control procedures and
that the format and content of the design changes were consistent
with design control procedures. The quality of the DCPs was good
,
)
Enclosure 2 I
!
_ l
- - . - . . - . - . . . - ~ . . . . - -------- - - - . - -
,
.
-
.
,
l 20
'
l
overall except for the minor discrepancies listed below in the
installation drawings for design change S95-2-8982. The discrepancies-
were as follows:
-
No general control dimensions were shown in the drawings. ;
-
-
No dimensions were given in the drawings for unequal sizes (or
dimensions) of structural members such as TS 8X4. L4X3. Plate 4X2,
, etc. At least one dimension is required to be shown for member
l orientation so installation matches the design analysis.
-
Sheet No. 228B of Support SCS-2H228 should be interchanged with
Sheet No. 229B of Support SCS-2H229.
'
l -
Section C-C on Sheet No. 228B of Support SCS-2H228 should be
rotated 90 degrees to be consistent with Support SCS-2H229.
i
The licensee's engineers took action to revise the drawings to correct
l the above discrepancies. The scope of each design change package was
l found to be adequate. The 10 CFR 50.59 Safety Evaluations were found to
be adequate. Except for the discrepancies noted in design change S95-2-
'
8982, the installation instructions were considered adequate to
implement the modifications. The inspectors also verified that the <
l
'
UFSAR and other documents e.g., drawings and procedures, had been
identified in the modification packages for revision.
'
- The inspectors walked down the completed modification for design changes
l S95-1-8857 and S95-1-8976. The inspectors examined the supports and
l compared the completed modifications with the design drawings. The
i
inspectors.found the.following discrepancies: ;
SUDoort No. Discreoancies
'
i
Cable Tray DCP Discrepancy:
Support 427
The fillet weld located at inside of the north-west
flange for the 5" wide flange post connected to 1/2"
top plate was measured to be 3/16". The drawing
required 1/4" fillet weld.
,
Pipe Support Pre-existing Discrepanci6s: '
l MS-R205
1) All four pre-existing fillet welds were found to
have 3/16" weld size. The drawings-showed the weld
size to be 1/4".
1
,
2) Two fillet welds shown on the drawings did not
'
exist at the top ends of Item 2 connected to Item 1
(both L3x3) (flush condition existed at the back of
Enclosure 2
i
!
_ _ _ ....___ _ ._-
- _ . - _ . - . _ . _ . . _ _ . . - _ . _
~
.
-
i
,
- -
j '
i 21 :
4
l
the two angles). The fillet welds could not be -
performed at these locations.
.
3) Item 4. 1/2"X7" Hilti Kwik anchor bolts, in the '
Bill of Materials on the drawings shows the number .
4
required to be four. The correct number is one. !
i DCP Discrepancy:
)
" i
'
One weld between the 4X4 tube steel and the base plate
was not recorded in Weld Data Form WD-0100 by the
welder and had not been inspected. This weld was .
'
required to be recorded and visually inspected per !
4
pre-determined weld data form by the modification ,
j planner-for ISI perpose, j
s
i Pre-existing discre)ancies refer to those that existed arior to
- implementation of t1e design changes (modifications). )CP discrepancies
i are those that occurred during implementation of the design change. The
i
discrepancies listed above are considered additional examples of VIO 50-
348, 364/96-10-01, applicable to Unit 1 only.
l In addition, the inspectors also found insufficient groove bevel welds
!-
between the bottom plate and the existing 5X5 tube steel for the
modification to cable tray support 427. The groove bevel welds should
' be made flush with the tube steel wall. The inspectors considered this
slightly insufficient weld practice to be a weakness in the weld quality
, and workmanship.
.The preliminary investigation for the undersized weld at cable tray
support 427 was that this weld was not selected for inspection within
the 10 percent weld surveillance sample inspected by the foreman for-
final acceptance of the weld products. The Weld Data Form WD-0100
contained in the work. order package and Section 3.15 of Procedure FNP-0-
I
'
SPP-WD-001 " Documentation of Welding and Related Activities", of FNP - i
Special Process Manual FNP-0-M-23. only requires inspection of 10
'
percent of the welds completed.in each modification, unless engineering
t specifies additional inspections. The weld at the pipe support MS-R205
I
was not recorded on the weld data sheet and therefore was not inspected.
, .
'
t c. Conclusions
The above discrepancies -combined with those identified in NRC IR 50- ,
-
325, 324/96-07, indicates that the licensee's welding inspection program !
may be deficient.
!
'
1 In general, the modification packages were judged to be of good quality.
y The modification packages contained sufficient specifications, drawings
, Enclosure 2-
4-
!
!
- _ ~
. . . . - - - - - _ - - - . ~ . . - - - - . - _ _ - - - - . . - . - . -
'
!
.
.
22 .
,
and 3rocedures for the installation. Two deficiencies were identified i
in t1e areas of the design drawing control and~ weld quality.
E1.3 Heavy Load Proaram
a. Insoection Scooe (37550)
The inspectors reviewed the licensee's procedures which control the
heavy load program.
b. Observations and Findinas
The inspectors discussed the Farley heavy load program with licensee 3
engineers and reviewed the current revisions of the procedures used for !
controlling heavy load lifts and movements. Procedures reviewed were: !
FNP-0-GMP-6.1 -6.4. -6.6. -57.0. -58.0: FNP-0-MP-11.0, FNP-1-MP-11.4: :
'
and FNP-2-MP-11.4. These procedures were used to control the operation.
maintenance, inspection, testing, qualification, load paths, and ;
operation of polar and mobile cranes, slings, load cells, lifting ,
devices, crane operators and riggers. Procedure Nos. FNP-1-MP-11.4 and
FNP-2-MP-11.4 for the reactor polar crane operations specify safe load
paths to limit the load traveling paths in order to prevent the loads ,
from accidently dropping into the reactor fuel and/or on safety-related ;
components. ;
,
The slings used for rigging were required to be visually inspected
annually and color marked after the inspections. Hoists were required
j
to be load tested annually and color marked also. The inspectors '
examined hand hoists stored in the cold tool room to verify the hoists
were marked with current color coded-red. No discrepancies were found.
c. Conclusions
The inspectors concluded that the procedures were adequate for control
and handling heavy loads. The licensee has established adequate
.
procedures for control of heavy loads for plant routine operations and
l
maintenance.
E1.4 Enaineerina Sucoort (ES) of Larae Scale. Comolex Evolutions (37551) ,
!
, The resident inspectors observed several major activities that involved I
l considerable engineering support, oversight, and interface by j
'
responsible engineers in the ES and Plant Modifications and Design (PMD) ,
departments. These actvities for the most !
first time at FNP and represented complex,part were
resource beingtasks.
intensive done for.the
[ ES and PMD involvement was very evident and effective throughout the
j implementation of these activities: ;
e Train A CRACS modification I
Enclosure 2 1
- _ _ .
.
_ . - - _. . - . - - - . . . _ - - . .- .- - -- - . -- --
.
-
. ;
.
23
e Augmented Kaowool walkdowns
e 1A and 2A CCW HX epoxy coating ,
E7 Quality Assurance in Engineering Activities !
'
E7.1 Quality Assurance Assessment and Oversiaht
a. Insoection Scooe (37550) .
The inspectors reviewed audits performed by the onsite Safety Audit and
Engineering Review group. i
b. Observations and Findinas ,
,
The inspectors reviewed the results of audits of engineering and
design activities. Audits reviewed were as follows:
-
Audit 94-PMD/09. Plant Changes and Modifications. An audit .
i finding was identified regarding failure to issue as-built i
!
notices for evaluation of equivalent parts for procurement. l
Several examples were identified. j
-
Audit 95-ISI/29-1 and 95-STP/95-1. Inservice Inspection and i
Surveillance Testing. An audit finding was identified
'
regarding an engineer who performed acceptance of a
containment leak test who was not properly certified to
accept and evaluate the results in accordance with licensee
procedures.
-
Audit 95-ISI/29-2. Inservice Inspection. Two audit findings
were identified regarding failure of a vendors performing
ISI activities to comply with site procedures. One finding
concerned an unqualified individual (a trainee) performing
eddy current testing. The other finding concerned improper
'
review of certificates of calibration for equipment used for
ISI work.
-
Audit 96-STPe/34-1. Engineering Support Surveillance Tests.
,
There were no findings in this audit. Two comments
!
concerning minor procedure discrepancies were included in
the audit report.
l -
Audit 96-CSP /04. Control of Special Processes. An audit
l finding was identified concerning nine examples of welding
l inspections which were not )erformed by independent
! inspectors. In all cases t1e foreman in charge of the work
performed the weld visual inspections. The licensee's weld
inspection program permits weld inspections to be performed
by foremen but not of work they were responsible for. The
i
Enclosure 2
l !
,
.
.
24
weld inspections are known as peer inspections and are not
performed by independent quality control inspectors.
c. Conclusions
The ins)ectors concluded that the audits were useful in providing
oversig1t to management regarding performance of engineering
activities. The audit finding regarding non-independent weld
inspections is indicative of failure of the mainteriance foremen in
understanding the licensee's welding inspection program regarding
independent inspections. This issue, although it has been
corrected, is another example of a deficient weld inspection
program (see Section E1.2).
E7.2 Safety System Self Assessment Proaram
a. Insoection Scooe (37550)
The inspectors reviewed the results from the safety system self
assessment (SSSA) program.
b. Observations and Findinas
The licensee performed self assessments on 13 safety related
systems. The assessments were performed by independent teams of
individuals with design engineering and operations experience who
performed an in depth review of system design and operation. The
teams were supplemented with contract personnel experienced in
performing these types of assessments. The assessments were
performed between 1990 and 1995 and included the following,
systems: control room ventilation service water, residual heat
removal, emergency diesel generators, reactor protection, safety
related electrical distribution, auxiliary feedwater, instrument
air post accident sampling. component cooling, containment '
isolation, containment spray, chemical and volume control, high
head safety injection and accumulators, and reactor cooling water.
Findings from the assessments were reported as strengths or
weakness. Weaknesses were identified by an action item which
required a response from the plant to resolve. More than 400 '
weaknesses were identified in the assessments of the 13 systems.
These weaknesses resulted in a number of LERs. changes to the
UFSAR. changes to design basis calculations, design changes, and
changes to operating procedures.
The inspectors reviewed the SSSA results for the auxiliary
feedwater, service water control room ventilation, and containment
isolation systems. The inspectors also reviewed corrective
actions performed in resport ! to weaknesses identified in the
SSSAs. The inspectors concluded that the corrective actions were
Enclosure 2
. _ _ _ _ _ _ _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _
l :
.
.
25
appropriate to resolve the identified weaknesses. As of the
inspection date only six action items remained open to resolve all
the weaknesses identified in the 13 SSSAs.
!
c. Conclusions
The safety system self assessments were an effective program to
identify and correct deficiencies in design and operation of the i
13 systems assessed. Discussions with the licensee disclosed that !
a new self assessment program to address balance of plant (non-
p safety related) systems is currently under review.
l
l IV. Plant Suooort i
R1 Radiological Protection and Chemistry Controls
,
R1.1 Tours of the Unit 1 and 2 Radioloaically Controlled Areas (71750) i
- ;
During the course of the inspection period the resident inspectors !
l conducted numerous tours-of the. radiologically controlled area (RCA) of '
! the auxiliary building for Units 1 and 2. In general. Health Physics
l (HP) control over the RCA and the work activities conducted within it. !
l were good. Material condition and housekeeping in the Unit 1 and 2 RCA
L were typically well maintained.
l !
R1.2 Review of Water Chemistry Controls and History
l
a. Insoection Scooe (79501)
By walk-down inspections, interviews and document review, the inspectors
l examined the licensee's primary and secondary water chemistry controls
'
and history to verify compliance with commitments and regulatory !
' requirements.
.
b. Observations and Findinas
The. licensee's on-line chemistry monitoring includes: cation conductiv-
.ity. monitors in 12 locations: specific conductivity monitors in six
_
'
locations: )H monitors in ten locations: sodium monitors in six loca- .
tions: two lydrazine monitors: and ten dissolved oxygen monitors. In _ l
addition the licensee has an in-line ion chromatograph with both units !
, samples being routed over to it. Future plans are to shut down these
sample' streams and rely upon the SG blowdown analysis, when the Unit 1
streams are connected to the Unit 2 monitor. There were no in-line real-
.
time chlorine monitors.
I
The licensee indicated that they have removed all the co)per alloy
materials from the secondary side, of both units, with tie exception of
l the aluminum bronze (CA614) condenser tube sheets. -As a result of
sludge lancing the licensee has removed as much as 1885.5 lbs. of copper
l
.
Enclosure 2
,
4-, , ,
_ __ _ _ . _ .. __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ . .
,
'
.
. ,
26
!
(Unit 1 refueling outage 2). more recently amounts under 300 lbs. were
removed. In-line instrumentation reported a total copper transport, for ,
a fuel cycle, at power levels above 30%, to be approximately 10 lbs, per
unit. The licensee has not identified any degradation to the aluminum
bronze tube sheets. The licensee has not chemically cleaned the SGs. ;
The licensee explains the difference between ten and something less than
300 lbs. to be a redistribution of the circulating load of copper
deposited in earlier cycles. .
The licensee has adequate procedures to identify locate, and repair
condenser tube leaks.
The licensee's program for SG layup is adequate. j
c. Conclusions
The licensee's water chemistry controls were appropriate for the circum-
stances.
Status of Radiological Protection and Chemistry Facilities and Equipment
'
R2
R2.1 Monitorina of Primary-to-Secondary Leakaae *
i
a. Insoection Scooe (79502)
By walk-down inspections, interviews and document review, the inspectors ;
evaluated the effectiveness of the licensee's procedures, equipment, and '
practices for monitoring primary-to-secondary leakage to verify compli-
ance with commitments and regulatory requirements.
b. Observations and Findinas
The licensee's actions related to NRC Bulletin 88-02. NRC Information
Notice Nos. 88-99. 91-43 and 93-56 were appropriate.
Primary-to-secondary leakage is monitored by real time N6 monitors
alarmed and annunciated in the control room at 5 gallons per day. In
addition, the licensee takes the following grab sam)les: gross
beta / gamma once per 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />; tritium once per monti: gaseous leak rate
once per month: gamma isotopic once per month: and steam jet air ejector
every 31 days, when no leakage is identified. The analysis periodicity
is progressively increased when primary to secondary leaks are
identi fied.
,
l The licensee does not perform condensate polishing. As a consequence, i
j the licensee was not concerned with regeneration or disposal of conden- l
! sate demineralizer resin and the associated effluent.
!
1
i
Enclosure 2
i
f-
'
1
_._ . . _ _ _ _ . . _ . . _ - . _.- _ . _ _. _ _ ._ . . _ _ _ _ _ . _ _ . - . . _ . -
'
i -
i
. :
- - , ,
.
- 27 j
l~ c. Conclusions
4
'
The licensee's current procedures. equipment, and practices for monitor- !
ing primary-to-secondary leakage are appropriate. j
.
i R3 Radiological Protection and Chemistry Procedures and Documentation
R3.1 Review of Radioloaical Protection and Chemistry Records and i
j Documentation l
a. Insoection Scooe (79501) l
.
The inspectors reviewed the TSs UFSAR and selected procedures to !
evaluate compliance of licensee's chemistry program with regulatory
, commitments procedural requirements and industry recommendations. 1
4
b. Observations and Findinas
, .
Except as noted below, the licensee's' procedures are consistent with the- .
r TSs. the UFSAR including in 3rocess changes, and Electric Power Research
1 Institute (EPRI) TR-102134. Revision 3. PWR Secondary Water Chemistry ;
i Guidelines, and EPRI NP-7077. Revision 2, PWR Primary Water Guidelines. :
i
Procedures for layup of SG are appropriate. ;
'
i
- -
'
e EPRI TR-102134 Revision 3 Table 2-1._ recommends, during cold
'
shutdown / wet layup (reactor coolant system (RCS) <200 F), once per
day dissolved 02 analysis. The licensee was unable to conduct
l- this analysis due to plant configuration. !
- . i
- e EPRI TR-102134. Revision 3. Table 2-2a recommends, during ;
. heatup/ hot shutdown (RCS>200 F.to 5% reactor power), feedwater be '
.
analyzed once daily for pH# c
<
ppb. The licensee's procecIur. Dissolved 0. ppb and Hydrazine,esprovid ,
,
did not provide any periodicity. ;
. e EPRI TR-102134. Revision 3. Table 2-2b recommends, during
heatup/ hot shutdown (RCS>200 F to 5% reactor power). SG blowdown. :
l
be analyzed once daily for Chloride., pab and Sulfate, ppb. Due to
work load considerations the licensee las chosen to conduct these !
analyses three times per week,
}
e. EPRI TR-102134. Revision.3, Table 3-5b. recommends, with the.
reactor critical -once per day analysis of_ Conductivity gS/cm @
, 25 C and pHa s.c. and once per week analysis.of Suspended Solids,
ppb and Silica, ppb. Due to work load and dose considerations the
licensee has chosen to conduct all these analyses once monthly-
- except for pH 2s c, which they conduct once per week.
e The licensee primary to secondary program was based on EPRI NP-
7077. Revision 2. The licensee was currently in the process of
Enclosure 2
!
$
"
, _- .. _ _ _ _ -- _ .
.
.
.
28
updating their program to Revision 3 dated November 1995. The
target completion date was December 31. 1996.
e A review of procedures 1-SOP-69.0 R2. 2-SOP-69.0 R2. CCP-201 R54
1-ARP-1.6 R28,62-ARP-1.6 R21. and CCP-31 R18 revealed an inconsis-
tency in the N monitors set points stated in CCP-31. The li-
censee made an immediate procedure change to correct the inconsis-
tency. The N6 monitors in the field had set points properly set.
e As described in NRC IR 50-348.364/96-02, the inspectors noted a i
number of wheeled instrument carts, chairs. and work platforms ,
unsecured and unattended in the control room back board area. To
address this issue the licensee issued a Night Order dated March
6, 1996, directing personnel to keep wheeled furniture out of the
rack area when not in use and to assure that the wheels are locked
when these items are in the rack area. During this inspection the
ins)ectors noted several wheeled carts and chairs with wheels not l
locted or blocked, unattended in the rack area. Apparently the l
Night Order was ineffectual. '
c. Conclusions
The licensee's water chemistry program is in compliance with regulatory
requirements and with industry guidelines, with some exceptions.
R7 Quality Assurance in Radiological Protection and Chemistry Activities
R7.1 Review of the Audit History for the Primary and Secondary Water
Chemistry Control Proarams i
a. Insoection Scooe (79502)
l
The inspectors reviewed the last two licensee Chemistry STP audits, the
last two biannual Chemistry audits as well as a spot audit in response
to EPRI PWR Primary to Secondary 1.eak Guidelines, to verify compliance
with commitments and regulatory requirements.
b. Observations and Findinos ;
Audit findings included weakness related to: alarm set points: control
of chemicals; improper response to air sample monitor alarm; improper
tag order implantation; shelf life:'and consistency with EPRI Guide-
lines. Appropriate corrective actions were taken or planned.
c. Conclusions
The area of primary and secondary water chemistry was subjected to
independent audits, with appropriate action taken for identified
weaknesses.
Enclosure 2
.. . _ . . __ _ __ _ _ _- .. _ _ _ - _ _ . _ _ _ ___ .__
.
'
.
29
j P4 Staff Knowledge and Performance in Emergency Preparedness
c P4.1 Emeraency Drill (71750)
i
Resident inspectors observed the conduct of two emergency plan (EP)
-
3ractice drills on September 11 and 18. and participated in the annual
i EP dress rehearsal conducted on October 9. 1996.
The inspectors observed licensee emergency response activities in the !
plant simulator (simulated control room), emergency offsite facility.
'
technical support center. Operations Support Center. The drills and )
, dress rehearsal was well coordinated by the EP staff. Emergency
4
"
response facilities ap] eared adequate, with no real significant ,
ecuipment problems. T1e emergency response crews were effective in !
acdressing onsite accident conditions and offsite consequences during !
l the drills and exercise. In preparation for the annual exercise on
i December 11, 1996, the EP staff went to considerable lengths to simulate 1
- NRC headquarters and site team play with additional plant em)1oyees. i
This effort added realism, and increased the communication clallenges '
, for the response crew, during the drill of September 18. and the dress :
rehearsal. The inspectors participated in the subsequent drill and ,
<
dress rehearsal critiques, and provided their observations. '
- S1 Conduct of Security and Safeguards Activities ,
S1.1 Routine Observations of Plant Security Measures (71750)
During routine inspection activities, resident inspectors verified that
. portions of site security program plans were being properly implemented.
1 This was evidenced by: 3 roper display of picture badges by plant
personnel: appropriate (ey carding of vital area doors: adequate
stationing / tours of security personnel: . proper searching of
packages / personnel at the Primary Access Point and SWIS: and adequacy'of
I compensatory measures (i.e., posting of guards) during disablement of
- vital area barriers. Security activities observed during the inspection
<
period were well performed and appeared adequate to ensure physical
protection of the plant. Guards were observed to be alert and attentive
while stationed at disabled doors and access covers to critical
-underground equipment (e.g., SWS valve boxes and FOSTs). Posted
positions were manned with frequent relief.
'
S1.2 Search of Trailers Enterina The Protected Area (71750)
On October 10. 1996, a resident inspector observed security guards
escort a Westinghouse sludge lance trailer into the protected area'that
was not searched. The trailer was posted as a RCA. The security guard
outside the protected area gate did search the truck, cab and driver
prior to entering the protected area. However, he'did not search inside
the trailer because he did not have his personnel dosimetry. Also, the
shift security supervisor instructed him that a search would not be
Enclosure 2
-
,. .- - - - . - . - - - . . - . - - - - - - .
-
,
.
'
t
30
necessary based on her communications with a senior HP technician. The
l inspector also noticed that certain cargo spaces under the trailer which
were not part of the posted RCA, were not searched. All the cargo door
pad locks had been unlocked, apparently to facilitate a search. Pending
the inspectors actions to ascertain additional information on the
circumstances surrounding this incident, interview responsible ,
individuals and meet with the Security Chief, this issue is identified
as Unresolved Item (URI) 50-348, 364/96-09-05. Failure to Search
<
Contractor Trailer Prior to Entry into the Protected Area.
! F2 Status of Fire Protection Facilities and Equipment
F2.1 Installation and Insoection of Kaowool One-Hour Fire Barriers
a. Insoection Scooe (71750)
A detailed discussion of the initial inspection findings associated with
Kaowool one-hour fire barriers was documented in Section F2.1 of IR 50-
348, 364/96-07. Due to resident inspector concerns regarding the
l adequacy of Kaowool installation practices and independent inspection
program. URI 50-348. 364/96-07-03. Inadequate Installation and
l Inspection of Kaowool Fire Barriers, was identified. Since this initial
i inspection, the resident inspectors have conducted additional walkdowns
l of Kaowool installations in the plant interviewed responsible
personnel, reviewed ap)licable regulatory and FNP program requirements,
conducted meetings wit 1 plant staff and management, and monitored
licensee implementation of compensatory and corrective actions.
b. Qbservations and Findinas
! On September 3.1996, in addition to the discrepancies described in IR
96-07, an inspector identified some cable trays with Kaowool
l
installation deficiencies (flammastic fire seal installation and
degradation). During the exit meeting on September.5. the inspectors
l expressed concerns about problems with Kaowool installations. The
licensee promptly establish fire watches to address the immediate
'
concern of adequate fire safety, but did not consider conducting their
own inspections to ascertain the scope of the problem.
l
l On September 25. and 30. the inspectors conducted further walkdowns of
l
Kaowool cable tray installations and identified additional installation
deficiencies similar to those of September 3. All inspector identified
deficiencies were identified to the Operations SS and shift clerk. The
l shift clerks promptly initiated the appropriate compensatory actions
(i.e. fire watches) pursuant to the Fire Protection Program (FPP)
described in UFSAR. Appendix 9B. On October 2. the inspectors
identified three raceways that were not completely wrapped with Kaowool
as depir.ted by " Safe Shutdown Raceway and Identification of Kaowool ,
,
Wrap" .dr awings 0-180536. Revision 9. and D-203281. Revision 14. These
raceways were BDE-15 and BHF21 in Room 160 (1C Charging Pump power
Enclosure 2
!
,'
.. - - . - ~ -- - . - - -- -- . . - - - - - .
.
'
.
t
.
,
31
cables and control wiring) and BDE-15 in room 2159 (2C Charging Pump
power cables), and appeared to be part of the licensee's FPP for l
ensuring safe shutdown capability. ,
The licensee performed a preliminary engineering evaluation of the
raceways with missing Kaowool and informed the inspectors that the
drawings were in error and the raceway sections were not required to be
wrapped. A further, more detailed review by the licensee determined
that raceway BDE-15 (1C Charging Pump power cables) in room 160 was
required to be wrapped with a one-hour fire barrier to meet 10 CFR 50.
Appendix R requirements of SIII.G.2 as described by the UFSAR. Appendix
9B.
1
On October 2. the inspectors met again with licensee management to
discuss the continuing concern regarding the adequacy of Kaowool fire
barrier installations throughout the plant. After the meeting. the
licensee immediately initiated one-hour roving fire watches in every
room of the plant that contained Kaowool. Later that same day,
management briefed the inspectors on their Kaowool Action Plan which
included: Revising FNP-0-FSP-43, " Visual Inspection of Kaowool Wraps,"
Revision 5: Training additional personnel to perform comprehensive
Kaowool inspections: Performing walkdowns of all Kaowool installations
recuired by the Fire Protection Program for safe shutdown: Evaluating
anc correcting any identified deficiencies;. and updating the FPP and
Appendix R Compliance Report as needed.
On October 3. an inspector interviewed one of the two electricians
responsible for performing the most recent Kaowool inspection per FNP-0- i
FSP-43. This inspection had been conducted March 4-7, 1996. The
'
responsible electrician-stated that the drawings and raceway metal
labels were difficult to read. He also stated that he had never
installed or inspected Kaowool before the March 1996 inspections, nor
had he received any training, was not knowledgeable of Kaowool
installation or design requirements. UFSAR, Appendix 9B..Section 9B.6.1
of the FPP, requires an independent inspection program to verify fire
protection systems conform with installation drawings and test
procedures. Paragraph 9B.6.1.E. requires that the inspection program
includes measures to assure that personnel performing these inspections
are knowledgeable in the design and installation requirements for fire
)rotection. The failure to ensure inspection personnel were
(nowledgeable in the assigned inspection area was identified as an
'
example of an apparent violation (EEI) 50-348, 364/96-09-07. Inadequate
Kaowool Inspection Program.
The inspector reviewed FNP-0-FSP-43 to determine if it provided adecuate
guidance to the personnel performing the inspections. UFSAR, Appencix
9B, paragraph 98.6.1.F.1 requires that inspection procedures,
instructions, and checklists provide identification of characteristics
- to be inspected. The inspectors found that the procedure only verified
that there were no breaks in the continuous run of Kaowool wrap. The
'
Enclosure 2
- _ _
_ _
.
.
32
inspectors determined that FNP-0-FSP-43 did not direct the inspection
personnel to examine the Kaowool for compression or the flammastic fire
seals for proper installation and deterioration. These are important
characteristics which ensure the Kaowool was installed, and maintained.
IAW original installation design requirements. Inspections conducted by
the NRC and the subsequent detailed walkdowns by the licensee identified
numerous deficiencies related to these characteristics. This failure of
the inspection procedure to identify all necessary design and
maintenance characteristics is another example of EEI 50-348. 364/96-09-
07. Inadequate Kaowool Inspection Program.
On October 4. the licensee trained approximately 24 Kaowool inspectors
and certified them as Level II inspectors to perform inspections of
Kaowool installation. The training focused on identification and
installation details. It consisted of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of classroom
instruction and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of on-the-job training in the plant. An
inspector observed the classroom portion of the training. At the
conclusion of the training the ins)ector observed that the trainees .
appeared to be confused. During oaservations by the inspector of some
of these personnel performing inspections, this confusion was apparent
during their inspections. The inspector also reviewed the proposed
inspection plan, ETP-3409. and determined it was adequate.
On October 5 the licensee commenced detailed Kaowool inspections of all
installations in the plant. All inspections were coordinated by the ES
Performance Review Group. A resident inspector observed three two-
person teams during the first four hours of inspection activities. The
first team observed by the resident inspector failed to identify an
apparent aroblem with the flammastic seal. This deficiency was on
raceway AED454 in Room 2316 which had no flammastic being visible eight
inches up the raceway. The resident inspector discussed this oversight
with the inspection team and the overall inspection coordinator. The
licensee coordinator agreed that this kind of deficiency should be
documented for further engineering evaluation. He then contacted each
team to ensure they understood this part of the installation criteria
and the necessity of documenting all deficiencies. The resident
inspector observed two more teams in the field and did not identify any
more performance problems. The resident inspector concluded that this
comprehensive inspection effort by the licensee should be adequate to
identify existing Kaowool installation deficiencies.
The licensee's inspection took about four days to complete and
identified the following:
l
1) The 1B and 1C Charging Pump (B train) power cables and room cooler
j cables in Room 175 of fire area 1-004 were not fully enclosed in a
i fire barrier having a 1-hour rating as required 10 CFR 50.
Ap)endix R and the FPP. In these four particular raceways (i.e.,
BDE-0A, BFDB06. BFDD15 and BHF-36) the Kaowool wrap was missing
,
Enclosure 2
i
l
1
.
33
for approximately 24 inches between the cable tray and the
opposing firewall.
2) Cables for Unit 1 main steam isolation and auxiliary feedwater
flow control in raceway BHFC33 that traversed across room 319 of
fire area 1-042 were not wrapped by Kaowool. The licensee had
failed to install the one-hour fire barrier required by the
Appendix R compliance report referenced by the FPP.
3) The licensee verified the resident inspectors finding that 18 feet
of raceway BDE-15 carrying power cables for the 1C Charging Pump
(B train) in Room 160 of fire area 1-004 was not enclosed in a
fire barrier having a 1-hour rating as required by 10 CFR 50.
Appendix R and the FPP.
4) Thirty-six deficiency reports (DR) were generated. The DRs were
broken down as follows:
Kaowool that has been compressed 21 DRs
Kaowool that is partially missing 5 DRs
Kaowool missing 5 DRs
Kaowool damaged or deteriorated 5 DRs
5) Over 100 instances were found where FNP-0-PMP-507 "Kaowool
Installation Procedure." Revision 5, was not strictly followed.
PMP-507. step 6.2.8. stated that "In areas such as floors, walls,
and ceilings where the blanket wrap ends, fire protection seals
shall be installed as specified using mastic coatings such as... ,
The sealant shall be troweled completely around the wra) ped cable
tray at the floor, walls, or ceiling. The seals shall 3e not less
than 1/4" thick and extend not less than 8" onto the Kaowool and
not less than 8" onto the floor, wall or ceiling." The licensee !
inspection identified numerous junctions where the flammastic was i
not troweled for 8" completely around the Kaowool wrapped cable i
tray and/or did not extend 8" onto the floor, wall or ceiling. In I
many instances, the flammastic was also cracked. damaged, or )
indeterminate due to the Zetex outer wrap. i
FOL NPF-2. Condition 2.C(4), and FOL NPF-8. Condition 2.C(6), prescribes '
i
that Southern Nuclear shall implement and maintain in effect all
provisions of the approved FPP as described in the UFSAR. Appendix 9B. )
Items 1) through 3) above represent examples where the licensee failed j
to install one-hour fire barriers as specified by the 10 CFR 50. l
Appendix R program described in the FPP. These noncompliant conditions
are identified as EEI 50-348/96-09-06. Failure to Install Required One-
hour Fire Barriers.
Items 4) and 5) above. are additional performance-based examples of the l
inadequacy of the licensee's independent inspection program to verify
Kaowool fire-barriers properly installed and maintained in the plant. '
Enclosure 2
l
- - - - .- -- - . - - - . . . - - - - - . . - .
.
.
-
i
.
34
Section 9B.6.1 of the FPP. requires an independent inspection program to
verify fire protection systems conform with installation drawings and
test procedures. Fire Surveillance Procedure. FNP-0-FSP-43. " Visual '
Inspection of Kaowool Wraps." Revision 5. provided the acceptance
criteria, instructions, and references for installation details for
conducting inspections of Kaowool wraps used to provide the 1-hour rated ,
fire barriers prescribed by the FPP. The number of longstanding 1
deficiencies identified in the ETP-3409 inspection plan demonstrated
that the Kaowool inspection program was ineffective. This is another
example of EEI 50-348. 364/96-09-07. Inadequate Kaowool Inspection
Program.
A resident inspector also )erformed reviews of the Kaowool Qualification
testing documentation whic1 the licensee provided in res)onse to
questions regarding what configuration was required for (aowool
terminations at walls and ceilings and the effects of uninsulated
conduits and support penetrating the Kaowool wraps. The ins)ector was
not able to conclude that plant configurations were within t1e scope of
the tested ccafiguration based on the documentation provided by the
licensee. The qualification test configurations consisted of a Kaowool i
wrapped raceway and conduit extending through a catenary furnace. In ;
all cases the Kaowool wraps extended twelve inches )ast the ends of the
furnace. The furnace openings were then sealed witi Kaowool and
firebrick dust. These tests do not appear to reflect the configuration l
' of a Kaowool wrap butting against a fire rated barrier, nor penetrating I
conduits and supports, which are used throughout the plant. The tests
indicate that any gaps in the Kaowool wrap will significantly degrade
its effectiveness. This issue is identified as URI 50-348, 364/96-09-
08. Adequacy of Kaowool Qualification Tests to Scope Installed
Configurations. Resolution of this item will require additional review
of the Kaowool Qualification test report and inspection of plant
specific configurations by the NRC.
c. Conclusions
1
'
The inspectors identified two apparent violations regarding 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated
fire barriers. One apparent violation was for failure to install
required one-hour fire barriers. The second apparent violation was for
a failure to have an adequate inspection program for Kaowool fire
barriers. The inspectors also opened URI 50-348, 364/96-09-08. Adequacy
of Kaowool Qualification Tests to Scope Installed Configurations,
pending further NRC review.
!
Enclosure 2
1
. _ _ _.._ _ _ _ _ _ __ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..~ . _ _
.
.
. !
35 i
F5 Fire Protection Staff Training and Qualification ,
F5.1 Fire Briaade Drill
a. Insoection Scoce (71750) i
On September 30. 1996, a resident inspector observed the actions of the ,
dayshift fire brigade during a scheduled drill. I
b. Observations and Findinas
Fire protection training personnel notified the control room of a
simulated fire at the onsite Water Treatment Facility. The dayshift .
fire brigade mustered at the emergency response van and drove to the )
Water Treatment Facility towing the fire hose trailer. Fire brigade ;
members donned appropriate protective clothing, including self-contained '
breathing apparatus, and laid out the necessary fire hoses from the i
closest hydrant station to the simulated fire. Brigade members, under !
direction of the brigade leader then demonstrated appropriate fire i
fighting techniques. Afterwards, a criti ue of the drill was conducted '
by the drill director with the entire bri ade.
c. Conclusions
,
Fire brigade response was timely. Fire fighting equipment was in good
!
condition. Brigade member actions demonstrated proficiency in fire
fighting capability. The drill scenario was adequate.
F8 Miscellaneous Fire Protection Issues (71750)
F8.1 (Closed) URI 50-348. 364/96-07-03. Inadeauate Installation and
lnsoection of Kaowool Fire Barriers
This URI is closed based on the written discussion and the EEIs
identified in Section F2.1.
V. Manaaement Meetinas and Other Areas
X1 Review of UFSAR Commitments
'
A recent discovery of a licensee o)erating their facility in a manner
contrary to the UFSAR description lighlighted the need for a special
focused review that compares plant practices, procedures and/or
parameters to the UFSAR descriptions. While performing the inspections
discussed in this report, the inspector reviewed the applicable portions
l
'of the UFSAR that related to the areas inspected. The inspectors
verified that the UFSAR wording was consistent with observed plant ,
practices, procedures and/or parameters. Certain exceptions were !
identified as follows. 1
!
'
Enclosure 2
.
m -
r- --
.
.
.
36
e UFSAR Appendix 98. Attachment B. 10 CFR 50 Appendix R Exemptions.
Section 29.2 states " intervening combustibles between redundant
service water valves for CCW are minimal, consisting of primarily -
of cable insulation." Licensee failed to control the. storage and
placement of intervening combustibles between redundant SW inlet !
valves to the CCW HXs during epoxy work on the 1A 2A.1B and 2B !
CCW HXs. Positive controls were put in place during work on the !
1C and 2C CCW HXs. Refer to section M1.2.
e UFSAR Appendix 9B. Attachment B. 10 CFR 50 Appendix R Exemptions,
described Kaowool wraps which should have been installed but were
actually missing. Refer to section F2.1.
X2 Exit Meeting Summary
The resident inspectors presented the inspection results to members of
licensee management on October 17, 1996, after the end of the inspection
period. The licensee acknowledged the findings presented.
The resident inspectors asked the licensee whether any materials
examined during the inspection should be considered proprietary. No
proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED
Licensee
W. Bayne, Chemistry / Environmental Superintendent
R. Coleman. Maintenance Manager
P. Crone. Training Supervisor !
S. Fulmer. Technical Manager
H. Garland. Assistant Maintenance Manager
D. Grissette. Operations Manager
R. Hill General Manager - Farley Nuclear Plant
C. Hillman Security Chief
R. Martin. Su)erintendent Operations Support ,
M. Mitchell, iealth Physics Superintendent !
C. Nesbit. Assistant General Manager - Support
J. Odom. Superintendent Unit 1 Operations ;
- J. Powell. Superintendent Unit 2 Operations '
R. Rogers. . Supervisor. Engineering Support
L. Stinson. Assistant General Manager - Plant Operations
J. Thomas. Engineering Support Manager
Enclosure 2
. . _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ - . _ _ _ _ _ _ . _ _ .
.
-
. i
l
.
t
37
,
B. Yance. Plant Modifications and Maintenance Support Manager
W. Warren. Engineering Support Supervisor - Performance Review
G. Waymire. Safety Audit and Engineering Review Site Supervisor
L. Williams. Training and Emergency Planning Manager >
l
NRC
!
J. Zimmerman. NRR Project Manager - Farley Nuclear Plant
INSPECTION PROCEDURES USED
IP 37550: Engineering !
IP 37551: Onsite Engineering
IP 40500: Effectiveness of Licensee Controls in Identifying. Resolving, and
Preventing Problems ,
IP 61726: Surveillance Observations .
IP 62707: Maintenance Observations :
IP-71001: Licensed Operator Requalification Program Evaluation *
IP 71707: Plant Operations >
IP 71750: Plant Support Activities
IP 79501: PWR Water Chemistry Control and Chemical Analyis - Audits i
IP 79502: Plant Systems Affecting Plant Water Chemistry
'
IP 73753: Inservice Inspection .
'
IP 92901: Followup - Operations '
IP 92902: Followup - Maintenance ,
IP 92904: Followup - Plant Support
ITEMS OPENED CLOSED, AND DISCUSSED
!
Ooened l
1
Iygg Item Number Status Descriotion and Reference
-VIO 50-348, 364/96-09-01 Open Multiple Valve Misalignments By '
System Operators (Section 01.5)
NCV 50-348.-364/96-09-02
-
Open Untimely Incorporation of TS
Amendment Into Plant Procedures
! (Section 03.1)
IFI 50-348. 364/96-09-03 Open Inconsistent Application of Remedial
l Training Documentation Guidance
'
(Section 05.1)
i IFI 50-348. 364/96-09-04 Open CCW HX Epoxy Coating And Broken
- Tubes (Section M1.2)
1
- Enclosure 2
L
!-
.
'
.
.
38
URI 50-348. 364/96-09-05 Open Failure to Search Contractor Trailer
Prior to Entry into the Protected
Area (Section S1.2)
EEI 50-348/96-09-06 Open Failure to Install Required 1-hour
Fire Barriers (Section F2.1)
EEI 50-348, 364/96-09-07 Open Inadequate Kaowool Inspection
Program (Section F2.1) '
URI 50-348. 364/96-09-08 Open Adequacy of Kaowool Qualification i
Tests to Scope Installed ,
Configurations (Section F2.1)
Closed
.
Iygg Item Number Status Descriotion and Reference
,
IFI 50-348, 364/95-12-01 Closed Lack of Measurable Performance
Indicators Due to Use of WOG ERG
(Section 08.1)
i
URI 50-348, 364/96-07-03 Closed Inadequate Installation and
Inspection of Kaowool Fire Barriers
(Section F8.1)
NCV 50-348, 364/96-09-02 Closed Untimely Incorporation of TS
Amendment Into Plant Procedures
(Section 03.1) l
l
VIO 50-364/96-03-03 Closed Steam Generator Tube Flaws Within F.
Distance (Section M8.1). l
l
LER 50-364/95-001 & Closed Steam Generator Tube Degradation and
LER 50-364/95-001-1 Closed Tube Status (Section M8.3).
LER 50-348/95-009 Closed Steam Generator Tube Degradation and
Tube Status (Section M8.4). q
LER 50-364/96-002 Closed Misapplication of Technical
Specification 4.4.6, Requirement
Regarding F*- (Section M8.2).
I
' Enclosure 2
.
- _ . _ _ . -_ .... - .... . _ _ . _ . _ . . _ . _ . . _ - _ _ . _ - _ _ . _ _ _ _ _ _ _ _ _. __ _ __ _ _
,
p .
.
v
39 ;
{ .
j LIST OF ACRONYMS USED l
t
- ARP Annunciator Response Procedure
a
AVB Anti-Vibration Bar
Corrective Action Report
'
CAR ;
1 CCP Chemistry-Radiochemistry Procedure
- . CCW Component Cooling Water '
1
CFR Code of Federal Regulations
I
CRACS Control Room Air Conditioning System
i
DBA Design Basis Accident *
- DCP Design Change Package
- DR Deficiency Report ,
! EDG Emergency Diesel Generator
EEI Escalated Enforcement Item
EPRI Electric Power Research Institute
r- EPB Emergency Power Board
ES . Engineering Support group ~;
ETP Engineering Test Procedure !
'
F* A TS-defined SG tube acceptance criteria
FHP Fuel Handling Procedure
FNP Farley Nuclear Plant
FNPIR Farley Nuclear Plant Incident Report :
FOL Facility Operating License
FOST Fuel Oil Storage Tank
HP- Health Physics
HX Heat Exchanger
IAW In Accordance With
IFI Inspector Followup Item
lbs. Pounds
IP Inspection Procedure
IR Inspection Report
ISI Inservice Inspection
L* A TS-defined SG tube acceptance criteria
- LCO Limiting Condition for Operation
LER Licensee Event Report
- MCB Main Control Board
MCR Main Control Room-
-MTG Main Turbine Generator
NCV Noncited Violation
NDR Nonconformance Disposition Report
NIS Nuclear Instrumentation System
- N" Nitrogen 16 isotope
NRC U.S. Nuclear Regulatory Commission
NRR- Office of Nuclear Reactor Regulation
0 Gaseous Oxygen _- i
ObSCC Outer Diameter Stress Corrosion Cracking l
00S Out of Service !
Enclosure 2 l
l
,
.__ _. ~. . _ . . . _ _ . - . . _ _ . . _ . _ . - . . - - - - _ . _ - _ .
,
.
.
>
.
40
i
PWSCC Primary Water Stress Corrosion Cracking
PDR Public Document Room
pH The negative logarithm of the hydrogen concentration.
- pH .c )H determined at 25 C
l ns 3 arts Per Billion
M Project Manager ,
PMD Plant Modifications and Design group
PMP Plant Maintenance Procedure
l PWR Pressurized Water Reactor
i
'
RCA Radiologically Controlled Area
REA Request For Engineering Assistance
l SCS Southern Company Services
l SFP Spent Fuel Pool
'
SGFP Steam Generator Feed Pump
SNC Southern Nuclear Operating Company
SO System Operator
<
S0P _ System Operating Procedure
l SR0 Senior Reactor Operator
l SRST S)ent Resin Storage Tank
i
SS S11ft Supervisor l
l SSSA Safety System Self Assessment
.
STP Surveillance Test Procedure '
l
SWS Service Water System ,
'
-SWIS Service Water Intake Structure
,
'
Tm Average Reactor Coolant System Temperature ,
'
T Reference Program Temperature
TEdFW Turbine Driven Auxiliary Feedwater
TO Tag Order
TS Technical Specifications l
U2RF11 Unit 2 eleventh refvaling outage
UFSAR Updated Final Safei.y Analysis Report -
<
URI Unresolved Item l
' UTEC Ultrasonic Testing )
VCT Volume Control Tank l
VIO Violation i
WO Work Order l
WEXTEX Westinghouse Explosive Tube Expansion i
U
L
L
.
4
J
! Enclosure 2
!
!
!
l