IR 05000395/1997012
| ML20199H234 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/17/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20199H222 | List: |
| References | |
| 50-395-97-12, NUDOCS 9711260054 | |
| Download: ML20199H234 (23) | |
Text
_ . - _ - . __ _ _ _ _ - _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ .
S
.
U.'S. NUCLEAR REGULATORV COMMISSION REGION 11 Docket No.: 50-395 License No.: NPF 12 Report No.: 50-395/97 12 Licensee: South Caro 11na' Electric & Gas (SCE&G)
- Ficility: V. C. Summer Nuclear Station
.- Location: P. O. Box 88 Jenkinsv111e. SC 29065 Dates: September 7-- October 18, 1997 Inspectors: B. Bonser.. Senior Resident Inspector T. Farnholtz. Resident Inspector P. Fillion, Reactor Inspector. RII (Section E1.2)
Approved by: R. C. Haag. Chief. Reactor Projects Branch 5-DivisionofReactorProjects ig[ 3
.
>
. '
L L-
.
EXECUTIVE SUMMARY V. C. Summer Nuclear Station NRC Inspection Report No. 50-395/97-12 This integrated inspection included aspects of licensee operations maintenance engineering, and plant support. Thereportcoversa8 week period of resident inspection: in addition, it includes the results of an announced inspection by a regional inspecto Qgerations
- During plant manipulations involving a moisture separator reheater test, the operators maintained positive control of the plant (Section 01,2).
- Power reduction, shutdown, and shutdown plant operations for the refueling outage were conducted safely with good control and conmunications. Excluding the unanticipated lowering of Tavg below the minimum temperature for criticality, operators demonstrated good control of changing plant conditions. Reactor defueling operations were condected in accordance with the approved procedure (Section 01.3).
- The licensee closely monitored the key safety functional areas and properly assessed potential risks to plant safety during the refueling outage. Operators were aware of the status of the safety functional areas (Section 01.4).
- Reviews of system lineups and plant configuration during plant shutdown identified no concerns. The licensee implemented satisfactory controls to ensure plant safety and reduce potential risk (Section 02.1).
- A Nuclear Safety Review Committee meeting met Technical S)ecification requirements and provided constructive review and feedbacq to plant management (Section 07.1).
- The Independent Safety Engineering Group (ISEG) review of the outage schedule and outage activities contributed to plant safety during the refueling outage (Section 07.2).
Maintenance
- All observed maintenance tasks were conducted in a corapetent and professional manner. Proper radiological controls were used when required (Section M1.1)
- A non-cited violation was identified for a failure to correctly implement the requirements of a maintenance procedure during diesel generator maintenance. An incorrect tool was used to clean carbon de)osits from a piston. The inspectors concluded that the licensee had tacen adequate corrective action (Section M1.2).
'
.
.-
- i e The work associated with the A diesel generator jacket water and ,
intercooler heat exchangers was effective in determining the overall t condition of the tubes and gave reasonable assurance that they were capable of performing their safety related functions (Section M1.3). l e The licensee's inspection of the~ emergency feedwater flow control valves -
revealed some pitting and foreign material. .The valves were machined and reassembled. (Section M1.4).
- The observed surveillance tests were conducted in accordance with !
approved procedures with no discrepancies or concerns identified <
(Section M2.1),
. A review of Reactor Coolant System boric acid leakage inspection results '
.
found that, in general, there was-no significant leakage. The licensee was meeting the requirements of the American Society of Mechanical Engineers code relief requests (Section M2.2).
Enaineerina a e A ' review of a safety evaluation for a thermal study of the service water ;
)ond concluded that installing a plug in the interconnecting pipe Jetween the service water pond and Lake Monticello, and installing 3
'
temperature monitoring devices in the pond did not constitute an unreviewed safety question (Section E1.1).
.- A modification prepared by Design Engineering. Implemented in the fall 1997-outage, was reviewed and found to meet the requirements for design
. control. The modification resolved the Thermo Lag issue, an industry -
o wide. issue related to fire protection requirements (Section E1.2). i e A review of the Final Safety Analysis Report fuel handling accident i analysis assumptions that applied during defueling of the reactor concluded that current operating practices and plant characteristics were consistent with the accident analysis assumptions (Section E3.1).
Plant Sucoort
'
e Radiological controls were adequately implemented and maintained during the increased work activity of the refueling outage. The significantly higher dose rates in the radiation controlled area resulted from a ;
chemically induced reactor coolant system (RCS) crud burst and the licensee's inability to reduce dose rates-to pre outage levels (Section ;
R1.1) . .
! e ' A pre-job briefing conducted prior to a r actor building entry to-l decontaminate the building was professional and complet Street .
L clothes accessibility to the reacte building was maintained at the l beginning of refueling outage (Section R1.2).
- .e Observation of an emergency drill conducted during the evening hours L concluded that the site Emergency Response Facilities could be activated
.
l = . _ _ _ = ._ _ .
i
! _ __ _ . _ . __ _, , . . ._ _ _... _ _ . _ _ _ . _ _ _ m.. _ _
e e
and staffed within one hour. Personnel responding to the site were being appropriately monitored for fitness for duty considerations (Section Pl.1).
. Observed security a:tivities including compensatory measures were found to be acceptable during the increased refueling outage activity (Section S1.1).
. Observation of fire protection inspection activities during the refueling outage concluded that adequate measures were implemented to control fire doors and transient combustibles (Section F1.1).
l l
. _ _ - - _
,
.
, :
Reoort Details j Summary of Plant Status ,
The Unit began this inspection period at full powe Power was reduced to .
approximately 95 percent on September 21'to perform secondary plant steam flow !
testing. The plant was returned to full power the following da From ;
September 26 to October 3. power coasted down from 100 percent to
'
approximately 92 percent due to core End 0f-Life (E0L). On October 3. power '
was reduced to 90 percent for moisture separator reheater relief valve testing. -On October 4. the plant was shutdown and the refueling outage-commence ,
I. Doerations ,
01 Conduct of Operations 01.1 General Comments (71707) !
Using inspection Procedure 71707, the inspectors conducted frequent
<
reviews of ongoing plant operations In general, the conduct of operations was professional and safety-conscious.
,
01.2 Hoisture Seoarator Reheater (MSR) Steam Flow Testina > Insoection Scoce (71707) -
'
The inspectors observed the control room operations during the conduct of Preventive Test Procedure (PTP)-230.001, MSR Steam Flow Setup and Verification," Revision .
3 Observations and Findinas
,
, On September 21. the licensee performed PTP-230.001 to determine the optimum steam flow through the MSRs and subsequently the sizing of the steam supply nozzles. To perform this test, reactor power was reduced to approximately 95 percent and steam flow through the MSRs was varied while data was taken. The inspectors attended the pre-job briefin The test and the expected plant response were discussed and all questions were addressed. The control room operators maintained .
'
positive control of the plant during the entire test duration. -Key plant made..parameters The requiredwere test data monitored closely and was obtained whilethe steam
)lantflow waschanges returnedwere to 100 percent power. -The-inspectors concluded that tie actions of the-control- room o)erators during the performance of the test were in accordance wit 1 approved procedures,
, Conclusions During plant manipulations involving a moisture separator reheater test ,
the operators maintained positive control of the plant (Section 01.2).
,
F t
. ~
, J, 4 -A . . , '- ~ . ' , -
-
. i-_ vJ 7 4, _, m. .. 4 L -- W . m . ..,_.,%,,m,, -,-....-.,,,,,,,m . - . . _ - . , , - - . - . - , . mr ,,,,~~1,c, <. . , -u
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __- _ __ ___ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
01.3 Power Reduction and Plant Shutdown in Preparation for Refuelina Outaae 10 (RF-10) Insoection Scope (71707)
The inspectors observed portions of the power reduction plant shutdown, and shutdown plant operations conducted to place the plant in the conditions required for RF-10 activities. These operations were controlled by the use of General Operating Procedures (GOPs). Observations and Findinas From September 26 to October 3. reactor )ower coasted down from 100 percent to approximately 92 percent as t7e core approached E0L. To maintain Reactor Coolant System (RCS) Average Temperature (Tavg) in the required band, the control room operators periodically adjusted steam flow to the main turbine. The inspectors observed good command and control of plant conditions during this time perio On October-3. power reduction was commenced in preparation for RF-1 The inspectors observed portions of the power reduction which included opening the main generator output breaker, main turbine overspeed testing, and Emergency Feedwater (EFW) system testin These evolutions were conducted in a professional manner using GOP-4.g." Power Operation (Mode 1)." Revision 11 and other applicable test 3rocedures. The inspectors observed good communication practices aetween the control room operators and between the control room and operators in the plan On October 4. the plant entered Mode 2 and power was decreased to between one and three )ercent. The inspectors observed the operators transition to G0P-5. "teactor Shutdown from Startup to Hot Standby (Mode 2 to Mode 3)." Revision 9. All initial conditions were met including RCS Tavg stable between 555'F and 559'F. Step 3.3 of this procedure states:
Using Manual Rod Control decrease Reactor Power to 10% and verify stable temperature control of the Reactor Coolant Syste During the performance of the step, the control room operator observed Tavg decreasing below the minimum temperature for criticality (551*F) as !
stated in Technical Specification (TS) 3.1.1.4. This was due to the number of steam drains in the secondary plant left in service during the power reduction. The action statement associated with TS 3.1. requires that Tavg be restored to within its limit within 15 minutes or be in hot standby within the next 15 minutes. The inspectors noted that the operators immediately took steps to restore Tavg to greater than or equal to 551*F. Reactor power was ircreased slightly to approximately one )ercent and steam drains were closed to reduce the heat transfer out of tie RCS. The inspectors verified that Tavg was below 551 F for only approximately eight minutes. The minimum Tavg during this time was approximately 548'F. A Condition Evaluation Report (CER) was written to
_ _ . .
.
3 document this event (CER 97-0931). The inspectors considered the actions of the operators to be appropriate during the recovery of RCS Tav The inspectors also observed portions of the plant cooldown, solid plant operations, and RCS drain down performed in accordance with the following procedures:
- GOP 6. " Plant Shutdown from Hot Standby to Hot Shutdown (Mode 3 to Mode 4)," Revision 7
- GOP-7. " Plant Shutdown and Cooldown from Hot Shutdown to Cold Shutdown (Mode 4 to Mode 5)," Revision 6
- GOP 10. " Core Refueling (Mode 5 to Mode 6. Defuel, and Refuel to Mode 6)." Revision 10 The inspectors verified that adequate RCS level indications were available durin The inspectors'g RCS drain walkdown of thedown to 9" below temporaril the installed reactor tubing vessel used to flang indicate RCS level found it to be correct 1 installe The inspectors also observed reactor defueling operations from the control room. the reactor building, and the fuel handling building. A full core off-load was performed in accordance with Reactor Engineering Procedure (REP)-107.002. " Core Off-load." Revision 7. The defueling operation was controlled by reactor engineering personnel in the control room who were in communication with aersonnel in the reactor building and in the fuel handling building. Each fuel assembly was effectively tracked from its location in the reactor vessel to its location in the spent fuel pool. Water clarity in the spent fuel pool and in the refueling cavity was excellent. Personnel involved in the defueling o)eration were knowledge 21e and professional in the performance of t1eir assigned tasks. The inspectors did not identify any deficiencies or concerns in this are c. Conclusions Power reduction, shutdown, and shutdown plant operations for the refueling outage were conducted safely with good control and communications. Excluding the unanticipated lowering of Tavg below the minimum temperature for criticality, operators demonstrated good control of changing plant conditions. Reactor defueling operations were conducted in accordance with the approved procedur .
.
.
01.4 Shutdown Safety Insnection Scone (71707)
The inspectors observed the licensee's monitoring of the key safety functional areas during plant shutdown to defueling. The areas included: reactivity control, core cooling, power availability, containment integrity, and RCS inventory contro Observations and Findinas The licensee monitored key safety functional areas to optimize plant defense in-depth during the refueling outage. This involved protecting specific trains of equipnient and power sources, and monitoring valve line ups during plant evolutions and different plant configuration The inspectors observed that during the twice daily plan of the day meetings, the Operations and Independent Safety Engineering Group (ISEG)
representatives discussed the safety functions, upcoming evolutions that could affect the safety functions, and areas of potential ris The safety functions were also discussed during Operations shift turnover meetings and were listed on a status board in the control room. The inspectors also observed that operators were aware of the status of the safety function Conclusions The licensee closely monitored the key safety functional areas and
<
properly assessed potential risks to plant safety during the refueling outage. Operators were aware of the status of the safety functional area Operational Status of Facilities and Eauipment 0 Enaineered Safetv Feature System Walkdown (71707)
Following the plant shutdown for the refueling outage the inspectors performed control room and plant tours to monitor adherence to shutdown and refueling TS requirements and to monitor the key safety functional areas. The inspectors reviewed system lineups and plant configuratio The inspectors identified no concerns during these tours and concluded that the licensee implemented satisfactory controls to ensure plant safety and reduce potential ris .
07 Quality Assurance in Operations 07.1 Nuclear Safety Review Committee Meetina Insnection Scoce (71707)
The inspectors attended a portion of a Nuclear Safety Review Committee (NSRC) meeting on September 17 to observe and assess NSRC Tunctions.
@i w! <
1_________.__._____._______._____________.________________________________________________________________
_
- .- - - . . - _ . - - . - . ---- - - .- - - .-.
t
,
t
..
i
. /
5 !
t bservations and Findinas j An NSRC meeting was held on September 17 at the licensee's Nuclear Training Center. The inspectors attended a portion of-the day long meeting to assess the activities of the committee. The inspectors-observed the committee review recent significant operational _ plant events, the results of management watchstanding activities in the ,
control room, the deficiency reporting system used on site, and the -t
!
engineering plant system performance status report preparation proces The inspectors observed that the required number of NSRC members were "
i present and that the committee was providing an independent review of the activities designated in the TS. The inspectors concluded that the t NSRC meeting was providing constructive review and feedback to plant:
managemen Conclusions ,
An NSRC meeting met TS requirements and provided constructive review and ,
feedback to plant managemen i 07.2 Indeoendent safety Enaineerina Grouo Activities Insoection Scoce (71707)
The inspectors reviewed the ISEG Pre Outage Schedule Safety Review and ISEG activities during the outag b. ' Observations and Findinas The ISEG Pre Outage Schedule Safety Review focussed primarily '
a review of outage activities from a nuclear safety perspective c on identifying potential adverse safety impacts. The ISEG review - at '
beyond verifying compliance with the TS for Modes 5 and 6. and
_ identified changes in the outage schedule that could enhance the key safety functional areas and reduce potential risk. During the outag ISEG provided daily updates and reviews of the key safety functional areas and discussed relevant recent industry events during the twice daily plan of-the-day meetings. The inspectors also observed frequent ISEG presence in the plant during the outage, i Conclusions ,
The ISEG review of the outage schedule and outage activities contributed
- -
to-plant safety during;the refueling outag .
.
I i
>
,+,.~.s-m..--,-..-,.c.---.v-.-,em.~...~,, - .m.,-. , - - , - , , , , + w--,-rm.+.,c. ,v.- . . - - , ,-v ,, + ~<*
_ . _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ - _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _
i o- l r ,
II. Maintenance M1 Conduct of Maintenanc !
M1,1 General Comments ;
' Insoection Scone (62707) f
The inspectors observed or reviewed all or portions of the following work activities: :
. Preventive Maintenance Task Sheet (PHTS).P0205161. Clean and l Inspect--the A Diesel Generator (DG) Jacket Water Heat-Exchange l
- - PMTS P0205162, Clean and Inspect the A DG Intercooler and Injecto .!
Cooling Water Heat Exchange ;
- PMTS P0204941, Main Turbine Overspeed Tes * PHTS P0205160, Clean and Inspect the A DG Lube Oil Cooler Heat Exchange ,
o PMTS P0205121, Replace "0" Rings on Intake Manifold and Cylinder Liner on the A D ,
o PMTS P0205123, Perform PreveHve Maintenance on the A D .
. . Work Request (WR) 96T3271. Repair Seat Leakage on-IFV03541 EF Flow Control. Valv __
e PHTS P0204999. Change Out of Operator Diaphragm on IFV03541-0-E * WR 96T3270. Repair Seat Leakage on IFV03531-EF Flow Control Valv * PMTS P0204997, Change Out of Operator Diaphragm on IFV03531-0 E . WR 96T3275, Inspect Seat and Repair if Necessary on IFV03556-E _
e- PMTS P0205002, Change _0ut of Operator Diaphragm on IFV03556-0-EF.
I e .PMTS P0204912, Perform Eddy Current inspection on A Component
. Cooling Water (CCW) Heat Exchanger Tube .
. ;WR_9716188,'8 Train. Centrifugal Charging Pump Inboard Seal-Rep 1acement,. ,
.
ll Observations and-Findinas
~
LThe-inspectorsfoundtheworkperformedundertheseactivitieswas .-.
professional and thorough. All-of the work-was performed with the work :
-
package present amt in use. Additional discussion for some of these !
! activities is provided in ti.e .following . maintenance paragraph ;
i IL !
-
1 . ~ ~ _ _ . - . - _ _ _ _-..._= -.-._. - .- --..
_ _ _ _ _ _ _ _._ _ _ _ _ .._.__ _ _ _. _ _ _ _.._ _
l
.
.
!
I 7 Conclusions ]t All observed maintenance tasks were conducted in a competent and j professional manner. With the exception of the maintenance effort to :
clean the A DG piston (paragraph M1.2b.) appropriate tools, equipment !
and procedures were in use. Proper radiological controls were used when .
require M1.2 partial Teardown of the A DG l: Insoection Scone (62707) ,
!
The inspectors observed work in progress on.the partial disassembly of l
the A DG (MMP Procedure being) performed-180.03 in accordance
" Emergency Diesel Generator with Mechanical Maintenance Miscellaneous !
Maintenance". Revision 9. Four cylinders were disassembled and '
inspected (numbers 5, 8, 11._and 12).
b Observations and-Findinas l
For the disassembled cylinders, the pistons were removed, cleaned, and 1 inspected. In addition, the cylinder liner and cooling water jacket assembly was removed, disassembled, cleaned, and inspected. New "0" -
.
rings wert installed on the liner and reassembled with the cooling water -
jacket. This assembly was then hydrostatically pressure tested to 87 ,
psig for five minutes to ensure that a positive seal was establishe On October 7. the inspectors observed work in progress on the number 5 cylinder. This work included the piston cleaning prior to reassembl MMP-180,033. Step 7.14.3 states: -
" Remove carbon deposits on piston crown and valve !
pockets with a bevel edged brass scraping tool."
'
The technician performing this work was using a steel gasket scrapper to }
clean the carbon deposits from the )iston instead of the bevel edged
,
brass scraping tool called for in tie procedure. Since the piston is made from aluminum it is-important to use a. soft metal tool for this task to prevent damage to the piston. The supervisor for this job noticed the use of the incorrect tool and corrected the technician.. The inspectors inspected the piston in the area where the steel tool had been used and did not identify any damag '
This-failure to properly implement procedural requirements during A DG maintenance is identified as a violation. This non-repetitive.-licensee
. identified and corrected violation is being treated as a Non-cited 1 Violation (NCV) consistent with Section VII B.1 of the NRC Enforcement Policy. This is identified as NCV 50-395/97012-01.
l
!
.
u ._ ._ _ __ __ _ _ _ _ .. _ _ _ , _ _ _ _ _ _ . - . .- ---- - _ - - - _ - _ - . - - - - - _ - .
e
,
'
i
. !
8 I CpDClusions :
.;
An NCV was itentified for the failure to correctly implement the !
requirementt, of a maintenance procedure during DG maintenance. An ;
,
incorrect tool was used to clean carbon deposits from a piston. The
-
.
Inspectors concluded that the licensee had taken adequate corrective :
action; !
M1.3 Heat Exchanaer Cleanina and Eddy Current Testina for A OG Insoection Scooe (62707) j The inspectors observed the work in progress to clean and eddy current i test the tubes in the A DG jacket water cooler, and the intercooler and ;
injector cooling water heat exchanger l
' Observations and Findinas The licensee performed tube cleaning and eddy current testing of the t jacket water and intercooler heat exchangers on the A DG. The )urpose ;
of this effort was to establish the overall condition of these leat .
exchangers and determine if replacement was warranted during RF-10. All !
tubes were cleaned using a brush and 20 percent of the total numuer of ;
tubes in each heat exchanger were eddy current tested. The acceptance criteria for plugging tubes in these heat exchangers was indications greater than 80 percent through wall. Three tubes not previously plugged in the intercooler heat exchanger had indications greater than ,
80 percent and were plugged. No additional tubes in the jacket water heat exchanger required plugging. The licensee determined that these heat exchangers did not require replacement during RF-1 , Conchsions The work associated with the A DG jacket water and intercooler heat exchangers was effective in determining the overall condition of the tubes and gave reasonable assurance that they were capable of performing .
their safety-related function .M1.4L Emeraency Feedwater (EFW) Flow Control Valve Maintenance Insoection Scoce (627071 ,
The inspectors observed )ortions of maintenance to rework the EFW Flow Control Valves-(FCV). Tie motor driven EFW pump FCVs had been 3reviously identified as leaking past their seats when the motor driven EFW pumps were running and the valves were close ,
b,. Observations and Findinas On June 13, 1996, the licensee issued Licensee Event Report (LER) 50- l 395/96004 to describe an unanalyzed condition regarding reactor building line break analyses. Specifically. the licensee determined that the
-
,
h
. ,.-,u, .2,,a-,,A_-.,~.;.-,.___,__._--.._- . . . - . , . - - , _ - . _ _ . . . . . . - _ - - - - , . . . .
. i
. i
motor driven EFW pump FCVs were leaking past their seats when the motor driven EFW pumps were running and the valves were closed. This caused steam generator water levels to increas It was determined that this condition was not covered by the Emergency Operating Procedures (EOPs)
and may not be addressed in the safety analysis report. The licensee instituted com)ensatory actions and committed to rework the leaking motor driven E W pump flow control valves by the end of RF-10. In addition, the licensee expanded these actions to include the turbine driven EFW pump flow control valves. The observed condition and the licensee's actions were described in NRC Inspection Report 50 395/9600 During the inspection period, the inspectors observed the work in progress on these valves. A small amount of foreign mater',al was found in each of the motor driven EFW pump flow control valves and some pitting was found on the seat ring sealing surfaces. The valves were machined and reassembled and the valve operator diaphragm was replaced, lhe repair work was conducted in accordance with the maintenance procedure requirements. No concerns were identified. The LER will remain open pending the completion of testing to determine if the rework was effective in returning these valves to within the analyzed condition, c. Conclusions The licensee's efforts to rework the EFW flow control valves revealed some pitting and foreign material. The valves were machined and reassemble H2 Maintenance and Material Condition of Facilities and Equipment M2.1 Surveillance Observation a. Insnection Scone (61726)
The inspectors obsers !d all or portions of the following surveillance tests:
- Surveillance Test Procedure (STP)-146.002. " Reactor Makeup Water
! System Refueling Alignment Verification," Revision 3 i
- STP-215.002A " Containment isolation Leakage Test for the AH. S IA and NN Systems." Revision 3
- STP-125.008. " Diesel Generator A Refueling Operability Test."
Revision 4
- STP-501.001. " Battery Weekly Test." Revision 9
- STP-345.044. "On-Line Transmitter Time Response Testing."
Revision 2
.
. .- ._ _ __ _ _ _ .. _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ . . _
o ;
~
_.- l t
'
b.- Observations and Findinas ,
The inspectors verified that the surveillance requirements of TS 4.9. were met by the performance of STP-146.00 This STP verified that in i Mode 6 with'the reactor vessel head closure bolts less than fully tensioned. Valves XVD08430-CS, XVD08454 CS, XVD08441 CS, XVD07053-T i and XVD08439 CS were locked closed in order to prevent an inadvertent '
!
dilution. The. inspectors also verified that the licensee's Surveillance Test Task Sheets (STTSs) were in place to ensure that this surveillance j requirement would be performed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as required by TS ;
4.9.1.3, No discrepancies were identifie '
The inspectors observed Local Leak Rate Testing (LLRT) in progress using procedure STP 215.002A. The technicians performing these tests demonstrated a good level of knowledge and understanding of the 1 procedure and test equipment. No concerns were identifie ;
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> run of-the A DG (STP 125.008) followingmajormaintenance i provided reasonable assurance that the DG was capable of performing its '
safety-related functio > Conc 19ttom
'
The observed surveillance tests were conducted in accordance with approved procedures with no discrepancies or concerns identifie .
M2.2 Reactor Coolant System Leak Insoection , Insoection Scoce (61726)
The inspectors reviewed the results of the licensee's boric acid
-
inspection performed in accordance with STP 250.001 " Reactor Coolant ,
System Leak Test." Revision Observations and Findinas Following plant shutdown, with the RCS cooled down and depressurized, '
the licensee performed a visual boric acid inspection of insulated The ins ;
components reviewed the with results pressure retaining bolted of the licensee's connections.In general,pectors Mpectio there was no significant boron leakage identified during the ins)ectio .
Several bolted connections had leakage which was found to )e unsatisfactory and work orders were written to repair the leakag The inspectors also reviewed two licensee inservice inspection relief requests that were used during this surveillance test. The first relief reques: involved performing the visual inspections with the RCS cooled down and depressurized instead of performing the inspections at normal The ins RCS operating conditions as required by the ASME code,The require removal ,
'
licensee requested relief from performing-these inspections at normal L RCS operating temperature and pressure due to the hazardous environment L '
L
-.!
>
o . _ _ . _ _ _ = _ _ _ _ _ _ ,_ m
O
.
O
during these conditions. The second relief request involved actions to be taken when leakage was detected at a bolted connection. The code requires that if leakage occurs at a bolted connection, the bolting shall be removed, visually examined for corrosion. and evaluated. As an alternative the licensee proposed performing an evaluation considering several specific variables. The ins)ectors reviewed a sample of the evaluations for bolted connections tlat were identified to have leakage *
and found that the evaluations considered all the variables listed in the relief reques c. Conclusions A review of RCS boric acid leakage inspection results found that, in general, there was no significant it:akage. The licensee was meeting the requiremei.ts of the ASME code relief request H8 Hiscellaneous Maintenance Issues (92902)
M (Closed) VIO 50 395/95002 01 : failure to provide adequate instructions and corresponding evaluation for an EDG maintenance activity. This violation was denied by the licensee in a letter dated April 24, 199 A NRC review of the violation concluded a violation did not occur and was documented in a July 21, 1995, letter to the license M8.2 (00en) IFl 50-395/97011-01: snubber reduction and testing program. The inspectors continued to monitor the licensee's snubber testing )rogra The licensee's testing of snubbers, removed as part of the snub)er reduction program, identified an unusually high number of degraded snubber In order to more closely chr'acterize the extent of the snubber problem, the licensee decided to change the snubber surveillance testing method from TS 4.7.7.e.(2) to TS 4.7.7.e.(1). The surveillance functional testing performed under TS 4.7.7.e.(1) samples 10 percent of each type of snubber in use. At the close of the inspection period the licensee had decided to test 100 Arrestor (PSA) 4". 4".1" and 3" size percent of the Pacific snubber Scientific This decision was based on identifying the extent of the snubber problem rather than on the snubber failure rat The inspectors questioned the licensee's plans to perform a portion of the TS required surveillance testing while the plant was operating. The surveillance requirement. TS 4.7.7.e. states that testing will be performed at least once per 18 months during shutdown. A review of this issue with NRC staff concluded that the intent of the surveillance requirement was clear. The snubber TS functional testing is required to be performed while shutdown. The licensee was informed of the staff's review and conclusion that snubber TS functional testing shall be performed while shutdown. The licensee accepted this position on snubber TS functional testin . _ _ _ _ -
,
O
.*
12 j 1 11. Enaineerina El Conduct of Engineering $
E1.1 Review of Safety Evaluation For Service Water Pond Thermal Study ; Insoection Secoe (37551)
The inspectors reviewed the licensee's safety evaluation prepared for the service water system pond thermal performance stud Observations and Findinas, {
i The licensee gathered SW pond thermal data during plant cooldown for the refueling outage to assess the thermal )erformance of the SW pond (see IFI 50-395/96009 04). The SW pond is tie ultimate heat sink for the ;
plant. In order to obtain valid data for the study, the licensee t isolated the SW pond from Lake Monticello. installed buoys in the pond to collect tem)erature data, and installed temperature monitoring equipment in tie SW exit pipin Lake Monticello and the SW pond are interconnected by a 36-inch non-safety class pipe. The section of piping in the SW intake building is safety class piping. The lake serves as a back-up water source to the SW pond. To conduct the thermal study the licensee installed a removable plug in the 36-inch line in the SW pump house to isolate the SW pond from the lak The interconnecting line maintains an ,
ecuilibrium between normal operating levels in Lake Monticello and the Sk pon Levels in the lake change frequently. The SW aond level ,
normally fluctuates with changing levels in the lake. T1e pipe is constructed such that a minimum SW pond level will always be maintaine <
The plug was held in place with t > : hydraulic pressure created by maintaining the SW pond level above the lake level during the test. The 31ug was designed to settle to the bottom of the SW pump house intake Jay should it come loos The inspectors reviewed the FSAR to verify the design basis of the interconnecting pipe. The FSAR stated the design basis for shut down of the plant utilizes the SW pond without reliance upon the lake. The minimum level in the SW pond is sufficient to mitigate potential *
accidents. The inspectors also reviewed the potential effects of the plug on the SW pumps should the plug have come loose. The inspectors reviewed the drawings for the interconnecting >1pe and the layout of the SW pump house and concluded that the plug in tie inconnecting pipe would not affect SW system operability. There would be no interaction between the SW pumps and the plug due to the SW traveling screen The inspectors also concluded that placing the buoys in the SW pond and -
placing the temperature monitoring devices in the exit piping would not
.
affect SW system operability. With the combination of trash racks in i the SW intake canal and the traveling screens in the SW pump house. i this equipment broke free. it would not affect SW pump operatio ,
__ . _ _ _ _ _
._-_____ ______________-__ _ _ _ _ _ _ _ -_ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ .
.
..
.
13 Conclusions ,
,
A review of a safety evaluation for a thermal study of the SW pond concluded that installing a plug in the interconnecting pipe between the SW pond and Lake Monticello, and installing temperature monitoring devices in the pond did not constitute an unreviewed safety questio El.2 Modification to Resolve Thermo taa Issue Insoection Stone (37550)
The inspector reviewed plant modification ECR 34498. Appendix R Thermo-Lag, which was implemented to resolve industry wide issues with the use of Thermo Lag as expressed in NRC Bulletin 92-01. Supplement 1. " Failure of Thermo Lag 330 Fire Barrier System to Perform its Specified Fire Endurance Function." The modification was prepared by Design Engineering. The review included a detailed documentation review. an and extensive walkdowns of the areas .
independent where the work assessment was ongoing. ofThe ampacity,llowing fo two supporting calculations were reviewed: DC 6500 0020. "Ampacity Correction factor for Gypsum Board Barrier." Revision 0: and DC 8500-021. Cable Ampacity Correction Factor for Thermo-Lag Fire Barrier. Even though modification ECR-34498 removed all the Thermo Lag from the plant. calculation DC-8500 0021 was necessary to address the period of time that cables were energized with the Thermo-Lag installed. Requirements which aDplied to this scope of inspection were 10 CFR 50. Appendix 8. Criterion 111. " Design Control."
and 10 CFR 50.48 " Fire Protection." Observations and Findinas Modtfication ECR-34498 removed all the Theuno Lag used to protect '
electric circuits described in the safe shutdown plan. Two raceways were reworked by replacing the cables of interest with cable designed and tested to operate in a fire. One raceway was re-routed around the fire zone of concern. One raceway was enclosed with a gypsum board barrie \nalysis and work instructions contained in the modification package were sufficiently detailed, and covered all the relevant consideration The ampacity calculations were performed in accordance with industry standards and widely accepted methods. The inspectors agreed with the conclusions of the calculatio The completed work was accomplished in accordance with the modification package drawings and instructions. Work was ongoing at the time of the inspection, and walkdowns were made on three consecutive day The inspectors asked for backup documentation which would be retained separate from the modification package. The licensee provided this documentation which had been generated during development of the modification. Examples were the selection of the connectors to use with
. . - - -_ _ .
_
O
.. ,
.
nickel coated copper wire and Underwriters Laboratories rating for the !
gypsum boara barrie Use of the fire environment rated cable was seen as a deviation to the requirements of Section 111 G of Appendix R to 10 CFR 50. The licensee stated that they had been in communication with NRC Headquarters concerning this matter, and that to date the NRC had not expressed any objection to the particular a) plicatio In a letter to V. C. Summer dated October 10. 1997, the NRC concluded that the use of a fire resistant cable would provide the same protection as that of a one-hour fire barrier. The deviation was determined to be acceptable, c. Conclusions A modification prepared by Design Engineering implemented in the fall 1997 outage, was reviewed and found to meet the requirements for design control. The modification resolved the Thermo-Lag issue, an industry wide issue related to fire protection requirement E3 Engineering Procedures and Documentation E3.1 Review of Fuel Handlino Accident Assumotions a. Insoection Scone (37551)
The inspectors reviewed FSAR section 15.4.5. Fuel Handling Accidents, to review the licensees compliance with the accident analysis assumptions that applied during defueling of the reacto Regulatory Guide (RG) 1.25. " Assumptions Used For Cvaluating The Potential Radiological Consequences Of A Fuel Handling Accident In The Fuel Handling And Storage facility For Boiling And Pressurized Water Reactors" was also included in the revie b. Observations and Findinas The inspectors reviewed FSAR section 15.4.5. Fuel Handling Accidents, for inside and outside of containment. The inspectors review of the assumptions and the parameters used in the FSAR analysis concluded that the fuel handling accident section was consistent with current o>erating practices and plant characteristics. The inspectors review of RG 1.25 concluded that the licensee was meeting the assumptions of the R c. Conclusions A review of the FSAR and RG 1.25 fuel handling accident analysis assumptions that applied during defueling of the reactor concluded that current operating practices and plant characteristics were consistent with the accident analysis assumption ________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.*
.
E7 Quality Assurance in Engineering Activities (37551)
E7.1 Review of Uodated Final Safety Analysis Report (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 compared plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the FSAR that related to the areas inspected. No discrepancies were identifie I Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 General Comments (71750) .
During inspection activities and plant tours ..; inspectors routineiy observed refueling outage radiological controls. Due to a large chemically induced RCS crud burst following plant shutdown and the licensee's inability to reduce dose rates to pre outage levels, many areas in the auxiliary building and containment had significantly higher than expected dose rates. The inspectors observed that radiological controls were adequately implemented and maintained during a period of increased work activity and significantly higher dose rates, than usual in the radiation controlled are R1.2 Reactor Buildina Decontamination Insoection Scope (71750)
The inspectors reviewed the licensee's efforts to decontaminate the reactor building prior to the stu t of RF-1 Observations and Findinas On October 2, the inspectors attended a pre-job briefing conducted with personnel who were )reparing to enter the reactor building to perform decontamination. T1e licensee has historically maintained the majority of the reactor building accessible in street clothes during scheduled outages. To continue this level of accessibility, the licensee made early efforts to decontaminate the reactor building prior to the start of RF-10. The pre-job briefing was conducted in a professional manner with all relevant aspects of the job discussed. No concerns were identifie l C
.*
.
16 Conclusions A pre-job briefing conducted prior to a reactor building entry to decontaminate the building was professional and complet Street clothes accessibility to the reactor building was maintained at the beginning of RF 1 P1 Conduct of EP Activities Pl.1 Emeraency Drill Observation Insoection Stone (71750)
The inspectors cbserved a site augmentation drill to test the licensee's ability to staff the Emergency Response Facilities (ERF) with the appropriate personnel needed to support the site in an emergency during evening hour Observations and Findinas On the evening of September 9. the site conducted an emergency response organization augmentation drill to test the ability to adequately staff the ERFs in an emergency during the evening hours. The drill was initiated at 7:30 p.m. by informing the on-shift communicator that the station had declared a Site Area Emergency and instructed him to activate the radio pagers. Emergency response personnel responded as expected and the site ERFs were activated by 8:10 p.m. with the minimum staffing presen The inspectors also observed that fitness-for-duty was being checked during the drill. Security officers at the Protected Area entry point were questioning personnel on their consemption of alcohol in the last five hours, if alcohol bad been consumed, equipment was available to administer a breath test, Conclusions Observation of an emergency drill conducted during the evening hours concluded that the site ERFs could be activated and staffed within one hour. Personnel responding to the site were being appropriately monitored for fitness-for-duty consideration S1 Conduct of Security and Safeguards Activities S1.1 General Comments (71750)
The inspectors observed security activities including compensatory measures during the conduct of plant tours. The inspectors observed that securitv had responded adequately to the increased activity during the refueling outag _ -. - _ _ _ - - _ - _ - - _
-- _
.. .
.
.*
.
F1 Control of Fire Protection Activities F1.1 General Comrnent; (71750)
The inspectors observed that the licensee's in plant monitoring of fire protection had increased due to the increased work activity in the plant. The inspectors observed fire protection technicians checking fire doors and checking for transient combustibles. The inspectors concluded that the licensee had implemented adequate measures to ensure fire protection during the outag V. Manaaement Heetinas X1 Exit Heeting Summary The inspectors )rescnted the inspection results to members of licensee managenent at tie conclusion of the inspection on October 24, 1997. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie _ _ _ - _ _ _ _ _ _ _
o l
e e
PARTIAL LIST OF PERSONS CONTACTED (
Licensee ,
F. Bacon Manager. Chemistry Services L. Blue. Manager. Health Physics S. Byrne. General Manager. Nuclear Plant Operations P. Clary. Manager. Quality Systems M. Fowlkes. Managei'. Operations S. Furstenberg. Manager. Maintenance Services J. LaBorde. Supervisor. Electrical Design Engineering D. Lavigne. General Manager Nuclear Support Services M. Lynn Electrical Design Engineer G. Moffatt. Manager. Design Engineering K. Nettles. General Manager. Strategic Planning and Development H. 0*0uinn. Manager. Nuclear Protection Services A. Rice. Manager. Nuclear Licensing and Operating Experience A. Robosky. Fire Protection Engineer (consultant)
W. Stuart. Supervisor. Mechanical Design Engineering G. Taylor. Vice President. Nuclear Operations R. Waselus Manager. Systems and Component Engineering R. White. Nuclear Coordinator. South Carolina Public Service Authority B. Williams. General Manager. Engineering Services ,
G. Williams. Associate Manager. Operations 3, INSPECTION PROCEDURES USED
-
IP 37550: Engineering IP 37551: Onsite Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92902: Followup Maintenance ITEMS OPENED. CLOSED, AND DISCUSSED Ooened 50-395/97012-01 NCV failure to correctly implement the requirements of a maintenance procedure during diesel generator maintenance (Section M1.2)
ClQfEd 50-395/95002-01 VIO failure to provide adequate instructions and corresponding evaluation for an emergency diesel generE or maintenance activity (Section M8.1)
50-395/97012-01 NCV failure to correctly implement the requirements of a maintenance procedure during diesel generator maintenance (Section M1.zi
!
. _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
o 8%
Discussed 50-395/97011-01 IFI snubber reduction and testing program (Section M8.2)
50-395/96004 LER unanalyzed condition regarding reactor building line break analyses (Section M1.4)
.e A
.
- - - - - - _ - - - - - - - - _ _ - - - - - - - _ - - - - -