IR 05000395/1997003: Difference between revisions

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{{Adams
{{Adams
| number = ML20140D452
| number = ML20149G902
| issue date = 06/02/1997
| issue date = 07/11/1997
| title = Insp Rept 50-395/97-03 on 970323-0503.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support
| title = Ack Receipt of 970702 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp 50-395/97-03
| author name =  
| author name = Belisle G
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| addressee name =  
| addressee name = Taylor G
| addressee affiliation =  
| addressee affiliation = SOUTH CAROLINA ELECTRIC & GAS CO.
| docket = 05000395
| docket = 05000395
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-395-97-03, 50-395-97-3, NUDOCS 9706100387
| document report number = 50-395-97-03, 50-395-97-3, NUDOCS 9707240119
| package number = ML20140D419
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 2
| page count = 26
}}
}}


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U., S. NUCLEAR REGULATORY COMMISSION
 
==REGION II==
Docket No.: 50 395 License No.: NPF 12 Report No.: 50 395/97 03 Licensee: South Carolina Electric & Gas (SCE&G)
Facility: V. C. Summer Nuclear Station l
Location: P. O. Box 88 Jenkinsville, SC 29065 Dates: March 23 May 3, 1997    j i
Inspectors: B. Bonser Senior Resident Inspector T. Farnholtz, Resident Inspector D. Jones, Reactor Inspector, RII (Section R1.2)
L. Garner, Project Engineer, RII (Sections 01.1, 04.1 and R2.1 [ partial], 04.2 and E1.3)
Approved by: G. Belisle, Chief, Reactor Projects Branch 5 Division of Reactor Projects i
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Enclosure 2
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9706100387 970602 PDR ADOCK 05000395 0 PDR i
 
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EXECUTIVE SLM4ARY V. C. Summer Nuclear Station NRC Inspection Report No. 50-395/97 03 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6 week period of resident inspection; in addition, it includes the results of an announced inspection by a regional inspector and a project engineer.
 
Doerations e The licensee took conservative actions to mitigate any potential reactor
)ower transient during corrective maintenance on a Moisture Separator Reheater pressure transmitter (Section 01.1).
 
. A manual reactor trip was initiated following a loss of electro-hydraulic control system fluid. All safety related components functioned as ex with secondary,non pecte safetySeveral relatedproblems wereThe component identified licenseeassociated took appropriate actions to correct these problems. Operator performance during these events was acceptable (Section 01.2).
 
* All safety systems functioned as designed following a turbine trip and reactor trip. The inspectors concluded that during these events the plant had operated safely. The response by operators was not based on a complete awareness of the plant's response to a high high Steam Generator (SG) level, P-14 actuation. A violation was identified for a failure to establish procedures appropriate to the circumstances. The feedwater system operating procedure and the general operating procedure in use did not provide adequate guidance to ensure that operators could maintain control of SG levels. The turbine trip abnormal operating procedure did not provide adequate operating instructions for response to a turbine trip on high high SG level (Section 01.3).
 
. Start ups on April 25 and 28 were performed safely, with good control and communication between the operating staff and reactor engineerin Reactivity additions were carefully controlled and monitored. On both start ups critical rod position was within the estimated critical rod position calculated by reactor engineering (Section 01.4).
 
* Pressure transients in the Deaerator (DA) had occurred due to the introduction of cool condensate water into a hot DA. The inspectors identified an Inspector Followup Item (IFI) to review the cause and corrective action for these pressure transients (Section 02.2).
 
* A repeat violation was identified concerning the failure to follow the applicable procedure to raise the aressure in the reactor building. The operator opened two containment ex1aust valves rather than the two containment supply valves called for in the procedure (Section 04.1).
 
. Lower Auxiliary Building Operator logs were taken in accordance with Operations administrative procedures (Section 04.2).
 
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.. The Plant Safety Review Committee (PSRC) monthly meeting met Technical Specification (TS) quorum requirements and focused on safety while reviewing agenda items (Section 07.1).
 
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At the PSRC meetings prior to plant restart, issues were adequately documented, discussed and resolved such that there were no open safety issues affecting a safe plant restart. Plant management ensured that-all issues affecting plant safety were resolved before plant startu The post trip reviews performed by the Independent Safety Engineering
. Group made a contribution to the plant restart effort and provided useful insights to the events (Section 07.2).
 
Maintenance e Maintenance activities were generally conducted in an appropriate and professional manner. Adequate actions were taken to identify and repair plant equipment as required (Section M1.1).
 
* A weakness was identified in the licensee's program for lifting loads in the yard. The licensee demonstrated poor work practices and non-conservative decision making on the part of the responsible supervisor when a Service Water (SW) pump motor was lifted over the top of the safety related SW pump building with no engineering analysis or evaluation performed to determine the consequences of a dropped load onto safety related structures or components (Section M1.2).
 
. Surveillance activities were conducted satisfactorily and in accordance with applicable procedures (Section M2.1).
 
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A Non cited Violation (NCV) was identified for the failure to document and report a deficiency for a Condensate Storage Tank (CST) nitrogen injection valve surveillance test which was perforved nine times over a two and a half year aeriod. The test procedure directed the test performer to open a autterfly valve which could not be fully opened due to an interference (Section M2.2).
 
Enoineerina
. The plant responded as designed when the high high SG level caused only one feedwater isolation valve to close on April 26 (Section E1.1),
a The licensee had made a proactive decision by declaring the A SW pump motor inoperable based on the similar design of all three motors and on the similar failures of the B and C motors (Section E1.2).
 
. The licensee implemented actions to minimize potential damage to safety-related equipment during the Industrial Cooling Tower replacement (Section E1.3).
 
* An Unresolved Item (URI) was identified concerning the cause of an as-found, out of-specification condition on some Emergency Feedwater (EFW)
system flow control valves and the consequences of higher than expected
 
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l    3    l EFW flow differentials. The as left condition of the EFW system was acceptable (Section E2.1).      -
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. As part of their Final Safety Analysis Report (FSAR) Review Project, the  ;
licensee has established a method for prioritizing and tracking  '
identified FSAR discrepancies from plant operating practices and design  i docunents (Section E7.1).      ,
Plant Support
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e The licensee's program for transportation of radioactive materials had  !
been effectively implemented pursuant to Department of Transportation and NRC regulations (Section R1.1).    ;
 
. An URI was identified concerning the failure to meet the requirements of  i 10 CFR 70.24, criticality accident requirements (Section R2.1).  '
. Appropriate compensatory measures were implemented while a modification  l to the protected area fence was in progress (Section S2.1).    ;
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I Report Details Summary of Plant Status Unit 1 began the inspection period at 100 percent power. On April 17 power was reduced to 90 percent to support maintenance on a Moisture Separator ;
Reheater pressure transmitter. Power was returned to 100 percent later on '
April 17. At 4:54 p.m. on April 22 the unit was manually tripped from 100 percent power due to an Electro-Hydraulic Control (EHC) system fluid leak on the main turbine number one Combined Intercept Valve (CIV) shutdown serv The unit remained in Hot Standby (Mode 3) until April 25. On April 25 at 1:45 '
p.m., the licensee commenced a reactor start-up. The reactor was taken critical at 2:48 p.m. and entered Power Operation (Mode 1) at 6:08 p.m. on April 25. On April 25 at 9:04 p.m., the licensee commenced a load decrease from 25 percent reactor power due to another EHC leak on the main turbine number one CIV shutdown servo. At 9:14 p.m., the turbine was tripped and reactor power was stabilized at seven percent. On April 26 at 4:20 a.m.,
after repairing the EHC leak, the licensee commenced raising reactor powe On April 26 at 8:43 p.m., the main turbine tripped on high high SG level from 43 percent reactor power. Four minutes later at 8:47 3.m., the reactor tripped from six percent power on low low SG level. T1e unit remained in Mode 3 until April 28. On April 28 at 4:40 a.m., the licensee commenced a reactor start up. The reactor was taken critical at 5:28 a.m. and entered Power Operation (Mode 1) at 10:15 p.m. on April 28. On May 1 at 6:45 a.m., reactor power reached 100 percent. The plant remained at full power through the end of the inspection perio Operations Conduct of Operations
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01.1 General Comments (71707. 40500)    4 Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety-conscious: specific events and noteworthy observations are detailed in the sections belo On April 17, licensee took conservative actions to mitigate any potential reactor power transient during corrective maintenance on a Moisture Separator Reheater pressure transmitter. The power reduction to 90 percent and return to full power was performed in accordance with a standing order and normal operating procedure :
01.2 Manual Reactor Trio Due to Hydraulic Oil Leak Inspection Scope (71707)
The inspectors reviewed the licensee's actions and the plant response following a manual reactor trip which was initiated because of an EHC system lea ,
 
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2  ; Observations and Findinas
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On April 22, at 4:54 p.m., control room operators initiated a manual reactor trip from approximately 100 percent power. A main control board annunciator alarmed indicating an EHC reservoir hi/lo level conditio Operators were dispatched to investigate the reservoir level and to I evaluate an EHC system leak on the main turbine number one CIV which was identified earlier in the week. Local indication of the EHC reservoir was pegged low and it was reported that the leak on the CIV was much larger than previously identified. The Shift Supervisor, in  ,
consultation with the Associate Operations Manager, ordered a manual '
reactor trip. The resulting turbine trip caused the CIV to close which isolated the EHC system lea The EHC fluid that leaked from the system was contained in the turbine building sump near the base of the main condenser. The licensee roped off the area and cleaned up the EHC flu 1d using appropriate personnel protection and fire prevention measures. No personnel were injured during this even Safety related components functioned as expected on the reactor tri All control rods inserted fully. A review of the sequence of events recorder data revealed that the rod drop times were within the TS requirement S3ecifically, the TS requires that all rods must fully insert in less t1an or equal to 2.7 seconds. For this trip, all rods fully inserted in 2.539 second Following the trip, a low low level condition in the A SG caused an automatic start of the A and B motor driven EFW pumps and subsequent l feeding of all three SGs. The SG levels were restored to the proper level. A concern was identified by the licensee involving a larger than expected EFW differential flow to each of the SGs immediately following
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the EFW pump start. This concern is discussed in detail in Section E2.1 l of this repor On the secondary side, several problems were identified following the trip. These problems, along with the resolutions are discussed below:
. Two of the three Feedwater Regulating Valves (FRVs) (FV-488 for-the B SG, and FV 498 for the C SG) did not indicate closed on the main control board as expected. The significance of this was minimal since all three feedwater isolation valves closed to prevent overfeeding the SGs.
 
l  The post-trip investigation revealed that FV 488 fully closed as expected following the reactor trip. The closed limit switch on
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the valve required adjustment to provide the proper indication on the main control board. The licensee 3erformed a diagnostic test
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on FV-498. This test indicated that t1e valve was not operating as smoothly as expected and that an increase in operating friction
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had occurred since the last time this test was performed. The
 
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valve was cleaned and lubricated to improve the operation and placed back in servic l
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. The B and C Main Feedwater Pumps (MFPs) failed to trip on the first attempt from the main control board. The pumps tripped on
.the second attempt. All three MFPs were tri) ped in accordance with Emergency Operating Procedure (E0P), E0)-1.1, " Reactor Trip Recovery," Revision 7. Step 4(c).
 
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Post trip testing did not indicate any problem with the trip circuit. The licensee determined that the most probable cause of this problem was that the operator did not hold the trip switch in the trip position long enough. A procedure revision to System Operating Procedure (SOP), S0P-210, "Feedwater System," Revision ,
13 was being developed to instruct the operators to hold the trip l
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switches until the trip light illuminate . The speed changer for the B MFP did not run to minimum speed
. following the pump tri The licensee determined that the motor associated with this speed i changer had an open winding. The motor was replaced and tested
, satisfactory, i
. Non return check valve (XVC 2014B) from the 2B FW heater to the Moisture Separator Reheater did not indicate closed on the main control board. This valve closes on a turbine trip to isolate the main turbine from the energy contained in the feedwater heate The post trip investigation determined that a solenoid
_ malfunctioned on the check valve booster. The main control board
; indication comes off the booster which did not function properly.
 
j However, the check valve disk did go closed as expected following the tri i All of the problems noted were on the secondary, non safety related equi) ment of the plant. The licensee attributed some of these problems to t1e long operating run (334 days) the plant experienced prior to the
: tri Much of the equipment which did not function properly following
: the trip had not been called on to operate since the last refueling L
outage in May 199 The licensee determined that the cause of the EHC system leak, which necessitated the manual reactor trip, was a ruptured 0-ring on the shut down servo on the number one CIV. The servo assembly and the four associated 0-rings were replaced. The ruptured 0 ring was being examined to determine the cause of the failur At 9:04 p.m. on April 25, with the )lant at 25 percent power, operators commenced a load decrease when anotler EHC leak was identified. The magnitude of this leak was less than the previous leak and was such that reactor power was reduced in a controlled manner to about seven percent-1
 
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and the main turbine taken off-line. The source of the leak was the ,
same number one CIV but from a smaller diameter 0 ring in the shut down servo. The leak was repaired and on April 26 at 4:20 a.m., the licensee !
, commenced raising reactor powe ;
' Conclusions
. A manual reactor trip was initiated following a loss of EHC system fluid. All safety related components functioned as expected. Several problems were identified with secondary, non safety related component l The licensee took appropriate actions to correct these problem t Operator performance during these events was acceptable.
 
01.3 Iurbine Trio and Automatic Reactor Trio  j Inspection Scope (71707)    l l
The inspectors reviewed the licensee's actions and the plant response !
following a turbine trip and an automatic reactor trip on April 2 '
4 Observations and Findinas i On April 26 at 4:00 p.m., the plant initiated a reactor power increase from 30 to 50 3ercent at approximately three percent per hour. At l 8:40 p.m., wit 1 reactor power at 42 percent, while areparing to start a second MFP and a third FW booster pump, operators o> served SG water levels trending downward. In response to the lowering SG levels, operators manually raised demand on the MFP master controller on the
 
control board to increase FW flow to the SGs. The adjustments to the ;
controller caused SG levels to rise quickly. At 70 percent level in the
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SGs, operators observed FRV automatic control to be responding slowly to ,
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control SG levels and took manual control of the FRVs to reduce flow to i the SGs. At 8:43 p.m., a high high SG level annunciator alarm (79 i percent SG level) P 14 was received on the C SG. This caused a main turbine trip, a MFP trip, a start of the motor driven EFW aumps, and a !
closure signal to the FW Isolation Valves (FWIV) and the F1V )erators rapidly lowered reactor power to about seven percent. Only tie A train FWIV closed on the P 14 signal. This FWIV response was l questioned. An investigation found the P 14 circuitry responded as ;
designed (see Section E1.1).
 
Operators promptly restarted a MFP in an attempt to feed the SGs and restore levels. At 8:47 p.m., before SG levels could be restored, the reactor tripped from about seven percent reactor power on low low SG water level in the A SG (27 percent). With the A FWIV closed and the A SG being supplied by only EFW flow, level in the A SG could not be restored before reaching the low-low level setpoint. On the reactor trip all safety systems functioned as designe The investigation of the high high SG level turbine trip showed that operators had lost control of SG 1evels. The General Operating Procedure, G0P 4, " Power Operation (Mode 1)," Revision ll, and S0P 210,
 
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i l "Feedwater System," Revision 13, require establishing a differential
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established according to a program level given in S0P 210 in order to
! maintain control of FW flow and control of SG levels. The inspectors * ,
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review of dp data subsequent to the turbine trip and reactor trip showed :
that as power approached 40 percent, FW and steam header dp had trended .
downward to below the program level. At 40 percent power the program dp i
; is 130 psid. At 8:40 p.m., according to plant computer data, dp was 60 l psid prior to the adjustment of the FW pump master speed controller by i the operator. Plant operators, however, stated that they were cognizant ;
of the program dp in SOP-210 and that their Main Control Board (MCB) l
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instrumentation showed it was being maintained correctly. To compensate !
for the actual dp and the increasing demand for FW flow the FRVs opened ,
beyond their normal operating position. This reduced the capability of i the FRVs to effectively control flow to the SGs. - With the reduced flow !
control capability of the FRVs the adjustment to the FW master controller to raise lowering SG 1evels caused a significant increase in FW flowrate and a quick rise in SG levels. The FRVs automatic control responded slowly to the rapidly rising SG levels. At about 70 percent-SG level when operators took manual control of the FRVs they were unable to control flow before the C SG reached the high high SG 1evel setpoint (P 14) which caused the turbine tri Operators interviewed after the event believed they were maintaining the correct dp between the FW and steam headers and stated they had been
, checking dp on their MCB instrumentation. This is accomplished by reading the aressure from the FW pump discharge header pressure instrument ()I-508) and steam header pressure instrument (PI-464C) and determining the difference to obtain dp. The licensee performed a calibration check of the HCB instrumentation used to determine dp and found the instrumentation to be in calibration. The instrumentation, however, has a 0 to 1300 psi scale with an allowable two percent margin of erro As a result, the dp maintained between FW header pressure and steam header pressure was not adequate to maintain control of SG levels. This resulted in a loss of SG level control and a high high SG level turbine .
trip from 42 percent power. The inspectors concluded that procedures, !
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S0P-210 and G0'P 4 were not adequate to ensure that operators could maintain adequate control of SG levels. Operators also did not  l recognize the potential loss of SG level control associated with the FRVs being close to full open, This failure to establish procedures to accurately control FW dp and maintain SG levels'is identified as a Violation, 50 395/97003 0 !
Four minutes after the turbine trip at 8:47 p.m., an automatic reactor i- trip occurred from six percent reactor aower on A SG low low level.
 
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Following the turbine trip the Control Room Supervisor directed <
l operators to lower reactor power to seven percent. Based on operator ;
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feedback, the operators believed that FW flow to the SGs could be '
. reestablished promptly to restore SG levels. Lowering reactor power !
l further to one to three percent for EFW to provide adequate SG level '
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  .o July 11, 1997 South Carolina Electric & Gas Company ATTN: Mr. Gary Vice President, Nuclear Operations Virgil C. Summer Nuclear Station P. O. Box 88 Jenkinsville, SC 29065 l
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SUBJECT: NRC INSPECTION REPORT NOS. 50 395/97-03
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control did not appear to be a priority. Operators failed to recognize all the component actuations that should be expected on a high high SG level actuation and did not observe that the A FWIV had closed. When FW flow was restarted immediately prior to the reactor trip only the B and l
; C SGs were receiving FW. All safety systems functioned as designed on i
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the reactor trip.


j The inspectors concluded that Abnormal Operating Procedure, A0P 214.1,
==Dear Mr. Taylor:==
" Turbine Trip," Revision 2, did not provide adequate operating instructions for a turbine trip from high high SG level. The inspectors
i Thank you for your response of July 2,1997, to our Notice of Violation,  i issued on June 2, 1997, concerning activities conducted at your V. C. Summer l facility. We have evaluated your response and found that it meets the '
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requirements of 10 CFR 2.201. We will examine the implementation of your  ;
found that o wrators were unaware that they had received a FW isolation signal and tlat owrators were not familiar with the expected plant response to a SG ligh high level actuation. This failure to establish procedures to respond to a high high SG level and resulting turbine trip i
corrective. actions during future inspections.   '
; is identified as a second example of Violation 50-395/97003-0 l
We appreciate your cooperation in this matter.
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c. Conclusion All safety systems functioned as designed following the turbine trip and reactor trip. The' inspectors concluded that during these events the 1 3 plant had operated safely. The response by operators was not based on a complete awareness of the plant's response to a high high SG level P-14 actuation. The inspectors also concluded that the FW system operating procedure and the general operating procedure in use did not provide adequate guidance to ensure that operators could maintain control of SG 1evels. The inspectors concluded that the turbine trip abnormal operating procedure did not provide adequate operating instructions for response to the turbine trip on the high high SG leve .4 Plant Restarts Following Plant Trips a. Inspection Scope (71707)
i The inspectors observed reactor start ups on April 25 and April 28 !
following the plant trip i Observations and Findinas    !
On April 25 at 1:50 p.m., operators commenced a reactor start up with rods following the manual reactor trip on April 22. The reactor was taken critical at 2:48 p.m. on control bank On April 28 at 4:41 a.m., operators commenced a second reactor start-up with rods following the automatic reactor trip on April 26. The reactor .
was taken critical at 5:28 a.m. on control bank D. The control bank D !
critical. position was within the tolerance allowed by the estimated critical rod positio The inspectors observed both start-ups and found that they were performed in accordance with GOP 3, " Reactor Startup From Hot Standby To Startup (Mode 3 to Mode 2)," Revision 10. The ins actors noted that ;
shortly before the actual reactor start ups, the slift crew performing :
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Sincerely, j
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Original signed by George A. Belisle George A. Belisle Chief Reactor Projects Branch 5 Division of Reactor Projects Docket No. 50-395 License No. NPF 12 cc: R. J. White Nuclear Coordinator (Mail Code 802)
S.C. Public Service Authority Virgil C. Summer Nuclear Station P. O. Box 88 Jenkinsville, SC 29065 J. B. Knotts, Jr., Esq.


each start up had practiced the evolution in the simulator. Prior to 1 commencing each start-up, a thorough control room pre job briefing was '
Winston & Strawn 1400 L Street, NW Washington,'D. C. 20005 3502 cc continued: See page 2   f O
conducted covering many aspects of the start-up. The start-ups were i deliberate, carefully controlled, and performed with clear i communications. During both start-ups, the reactor operator and Control !
9707240119 970711  hk,kkkk .
Room Supervisor were focused on reactor controls and closely monitored l core response during reactivity additions and the approach to criticality. Rod withdrawal and core response were coordinated with the reactor engineer present in the control room. Reactor engineering also closely monitored both start ups and performed an inverse count rate ratio plot to monitor the approach to criticality. All systems ;
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functioned as expected during the start u ;
, PDR ADOCK 05000395 P PDR ftOf
l Conclusions The inspectors concluded that both start-ups were performed safely, with good control and communication between the operating staff and reactor
, engineering. Reactivity additions were carefully controlled and monitored. On both start ups. the critical rod position was within the tolerance of the estimated critical rod position calculated by reactor engineerin Operational Status of Facilities and Equipment 02.1 Enoineered Safety Feature (ESF) System Walkdown (71707)
The inspectors used Inspection Procedure 71707 to walk down accessible portions of the A and B diesel generators * air start systems. There were no concerns identified during these walkdowns.
 
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02.2 Feedwater System Waterhammers Inspection Scope (71707)
The inspectors reviewed potential waterhammer/ pressure transient events associated with the condensate system and DA during the plant restart from April 25 to April 2 ,
      . Observations and Findinas During the plant restarts following the plant trips, water hammer / pressure transient events occurred in the condensate system and the DA. The DA and DA storage tank are the dividing line between the condensate and FW system and provide deaeration, heating and storage of condensate prior to use by the FW system. The DA storage tank also provides sufficient static head to meet FW booster pump net positive suction head requirement Following these pressure transients the inspectors walked down the main condensate piping on April 29 and did not identify any obvious damage due to these pressure transients. The inspectors also reviewed the potential causes of these events with Engineering. The transients
 
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f l appear to have been caused by the injection of relatively cool  .
l condensate water into a hot DA steam space causing flashing of the !
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condensate, immediate pressure and temperature depression in the DA, !
bulk boiling and flashing of the DA storage tank water, and high stress l loads on the DA. At the end of the inspection >eriod Engineering had not completed their review of these events and lad not proaosed ;
L corrective actions to prevent similar future transients. Jntil the !
inspectors review the pending engineering evaluation and proposed l corrective. actions by operations, this issue is identified as Inspection i Followup Item. IFI 395/97003 0 l
       , Conclusion l The inspectors concluded that the pressure transients had occurred due to the introduction of cool condensate water in to a hot DA. The inspectors identified an IFI to review the cause and corrective action for these pressure transient Operator Knowledge and Performance 04.1 Reactor Buildina Ventilation System Operatina Error
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. Insoection Scope (71707)    '
The ins >ectors reviewed the circumstances associated with an operator error t1at occurred while attempting to raise Reactor Building (RB)
pressure using the RB Ventilation Syste I Observations and Findinas    j s
On April 13. -the control building operator attempted to raise RB l pressure after receiving a containment low differential pressure alar j This is a relatively routine operation that is to be performed in i accordance with 50P-114. " Reactor Building Ventilation System," Revision 15 Section III.N. While performing Step 2.2 the operator opened valves l PVG 6066 and PVG 6067, containment purge exhaust isolation valves l instead of PVG 6056 and PVG 6057, containment alternate purge supply '
isolation valves. This resulted in a slow decrease in RB pressure
; instead of the expected increase. No TS limits were exceeded during ,
this event and the misaligned valves were identified by a second
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i operator and the correct valve lineup was establishe '
l This event was a repeat of an identical error which was documented in Inspection Report No. 50 395/96013. The earlier event occurred on
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November 12, 1996, and was one of two examples that was identified as an L NCV 50 395/96013 01. The licensee's corrective actions for the 1996 l event were documented in Condition Evaluation Report (CER)
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96 0345. At that time, the individual involved was counselled as to the !
! proper use of the approved procedure and self checking. Also, all operations personnel were briefed on the event by the individual to
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stress the importance of procedural compliance. The inspectors verified this action by reviewing the training attendance records. The
 
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l inspectors concluded that these corrective actions could have reasonably prevented the occurrence of the most recent event. The failure to ,
correctly implement procedure SOP 114 is a violation of TS 6.8.1.a. and is identified as Violation 50-395/97003 0 On A)ril 15, the inspectors observed an operator increasing the pressure in tie RB in accordance with S0P-114. Section III.N of SOP 114 which l provided instructions for the evolution was identified as " Reference Only, Procedure Segments May Be Performed From Memory. Must Verify Work Following Each Segment." The evolution was performed without difficulty. The inspectors observed that the valves involved in the April 12 improper valve lineup were identified in the procedure exactly as they were labeled on the control panel. The inspectors concluded that the arocedure format and content was adequate to successfully perform t1e evolution. Several potential procedural enhancements were discussed with the licensee. For example, the inspectors noted that Steps 1.6 and 1.7 identified gaseous and iodine radiation monitors RMA0004 and RMA0002 in a manner different than the nomenclature on the control panel equipment identification labe Conclusions A repeat violation was identified concerning the failure to follow the applicable procedure to raise the pressure in the RB. The operator opened two containment exhaust valves rather than the two containment supply valves called for in the procedur .2 Lower Auxiliary Buildina 00erator Rounds Inspection Scope (71707)
The inspectors observed the licensee performing lower auxiliary building operator round Observations and Findinas On April 14, the inspectors observed the night shift Lower Auxiliary Building Operator recording his TS logs. The inspectors verified that tha log was taken in accordance with Operations Administrative Procedure (OAP), 0AP 106, " Operating Logs " Revision 6. Change A. The log readings were recorded on a hand held data entry device. The inspectors verificd that selected values were correctly entered and reviewed the recorded log with the operator upon completion of his rounds. All data taken met acceptance criteria and channel checks were performed as required by TS.
 
l Conclusions l Lower Auxiliary Builaing Operator logs were taken in accordance with
'
OAP _ a 4 _a .--s a 6, 4* +4.-R -w & ~-A = = W
.
 
07 Quality Assurance in Operations 07.1 Plant Safety Review Committee (PSRC) Monthly Meetina InsDeCtion SCoDe (71'707)
The inspectors observed the conduct of the monthly PSRC meetin Observations and Findinas The PSRC monthly meeting on April 15 met TS quorum requirements and focused on safety while reviewing agenda items. Focus on safety was demonstrated by deferring ap3roval of a proposed change to Station Administrative Procedure (SA)) SAP 123. " Procedure Use And Adherence,"
Revision 1, Change B, until questions concerning the 10 CFR 50.59 process and FSAR revision process were addresse Conclusions The PSRC monthly meeting met TS quorum requirements and focused on safety while reviewing agenda items.


07.2 Observation of PSRC and Independent Safety Enaineerino Group (ISEG)
    - . .. . - - - - . .-
' Activities Insoection Scope (71707)
  *
The inspectors observed the_ conduct of PSRC meetings on April 25 and April 27 to discuss plant restart issues and make a decision on plant restart. The inspectors also observed ISEG involvement in the evaluation of post trip documentation and their evaluation of the cause of the reactor trip Observations and Findinas Following both reactor trips, PSRC meetings were convened prior to plant restart to review the resolution of restart issues and determine plant readiness for restart. The PSRC meetings discussed each restart issue in detail. The cause and the resolution of the significant issues were presented to the PSRC by the responsible engineer and then discusse The questioning and discussion during the meeting was penetrating and reflected a desire to reach a safe resolution of the issue The inspectors also observed the ISEG involvement in the review of post trip documentation and data, in the PSRC meetings, and in the resolution of post trip issues. ISEG made several observations that'provided useful insights to the event . ~. _ - - - . . - . . . .
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      : Conclusions
SCE&G   2 cc: Continued Chairman Fairfield County Council P. O. Drawer 60 Winnsboro, SC 29180
.
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      ,
Virgil R. Autry, Director Radioactive Waste Management
The inspectors concluded that all issues were adequately documented, discussed and resolved at the PSRC meetings such there were no open i safety issues affecting a safe plant restart. The inspectors noted that i plant management ensured all issues affecting plant safety were resolved before plant startup. The inspectors also concluded that the post trip
  !
      .
i reviews performed by ISEG had made a contribution to the plant restart effort and provided useful insights to the event II. Maintenance  i M1 Conduct of Maintenance M1.1 k neral Comments Inspection Scope (62707)   ,
i
.The inspectors reviewed or observed all or portions of the following )
work activities
      :
* Maintenance Work Recuest (HWR) 9700064, Investigate and repair i cause of C Service Water'(SW) pump motor trippin l l
      '
. MWR 9709982, Perform flow scan on feedwater regulating valve FV 49 . MWR 9710078. Feedwater pumps main speed setpoint station failure detectio Observations and Findinas On April 1 the licensee attempted to place the C SW pump in service on train A to allow the A SW pump to be tagged out to perform maintenance on the A SW screen wash isolation valve. Several seconds after the C SW pump start, the motor tripped on a 50G ground relay. Electrical maintenance personnel performed a megger test on the motor which indicated a possible fault to ground and damage to the motor. The operator present at the motor monitoring the start stated that there was a smell of burned insulation when the pump start was attempted. The ,
licensee tagged out the motor and shipped it off site to be repaire '
The inspectors did not identify any concerns with observed maintenance associated with the SW yump with the exception of a concern identified during the lifting of tie pump motor. This concern is described in detail in Section M1.2 of this repor Followirig the manual reactor trip on April 22, it was observed that the FRV for the C SG (FV-498) did not close fully as expected. The licensee wrote an HWR to perform a diagnostic test on this valve to determine the
 
      .
  .
 
cause of the failure. A slight increase in operating friction was observed. The valve was cleaned and lubricated and returned to servic Conclusions Maintenance activities were generally conducted in an appropriate and professional manner. Adequate actions were taken to identify and repair i
>
plant equipment as require !
M1.2 Liftino of C SW Pumo Motor Insocction Scope (62707)
The inspectors observed the removal of the C SW pump motor from the SW ,
pump building to a flat bed trailer for shipment off site for repair l I Observations and Findinos On April 2, the inspectors observed the licensee remove the C SW pump motor from the SW pum) building and load it on to a flat bed trailer for shipment off site to ]e repaired. This required two separate lifts and two different lifting devices. The first lift was done inside the SW building using an installed overhead crane to lift the motor from the motor base to a track mounted skid that carried the motor to the opening i of the SW pump buildin From the SW building opening, an outside crane '
lifted the motor from the skid and lowered it onto a trailer. The motor weighs approximately 16,000 pound While observing this activity, the inspectors noted that the outside load path for the motor was over the SW pump building which is a safety-related structure. Specifically, when the motor was lifted from the skid on the south side of the SW pump building using the outside crane, it was swung over the southwest corner of the building and lowered onto the flat bed trailer which was parked on the west side of the pump buildin The inspectors reviewed General Maintenance Procedure (GMP), GMP-100.015. " Miscellaneous Crane Operators," Revision 3. This is a safety-related procedure which was written as part of the licensee's commitment to NUREG 0612. " Control of Heavy Loads at Nuclear Power Plants." This procedure addresses the use of all cranes and hoists used in the yar Section 7.4 of this procedure provides the requirements for performing critical lift Paragraph 7.4.1 describes a critical lift as follows:
  " Selected items due to their size, configuration or susceptibility to damage require controls in excess of those previously identified. These items are to be determined by the responsible supervisor."
 
Attachment II of GMP 100.015 " Critical Lift Checklist," is required to be used for all critical lift Part I of this checklist is to be filled out by the responsible job supervisor with assistance from an
 
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Bureau of Solid and Hazardous Waste Management S. C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 R. M. Fowlkes. Manager Operations (Mail Code 303)
13  ;
South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station P. O. Box 88
 
engineer to determine if a safe load path is required and the planned
, route. The licensee did not consider the lift of the C SW pump motor to '
; be a critical lift and therefore did not use Attachment II to plan the :
path of the lift. For this lift, no engineering analysis was performed l to determine the consequences of a dropped load to the SW pump building or to the safety-related components inside the building.
 
,
The licensee has designated the responsible job supervisor with the responsibility to determine when and if engineering personnel are to be r
, consulted as to the consequences of a lift in the yard. The inspectors !
: considered this responsibility misalaced since the supervisor may not be familiar with the construction met 1ods used for safety related'
 
structures and is not in a position to perform an adequate evaluation for the consequences of a dropped load. The inspectors discussed this i with plant management. They acknowledged the inspectors' comment )i l The inspectors concluded that the lift of the C SW pump motor over the, )
, safety-related SW pump building to be an example of a poor work practice j i and non conservative decision making. This is identified as a weakness i
'
in the licensee's program for lifting loads in the yar l
      , Conclusions
-
A weakness was identified in the licensee's program for lifting loads in the yard. The licensee demonstrated poor work practices and non-
, conservative decision making on the part of the responsible supervisor
.
when a SW pump motor was lifted over the top of the safety related SW l
: pump building with no engineering analysis or evaluation performed to 1 i determine the consequences of a dropped. load onto safety related structures or components.
 
5 M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Surveillance Observation
' Insoection Scope (61726)
':
The inspectors reviewed or observed all or portions of the following
; ' surveillance tests:
  . STP 130.003, " Valve Operability Testing (Modes 1, 2 and 3),"
  . Revision 4
,
  . STP-142.005, " Turbine Trip Actuation Device," Revision 3
  . ICP-310.008, " Reset High Flux At Shutdown Alarm," Revision 3
,
  * STP 134.001, " Shutdown Margin Verification," Revision 8 I
  * STP-102.001 "NI Analog Channel Operational Test," Revision 5
      !
 
      .
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b. Observations and Findinas On April 23, the licensee performed STP-130.003, Section 6.5, "Feedwater System Valve Testing," Step 1. The purpose of this section was to test the K648 and K649 train A output relays to ensure that they function to close the main FW bypass valves. Step 6.5.i.2 directs the test performer to energize the K648 output relay by turning test switc S 841, to the right hand position, depress and hold. During the performance of this step, the relay failed to energize. The licensee determined that the test switch may not have been operated correctly or that the test switch was mechanically bound preventing the switch from making good contact. The test was reperformed twice with satisfactory ,
results following the initial failure. When the circuit is lined up for normal operation, the test switch is bypassed such that a faulty test switch would have no effect on the operation of the K648 output rela The inspectors did not identify any concerns with the licensee's actions regarding this test failur '
c. Conclusions Surveillance activities were conducted satisfactorily and in accordance with applicable procedures.
 
M2.2 CST Nitroaen In.iection Check Valve Test a. Inspection Scooe (61726)
j
 
The inspectors reviewed the performance of STP 220.009, " Closure Testing I of XVC01067 EF, CST Nitrogen Injection Check Valve," Revision l
      .
b. Observations and Findinos On April 25, licensee test unit personnel attemated to perform STP 220.009 to demonstrate the full closed capa)ility of the CST nitrogen injection check valve, XVC01067-EF. Step 6.1.3 directs the test performer to open XVB01035 EF, CST drain valve. The test performer could not fully o)en this butterfly valve because of interference which would not allow t1e valve o>erating handle to rotate through a full circl Because of the ina)ility to fully open this valve, the test performer wrote a test deficiency for this test and declared the test j unsatisfactor j
      !
This test procedure had been in place since November 1994. A records !
review revealed that the test had been performed nine times and signed '
off as satisfactory. During these previous tests, the test performers did not identify the aroblem with XVB01035 EF but instead opened the valve as far as possi)le given the interference and continued with the tes i
      !
Procedure SAP 134. " Control of Station Surveillance Activities "
Revision 8, Section 5.5.1, describes the responsibilities of the test
 
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Jenkinsville, SC 29065 April Rice, Manager
e performe Paragraph 5.5.1.A.7 describes one of the test performers
. Nuclear Licensing & Operating
-
,  Ex wrience (Mail Code 830)
responsibilities as follows:
Sout1 Carolina Electric & Gas Company Virgil C. Summer Nuclear Station '
  " Documenting and reporting Test Deficiencies, i i  immediately to the Shift Supervisor and Responsible Supervisor for corrective action."
P. O. Box 88 4 Jenkinsv111e, SC 29065'
 
    (o -
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Distribution:
The test performers for the previous nine tests between November 1994  ;
.
and April 1997 did not identify as a test deficiency the inability to
;  fully open the CST drain valve as directed by the test procedure.
 
;        1 The inspectors reviewed the test methodology and the acceptance criteria i
;  to determine if the previous tests were satisfactory even though the CST drain valve could not be fully opened. The inspectors concluded that the test results were acce) table despite the inability to perform the test as written since the Jackleakage past the nitrogen injection check valve being measured was a relatively small amount (less than one gallon !
per minute (gpm)). This amount of water could be 3assed through the  '
partially open butterfly valve and prcub accepta)le test data.
 
. Once the problem was identified, the licensee took appropriate corrective actions. Specifically, the test procedure was revised to
>
instruct the test performer to throttle open XVB01035 EF. Also, test  .
>
unit personnel were counseled as to the expected actions to be taken  1
:  when a test 3rocedure could not be performed as written. The inspectors 4 considered t1ese actions to be adequate.
 
-
The failure to document and report a long standing test deficiency for the CST nitrogen injection check valve test is identified as a
. violation. This licensee identified violation is being treated as an i  NCV consistent with Section VII.B.1 of the NRC Enforcement Policy. This
:  is identified as NCV 50-395/97003-04.
 
. Conclusions    ,
An NCV was identified for the failure to document and report a deficiency for a CST nitrogen injection valve surveillance test which was performed nine times over a two and a half year period. The te.st i
procedure directed the test performer to open a butterfly valve which could not be fully opened due to interferenc III. Enaineerina El _ Conduct of Engineering E1.1 Seal-in Function of Hioh Hiah SG Level Turbine Trio (P-14) Inspection Scope (37551)
On the high high SG level actuation on April 26, the FW isolation signal closed only the A train FWIV. The inspectors reviewed the function of
 
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16  i the P 14 signal to determine if this had been the correct plant respons Observations and Findinas i
A review of the TS Bases following the event found that the automatic i functions of P-14 are not necessary for ESF system actuation but enhance the overall reliability of the ESF functions. The licensee's review of design information found that P 14 does not require a seal in functio It is required to prevent overfilling the SG and has no specific requirements if the SG level is below the P 14 setpoin The P 14 FW isolation function on April 26 closed only the A train FWI The licensee's review found that if the P-14 signal' actuates and resets before a FWIV has cleared its full o>en limit switch the FWIV.will remain open. The licensee also checced the full open limit switches on the FWIVs to verify no obvious misalignment or switch failure. There were no problems identified. The licensee is also reviewing from an operating standpoint if it is wise to not have all FWIVs close on a FW isolation signal. The inspectors had no further question j Conclusions    j The ins)ectors concluded that the plant had responded as designed when the higdhigh SG level had caused only one FW isolation valve to clos ,
El.2 SW Pumo Motor Failures Inspection Scope (37551)
The inspectors reviewed Engineering's resolution of two SW pump motor
  . failure , Observations and Findinas    l
      !
As described in Section M1.1 the C SW pump motor failed on April The l investigation of the motor failure found that the motor had physical damage to the stator coils. The motor was sent off site for repai Visual inspection at the vendor confirmed that there was physical damage to three coils off a line connection. This damage had created a carbon track to the frame of the motor which was detected by a ground fault relay. The testing at the vendor confirmed that the motor failure was caused by one or more turn to turn faults in the coils and that rewinding of the stator was required. Late in 1996, while in a vendor's repair facility for maintenance, the B SW pump motor had experienced a similar failur Based on the similar failures of the B and C SW pump motors, the licensee declared the A SW pump motor inoperable on May 2 until it could be sent off site for rewind of the coils. This licensee's decision was based on the fact that all three SW pump motors were the same age and design and have been exposed to the same environmental and operational
 
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condition It appeared to the licensee that the insulation used on the i '
j . coils had reached the end of its useful lif Conclusions The inspectors concluded that the licensee had made a )roactive decision by declaring the A SW pump motor inoperable based on t1e similar design of all three motors and on the similar failures of the B and C motor E1.3 Industrial Coolina Tower Modification Insoection Scope (37551. 71707)
The inspectors reviewed Engineering Information Request (EIR), " Review of Lifting and Rigging Operations Associated With ECR 50028. Industrial Cooling Tower Replacement, For Compliance With NUREG 0612 " Revision 1 dated March 26, 1997. The inspectors also attended the pre job brief prior to the initial crane lif Observations and Findinas The inspectors reviewed crane operations associated with the Industrial Cooling Tower Modification. Although the Industrial Cooling Towers are not safety related, the towers are located on the safety related Auxiliary Building roof. Replacement of the towers involved moving ;
equipment and components via crane onto and off the Auxiliary Building
      '
roof and in proximity to other safety related structures and component The EIR adequately addressed questions associated with crane operation The EIR demonstrated that with restrictions on load height and crane boom configuration the consequences of a dropped load or a failed crane boom would not place the facility in an unanalyzed conditio The inspectors concluded that the licensee had implemented actions to minimize potential damage to safety related equipment during the Industrial Cooling Tower replacement. Specifically, safe load paths were established for crane operations, safety related equipment and structures were identified to the crane operator and riggers, and precautions were identified to safely aerform the crane operations. In addition, the pre job brief prior to t1e initial crane lift adequately addressed potential safety concerns and precautions to address those concerns: discussed individual res)onsibilities for operations, maintenance, engineering, health p1ysics, security, and contract personnel: and established appropriate communication link Conclusions The inspectors concluded that the licensee had implemented actions to minimize potential damage to safety related equipment during the
; Industrial Cooling Tower replacement.
 
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18  ,
.E2 Engineering Support of Facilities and Equipment  !
E2.1 EFW Flow Balance
      , Inspection Scope (37551)
The inspectors reviewed the licensee's actions regarding a larger than expected differential flow to each of the three SGs following the manual reactor trip on April 2 Observations and Findinas    i Following the reactor trip on April 22, the A and B motor driven EFW pumps started automatically.due to a low low water level condition in i the A SG. It was noted on the EFW actuation that there appeared to be a !
large differential in EFW flows to each SG. Six minutes after the i reactor was tripped, the approximate EFW flow rate to each of the SGs '
was as follows:
Steam Generator A 380 gpm  i Steam Generator B 302 gpm  l Steam Generator C 263 gpm This flow rate data was taken when the SG pressures were essentially equal. These numbers give flow differentials of approximately 78 gpm between the A and B SGs, 39 gpm between the B and C SGs, and 117 gpm between the A and C SGs. These differential flow rates were higher than those observed during prior trips. The maximum differential flow rates observed during the last three trips were as follows:
August 25, 1989 56.2 gpm May 21, 1992 45.2 gpm January 12, 1993 56.9 gpm The licensee concluded that the change in the maximum differential flow rates observed during the April 22 trip was as a result of a physical change in the EFW syste Post trip testing and inspections determined that the motor driven EFW pump flow control valve mechanical stops for the A SG (IFV-3531) and the B SG (IFV 3541) were not within s)ecifications. The vendor drawings for 1 these valves specify that the meclanical stops should be adjusted to '
limit the valve travel to 2-9/16 inches. The as-found valve settings j were 31/4 inches on the A SG valve and 2 45/64 inches on the B SG i valve. The licensee wrote Non conformance Notice (NCN) 97-0366 and 97- l 0366A to address this condition. The valve mechanical stops were ;
adjusted to within specification. Subsequent testing indicated that the '
EFW flow differential was consistent with past test results. The !
inspectors agreed that the as left condition of the EFW system was j acceptable for continued plant operatio l
 
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A. Johnson, NRR R. Gibbs, RII P. Fillion, RII D. Jones, RII
The licensee initiated a root cause evaluation effort to determine the i cause of the out of specification valve stop adjustments and the
'W. Stansberry, RII 4 R. Aiello, RII PUBLIC NRC Resident Inspector U.S. Nuclear Regulatory Commission
-  consequences of the higher than ex)ected EFW flow differential ;
. Route 1, Box 64 Jenkinsville, SC 29065 i
Pending the inspectors review of t1e results of this evaluation, this {
*        l'
. will be considered an Unresolved Item (URI) 50-395/97003 0 j
OFFICE Rif:DRP RffiffCM SIGNATtRE # v'
      . Conclusions A URI w6s identified concerning the cause of an as found, out of specification condition on some EFW system flow control valves and the consequences of higher than expected EFW flow differentials.
   (
 
1'
The as-left condition of the EFW system was acceptable.
 
,
E7 Quality Assurance in Engineering Activities (37551)
E7.1 Review of Vodated Final Safety Analysis Reoort (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 i discussed in this report, the inspectors reviewed the applicable !
ortions of the FSAR that related to the areas inspected. No iscrepancies were identifie On April 15, licensing and engineering personnel arovided the inspectors )
with an overview of their FSAR Review Program. T1e inspectors verified that the licensee has established a method for tracking identified FSAR discrepancies from plant operating 3ractices and design document Discrepancies are assigned one of t1ree priorities. Priority A items require immediate evaluation and resolution. Review of Emergency Diesel !
Generator discrepancies identified no priority B and C items that should i have been classified as priority IV. Plant Suppor_t R1 Radiological Protection and Chemistry (RP&C) Controls l
R1.1 General Comments (71750)
The inspectors observed radiological controls during the conduct of I tours and observation of maintenance activities and found them to be acceptabl R1.2 Transoortation of Radioactive Materials
      ; Insoection Scope (86750. TI 2515/133)
The inspectors reviewed selected elements of the licensee's program for ;
transportation of radioactive materials to determine whether the ;
licensee properly processes, packages, stores, and ships radioactive ;
 
.
20  l materials and whether the changes to the Department of Transportation (D0T) and NRC regulations, which became effective on April 1,1996, had ,
been implemented. The review included records for training of personnel on the changes to the regulations, procedures for preparing radioactive material for shipment, and shipping papers for selected recent ,
shi>ments. Those procedures and records were evaluated for consistency :
wit 1 the recuirements delineated in 49 CFR Parts 170 - 179 and 10 CFR 71 !
for licensec material transported outside of the confines of the plan ]
b. Observations and Findinas The inspectors reviewed the training records for three individuals authorized to sign shipping papers and determined that training on the changes to the regulations had been 3rovided during March 1996, i.e.,
prior to the effective date of the c1anges. Refresher training was also provided during March 1997. The manuals for that training were also reviewed and found to have specifically addressed the new rules for the following topics: Low Specific Activity (LSA) and Surface Contaminated Object (SCO) hazards, definitions, and requirements: placarding, labeling, and marking of vehicles and packages: use of Systeme Internationale (SI) units on shipping papers, labels, and emergency i response instructions after April 1,1997: fissile classification; waste classification; and shipping papers. The inspectors reviewed the currently effective Health Physics Procedures (HPPs) 702, 703, 712, 716, 717, and 724, and determined that the instructions therein were consistent with applicable DOT and NRC requirements for selection of an acceptable container for various types of materials, fissile and LSA classifications, vehicle placarding, package marking and labeling, and contamination and radiation levels. The licensee also used computer l programs for guidance in preparing radioactive materials for shipment and for generating shipping papers. Those programs included libraries of A and A2 values, i.e., radionuclide activity levels used for
 
selection of proper shipping packages. The inspectors verified that the A 1 and A 2valuer, for 6 selected radionuclides listed in those libraries were accurate. The licensee's shipment log indiccted that, as of May 1, 1 the licensec had made 23 shipments of radioactive material this year, the majority of which were limited quantities in excepted packages of contaminated protective clothing shipped to a laundry and samples shipped for analysis. The inspectors reviewed the shi) ping papers for 3 recent shipments consisting of: an underwater vacuum slipped as a SCO: a cask of resin shipped to Chem Nuclear Systems' Consblidation Facility; and contaminated protective clothing shipped to a laundry. The information on the shipping papers was found to be consistent with applicable DOT requirements and the licensee's procedure ,
On April 30, 1997, the inspectors observed a package of contaminated protective clothing being loaded onto a truck for shipment. The inspectors noted that the package and the vehicle were properly surveyed, the package was pro]erly marked, and the appropriate information was recorded on tie shipping papers. The inspectors, accompanied by the licensee, toured interior and exterior storage areas used for temporary storage of packaged low-level radwaste awaiting
 
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I  shipment, radwaste awaiting further processing, or slightly contaminated i equipment (tools, scaffolding, etc.) held for reuse. The inspectors i noted that the containers were appropriately labeled. At the  ;
inspectors' request, the licensee performed dose rate surveys of several
'      '
of the containers. The observed dose rates were consistent with the dose rates recorded on the container's label .
L Conclusions l
t
      !
'
'
Based on the above reviews, it was concluded that the licensee had effectively implemented a program for transportation of radioactive ;
NAME LGarner alt DATE 07 / 1 / 97 07 / /_97 07 / / 97 07 / / 97 07 / / 97 07 / / 97
materials pursuant to DOT and NRC regulations.
. COPY 7 [ES) NO YES [ND) YES NO YES NO YES NO YES N0 l UttICIAL(Rt. CORD COPY DOCUMitti NMIL: G:\5UM9703.RE5 i
 
        :
l s
R2 Status of RP&C Facilities and Equipment  :
      !
R2.1 Comoliance with 10 CFR 70.24. Criticality Accident Reouirements  i Inspection Scope (71750)
      ,
The inspectors reviewed the licensee's facilities, equipment, and procedures to ascertain if they complied with the requirements of 10 CFR 70.24, criticality accident requirement '
      ! Observations and Findinas    l
      ;
The portions of 10 CFR 70.24 which are applicable to the V. C. Summer i station state that a licensee authorized to possess special nuclear J material in sufficient quantities is required to maintain a monitoring system in the area where this material is handled, used, or stored. The monitoring system shall be capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of 2 meters from the reacting material within one minute. Coverage of all areas shall be provided by two detectors. The detectors must energize clearly audible alarm signals if accidental criticality occur Also, the licensee must maintain emergency procedures for each of these i areas to ensure that all personnel withdraw to an area of safety upon !
the sounding of the alarm. These procedures must include the conduct of j drills to familiarize personnel with the evacuation plan and designate a i responsible individual to determine the cause of the alarm. In  '
addition, the licensee ,must have radiation survey instruments in accessible locations for use in an emergenc .
l The inspectors reviewed NUREG 0717. " Safety Evaluation Report (SER)  !
Related to the Operation of Virgil C. Summer Nuclear Station, Unit 1," i issued in February 1981. Section 12.3 of this document states:
The applicant has provided area radiation monitors around the fuel storage areas to meet the requirements of Section I
70.24 of 10 CFR Part 70 and to be consistent with the I
 
guidance of Regulatory Guide 8.12 " Criticality Accident Alarm Systems."
 
An initial search of licensee records failed to locate documents submitted to the NRC that supported this statement. As a result, the inspectors attempted to independently verify that the licensee was in compliance with the requirements of 10 CFR 70.24. The fuel har.dling bridge crane radiation monitor (RM G8) is located inside the Fuel Handling Building and provides a local audible evacuation alarm. The RM G8 radiation monitor range starts at 0.1 mR/hr. This range is sensitive enough to detect an inadvertent criticality anywhere in the Fuel Handling Building which would produce an amount of radiation described in 10 CFR 70.24. The Fuel Handling Building exhaust radiation monitor (RM A6) is located immediately outside the Fuel Handling Building in the Auxiliary Building. The RM A6 radiation monitor is attached to the ventilation duct associated with the Fuel Handling Building and would be capable of detecting airborne radiation which would be produced by activation of particles and gases during a criticality event. The licensee was unable to provide documentation that the RM A6 radiation monitor was capable of meeting the sensitivity requirements of 10 CFR 70.24. The RM A6 radiation monitor energizes an alarm in the main control room but provided no audible alarm in the Fuel Handling Building. Thus if a criticality event would occur, the control room operators would be required to respond to the alarm, evaluate the condition and then announce evacuation of the Fuel Handling Building.
 
One purpose of a criticality monitoring system is to allow rapid personnel evacuation from potentially lethal radiation dose rates.
 
Since the RM A6 radiation monitor did not provide an alarm inside the 1 Fuel Handling Building, the inspectors concluded that the RM A6 i radiation monitor did not meet the 10 CFR 70.24 requirement for l providing a clearly audible alarm signal if an accidental criticality were to occur. The inspectors verified that portable radiation monitoring instruments would be available to assess the radiation hazards during a criticality even The inspectors discussed with the Manager of Nuclear Licensing and Operating Experience the conflict between the results of this inspection and NUREG 0717. The discussion also addressed the unavailability of information to demonstrate that the RM A6 radiation monitor meets the sensitivity requirements of 10 CFR 70.24. The inspectors were informed that the licensee plans to request an exemption from 10 CFR 70.24.
 
The inspectors reviewed the licensee's procedures regarding radiation emergencies. Several procedures were in place which direct personnel on the appro)riate actions to be taken in the event of a radiation monitor alarm. T1ese procedures included Emergency Plan Procedure (EPP) 012, i
"0nsite Personnel Accountability and Evacuation," Revision 11: HPP 407,
" Controls for Receipt of New Fuel," Revision 3: A0P-123.3, " Potential Fuel Assembly Damage During Refueling," Revision 1: and Annunciator Res)onse Procedure (ARP)-019. " Fuel Handling Building Bridge Area RM G8 Hi lad," Revision Also, each employee and contractor on site receives station orientation training which includes instruction on
 
i
*      l
 
emergency evacuatio Health Physics personnel have been designated to determine the cause of the alarm and conduct surveys using portable radiation monitoring instruments which are maintained in an area which would be accessible in the event of an inadvertent criticalit The inspectors reviewed the status of drills to familiarize personnel on the evacuation procedures. On March 7, 1984, a drill was conducted 1 involving RM G8 and RM A6. The scenario involved the evacuation of !
personnel and was initiated by the. failure of a reactor building purge i line isolation valve. Actions taken during this drill were similar to those required for an inadvertent criticality event. The inspectors ,
considered this drill to meet the requirements of 10 CFR 70.24. A !
recent drill was conducted on March 12, 1997, after the licensee became i aware that the NRC was conducting inspections for compliance with 10 CFR- l 70.24 at other nuclear power plants. The inspectors considered the time I between these two drills (approximately 13 years) to be excessive. The ;
requirements of 10 CFR 70.24 state that the licensee shall maintain l emergency procedures to ensure that personnel withdraw to an area of safety upon the sounding of the alarm and to familiarize personnel with i the evacuation plan. The inspectors concluded that the emergency !
: procedures and personnel familiarization with the evacuation plan were ,
not being maintained because of the excessive time since appropriate !
drills were conducte '
The installed radia*.>on monitoring system and maintenance of emergency procedures (drills) o not meet the requirements of 10 CFR 70.2 Issues involving compliance with the requiremer.ts of 10 CFR 70.24 are currently being reviewed by the NRC. Until this review is completed, ,
this issue is identified as URI 50-395/97003 0 l
      <
c. Conclusions An URI was identified concerning the failure to meet the requirements of ;
10 CFR 70.24, criticality accident requirement S2 Status of Security Facilities and Equipment l
S2.1 Security Walkdown On April 30, the inspectors walked down modifications being made to the protected area fencing with the plant Security Manager. The inspectors verified that appropriate compensatory measures were implemented while the modification was in progress. The inspectors also toured the ;
Central Alarm Station and toured other plant areas to review security i measures. The inspectors identified no concern !
,
 
_ _ _ __ _ _ __ _ . _ _ _ _ _ _ _ , _ __ _.._ . .
24 Manaaement Meetinas X1 Exit Meeting Summary The inspectors ) resented the inspection results to members of licensee management at t1e conclusion of the inspection on May 12, May 20, and May 30, 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 PARTIAL LIST OF PERSONS CONTACTED Licensee F. Bacon, Manager, Chemistry Services L. Blue, Manager Health Physics and Radwaste S. Byrne, General Manager, Nuclear Plant Operations R. Clary, Manager. Quality Systems M. Fowlkes, Manager, Operations    4 S. Furstenberg, Manager, Maintenance Services  i D. Lavigne General Manager, Nuclear Support Services  i G. Moffatt, Manager, Design Engineering K. Nettles, General Manager, Strategic Planning and Development H. O'Quinn, Manager, Nuclear Protection Services  l A. Rice, Manager, Nuclear Licensing and Operating Experience 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 INSPECTION PROCEDURES USED l
      '
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 71707: Plant Operations    !
IP 71750: Plant Support Activities IP 84750: Radioactive Waste Treatment, and Effluent and Environmental Monitoring    i TI 2515/133: Implementation of Revised 49 CFR Parts 100-179 and 10 CFR Part 71
      !
j I
I
 
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. .
:
:    25 ITEMS OPENED AND CLOSED Opened 50 395/97003 01 VIO failure to establish procedures appropriate to the
 
circumstances (Section 01.3).
 
! 50 395/97003 02 IFI licensee corrective action to prevent deaerator
.
pressure transients (Section 02.2).
 
'
50 395/97003 03 VIO failure to follow procedure to raise RB pressure (Section 04.1).
: 50 396/97003-04 NCV failure to follow procedure to document and report a
;  long standing test deficiency (Sect. ion M2.2).
 
50 395/97003 05 URI evaluation of high emergency feedwater system i  differential flow rates (Section E2.1).
 
! ~50 395/97003-06 URI failure to meet the requirements of 10 CFR 70.24,
 
criticality accident requirements (Section R2.1)
Closed 50 395/97003 04 NCV failure to follow procedure to document and report a long standing test deficiency (Section M2.2).
 
)
}}
}}

Revision as of 12:38, 26 October 2020

Ack Receipt of 970702 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp 50-395/97-03
ML20149G902
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 07/11/1997
From: Belisle G
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To: Gabe Taylor
SOUTH CAROLINA ELECTRIC & GAS CO.
References
50-395-97-03, 50-395-97-3, NUDOCS 9707240119
Download: ML20149G902 (2)
v • d • e


Text

_ _ . _. .

,

.o July 11, 1997 South Carolina Electric & Gas Company ATTN: Mr. Gary Vice President, Nuclear Operations Virgil C. Summer Nuclear Station P. O. Box 88 Jenkinsville, SC 29065 l

SUBJECT: NRC INSPECTION REPORT NOS. 50 395/97-03

Dear Mr. Taylor:

i Thank you for your response of July 2,1997, to our Notice of Violation, i issued on June 2, 1997, concerning activities conducted at your V. C. Summer l facility. We have evaluated your response and found that it meets the '

requirements of 10 CFR 2.201. We will examine the implementation of your  ;

corrective. actions during future inspections. '

We appreciate your cooperation in this matter.

Sincerely, j

Original signed by George A. Belisle George A. Belisle Chief Reactor Projects Branch 5 Division of Reactor Projects Docket No. 50-395 License No. NPF 12 cc: R. J. White Nuclear Coordinator (Mail Code 802)

S.C. Public Service Authority Virgil C. Summer Nuclear Station P. O. Box 88 Jenkinsville, SC 29065 J. B. Knotts, Jr., Esq.

Winston & Strawn 1400 L Street, NW Washington,'D. C. 20005 3502 cc continued: See page 2 f O

9707240119 970711 hk,kkkk .

.

, PDR ADOCK 05000395 P PDR ftOf

- . .. . - - - - . .-

,

SCE&G 2 cc: Continued Chairman Fairfield County Council P. O. Drawer 60 Winnsboro, SC 29180

Virgil R. Autry, Director Radioactive Waste Management

!

!

Bureau of Solid and Hazardous Waste Management S. C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 R. M. Fowlkes. Manager Operations (Mail Code 303)

South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station P. O. Box 88

Jenkinsville, SC 29065 April Rice, Manager

. Nuclear Licensing & Operating

, Ex wrience (Mail Code 830)

Sout1 Carolina Electric & Gas Company Virgil C. Summer Nuclear Station '

P. O. Box 88 4 Jenkinsv111e, SC 29065'

(o -

Distribution:

'

A. Johnson, NRR R. Gibbs, RII P. Fillion, RII D. Jones, RII

'W. Stansberry, RII 4 R. Aiello, RII PUBLIC NRC Resident Inspector U.S. Nuclear Regulatory Commission

. Route 1, Box 64 Jenkinsville, SC 29065 i

  • l'

OFFICE Rif:DRP RffiffCM SIGNATtRE # v'

(

'

NAME LGarner alt DATE 07 / 1 / 97 07 / /_97 07 / / 97 07 / / 97 07 / / 97 07 / / 97

. COPY 7 [ES) NO YES [ND) YES NO YES NO YES NO YES N0 l UttICIAL(Rt. CORD COPY DOCUMitti NMIL: G:\5UM9703.RE5 i

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