IR 05000395/1998004
| ML20236H016 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236G985 | List: |
| References | |
| 50-395-98-04, 50-395-98-4, NUDOCS 9807070048 | |
| Download: ML20236H016 (21) | |
Text
...
.
.
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No.:
50-395 License No..
,
Report No.:
50-395/98004
'
Licensee:
South Carolina Electric & Gas (SCE&G)
Facility:
V. C. Summer Nuclear Station Location:
P. O. Box 88 Jenkinsville, SC 29065 Dates:
April 5 - May 16,1998 Inspectors:
B. Bonser. Senior Resident Inspector M. King, Resident Inspector (In-Training)
P. Hopkins, Reactor Inspector (Sections 01.2. 08.1, 08.2.
and E8.2)
Approved by:
R. C. Haag, Chief Reactor Projects Branch 5 Division of Reactor Projects 9807070048 900615 PDR ADOCK 05000395 O
PDR Enclosure 2
,
_
. _ _ -
_ ___ - _-
-__-__________-_-__ - _ _
.
EXECUTIVE SUMMARY V. C. Summer Nuclear Station
'NRC Inspection Report No. 50-395/98-04 This integrated inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a six-week period of resident inspection: in addition, it includes the results of an
. announced inspection by a regional inspector.
Ocerations Communications during shift turnovers were clear with a free exchange of e
ideas.
Management was present for several of the turnovers (Section 01.2).
The licensee adequately compensated for a postulated accident scenario
.
involving a secondary system pipe break outside containment.
Control room operators were adequately prepared, by means of hardware changes, revisions to operating procedures and additional training to respond to the postulated accident within the assumed time (Section 01.3).
..
A pre-test briefing for,a feedwater flow control valve test was
' thorough, clear and detailed. Operators were aware of plant response
-
and specific parameters to monitor during the test-(Section 01.4).
e'
Annunciator response procedure actions were not adequately reviewed when a liquid radiation monitor alarmed and diverted reactor makeup water storage tank flow to a recycle holdu) tank, resulting in an unexpected high level.in the recycle holdup tanc. Based on previous observations of operator response to annunciators, this appeared to be an isclated occurrence (Section 01.4).
]
A detailed system walkdown of' the reactor building spray system found
.
that the system was able to perform its design functions for accident
-
conditions.
Valves were properly aligned per system drawings and
+
operating procedures.
Component labeling was correct (Section 02.1).
The knowledge level and performance of the auxiliary building upper
.
level-operator during routine rounds and minor evolutions were good.
The observed scope of the operator rounds was effective to ensure that potential equipment problems were identified (Section 04.1).
Observed licensed operator simulator requalification exam scenarios were
.
challenging and well conducted.
Examination criticues were thorough and provided a comprehensive assessment of crew and incividual competencies (Section 05.1).
'
Maintenance i
Observation of service water piping ultrasonic testing and maintenance i
e on a. reactor trip breaker identified no concerns. Good work practices and' techniques were observed. The technicians were knowledgeable of their tasks (Section M1.1).
L l
s
'
_. _. _ _ _ _ _ _ _. _.. _ _ _ _... _ - _ _ _ _ _ _
___.__
_
_
. - - _ _
- _ _ _ _ _ _ _
_
- - - _ _ _ _ - _ - - _ _ - _ _ - _ - -
.
.
In response to previous failures of the-A diesel generator governor, the
.
-licensee appropriately considered diesel generator testing' frequency and ~
established suitable goals to demonstrate diesel generator reliability (Section M1.2).
Observed battery testing was satisfactory and performed in accordance
.-
with surveillance requirements.
A weekly battery cell inspection due
= to a damaged post' seal, met the monitoring criteria specified by engineering (Section M1.3).
. A redew of Turbine-Driven Emergency Feedwater (TDEFW) pump surveillance
.-
procedure revisions and test results concluded that the revised
.~ surveillance was adequate to verify that the TDEFW pump would meet
1 design requirements. A review of previous testing results concluded that the pump had been meeting design requirements (Section M3.1).
_
"
A ' violation was identified involving' a failure to promptly. identify that
L-actuation logic testing was not' adequately testing the high-high steam generator level circuitry and for a failure to promptly' perform the test-
?
on each solid state protection system train (Section M8.1).
l l
Enaineerina
!
..
The observed. licensee response to a small service water. system leak l
' identified no items of concern, ' Good work' practices.. engineering.
-,
i L
evaluation and licensingLsupport were noted. The licensee's service
-
l:
- water system monitoring and control programs and tem)orary non-code-L
- repair performance were found to be in agreement wit 1 subject Generic-Letters (89-13 and 90-05) (Section E1.1).
During a )rocedure review the licensee identified that the Final Safety L
e Analysis Report (FSAR)' assumed the use of one letdown orifice during
'
. normal-plant operation'while two orifices were used during normal plant operations. A non-cited violation was identified for a failure to
.. establish adequate operating )rocedures to ensure that the plant was operated in accordance with t1e assumptions in the FSAR (Section E1.2).
I-A'non-cited violation was identified for a failure to prepare an
- -
,
p adequate safety evaluation. The safety evaluation for a test did not adequately assess the effects of increasing turbine first stage pressure
'
on the high steam flow setpoint (Section E8.1).
!
i
i l
l
!
..
.
-_ - _ _ _ - _ _ - _ -____ - _
.
d Report Details u
- Summary'of Plant Status-The. unit began the inspection period returning to full power following a power
. reduction to 87-percent for main steam safety valve testing. The unit reached
~
~ full power on the morning of April 5 and remained at full power for the remainder of the inspection period.
I.
Doerations M
01.
~ Conduct of Operations'
~
.
01 1 General Comments (71707)
,
l 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 below.
01.2 Review of Routine Control Room Activities a.
-Insoection Scoce-(71707)
'
The inspectors observed daily' shift turnovers and reviewed shift logs-and control room recorders.
.
b.
Observations and'Findinas During morning and: evening operational shift; turnovers, the ins)ectors l
observed that offgoing shift personnel' had prepared an outline 3riefing sheet with the shift activities ~ performed'along with any required follow-up activities. These items were discussed during the on-coming
,
shift briefing where representatives of. Other departments.were made b
aware of required' support and.the plant operational status.
Communications in the shift briefing were clear with a free exchange of-ideas. On several occasions' upper level managers were present.
The: inspectors observed.that shift iogs and control room recorders were frequently reviewed by operations personnel. -Operations personnel made-l accurate plant data entries to the control room logs and to control room l
recorders in accordance with procedures, c.
Conclusions o(
c i
Communications during shift turnovers were clear with a free exchange of
,
ideas,. Management was present for several of the turnovers.
,
E 01;3 ' Isolation Of Emeraency Fepdwater m
- a.
Insoection Scoce (71707)
j
)The inspectors reviewed the> licensee's preparations for isolation of steam to the Turbine Driven Emergency Feedwater (TDEFW) pump from the
'
-control room.
-
h~
- _-__- - -- -- _ - _ _ _ - __ _ _ _ _ _ _ _ _-_ - __ - - ___ - _ - -
.
.
- b. Observations'and Findinas The licensee' identified'a )ostulated accident' scenario involving a secondary' system pipe breac outside containment where Emergency Feedwater (EFW) flow from the TDEFW pump to a faulted Steam Generator-(SG) could not be. terminated from the control room-(see LER 395/98-04 and NRC Inspection Report No. 50-395/98003).. The-ability to-isolate EFW=
postulated accident scenario. The. licensee initially compensated for this' postulated accident by stationing an operator.at the Control Room Evacuation Panel-(CREP) such that EFW could be-isolated by securing the-TDEFW pumf from the CREP.
The licensee was'able to relieve the operator
"at the CREP by installing a jumper in a Main Control Board (MCB) panel, revising procedures:and training operators.
~ After the installation of a fuse in a spare fuse block, the jumper in
-_the.MCB panel ~provided the operators with the capability to isolate the steam supply to the TDEFW EFW flow to a. faulted SG. pump from the control room and thus isolate Additional actions that the licensee ~
performed as-part of the problem resolution included labeling the jumper the. spare fuse. and spare fuse block.
The inspectors reviewed Work Request (WR) 9807837 forthejumper installation and examined the installed jumper, spare fuse block.'and spare fuse. The inspectors also reviewed.the operations' procedural-change.andithe training operators received for the procedural revision and installation-of'the jumper. Operations Administrative Guideline.
(OAG)-103.4. "E0P/A0P User's Guide," Revision 2. was revised to include;
.
guidance on isolation of a faulted SG-from the control. room. A station order was also prepared to notify the o)erators of this change. The inspectors concluded that the licensee lad adequately addressed'the
-
isolation of steam to the TDEFW pump from the control room within the assumed time for the operators to identify and isolate EFW flow to a:
faulted SG.
c.
Conclusions-The licensee adequately compensated for a postulated accident scenario involving a secondary system pipe break outside containment-,
Control room operators were adequately prepared,-by means of hardware changes.
revisions to operating procedures and additional training, to respond to the postulated accident within the assumed time.
01 4 Plant Ooerations-g
- a.
Insoection Scoce (71707):
H The inspectors observed a shift briefing and the conduct.of a special test to reduce differential pressure across the Feedwater Control. Valves
~
L (FCVs). 'The inspectors'also reviewed a failure to' adequately respond to l
~a antrol-' room liquid radiation monitor annunciator.
l bW 1-
__
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ - -
_
l1
_ _ _ - _ _ _ _ - _ - - _ _ - _ _ _ - _ - _ _ _ _ - - _ _ _ - _ _ - - _ _ _ - _ - _ - - _ _
'
3~
L
' b. : Observations 'and Findinas -
p L
On May 12,'the inspectors observed an operations shift briefing for
-
- Preventive Test Procedure. PTP-125.016, "Feedwater Control Valve DP
._
-Reduction Test." Revision O'.
The purpose of the special test was to gather data to determine FCV response characteristics while reducing the differential pressure across the FCVs..Thf shift briefing covered the purpose and conduct of the test, test precautions and termination criteria, and applicable industry experience. During the test, which covered several days.'the inspectors observed plant response and operator awareness of plant parameters affected by the test. The
,
^
-inspectors found that operators were aware of plant response and l
specific parameters to monitor.
No concerns were -identified.
At 12:40 a.m. on May:13. the; control room received an unexpected high.
level alarm on Recycle Holdup Tank (RHT) #1. The inspectors reviewed the cause of the' RHT high level with the licensee. Water flow from the Reactor Makeup Water Storage Tank (RMWST) that was being processed through demineralizers and returned to the RMWST had been diverted to-the-RHT several hours earlier on the previous day shift.
Flow was diverted when the boron recycle system liquid radiation monitor. RML-16 momentarily alarmed and diverted the RMWST flow to the RHT. A Control room operator failed to identify the flow diversion due to the annunciator response procedure actions not being reviewed adequately.
This' condition continued for several hours until the'RHT filled to the high level alarm setpoint. Based on previous observations'of operator
. response to CR annunciators the inspectors concluded that control room
<
annunciators' are normally responded to appropriately and that.this error appeared to be an isolated occurrence.
'
c.
Conclusions'-
A pre-test briefing for a feedwater flow control valve test was thorough, clear'and. detailed.
Operators were aware of plant response 5.a specific parameters to monitor during the test. Annunciator response. procedure actions were not adequately reviewed when a liquid radiation monitor alarmed and diverted reactor makeup water storage tank flow to'a' recycle holdup tank. resulting in an unexpected high level in the recycle holdup tank.
Based on previous observations of operator response to annunciators this appeared to be an isolated occurrence.
i
!
'02:
' Operational Status of Facilities and Equipment
-
'02.1Enaineered Safety Feature System Walkdown (71707)
a.
Insoection Scooe (71707)
l The inspectors-conducted a detailed system walkdown of the Reactor u
Building Spray'(RBS) System.
!
!
l
.
.
!
..
. _ _ _
_ - - _ _
.
.
b.
Observations and Findinas The inspectors conducted a detailed system walkdown of the RBS system to assess the general condition of system com)onents including labeling, to verify that system valve positions match tie systems drawings and station operating procedures and to assess plant housekeeping around system components.
No misaligned valves were identified and com)onent labeling was adequate.
The inspectors also reviewed the ap)lica31e sections of the Final Safety Analysis Report and reviewed tie system configuration for applicability of a potential concern identified at another plant involving the entry of sodium hydroxide tank nitrogen cover gas into the RBS. The inspectors found that the RBS system was able to perform its design function for accident conditions and no concerns were identified.
c.
Conclusions A detailed system walkdown of the reactor building spray system found that the system was able to perform its design functions for accident conditions.
Valves were properly aligned per system drawings and operating procedures.
Component labeling was adequate.
Operator Knowledge and Performance 04.1 Auxiliary Buildina Goerator Rounds a.
Insoection Scooe (71707)
The inspectors accompanied an Auxiliary Building (AB) upper level operator during the performance of a routine tour and TS required log taking and the performance of minor evolutions.
b.
Observations and Findings On May 5. the inspectors observed the routine activities of the AB upper level operator which included a complete tour of the assigned spaces in the auxiliary building and the recording of logs. The operator demonstrated a good level of knowledge and familiarity with his duties and responsibilities. The thoroughness of the tour and the level of detail demonstrated by the operator when examining plant equipment indicated that this ty)e of operator round would identify potential equipment problems.
T1e inspectors also observed the AB operator place a Chemical Volume and Control System (CVCS) cation resin bed in service and pump the steam generator blowdown monitor tank to the blowdown holdup tank. The operator was knowledgeable and used the applicable system operating procedures to perform these evolutions.
l
!
l
_
- _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
_ _ - - - - _
_ _ _ _ _ - - - - - _ _ _ _.
.
. _ _ _
.
.
c ',
Conclusions The knowledge level and performance of the auxiliary building upper level operator during; routine rounds and minor ' evolutions were good..
The observed scope of'the operator rounds was effective to ensure that potential equipment problems were identifMd.
I
_ 05 Operator Training and Qualification-05.1 Licensed Ooerator Reaua-ification Trainina Annual Simulator Examinations.
a.
In'soection Scooe (71707)
L The' inspectors observed portions of-three simulator examinations and the l!
critiques of each examination.
f b.- Observations and Findinos
~
' On. A ril 28.- the inspectors observed licensed operator requalification.
simu ator examinations. -The simulator scenarios were challenging and -
!
well conducted.
The use of an auxiliary operator in the simulator control' booth was-noted as a. positive' practice to help ensure realistic q
simulated field operator res)onses during the scenarios.
The
',
-examination critiques were tlorough and provided a' comprehensive
. assessment of individual and crew performance. All the concerns
,
!
-identified by the inspectors were also identified during the instructor-d examination critiques.
.
c; Conclusions-
'
J Observed licensed ' operator simulator requalification exam scenarios were -
f L
challenging and well conducted.
Examination criticues were tnorough and l<
provided'a comprehensive assessment of' crew and incividual competencies..
,
Miscellaneous Operations Issues (92901)
108.1 (Closed) Violation 50-395/97007-01: failure to comply with requirements
.of Technical -Specification 3.6.4. Containment Isolation. This violation was the result of a deficiency in STP-301.004. " Train A Containment Hydrogen Monitor Calibration." Revision 2.
The procedure did not'
'
contain adequate administrative controls to assure operability of the hydrogen monitor containment isolation valves during hydrogen monitor calibrations.
.
The-inspectors reviewed the procedures and verified that. procedures l-STP-301.004 and STP-301.005. " Train B Containment Hydrogen Monitor
!-
Calibration." were revised to reflect that calibrations will be performed during mode 5.. cold shutdown or mode 6. refueling when
'
Technical Specification 3.6.4 does not apply. The inspectors concluded
]
that the licensee's corrective. actions were adequate.
'
.
,
-_
_ - _ _ _ _ _ - _ _ _ _ _
-_
_ _ _ - _ - _ _
-
,
'08.2 (Closed) Violation' 50-395/96015-02: failure to establish containment integrity. prior to core alterations. The licensee stated that failure to comply with Technical Specification 4.9.4 Reactor Building Penetration. was a result of misinterpreting the definition of Core Alteration. This'resulted in performing Core Alterations before verifying _ that reactor building penetrations were in their:
closed / isolated positions.-
The licensee submitted a Technical Specification Change Request to the
~NRC on March 26 1997. The NRC approved the request to the current definition of core alteration as found in NUREG-1431 on September 17, 1997.
II. Maintenance M1 Conduct'of Maintenance M1.1' Observation 'of Work Activities
.a.
Insoection Scooe (62707)
The ins'pectors observed portions of maintenance and surveillance testing activities.
b.
Observations and Findinos j
0n April 28. the inspectors' observed a portion of WR 9807828.'" Perform
' Work Required By NCN-980369 On "B" Train Service Water Piping."
.The.
ultrasonic testing of a section of 20-inch B train service water piping was observed.
The testing was performed in response to a service water leak identified on a four-inch service water pipe (see Section E1.1).
i The.. work was performed in.accordance with Quality System Procedure.
(QSP)-516, "Ultraso6ic Thickness Determination," Revision 5.
The
,
technician performing the test followed the procedure and was i
knowledgeable of the task.
.
.
On May 5, the inspectors observed the performance of a portion of Surveillance Test Procedure (STP)-506.009, " Reactor Trip Breaker i
. Testing," Revision 14 (STTS 9802571).
The task involved testing and preventive maintenance on a spare trip breaker in the electrical
&
maintenance shop. The technicians were familiar with the operation of
!
the trip breaker mechanism, and were using the procedure and the
,
i technical manual. The inspectors observed the technicians clean,
)
'
lubricate and ins)ect the trip breaker in accordance with the technical manual..The breater internals appeared clean and in good condition.
c.
Conclusions Observation of service water piping ultrasonic testing and maintenance on'a reactor trip breaker identified no concerns. Good work practices and techniques were observed. The technicians were knowledgeable of their tasks.
l.
bi
_ - - _ - _ _ _ _ _ -
. _ _ _ _ _ - _ _ _ _ -
.
.
M1.2 Review of Diesel Generator Maintenance 2.
Insoection Scooe (62707)
The inspectors reviewed the licensee's implementation of goal setting and monitoring for the A Diesel Generator (DG) as required by paragraph (a)(1) of the Maintenance Rule, 10 CFR 50.65. The inspectors also reviewed implementation of TS Amendment No.139 which removed emergency diesel generator accelerated testing requirements.
b.
Observations and Findinas The A DG was recently placed in 10 CFR 50.65 (a)(1) status due to maintenance preventable functional failures on the A DG governor system.
.On March 30 a TS amendment was approved which removed DG accelerated testing requirements. The inspectors reviewed the licensee's implementation of the TS amendment with regards to testing frequency, and the goal setting and monitoring established for the A DG.
The inspectors reviewed the meeting minutes of the Maintenance Rule Expert Panel held on April 2.
The panel discussed continuing accelerated testing of the A DG and concluded that testing the A DG more frequently than monthly would not be appropriate since the A DG governor problem had been corrected and it had been acknowledged by the industry and the NRC that accelerated testing could be detrimental to DG reliability.
The. expert panel agreed on goals to verify DG frequency stability prior to exciter shutdown following DG surveillance testing, and to achieve satisfactory monthly surveillance tests with no failures associated with the governor over the next six months.
The inspectors concluded that the licensee had appropriately considered DG testing frequency and established appropriate goals to demonstrate the reliability of the A DG.
c.
Conclusions In response to previous failures of the A diesel generator governor, the licensee appropriately considered diesel generator testing frequency and established suitable goals to demonstrate diesel generator reliability.
M1.3 Battery Inspections a.
Insoection Scoce (61726)
The inspectors observed battery testing and inspection.
. b.
Observations and Findinas On May 4, the inspectors observed the performance of weekly A train battery surveillance test procedure, STP-501.001, Battery Weekly Test,"
Revision 9.
The inspectors observed the technicians perform the test and observed that the battery pilot cell met the acceptance criteria.
The inspectors also observed electricians perform WR. 9807795. " Cell #45 C
_ _ _
_ _.
_
___
_
_
_ _ _ _
. _ _ _
_
_ _
_ _
- _
__
- _ -___ ____ _ __
_ __ _
.
.
Has Evidence of Acid Leaking On Surface of Terminal and Possible Crack In Seal. Monitor Cell #45 Weekly and Replace Cell During RF11 Per NCN 98-0272.~ On March 23. during a routine visual battery inspection electricians had observed positive post seal corrosion. One of the two cracked. post seals had swollen to the point that the seal ring had positive The WR was to monitor battery cell #45.
The inspectors reviewed the licensee's disposition of Non-Conformance Notice (NCN) 98-0272 written to address the battery problem.
The NCN concluded that the current carrying pati would not be affected by the seal corrosion as long as the bolted connection and the integrity of the copper insert in the battery post were not compromised. The NCN disposition recommended, on a weekly basis, monitoring cell voltage and specific gravity, monitoring intercell connection resistances associated with the positive post and connectors, and checking for copper contamination and accumulation of electrolyte around the post seal area.
The inspectors observed technicians perform these recommended inspections on A train bactery cell #45 in accordance with EMSI-NCN-980272. " Weekly Monitoring of XBA1A Cell #45." The results of the inspection were satisfactory and no degradation was identified during the inspection.
c.
Dnclusions Observed battery testing was satisfactory and performed in accordance with surveillance requirements. A weekly battery cell inspection, due i
to a damaged post seal, met the monitoring criteria specified by
{
engineering.
M3 Maintenance Procedures and Documentation M3.1 Turbine Driven Emeraency Feedwater Pumo (TDEFW) Surveillance Test a.
Inspection Scope (61726)
The inspectors reviewed changes to a TDEFW pump surveillance test procedure.
b.
Observations and Findinas The inspectors reviewed changes to surveillance test procedure STP-220.002. " Turbine Driven Emergency Feedwater Pump and Valve Test."
Revision 2.
The licensee identified that the existing TDEFW pump TS surveillance requirement TS 4.7.1.2.a.2 was incorrect in that testing parameters were lower than the minimum aump design requirements.
This l
1ssue was documented in NRC Inspection leport No. 50-395/98-03.
As part of their corrective action the licensee changed STP-220.002 to ensure that the test acce)tance criteria would verify that the minimum design requireemts for t1e pump were satisfied.
The previous testing
<
had not verified that the pump would develop sufficient head to su) ply 380 gpm to 2 of 3 SGs with concurrent recirculation flow through tie
__
-
_
_ _ _ _ _ _
_
.
!
!
recirculation line.
Results from previous surveillance tests were reviewed. The results were found to meet the design criteria of the TDEFW pump after an engineering review. The inspectors reviewed previous test results, the revisions to the test procedure, and the results from the revised test aerformed on May 6.
The ins)ectors concluded that the TDEFW pump lad been operable prior to t1e testing change and that the revised surveillance was adequate to verify that the TDEFW pump would meet design requirements. The inspectors questioned the licensee on their plans to correct the TS surveillance requirements in TS 4.7.1.2.a.2.
The licensee stated that they plan to submit a TS change request to correct the surveillance requirements.
c.
Conclusion.s A review of Turbine Driven Emergency Feedwater (TDEFW) pump surveillance procedure revisions and test results concluded that the revised surveillance was adequate to verify that the TDEFW pump would meet design requirements. A review of previous testing results concluded that the pump had been meeting design requirements.
M8 Miscellaneous Maintenance Issues (92700, 92902)
M3.1 (Closed) Unresolved Item (URI) 50-395/98001-01: review solid state protection system TS operability and testing requirements. The inspectors verified that procedures STP-345.037. " Solid State Protection System Actuation Logic and Master Relay Test Train A." Revision 14 and STP-345.074. " Solid State Protection System Actuation Logic and Master Relay Test Train B." Revision 9, were revised.
Procedure STP-345.037 was revised on January 29 and performed on January 30.
Procedure STP-345.074 was revised on February 17 and performed on February 20.
The procedures were revised to verify that the parallel inputs for high-high SG level and SI were tested in the feedwater isolation circuitry.
Previous surveillance had not adequately tested each possible interlock logic state as required by TS.
The issue was first identified to the licensee in a Westinghouse Technical Bulletin dated December 30, 1997.
On January 23. the licensee initially considered these procedural changes as an enhancement to their SSPS testing and considered the surveillance tests to be adequate prior to making the changes in the test procedures.
The inspectors questioned the licensee's position on this issue based on the definition of Actuation Logic Test in the TS. The TS definition states that an Actuation Logic Test shall be the ap)lication of various simulated input combinations in conjunction with eac1 possible interlock logic state and verification of the required logic output.
The SSPS testing for the P-14. steam generator high-high level signal, was not accomplishing this test function before the procedures were revised.
On March 19. the licensee reevaluated their position on this issue and concluded that the SSPS testing had not tested each logic state to verify the required logic' output. On March 23. during a telephone conference with NRC l
_ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _.
,
-
.
.
...
l
~
staff. the licensee stated that they had reevaluated their position on this issue and understcod they had not been-meeting the TS surveillance requirement.
The inspectors concluded that the licensee failed to assure that a condition adverse:to quality was promptly identified'and corrected. The licensee.. failed to promptly identify that the actuation logic tests were notl adequately testing the high-high SG level circuitry as required by
,TS; The licensee's failure ~to recognize that the SSPS testing deficiency identified in the Westinghouse Technical Bulletin constituted a failure on their part to adequately perform a TS surveillance requirement influenced the subsequent corrective action. The inspectors
considered that the recognition and understanding of the significance of l
a potential problem are key parts of the identification process. The timeliness of the' licensee's actions to perform the revised tests on each.SSPS train to verify these circuits were operable was not commensurate with the expectations for an adequate TS surveillance. test.
While discussing this-issue, the, licensee stated that the timeliness of 1,
their actions were directed, in part, by TS requirements for the turbine-trip and feedwater isolation function based on high-high SG level. The
'
L licensee noted that TS Table 4.3-2. Functional Unit 5.b requires testing
!
of the automatic actuation : logic.
However.-corresponding Table 3.3 3 provides'no ACTION requirements for the automatic actuation logic should the surveillance test fail or the logic becomes inoperable for some
l other. reason. This issue is identified as Violation 50-395/98004-01.
,
failure to prorr,)tly identify and correct a condition adverse to quality.
m M8.2 -(Closed) Licensee Event Reoort (LER) 50-395/96-007: ITT Barton model 763 pressure transmitter-strain gage failures.
This LER. which included two.
I subsequent revisions. was issued as a Part 21 notification concerning the~ failure of ITT Barton model 763 pressure transmitters.
The issue L
involved soldering induced embrittlement of strain gage lead wires.
ITT
Barton stated they were initially concerned that instruments used for safety. related applications with excessive wire embrittlement could fail'
I when. subjected to high vibrational stresses experienced during nuclear l
power plant design. basis events.
ITT Barton subsequently concluded that I
the wires had sufficient strength. On June 13. 1997 ITT Barton l
reported that they were formally withdrawing the Part 21 notification l
and considered the issue closed.
L III.
Enaineerina El Conduct of Engineering
. E1.1-Review of Enaineerina Evaluation For Service Water Leak
'a.
Insoection SCoDe (37551)
The inspectors reviewed the licensee's engineering evaluation and actions in response to a small service water system leak, temporary non-code repair and associated ASME Code Class 3 pipe repair relief request
- submittal.
L
_ _ --_-_-- ____-__
. _ - -
.
L
]
-
i l
b.
Observations and Findinos On April 15. the licensee identified a through-wall pipe defect above a weld on a four-inch diameter Service Water (SW) system piping branch connection off an adjacent 20-inch diameter pipe. The four-inch pipe i
functions as a make-up source from the A train SW system to the Component Cooling Water (CCW) system surge tank.
Leakage was estimated at approximately five drops per minute. The SW piping is ASME code class 3 moderate energy piping. The licensee performed an operability evaluation and determined the appropriate repair method.
The inspectors reviewed Generic Letter (GL) 90-05 " Guidance for Performing Temporary Non-code Repair of ASME Code Class 1. 2 and 3 Piping." Generic Letter 90-05 included guidance on flaw characterization evaluation; augmented inspection guidance using Ultrasonic Testing (UT): and additional scope, limitations and specific considerations to allow a temporary non-code repair of ASME Code Class 3 moderate energy (SW design conditions are 65 psi and 95 degrees F)
piping. The generic letter stated that consideration of system interactions such as flooding water spraying on equipment. loss of flow, and design loading should be evaluated prior to performing temporary non-code repairs. The inspectors also reviewed additional
.
guidance the licensee obtained for stop-gap measures to limit leakage
from flawed piping while preparing a relief request.
This guidance was issued in a 1990 NRC Office of Nuclear Reactor Regulation (NRR)
Memorandum.
The memorandum allowed installation of a rubber gasket patch attached with hose clamps to limit leakage prior to the relief request for a temporary non-code repair being submitted for NRC approval.
The inspectors reviewed the licensee's engineering evaluation that included a review of five augmented ultrasonic tests performed on the most susceptible SW locations and the effects of leakage. spraying and potential flooding concerns.
The UT results indicated all locations inspected were acceptable based on meeting minimum wall thickness i
requirements. The SW leakage was found to be minor and did not impact system flow margins. There were no spraying or flooding concerns. The licensee's evaluation concluded that the service water system could meet its design basis requirements and was operable. The evaluation also concluded that basec on acceptable SW system structural integrity and the impracticality of performing a code repair within the SW 72-hour action statement (TS 3.7.4) a non-code repair relief request would be submitted.
The inspectors review of the SW evaluation concluded that the licensee's response'to the SW leak was conducted in accordance with current guidance and that the engineering evaluation adequately justified SW system operability and system integrity.
As a stop-gap leakage limiting measure the licensee installed a removable patch over the leak.
A Class 3 Pipe Repair Relief Request was submitted to the NRC on May 13.
The licensee plans to perform the ASME Code repair, pipe replacement.
during the next. scheduled Refueling Outage (RF-11) scheduled for Spring
- _ -
- _ _ _ _
_ _.
_
_ - - _ _ _ _ _ _ _ _
--
_ _ _ - -
_ _
-
-
l l
~
!
)
of 1999.
Overall good work practices and knowledge of engineering and I
licensing support requirements were demonstrated.
j i
Additionally. the recommendations of Generic letter 89-13 " Service Water System Problems Affecting Safety-Related Equipment" and the licensee's J
l commitments in response to Gl. 89-13 were reviewed and discussed with the j
SW system engineer. The SW system monitoring is being conducted in a systematic manner in accordance with Engineering Services (ES) 3rocedure l!-
ES-505. " Service Water System Corrosion Monitoring and Control 3rogram."
Revision 0.
An objective of the monitoring and control program is to j
allow detection of problems prior to development of through wall i
leakage. The system engineer indicated previous inspections had identified thinning and the need for replacement of some three-quarter inch piping. The four-inch piping ieak was the first larger size piping to be identified as needing replacement. The leak on the four-inch pipe is suspected to have been caused by Microbiologically Induced Corrosion (MIC). The piping will be examined when it is replaced. The system engineer also indicated additional inspection points would be added to
!
the long-term monitoring program as a result of this leak discovery.
Pending completion of the permanent code re) air the licensee plans to qualitatively assess at least once a week tie non-code repair patch and base piping for signs of degradation. At least once every three months the patch will be removed and the affected pipe evaluated for structural integrity. An engineering evaluation of the results will be performed
'
to estimate the degradation and prescribe remedial actions if necessary.
The inspectors concluded that licensee actions in response to a SW system piping leak were appropriate and that SW system operability and integrity were maintained.
T1e inspectors also concluded that the licensee's SW system monitoring program was sound and was systematically monitoring SW system _ piping corrosion.
c.
Conclusions j
The licensee's actions in response to a small service water system leak were appropriate.
Service water system operability and integrity were maintained. Good work practices, engineering evaluation and licensing i
support were observed. The service water system monitoring and control programs and temporary non-code repair performance were found to be in
)
agreement with commitments to the applicable Generic Letters.
'
,
E1.2 Chemical and Volume Control System (CVCS) Letdown Orifices a.
Insoection SCoDe (375s1)
The inspectors reviewed the use of CVCS letdown orifices during normal plant operation.
l l
'
_ _ _ _ _ _ - _ - _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _
_ -. _ _ _ _ _.
-
_ _ - - - _ _ - _
___-___-_--___- - - - -
.
..
b; Observations and Findinas
,
On' April 22. while 3erforming a 10 CFR 50.59 screening review for a
,
procedure change. tie licensee _ identified that the )lant was not being op'erated:in accordance with the oescri) tion in the :SAR.
The CVCS
' letdown orifice description in the.FSAR (Section 9.3.4.2.5.24) stated that one orifice-is designed _for normal letdown flow with the other two
. orifices serving /as standby.
During actual normal plant operation the -
- CVCS letdown system was operated with two orifices in service. A 60 gpm and a 45 gpm orifice were normally used. When this issue was identified the licensee reduced the operating letdown orifices to one 60 gpm-
' ori fice. A station order-was also prepared to explain to operators why
.this had occurred.
'
A review of the FSAR accident analyses also identified that operating with more than one letdown orifice had placed the plant in a condition
-
that was not analyzed in the FSAR accident analyses. The FSAR-(Section 15.3.7) accident: analysis for a break in lines from the reactor coolant pressure boundary that' penetrate containment stated that the'most severe pipe rupture, with regard to release of radioactivity during normal
- plant' operation, would be complete severance of the three-inch CVCS letdown line just outside containment upstream of the outer containment isolation valve.. The FSAR stated that complete severance of the letdown line would result in loss of reactor coolant'at the rate of 100 gpm This flow rate bounds the maximum flow rate ~ that would occur for a break
'through one 60 gpm orifice to atmospheric pressure. -For the 100 gpm flow rate the radiological doses resulting from the accident were evaluated and:found to be well within the 10 CFR 100 limits.
The guidelines for:the radiological consequences of the failure of small.
lines _ carrying primary coolant outside containment are in NUREG-800.
Standard Review Plan (SRP). Section 15.6.2.
The NUREG states.that the radiological consequences -are acceptable if the calculated whole-body and thyroid doses at the exclusion area and the. low po)ulation zone
- outer. boundaries-do not exceed 10 percent of the-10 CF1 100 dose criteria of 25 rem for.the whole-body and 300 rem for the thyroid. The licensee's review for the activity released and the subsequent offsite.
.
dose for two letdown orifices in operation remained well within the SRP
' acceptance criteria.- The inspectors review of this issue concluded that
,
.
-although the radiological consequences would have increased for a line break outside containment with two letdown orifices in operation the licensee would not.have exceeded 10 CFR 100 limits. The licensee has
,
. initiated an engineering evaluation to allow the use of two letdown
'
- ori fices.
n The inspectors also concluded that the licensee had not established
adequate operating procedures to ensure that the plant was operated in l
accordance with the assum)tions in the FSAR. The CVCS system operating procedure did not limit t1e number of letdown orifices to one during normal plant operation. This non-repetitive, licensee identified and t
'
<
'
'
__L
_ _ _. _. - _ _ _ _ _ _. _ _. _ _ _ _. _ _ _ _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _. _.. _ _ _ _ -... _
_ _ _.. _
_ _ _ _ _ _
_ _ _
__
_ _.. _ _ _ _ _ _ _.. _.
____
_ _ _ - - _ - - - - --
.
,
'
f corrected violation is being treated as-a Non-Cited Violation (NCV)
consistent with Section VII.B.1.of the NRC EnforcementiPolicy. This is identified as NCV 50-395/98004-02-
-
c.
Conclusions.
' Ailicensee brocedure review -identified that the Final Safety Analysis
- Report (FSA1): assumed the use of one. letdown orifice during normal plant-
_ operation while two orifices were used during normal plant operations.
s-A non-cited violation was identified for a failure to establish adecuate operating procedures to' ensure that.the plant was operated in accorcance with the assumptions.in,the FSAR..
<
'E8;
. Miscellaneous Engineering Issues (92903)
E8.1 (Closed) URI 50-395/98002-01:' assess safety significance of raising main
.'
'
turbine first stage pressure..The licensee reviewed the adequacy.of the high steam flow setpoint' for the plant conditions.that existed during
- the conduct of procedure PTP-230.001. "MSR Steam Flow Setup and
,
Verification,"; Revision 3.
The licensee also addressed the. potential.
effects that-otherllant. operations had on-the high stear flow setpoint during the test. Tle original safety evaluation for the test procedure-did not include a discussion of the effects of increasing turbine first
)
stage pressure on the.high steam flow setpoint.
.The inspectors reviewed the main steam isolation function.
Trip of the:
main steam line isolation valves on high steam line flow in coincidence
- with low-low RCS average temperature is one of the functions described i
.in-the FSAR (Section 15.4.2.1) that provides the necessary protection for a~ major-steam line rupture. The high steamline flow setpoints use
turbine first-stage pressure as a diverse indication of reactor power.
.
There 'are'normally expected variations in secondary plant operations that will affect: secondary plant characteristics including first stage
!
pressure. -The licensee establishes the high steam flow setpoint by measurement of nominal full power conditions and.then' application of the guidelines in TS-3.3.2; Table 3.3-4, 4.d.
The licensee s review of this L
issue concluded that this setpoint which is based on a measurement at L
nominal full power conditions provides an adequate margin of safety and
.there.is no reduction in this margin'of safety for normal plant operation, expected plant transients, a'nd off normal operations as
,
'
described in plant procedures. The inspectors reviewed this issue, held-discussions with the licensee and concluded that there had not been a reduction in the margin of safety provided by.this protective function.
The. inspectors reviewed the licensee's 10 CFR 50.59 screening review and safety evaluation for this test. The licensee performed a safety
,
. evaluation as required by'their program to justify that there was no unreviewed safety question. The safety evaluation, however, did not i
contain-an evaluation of the effects of raising turbine first stage
!
~ turbine pressure on the main steam-line isolation. protective function.
-
-
The safety evaluation did. include a discussion of other effects of
'
l
_w-
.
_ - _ - _ _ --_ - _
._
-.
.
.
raising first. stage pressure.
Included were the effects on the steam dump load rejection controller and.the rod control system.
The inspectors concluded that the safety evaluation had failed to meet the requirements of 10 CFR 50.59. " Changes, tests and experiments." in that'the licensee was conducting a test not described in the safety
'
analysis report and had not provided a satisfactory basis in their safety evaluation for the determination that the test did not involve an unreviewed safety question. The licensee's station administrative
)rocedure. SAP-107 "10 CFR 50.59 Unreviewed Safety Question Review 3rocess " Revision 1.--includes guidance on performing safety evaluations. The procedure states that explanations in response to the safety evaluation questions shall-be complete and provide justification for the answers.
One.of the areas the procedure includes for consideration is what system and com)onents are affected by the proposed change and what is the function of tiose systems affected. The licensee failed to establish in the safety evaluation that changing turbine first stage pressure would not affect the main steam isolation protective
. function. The subsequent review by the licensee after the test was terminated concluded there would be no reduction in the margin of safety
. provided by.this protective function.
The licensee has recognized that their 10 CFR 50.59 program needs strengthening and has initiated several enhancements to their program.
,
These enhancements include revision of SAP-107, completion of site specific procedure training for 10 CFR 50.59, and completion of computer based refresher training which covers basic philosophy and methodology of 10 CFR 50.59.
'
The safety evaluation for raising turbine first stage 3ressure failed to establish a satisfactory basis for the determination tlat a test did not involve an unreviewed safety question. This non-repetitive, licensee identified and corrected violation is being treated as a-Non-Cited i
Violation (NCV) consistent with Section.VII.B.1 of the NRC Enforcement
!
Policy. _This is identified as NCV 50-395/98004-03.
The inspectors also concluded that the licensee had exhibited a questioning attitude when a shift engineer raised the concern regarding the effect of changing first stage turbine pressure during the test.
,
E8.2 '(Closed) Violation 50-395/97011-04: Failure to install reactor makeup
,
line guard pipe in accordance with applicable drawing.
Corrective I-actions included properly installing the guard / spray shield, correcting i
plant as-built drawings and initiating an FSAR amendment.
]
~
.The inspectors verified that the reactor makeup.line spray shield had been properly aligned and welded into position as designated in the
.
, plant drawings'. The as-built drawings were revised to reflect the i
changes and provide proper detail and amplification for the reactor
'
makeup. water system. spray shields. The inspectors also verified that i-
-_ -
- ------ -- _ _
_
I (
the FSAR Revision Notice 97-0103 was prepared for inclusion into the j
next FSAR amendment to clarify the use of guard pipes.
The inspectors l
verified that the licensee's corrective actions were implemented.
IV.
Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 General Comments (71750)
The inspectors observed radiological controls during the conduct of tours and observation of maintenance activities and found them to be acceptable.
S1 Conduct of Security and Safeguards Activities S1.1 General Comments (71750)
The inspectors observed security activities during the conduct of plant tours and plant activities and found them to be acceptable.
V.
Manaaement Meetinas X1 Exit Meeting Summary The inspectors 3 resented the inspection results to members of licensee management at t1e conclusion of the inspection on May 22. 1998.
The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED Licensee F. Bacon Manager Chemistry Services L. Blue. Manager. Health Physics S. Byrne, General Manager. Nuclear Plant Operations R. Clary. Manager. Quality Systems M. Fowlkes. Manager. Operations S. Furstenberg, Manager Maintenance Services D. Lavigne. General Manager. Nuclear Support Services G. Moffatt. Manager. Design Engineering L. Hipp. Manager. Nuclear Protection Services 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
-
.e
-
...
17 -
INSPECTION PROCEDURES USED
--IP 37551i Onsite Engineering.
IP.61726: - Surveillance Observations IP 62707: ' Maintenance Observations
'IP.71707:
P1 ant Operations IP 71750:
Plant Support Activities IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor-facilities
IP 92901: -Followup _-'P1 ant Operations IP 92902: Followup - Maintenance
-
,
IP 92903: Followup. Engineering
' ITEMS OPENED AND CLOSED Ooened-50-395/98004-01'
VIO failure to promptly identify and correct-a -
. condition adverse to quality-(Section M8.1)
50-395/98004-02-NCV'
licensee-identified that the )lant was 'not being operated in accordance with tie descri) tion in Final Safety Analysis Report (Section El.2)
50-395/98004-03~
NCV.
failure to prepare an adequate safety evaluation (Section E8.1)
Closed 50-395/98001-01 URI review solid state _ protection system TS operability and testing requirements (Section M8.1)
'
.50-395/98002-01 URI-assess safety significance of raising turbine
~
first stage pressure (Section E8.1)
50-395/97007-01-VIO failure to comply with requirements of Technical l-Specification 3.6.4. Containment Isolation
!_
(Section 08.1)'
50-395/96015-02 VIO failure to establish containment integrity prior to core alterations (Section 08.2)
50-395/98004-02
- NCV-licensee identified that the )lant was not being operated in accordance with tie descri) tion in Final Safety Analysis-Report (Section E1.2)
50-395/98004-03 NCV'
failure to pr pare an adequate safety evaluation (Section E8.1
_
_-,,.__a_
_,__---------_----------a----------------a----
<-----1--
- - - - - - - - ^ - - '
- - -
" ' - -"
' ' ' - " '
.
.
50-395/96-007 LER Part 21 for ITT Barton model 763 pressure transmitters (Section M8.2)
50-395/97011-04 V10 failure to install reactor makeup line pipe guard (Section E8.2)
u_