ML20236C169
| ML20236C169 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 10/07/1987 |
| From: | Lesser M, Peebles T, Vandoorn P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236C148 | List: |
| References | |
| 50-413-87-30, 50-414-87-30, NUDOCS 8710270042 | |
| Download: ML20236C169 (16) | |
See also: IR 05000413/1987030
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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101 MARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.:
50-413/87-30 and 50-414/87-30
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Licensee: Duke Power Company
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422 South Church Street-
Charlotte, NC 28242
Docket Nos.:
50-413 and 50-414
License Nos.:
Facility Name:
Catawba 1 and 2
Inspection _Co'nducted: August 26 - September 25, 1987
Inspectors'
4/A
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/o/7N 7
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P.
K. Van 0oorn
Date Signed
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/O/7l? 7
M. S. Lesser
Date Signed
/v-7- D
Approved by:
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T..
A. Peebles, Section Chief
Date Signed
Projects. Branch 2
Division _of Reactor Projects
SUMMARY
Scope:
This routine, unannounced inspection was conducted on site inspecting
in the areas of review of plant operations; surveillance observation;
maintenance observation; review of licensee nonroutine event reports; Part 21
reports and followup of previously identified items.
Results:
Df the six (6) areas inspected, four apparent violations were
identified in two areas (Failure to Follow T.S. 2.2.1 in Determining Equation
2.2-1 Was Satisfied Following Nonconservative Reactor Trip Setpoint Adjustment
paragraph 3.d. , Unauthorized Isolation of an Auxiliary Feedwater Pressure
Switch Rendering the Auxiliary Feedwater System Unable to Function as Designed
Under Certain Conditions
paragraph
3.e.,
Failure to Establish Adequate
Measures to Periodically Calibrate All Safety Related Instruments
paragraph
3.e., and Failure to Ensure Procedures are Adhered to Concerning Locked Valves
paragraph 5.b.)
B71027004y g,39,4
{DR
ADOCK 05000411
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
J. W. Hampton, Station Manager
- H. B. Barron, Operations Superintendent
W. F. Beaver, Performance Engineer
W. H. Bradley, QA Surveillance
S. Brown, Reactor Engineer
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B. F. Caldwell, Station Seivices Superintendent
R. N. Casler, Operating Engineer
R. H. Charest, Station Chemistry Supervisor
- M. A. Cote, Licensing Specialist
T. E. Crawford, Integrated Scheduling Superintendent
W. P. Deal, Health Physics Supervisor
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- C. S. Gregory, I. & E. Support Engincar
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- C. L. Hartzell, Compliance Engineer
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J. Knuti, Operating Engineer
F. N. Mack, Project Services Engineer
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- W. W. McCollough, Mechanical Maintenance Supervisor
- D. S. Miller, QA Engineer
- G. S. Mitchell, Operations Production Specialist
C. E. Muse, Operating Engineer
F. P. Schiffley, II, Licensing Engineer
G. T. Smith, Maintenance Superintendent
J. Stackley, I. & E. Engineer
. .
D. Tower, Shift Operating Engineer
R. F. Wardell, Technical Services Superintendent
J. W. Willis, Senior QA Engineer, Operations
Other licensee employees contacted included technicians, operators,
mechanics, security force members, and office personnel.
- Attended exit interview.
2.
Exit Interview
The inspection scope and findings were summarized on September 25, 1987,
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with those persons indicated in paragraph 1 above.
The inspector
described the areas inspected and discussed in detail the inspection
findings.
No dissenting comments were received from the licensee.
The
licensee did not identify as proprietary any of the materials provided to
or reviewed by the inspectors during this inspection. The following new
items were identified.
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- Violation = 414/87-30-01:
Failure 'to Follow T.S. 2.2.1 in Determining >
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- Equation 2.2-1 Was Satisfied Following ' Nonconservative . Reactor Tiip;
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Setpoint Adjustment.
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. Violation. 413/87-30-01: ' Unauthorized Isolation of an Auxiliary Feedwater-
Pressure Switch Rendering .the Auxiliary Feedwater. System' Unable to
Function as Designed Under Certain. Conditions.
Violation '413, 414/87.-30-02:
Failure to: Establish' Adequate Measures'.to-
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Periodically Calibrate 'All Safety Related Instruments.
Violation 413, ' 414/87.-30-03: Failure to Ensure Procedures are Adhered to
-Concerning. Locked' Valves.
' Inspector Followup Item 413/87-30-04:
Evaluation of Subcooling Nuisance
Al a rm ~.
In spector. . Foll owup .' It'em 413, 414/87-30-05:
Verify AFW ' Lineups Meet T.S.
Requireme_nts.
Inspector. Followup ~ Item 413, 414/87-30-06:
Review of Passive Safety '
Structures 1n Containment (LtR 413/87-05).
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Inspector Followup . Item 414/87-30-04:
Model D-5 Steam Generator Level
Control Upgrade (LER 414/86-19).
Inspector? Followup Item 413, 414/87-30-07:
Upgrade of CMD Personnel
Maintenance Training (LER 414/86-25).
Inspector Followup Item 413, 414/87-30-08:
Design Review to Assure that
Drawings for ' Solenoid . Valves Correspond with Manufacturer Model No. and
Part-Designations (LER 414/86-45).
Inspector Followup . Item 413, 414/87-30-09: Review of Repetitive Failures
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. of Borg-Warner Air Actuators (LER 414/86-46).
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Inspector Followup Item 413, 414/87-30-10: Review of Cause for CA Valves-
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Sticking Closed (LER 414/87-19)
Licensee Identified Violation 413/87-30-11:
Missed Surveillance of IB
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Diesel Generator.
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3.
Licensee Action on Previous Enforcement Matters (92702)
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(OPEN)
Unresolved Item 413/87-08-03:
Adequacy of System Upgrade
Prog ram.'
The licensee documented review of this item in a Memo to
File from R. L. Medlin dated June 15, 1987. The licensee documented
that the program was generally adequate but
-orovements in the
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formal program were warranted.
This item remains open pending
implementation of the improvements.
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b.
(OPEN);_ Unresolved Item 413/87.10-01,3414/87-10-01:
Single Failure
Vulnerability of <the Nuclear: Service. Water System.
Further review
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of'Jthe previously1 described concern showed that use of human
' operators was viable.
However, the licensee - was susceptible to a
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single failure in the Nuclearf ServiceLWater (RN)' System under certain
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. postulated ' conditions.
NRC/RII
had
previously. reviewed
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- licensee's interim actions to. remove the singleLfailure. vulnerability
and agreed with these actions. However, further' review was necessary
Lto determine.if RN.had been degraded prior to the 1:censee'~s interim ~
actions and to determine the' technical significance of'this problem.
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Further-NRC/RII and NRC/NRR review of this problem.has confirmed-that
licensee interim actions were acceptable, however, l single failure
criteria' was violated if the' licensee :has had a Diesel' Generator (DG)
fout of service greater than '72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> on either unit with one unit
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in operation.
Technical Specifications .(TS) allow one DG to be
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- inoperable for greater.than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if a unit is shutdown (Mode 5 or
-6).,
The situation was previously described as follows:
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On January 21, 1987, Station Problem Report #CNPR-2530 was generated
by site personnel' identifying a postulated single failure of the RN
system and requested evaluation from Design Engineering (DE). Part
1 of-Design Study-(CNDS-080/00) of March 27, 1987, concluded that an
unanalyzed condition exists and needs to be addressed. The scenario
involves-a. diesel generator or RN pump on either unit. inoperable for
greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, requiring shutdown of that unit. Under LOCA
conditions of the operating unit or ' loss of Lake Wylie and station
blackout, RN. suction and discharge would automatically shif t from
Lake Wylie to the Standby Nuclear -Service Water Pond (SNSWP), the
seismically qualified ultimate heat sink.
A single failure of a
SNSWP supply valve would block cooling water to two RN pumps.
Since
.one RN pump was already inoperable, the fourth RN pump would alone
be left to supply cooling to both units. With one unit in a LOCA
condition and one unit shutting down, one RN. pump is inadequate to
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cool all heat exchangers including remaining operating diesel
generators.
The licensee interim action .was to realign RN to SNSWP
and failing the SNSWP supply valve open if a DG was going to be
inoperable more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The licensee has recently performed an analysis idlich demonstrates
the adequacy of one RN pump. However, the NRC has not yet reviewed
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this analysis.
The licensee further has identified that from
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August 22 - September 1,1986 a DG was out of service for greater
than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Therefore, an unreviewed safety question existed at
this time.
In addition, it appears that procedure changes are
necessary for the recirculation phase and additional flow balancing
may be necessary to utilize the one pump operation.
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10 CFR 50,. Appendix A,
Criterion 44 requires that a system to
transfer heat from structures,. systems and components important to
safety, to an ultimate heat sink shall be provided.
This criterion
further requires operation under normal and accident conditions and
suitable redundancy to assure the safety system function can be
accomplished under blackout conditions, assuming a single failure.
Therefore, it appears that this is a violation. The licensee had not'
submitted their analysis to NRC at the close of this inspection,
therefore, the technical significance of this issue is yet to be
determined. This item remains open pending licensee submittal of the
analysis and further NRC review.
c.
(CLOSED) Violation 413, 414/87-14-01: Failure of Written Operations
Procedures to Adequately Require Implementation of the Security Plan
Under Degraded SSF Conditions.
The licensee responded to this item
in correspondence dated July 9,
1987 and August 27, 1987.
The
inspector reviewed the corrective action taken and considers this
item closed.
d.
(CLOSED) Unresolved Item 414/87-14-02:
Nonconservative Calibration
of Power Range Nuclear Instruments (PRNI).
The licensee applied
TS 2.2.1, Reactor Trip Instrumentation Setpoints, to an event where
the power range nuclear instruments were non-conservatively adjusted
based on an erroneous Reactor Thermal Power Calculation.
The
licensee incorrectly applied equation 2.2-1 to determine if the trip
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setpoint exceeded the total allowable value.
In a memorandum dated
May 14, 1987 from M. W. Hawes, the licensee determined the factors in
equation 2.2-1 to be as follows: Z (Statistical summation of errors
assumed in analysis) =5.96,
R (Rack Error) =0.1,
S (Sensor Error)
=0.0.
In a memorandum dated May 19, 1987, R was revised to 0 4
thus equation 2.2-1 would yield Z + R + S = 5.96 + 0.5 +0.0 = 6.46.
This value is less that the Total Allowance (TA) of 7.5.
The
licensee therefore concluded that equation 2.2-1 was satisfied and
the threshold for a reportable event was not met.
NRC Inspection
Report 50-413/87-16 and 50-414/87-16 paragraph 8 documented Region
II and NRR disagreement with the licensee's conclusion.
The basis
for Z excludes any errors associated with the sensor and rack drif t.
TS 2.2.1 specifies that Z is the value from Column Z of Table 2.2-1
which is 4.56.
The sensor value S should be increased by the percent
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of span the system was in error (112.3 - 109) /1.2 = 2.75.
The terms
of equation 2.2-1 then become 4.56 + 0.5 + 2.75 = 7.31.
This is
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greater than TA of 7.5 thus indicating the PRNI trip setpoints were
outside the safety analysis. The licensee has stated they feel the
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event was within the scope of the safety analysis based upon a
Westinghouse analysis, or based upon credit taken for up to 2% of the
calibration error in the Z term.
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The licensee's application of Equation 2.2-1 to the. event, however,
was not in accordance with the TS, thus was an invalid method to
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' determine that Equation 2.2-1 was satisfied and the event was within
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the safety analysis.
This is identified as Violation 414/87-30-01:
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Failure to' Follow TS 2.2.1 in Determining Equation 2.2-1 -Was
Satisfied Following Nonconservative Reactor Trip Setpoint Adjustment.
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(CLOSED)
Unresolved Item 413/87-25-01:
Inoperable Motor Driven
Auxiliary _ Feed Pump Due to Isolated Pressure Transmitter. On July 6,
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1987 at 3:16 p.m. the Unit 1 reactor was manually tripped from 70% -
. power following a loss of one main feed pump and decreasing steam
generator' levels. This incident is reported in Licensee Event Report
(LER) 413/87-26. Low low water levels in 2 out of 4. steam generators
(S/G) caused an Auxiliary Feedwater System (CA) auto start of both
the motor driven CA (MDCA) pumps and the turbine driven CA (TDCA)
pump.
(Low low level in one S/G would result in only the motor
driven CA pumps to start.) All three pumps started, however ICA-46B,
Motor Driven CA Pump 18 Discharge Isolation to 1C S/G, inappropriately
closed shutting off flow.
Control room personnel noticed that
1CA-46B was closed approximately 10 minutes later and were
unsuccessful at opening it from the control room.
At 3:37 p.m.
ICA-46B was manually opened.
The licensee initiated a separate investigation to determine why
1CA-46B went closed and the results were reported in LER 413/87-27.
The licensee discovered the motor driven pump inoperability interlock ~
logic had actuated and caused 1CA-46B to close.
The logic i s.
designed to protect against a common mode failure where the TDCA
pump and the remaining MDCA pump (following a single failure of one
MDCA pump) would feed the same depressurized S/G.
Under accident
conditions the CA system functions as follows: MDCA pump 1A supplies
1A S/G via 1CA-62A and S/G 1B via 1CA-58A. MDCA pump 1B supplies 1C
S/G via 1CA-46B and 10 S/G via 1CA-428. The TCDA pump feeds S/G's 18
and 1C.
Normally closed valves are provided to cross connect the
MDCA pumps or to allow the TDCA pump to supply S/G's 1A and 10,
however this requires operator action.
Section 7.4.1.3.2, section
10.4.9 and 15.2.8 of Catawba Nuclear Station Final Safety Analysis
Report (FSAR) requires that the CA system be able to supply at least
2 intact steam generators under accident scenarios assuming a single
failure.
No operator action is assumed for 30 minutes.
The motor
driven pump inoperability interlock logic actuates as follows:
If
MDCA pump 18 and the TDCA pump are operating following an automatic
CA start signal, and MDCA pump 1A fails to start as sensed by
discharge pressure switch ICAPS 5131 with a 30 second time delay,
ICA-46B closes to prevent MDCA pump 18 and the TDCA pump from
simultaneously supplying the 1C S/G.
Similar logic exists to shut
ICA-58A in the event of MDCA pump 1B failure.
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On' July 6,1CA-46B closed because pressure switch ICAPS 5131 was
isolated and provided a false signal to the logic that MDCA pump 1A.
had failed. The-licensee's investigation to date has been unable to
determine how and when ICAPS 5131 became isolated.
Attemptin'g to
pinpoint an activity which potentially could have isolated the
pressure switch, the licensee was unable to locate any calibration
documentation on this instrument.
No other maintenance activities
documented by a work request have ever been performed on 1 CAPS 5131.
The licensee has not been able to demonstrate operability of this
logic since July 17, 1986 when a reactor trip occurred in which
all three CA pumps started and the logic responded properly
Hence MDCA pump 1B was potentially inoperable from
July 17, 1986 to July 7,1987 as it would have supplied only 1 of 2
required S/G's urder certain conditions.
The LER safety analysis appears inadequate in that it does not
address whether or not the CA system would have functioned as
designed under various accident scenarios during the time MDCA pump
1B was' inoperable.
The analysis does state that "two redundant CA-
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pumps were available throughout the event".
This is apparently
referring only to the trip event, not the time frame from July 17,
1986 to July 7, 1987 in which one of the two redundant CA pumps was
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inoperable on numerous occasions.
The inspector postulated an accident scenario very similar to the
bounding accident described in FSAR section 15.2.8, where the CA
system would apparently fail to function as designed.
The scenario
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is ' as follows:
Assume a feedline break depressurizes S/G 1A, all
three CA pumps start, MDCA pump 1A feeds the break, rendering it
useless for supplying 1B S/G.
30 seconds later ICA-46B closes
because ICAPS 5131 is valved out, (the motor driven pump inoperability
interlock logic is made up). This isolates that line to 1C S/G. The
postulated single failure is that the TDCA pump starts but fails
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anytime after 30 seconds. This leaves S/G 1D being fed by MDCA pump
18 and is less than the FSAR requirement that a minimum of 491
gallons per minute (gpm) is supplied to two intact S/G's.
The
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licensee questioned the postulated failure scenario of the TDCA pump,
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stating that they did not feel this was a credible failure, .i.e.
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failing at least 30 seconds af ter pump start.
defines single failure as "an occurrence which results in the loss of
capability of a component to perform its intended safety functions."
Therefore the postulated failure scenario is not exempted from being
a credible single failure and must be considered.
This conclusion
was reinforced by discussions with K. Jabbour and Bill Lefave of
NRC:NRR.
It is therefore concluded that the auxiliary feedwater
system would not have been able to perform its intended safety
function of supplying two intact steam generators under certain
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conditions.
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The -isolationL valve for pressure switch -ICAPS 5131 was found. closed, .
out of position, rendering the. switch inoperables and thus rendering .
the auxiliary feedwater system' inoperable under certain. conditions.
The ~ time period of the inoperability cannot be -determined but-
potentially could have existed from July 17,1986J to July 7, !1987.
' Catawba - Nuclear . Station Directive. 3.1.15 prohibits operation. of
components or.removalfof components from service that affect station
operation without knowledge and approval of. control room personnel.
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. Instrument 1 CAPS 5131 was removed from service without ' knowledge or.-
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approval- of control room personnel sometime .between July 17, 1986
and . July 6, -1987.
Therefore this is a violation of TS 6.8.1.
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- Violation 413/87-30-01: Unauthorized -Isolation of an Auxiliary
Feedwater Pressure Switch Rendering the ' Auxiliary Feedwater LSystem
Unable" to - Function as Designed Under Certain' Conditions. -This
' violation ' is being considered for escalated enforcement . and is
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subject to . modification based . on further review. and information to
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be discussed at the NRC's RII office and therefore' a Notice of
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Violation is not issued with this report.
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The . following . additional concerns have been raised as a result of .
further review by the inspector:
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ICAPS 5131 was installed in April 1984 under F-13A #6764. . On.
August 5,1985 change 3 to IP/1/A/3140/03B, Auxiliary Feed System
Safety Related Instruments for Motor Driven Pump 1A, added instrument
ICAPS 5131 for calibration.
On March 26, 1986 IAE forwarded the
. Standing Work. Request, 006091 SWR, to planning.which would calibrate
ICAPS 5131. along with the ICA5000 loop.
The calibration was
eventually attempted on March 10, 1987, however was not performed
because the. uni.t was in the process of changing modes and operations
would notJ authorize the work.
The instrument was discovered to be
isolated' on July 7,
1987.
A calibration check was performed to.
demonstrate operability of the instrument and found to be in
.-tolerance as documented under work request 5749PRF.
On August 19,1987 010120 SWR was created to separate ICAPS 5131 from-
the ICA5000 loop calibration as the potential existed to pl:ce both
trains of Auxiliary Feed inoperable if both loops were calibrated
under the same work request.
On August 28, 1987 1 CAPS 5131 was
calibrated.
It was found. to be out of tolerance and was adjusted.
(The required setpoint is 200 +/- 20 psi and the as found setpoint
was 140 psi .)
The licensee is investigating how the instrument
drifted so significantly in only five weeks.
The inspector was concerned that the instrument was not calibrated
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in over 3 years after installation.
10 CFR 50 Appendix B requires
periodic calibration of safety related instruments.
DPC Topical
Report, Quality Assurance Program Section 17.2.12 specifies that a
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program for control and calibration for measuring devices affecting
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the proper functioning of safety related components is provided. As
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implemented by Station. Directive 3.3.7, ' Work Request Preparation,
006091 SWR specified a calibration. frequency of 18 months.
The
licensee's program appeared to meet Appendix B requirements by
assigning 1 CAPS 5131 to an 18 month periodicity.
The licensee's
program appears weak in that it did not require certain safety
related instruments to be subjected to assigned periodicities until
the first calibration was done, in some cases beyond the established
periodicity and grace period.
In this case the first calibration
was not done until July 7, 1987. No engineering evaluation was done
to justify this.
The licensee has stated that there currently exists approximately 20
Standing Work Requests (SWR's) which calibrate safety related
equipment that have yet to be implemented on a periodic basis.
Up
until this point the licensee had been issuing the. SWR's as manpower
and operating schedules allowed.
The licensee was asked to submit
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a schedule for completion of the SWR's.
The proposed schedule
submitted . indicated that the instruments would be calibrated by
October 20, 1987. The inspector additionally pointed out to the
licensee that station directives do not provide calibration require-
ments. for safety related instruments which are not covered directly
by TS or used for monitoring TS parameters. The licensee stated they
would evaluate this for possible upgrading.
The failure to perform
periodic calibrations on these safety related instruments does not
meet the requirements of Criterion XII of Appendix B and constitutes
a Violation 413, 414/87-30-02:
Failure to Establish Adequate
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Measures to Periodically Calibrate All Safety Related Instruments.
The licensee missed another opportunity to discover the isolated
pressure switch in addition to not performing the calibration.
The
licensee performs instrument valve lineups after refuelings to verify
instrument operability, in this case IP/1/A/3820/17A, Instrument
Valve Start Up Checklist for Auxiliary Feedwater System.
Unit 1
completed refueling activities in November 1986 and IP/1/A/3820/17A
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was performed, however the isolation valve for ICAPS 5131 had never
been added to the procedure after installation.
The procedure was
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revised October 1986 in which the required two year procedural review
was taken credit for, however, the review was inadequate as it failed
to add the isolation valve. This was another opportunity to discover
the isolated instrument that was missed by the licensee.
The
licensee believes that their guidance for maintaining the checklists
was not clear and stated they intend to review all checklists for
completeness.
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The licensee's investigation into this incident is weak in several
respects. The answer to why/when the instrument became isolated has
never been determined. Corrective action to prevent reoccurrence
is incomplete in that calibration program weaknesses were not
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recognized.
The consequences and safety significance of the
inoperable pressure switch i.e.
inoperable CA system under certain
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- co' ditions was not -recognized by the -licensee. Further the' adequacy,
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of the two year review ofLthe Instrument Valve Startup Checklist was-
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not questioned by the investigation.. These ' apparent ' weaknesses . .
should be'. addressed in the responses to the two v.iolations identified
- in this. paragraph (3.e.) and your evaluation and corrective l actions
~ h'ould also be discussed in the meeting to be held in the NRC's RII
s
loffice.
iThree violations were identified as described in paragraphs 3.d. and 3.e.
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above.
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4.
Unresolved Item
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No unresolved items were identified as a result of this inspection.
5. ' Plant Operatio'ns Review (Units 1 & 2) (71707 and 71710)
a;
The inspectors reviewed. plant operations throughout the reporting-
period to verify conformance with regulatory requirements. Technical
' Specifications- (TS), and administrative controls. . Control room logs,
danger . tag logs, Technical Specification Action Item Log, and the
removal and restoration log were routinely reviewed. Shift turnovers
were observed to verify that they were conducted in accordance with
approved pro'cedures,
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The inspectors verified by observation and interviews, the measures
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taken to' assure physical protection of 'the facility met current
requirements. Areas inspected included the security organization,
the establishment and maintenances of gates, doors, and isolation-
zones in the proper condition, that access control and badging were
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proper and procedures followed.
In addition to the areas discussed above, the areas toured were
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observ'ed for fire prevention and protection activities.
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included such things as combustible material control, fire protection
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systems and materials,
and fire protection
associated with
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maintenance activities.
b.
The inspectors conducted a detailed walkdown of portions of the
Nuclear Service Water System for both units.
During this and other plant area tours, the inspectors observed
selected valves required to be locked in position by station
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procedures. Although no valves were identified to be out of their
required position, some were observed not to be. locked or improperly
locked. The following valves were observed to be unlocked: 2RN938,
2RNC54, 1NV240, and ICA34.
Procedures requiring these valves to be
locked are: OP/0/A/6400/06, OP/1/A/6200/01, and OP/1/A/6250/02.
At
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the time of discovery, other established procedures or administrative
controls had not authorized the valves to be in a condition other
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~than that' required by the above referenced procedures. In each case
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a chain and lock was present in the vicinity of- the valve ~, however,
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noLattempt: had; been' made : to lock the chain: around the handle. .The.
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licensee stated they would review locking controls for:all groups who
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'might manipulate' valves. The licensee ~ was previously. issued'a. Notice
of Violation on-June.13,'1986, Violation- 414/86-18-01,. for failure to.
e
lock valves.
In' revision 2~ of its response to thel violation dated
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March 13, 1987, Duke Power' Company ' outlined corrective action to' be
taken for improperly:lockedLvalves which included:the use ofilocking-
devices, handwheel modification,- and review of locking criteria. The
actions were in. tended to; correct problems with specific valves which-
,
were J particularly difficult to lock and would be completed by
December 15, 1987, on Unit 1 and March 2,1988, on Unit 2.
The
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inspectors ' reviewed Station Problem Report (SPR) CNPR02471 and
concluded that the observed valves ' did not . fall into the scope 'of.
long term corrective action:being performed under-the SPR since the
valves were unlocked. rather - than improperly ~ locked.
Therefore~
corrective action in the area _of locking valves has been ineffective:
and.' a violation of T.S. 6.8.1 is' identified. This is Violation 413,
414/-87-30-03:
- Fail ure to Ensure Procedures are Adhered to-
Concerning. Locked Valves.
The inspectors additionally observed 2 valves which were improperly
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locked (chain wrapped loosely around the handle), 2SM70 and 2SM72.
This problem was determined to be within the scope of the SPR's
corrective action and therefore these valves are -not included as
part of the violation,
c.
Unit 1 Summary-
-The unit started the reporting period in Mode 3 after shutting down
to adjust reactor coolant pump seal leak off' and started up on
August 26.
On September 9,
the unit entered
T..S. 3.0.3 at 1430 when it was
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discovered that both . trains of Nuclear Service Water (RN) were
"A" train had been declared inoperable the day before
for routine maintenance.
At 1430 on 9/10/87 an equipment operator
heard the RN pump 18 discharge strainer motor operating with no
corresponding rotation of the strainer. The strainer is of the self
,
cleaning type which rotates periodically on a high differential
pressure to backwash itself.
It was discovered that a shear pin,
which fastens the motor drive shaft to the strainer shaft, had
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broken.
This constituted inoperability of the "B" train RN system
as the strainer was unable to rotate. The 1B RN pump was operating
at the time, supplying adequate cooling to loads.
Unit 1 exited
,
T.S. 3.0.3 at 1520 when functional testing of
"A"
train was
completed.
The licensee discovered a similar broken shear pin in
the Unit 2 "2B" RN strainer, preventing its automatic rotation (it
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.could however be rotated . manually). . Although the two shear. piri
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~ failures . are' on ' opposite - units, the: RN - sys. tem 'i s . shared and
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l compensatory actions were established by the ~ licensee.
Unit 1
operated'at or about 100% for the remainder of the. period.
d.
Unit 2 Summary
The' unit'. started the-reporting period in Mode-5.after having shutdown
-to replace the 28 reactor coolant pump seal' and. entered Mode:1 on
September 1. . On September 3 the unit tripped from 21% power on Steam
Generator Low Low Level while swapping feed to the lower nozzles. On
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September 4 the unit again started up.
The unit was shutdown on
. September 12'.when.the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement ended for the Turbine
Driven Auxiliary Feed Pump inoperability. Discretionary Enforcement
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was granted by_NRC:RII'for an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to remain in Mode 3
to replace a bearing on the pump. The pump was repaired and the unit
started!up on September 14. .0n September 15 the . unit was' manually
L tripped from 48% when a' Main- Feed Regulating Valve ~ failed shut'.'
The
-unit was restarted the same day and operated at or about 100% for the
rest of the period,
e.
During an NRC Operator Examination visit on September 14-17, 1987,
several questions were raised by the inspectors. Two of these issues
warrant further followup by the Resident Inspectors.
The examiner
,
identified that a computer subcooling margin alarm annunciator was
constantly' alarming on Unit 1.
Further discussions with the licensee
indicated that a conservative set point has been established and
other.subcooling information is available to the operators, i.e..this
alarm .is apparently not required and is set too conservatively.
Since nuisance alarms are a distraction to operators, the licensee
was asked to correct this problem.
This is Inspector- Followup Item
413/87-30-04: Evaluation of Subcooling Nuisance Alarm.
The examiner also identified that Auxiliary Feedwater System (CA)
valve ICA-6, suction to the Condensate Storage Tank (CST), was
isolated and
questioned whether this met
T.S. 4.7.1.2.1.a.4
. requirements which requires all automatic valves in the CA flow path
to be verified open.
The licensee reasoned that the T.S.
only
applied to the discharge side of CA and the automatic function of
ICA-6 was to fail shut upon low level in the CST. The CST is one of
four suction sources to CA and is non-safety-related.
The Resident
Inspector ' asked if the T.S. was being met while operating CA pumps
to cool piping due to leaking check valves since discharge flow
control valves are throttled during this evolution.
The licensee
indicated that the T.S. is meant to apply to a standby readiness
condition for CA, not at all times and the valves were capable of
automatic operation at all times.
The licensee is issuing a T.S.
,
interpretation relative to ICA-6 and is processing a T.S. change
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relative to the other valves.
The licensee's reasoning appears
acceptable but further review is necessary of the interpretation and
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T.S. change.
This is Inspector Followup Item 413, 414/87-30-05:
Verify AFW Lineups Meet T.S. Requirements.
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Oni September .23, .1987, at 6:00 a.m. , the inspector noted a . security,
guard apparently ' dosing while on a compensatory assignment for. a
vital- areandoor. . The licensee has investigated the problem and -
appears.to be taking appropriate disciplinary-action.
One Violation was identified as described-in paragraph 5.b. above.
6.
Surveillance.0bservation(Units 1 & 2) (61726).
a.
During the inspection period, the inspector verified plant operations'
were 1n compliance with- various TS . requirements.
Typical of these
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requirements were' confirmation of compliance with the TS for reactor
coolant chemistry, . refueling water tank, emergency power systems,
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safety : injection, emergency ; safeguards systems,
control
room.
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ventilation,' and direct current electrical; power . sources.
The.
inspector verified tha't surveillance testing was performed in
accordance~with the. approved written procedures, test. instrumentation
was- calibrated,
limiting conditions for operation were. met,
appropriate removal and restoration of the affected . equipment was
y
. accomplished, test results met requirements and were_ reviewed by
personnel other than the individual directing the test, and that any
deficiencies identified during the testing were properly reviewed and
resolved by appropriate management personnel.
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b.
The .fol. lowing surveillance activities were either reviewed or
observed wholly or in part.
007866 SWR
Visual Inspection of FNA Fuses
PT/2/A/4350/03
Electrical Power Source Alignment Verification
IP/2/A/3240/11
Power Range Nuclear Instrument Calibration at
Power
No violations or deviations were identified.
7.
Maintenance Observations (Units 1& 2) (62703)
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a.
Station maintenance activities of selected systems and components
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were observed /reviawed to ascertain that they were conducted in
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accordance with requirements.
The inspector verifled licensee
conformance to the requirements in the following areas of inspection:
the activities were accomplished using approved procedures, and
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functional testing and/or calibrations were performed prior to
returning components or systems to service; quality control records
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were maintained; activities performed were accomplished by qualified
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personnel; and materials used were properly certified. Work requests
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were reviewed to determine status of outstanding jobs and to assure
that priority is assigned to safety-related equipment maintenance
which may effect system performance.
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.The following maintenance activities were either reviewed or observed
wholly or in, parte
,
-5749L PRF.
Inspect / Repair 1CA46B Spuriously ~ Closing
138013'0PS.
Inspect / Repair 2BB57B Failing to Close
-5769. PRF
Inspect / Repair.1S86 Failing to Open
,
NSM 10857
-VP' Test Hardware Installation.
-
38149 OPS
. Inspect / Repair Main Feed Flow Oscillations
25508 OPS:
Replacement of. valve-1RN47B
No violations or deviations were identified,
8.>
Review of Licensee Nonroutine Event-Reports-and Part 21-Reports-
(Units l'& 2)-(92700)
a.
- The below ' li sted Licensee Event L Reports (LER) were reviewed to
determine 'if the information provided met NRC ' requirements.
The
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determi nation. - included: adequacy of description, verification. of.
.
compliance with _ Technical Specifications and regulatory, requirements,
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corrective actionL taken . existence of potential . generic problems,
reporting: requirements satisfied, and ' the relative safety signifi-
cance.of each event, Additional inplant reviews and discussion with
-plant personnel, as appropriate, 'were conducted for. those' reports
indicated by an (*).
The following LERs are closed:
'413/86-48, Rv.1'
Three Containment Purge-System Isolations Due'to.
Procedure Deficiency
413/86-57, Rv 3-
Inadequate Valve Operator Torque Settings Due to
Manufacturer Deficiency
- 413/87-05, Rv.2
' Unit Shutdowns Due to Design Deficiency With
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Containment Air Return Fans
413/87-17, Rv.1
Inoperable Fi re Barrier Penetration Due to
Defective Procedure
- 413/87-27
Auxiliary Feedwater Pump Inoperable Due
to'
Instrumentation
Being
Unknowingly
Isolated
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(violation issued para. 3)
413/87-31
Unit
1
Vent Radiation Monitor Unknowingly
Inoperable Due to Inadequate Logging Policy
- 413/87-33
Failure to Verify Operability of Diesel Generator
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1B Due to a Personnel Error (LIV issued, para.
8.c.)
.
- *414/86-19, Rv.2
Main Feeawater Isolation Due to Overfeeding of a
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- 414/86-45, Rv.1
Turbine
Driven
Auxiliary
Pump
Inoperable From Standby Shutdown Facility Due to
Design Deficiency
'*414/86-46
Main Feedwater Isolation Due to Failure of Main
Feedwater Bypass Control Valve
- 414/87-19
Manual Reactor Trip Due to Feedwater Control
Valve. Process Card Failure
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- 414/87-20
Failure to Verify Operability of Offsite Power
Sources Due to Inadequate Administrative Controls
414/87-22
Main Feedwater Isolation and Auxiliary Feedwater
Auto Start Due to Debris in a Feedwater Control
Valve Positioner.
-b,
Long term corrective actions were identified in six of the LER's
a
listed above.
The LER's are being closed, however, in order to
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assure NRC followup of long term actions, the following Inspector
Followup Items (IFI's) are being established:
IFI 413, 414/87-30-06:
Review of Passive Safety Structures in
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Containment (LER 413/87-05)
IFI 414/87-30-04:
Model D-5 Steam Generator Level Control
Upgrade (LER 414/86-19)
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IFI 413, 414/87-30-07:
Upgrade
of
CMD
Personnel
Maintenance
Training (LER 414/86-25)
IFI 413, 414/87-30-08:
Design Review to Assure that Drawings
for Solenoid Valves Correspond with
Manufacturer Model No. and Part Designations
(LER 414-86/45)
IFI 413, 414/87-30-09:
Review of Repetitive Failures of Borg-Warner
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Air Actuators (LER 414-86-46)
IFI 413, 414/87-30-10:
Review of Cause for CA Valves Sticking
Closed (LER 414/87-19)
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c.
The licensee identified that it had not been testing Diesel Generator
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18 at the required frequency specified by T.S. 4.8.1.1.2 and reported
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missing eight T.S.
surveillance
in LER 413/87-33.
The violation
stemmed from numerous errors in accounting for all valid failures of
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the diesel generator within the last 100 tests.
During the period
between April 6,
1987 to August 3,
1987, the diesel generator
was not being verified operable at the required frequency, hence
was technically inoperable for those periods in excess of 1 week
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following a test.
At no time however did a valid failure occur
during this interval'.
The licensee's evaluation of this event was
<
complete and appropriate corrective action was taken. As permitted
by Appendix C of 10 CFR 2, no Notice of. Violation is proposed and
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this incident is classified as a Licensee Identified Violation (LIV
413/87-30-11): Missed Surve111ances of IB Diesel Generator.
d.
Review of Part 21 Reports (92700)
(CLOSED) P21-86-04 (Unit 1) Transamerica Delaval Time Delay Relay.
Installation Problems.
The licensee contacted the vendor who
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indicated that the . problem was not applicable to Catawba and a Part
21 report had not been issued.
Therefore, this item is closed.
One Licensee Identified Violation was identified as described in paragraph
8.c.
9.
Previously Identified Inspector Findings (92701)
(CLOSED)
Inspector Followup Item 413/86-51-03, 414/8'6-54-03: Review of
Evaluation of Auxiliary Building Radiation Monitor.
The licensee has
evaluated this issue and determined that the time of 3.75- minutes per
sample point could be reduced to 1.25 minutes without sacrificing
2 representative samples from each location.
This improvement allowing
more timely sampling has been implemented and, therefore, this item is
.
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closed.
No violations or' deviations were identified.
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