ML20235Y756
| ML20235Y756 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/03/1989 |
| From: | Seidle W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20235Y738 | List: |
| References | |
| 50-285-89-05, 50-285-89-5, GL-85-06, GL-85-6, IEB-88-004, IEB-88-4, NUDOCS 8903140551 | |
| Download: ML20235Y756 (11) | |
See also: IR 05000285/1989005
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION'
REGION IV
NRC Inspection Report: 50-285/89-05
Operating License: DPR-40
Docket: 50-285
Licensee:
Omaha Public Power District (OPPD)
1623 Harney Street
Omaha, Nebraska 68102
Facility Name:
Fort Calhoun Station (FCS)
Inspection At:
Inspection Conducted: January 30 through February 3,1989
Inspectors:
M. E. Murphy, Reactor Inspector, Test Programs Section, Division-
of Reactor Safety
D. R. Hunter, Senior Reactor Inspector, Operations Program Section.
Division of Reactor Safety
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C. E. Johnson, Reactor Inspector, Plant Systems Section,
Division of Reactor Safety
Accompanying
Personnel:
T. Stetka, Chief, Plant Systems Section, Division of Reactor
Safety (February 2-3,1989)
W. C. Seidle, Chief, Test Programs Section, Division of Reactor
Safety (February 2-3,1989)
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Approved:
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. C. Seidle, Chief. Test Programs Section J
Date
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ivision of Reactor Safety
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Inspection Summary
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Inspection Conducted January 30 through February 3, 1989 (Report 50-285/89-05)
Areas Inspected:
Routine, announced inspection of previously identified
inspection findings, implementation of the ATWS requirements - TI 2500/20
(10 CFR Part 50.62), and licensee event reports.
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Results: <Within the three areas inspected, one. violation was identified
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(inadequate design activity, paragraph 2.8).
Followup was conducted on 12 open
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items involving previous inspection findings; 9 items were, closed. The
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remaining three items require completion of the licensee's committed actions.
- The licensee's. actions in compliance with ATWS requirements contained in 10 CFR Part 50.62 were found satisfactory. Two LERs.were reviewed for adequate
corrective action and proper reporting and found satisfactory.
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The licensee's corrective actions regarding~ the cracks, identified in the
Limitorq)ue valve operator. gear housings (including the root cause and generic
actions ; initiation of fomal incident reports regarding the reactor coolant
system weld radiographs; and formal documentation of evalutions regarding the.
- dispositions for conditions adverse to quality were not as comprehensive as
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expected by the NRC inspectors.
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DETAILS
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Persons Con'tacted.
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- W. G. Gates, Plant Manager
- G. Peterson, Assistant Plant Manager
- J. H. McKinnon, Acting Division Manager, Nuclear Operations
- S. K. Gambhir, Division Manager, Production Engineering
'*R. L. Jaworski,~ Manager, Station Engineering
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- A. W. Richard, Manager. Quality Assurance and Quality Control
- S. Peterson, Manager, Mechanical Engineering
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- J. K. Gasper, Manager. Training
- L. T. Kusek, Manager, Safety Review Group
- F. C.-Scofield,. Manager, Nuclear Planning
- K..Holthaus, Manager, Nuclear Engineering
- M. Guinn, Supervisor, Reactor Physics
- D. J. Mathews, Supervisor, Station Licensing
- C. J. Simmons, Licensing Engineer
.S. Miller, Electrical Engineer
- R.-Lewis, Supervisor, Mechanical Engineering
D. O. Bye, Electrical System Engineer
- J. T. Smith,~ Assistant Manager, Security Services
- K. L. Henry, Lead Systems Engineer
- M. A. Ferdig, Nuclear Management Development
D. Hendry, Engineer, Special Services
C. N. Bloyd, Supervisor, Inservice Testing
NRC'
- P. H. Harrell, Senior Resident Inspector
The NRC inspectors also contacted other operations, technical, and
administrative personnel.
- Denotes personnel attending exit interview.
2.
Followup on Previously Identified Inspection Items (92701, 92702)
2.1 (Closed) Violation (285/8705-08): This violation dealt with the failure
to implement the appropriate procedure to test tie breakers; instead a
procedure for main breakers was used. Review of the licensee's response
indicated the following:
1) correct calibration procedure was obtained
and performed satisfactorily on the tie breakers, and 2) the preventive
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maintenance (PM) sheets associated with the main breakers and tie breakers
were revised to specify the calibration procedures to be used.
The response appears to be adequate.
This item is closed.
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2.2 (0 pen) Violation (285/PR32-01):
Failure to Submit Report on Inoperable
Fire Barrier - The licesee is revising Standing Order G-58 to address
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fire door matters, in particular, and fire system impairments. The
planned training and procedure issue are not complete.
This item remains open pending completion of the training and issue of the
revised procedure.
2.3 (Closed) Violation (285/8832-03):
Failure to Have Written Instructions
for Removal of Seismic Support - The licensee analyzed the auxiliary steam
line supported by AX5-10 in aTcordance with 10 CFR 50.59 to determine the
effect of removing the support. The analysis concluded that removal of
AXS-10 did not reduce the margin of safety. Maintenance Order 885358 was
issued to complete the removal of AXS-10. This work has been completed.
The plant manager issued Memorandum FC-260-88 to reemphasize the FCS
policy on verbatim compliance. Pemorandum FC-2188-88 on procedure use and
common sense was also issued by the plant manager.
This item is closed.
2.4 (Closed) Violation (285/8821-01):
Failure to Establish Procedural Controls
in Regard to Conditional Release of Nonconformance Items - The licensee
has completed a general programmatic upgrade of procurement and storage
practices for safety-related material and services. The NRC inspectors
reviewed the following revised procedures:
S.O. G-18, "Nonconformance
Control," Revision 12, now includes a separate dedicated section for
processing nonconformances for operational " Conditional Release Basis"
material; S.0. G-22, " Receiving, Shipping, Stores Control and Storage of
Critical Quality Elements, Radioactive Material Packaging, Fire Protection
Material, and Limited CQE Items," Revision 35, and QADP-13, " Control of
Nonconformance Items and Materials," Revision 6, now include provisions
for engineering review. This review also verified that these procedures
also cover the control of nonconforming material identified during receipt
inspection, provide administrative controls to limit the scope of
nonconformance reports, and provide clear authority as to who shall
authorize the release of the material for use.
This item is closed.
2.5 (0 pen) Violation (285/8529-11.H.2; Deficiency 285/8529/2.5-1):
Inadequate
Welding, Preparation, and Inspection Associated With the Replacement of
Valve No. MS-100 - The licensee has not completed addressing the generic
corrective actions addressed in NRC Inspection Reports 50-285/88-25 and
50-285/88-46.
This item is open.
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2.6 (Closed) Violation (285/8529-II.F.2.d & e:
Deficiency 285/8529/2.5-2):
Installation Procedure Did Not Contain Adequate Instructions Regarding
Tubing Configuration and Accumulator Tank Locations - The licensee
evaluated the installed tube spacing by alternate calculations. These
calculations are complete, verified, and were reviewed by the NRC
inspectors. These calculations confirmed that the installed tubing is
adequately and seismically supported.
A training program for the design groups in engineering was developed and
presented by the licensee. This program was intended to improve the
quality of instructions in the areas of routing and supporting seismic
instrument tubing, sizing and providing seismic support for air
accumulators, and preparing and checking calculations.
This item is closed.
2.7 (Closed) Unresolved Item (285/8825-02): This issue was a concern that
OPPD had no procedures in place that required a review of identified
inadequate or incorrect design base data for deportability. This item was
left as an unresolved issue pending development by the licensee of
procedures for review of design basis reconstitution (DBR) documents for
operability and deportability considerations. Review of Procedure OP-29,
" Evaluating, Reconstituting, and Closing Design Basis Document Open
Items," Revision 0, indicates that the licensee has developed a procedure
to address DBR documents for operability and deportability considerations.
This unresolved item is closed.
2.8 (Closed) Unresolved Item (285/8839-01): This item identified three
concerns regarding the adequacy of the design of Pipe Support SIS-8 and
any similar pipe support designs.
The NRC inspectors had discussions with the licensee on addressing the
adequacy of similar supports. The licensee stated that Commonwealth
Associates, Incorporated had performed their modification and review in
accordance with their approved quality assurance (QA) program and no
discrepancies, flaws, or damage to SIS-8 have occurred since the time it
was modified. The licensee also stated that Stone and Webster considered
SIS-97 (a similar design) to be adequate.
Therefore, the licensee did not
consider the review of other similar supports to be necessary.
The NRC inspectors reviewed Calculation FC 02923 for Pipe Support SIS-8.
As the result of this review, the NRC inspectors identified a deficiency
in which an incorrect dimension was used to calculate the maximum moment,
even though the final results were within acceptable limits. As the
result of this finding, the NRC inspectors expressed concern that there
are similar pipe support designs at FCS which may have similar dimensional
errors in the calculation and, therefore, may not be within acceptable
design limits.
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Failure to have adequate design control measures is contrary to the
requirements of Criterion III of Appendix B to 10 CFR 50 and is considered
to be an apparent violation. (285/8905-01)
For record purposes, this unresolved item (285/8839-01) is considered to
be closed.
2.9 (0 pen) Open Item (285/8836-03): Cracks in Limitorque Valve Operator Gear
Hot. sings - The licensee identified numerous cracks in the motor operator
gear housing of four of eight safety injection (SI) valves as a result of
inspections of the valves during the 1988 refueling outage. The licensee
replaced the eight SI valve Limitorque operators with an upgraded valve
operator housing, which also included a new, different type torque switch.
Also, the new motor operator spring packs contained a grease relief path
to ensure proper operation when grease reached the valve spring pack.
The SI valves had been inspected to a lesser degree during the 1985 and
1987 outages with no cracks noted. Discussions and records review revealed
that the eight SI valves had been operated historically at the high thrust
condition on the valve stem, disc, and seat. The high thrust was also
translated to the upper gear housing of the motor operators.
During the
1985 inspection, five of the eight SI valves had been discovered with
thrust values in the range of 20,000 psi. This value exceeded the maximum
target thrust value of 14,000 psi plus 10 percent. The licensee reset the
thrust values (torque switch setpoint) on the valves and commenced an
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evaluation of the as-found conditions.
Discussions and record reviews revealed that similar valves
(Limitorque SMB-00) were not operated routinely near the motor operator
thrust limits.
The valves have been inspected routinely to ensure proper
operation, including thrust value setpoints and lubrication.
The cracks in the motor operator gear housings resulted from the high
thrust conditions applied routinely to the eight SI valves. Additionally,
the older model Limitorque SMR-00 motor operator housings were
manufactured from a lesser ductile steel than the newer valves.
It is
speculated that the overstress conditions could have been caused by
operation of the valves near the limiting conditions, torque switches
which were difficult to adjust, valve grease migration to the spring pack
and hardening (prior to 1985), or valve grease migration to the spring
pack with no grease relief provided (after 1985).
The NRC inspectors have no further ouestions regarding this matter at this
time; however, the item remains open pending the completion of the
licensee's documentation of the final root cause analysis and reason for
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the valve operator gear housing cracks. Further, the evaluations,
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actions, and completion dates regarding the other SMB-00 valves need to be
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addressed and documented.
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2.10 (Closed) Open Item (285/8836-02):
System Operability Determination Because
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of Weld Problems - The NRC inspector reviewed the licensee's 10 CFR Part 21
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evaluation and OSAR 88-53 evaluation. The NRC inspectors discussed the
determination that no substantial safety hazard was associated with the
specific welds and therefore, the system operability was not placed in
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question. The licensee had performed additional ultrasonic tests on
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selected welds and reviewed weld package documentation. The NRC inspector
review of the licensee evaluation identified that the radiographs for
welds FF-3BR and F-3A on the pressurizer spray line (reactor coolant
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system) remained in question (film density and penetrometer position).
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Further, the NRC review revealed that the supporting documented evaluations
were not included in the quality records program. This was an example
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where an incident report (IR) was not written for the specific condition (s)
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adverse to quality and the supporting (s) documentation providing the detailed
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evaluation of the identified problem
could have been more extensive.
This item is further addressed in NRC Inspection Report 50-285/89-08.
This open item 1s closed.
2.11 (Closed) Open Item (285/8836-04):
Resolution of Safety Injection /
Containment Spray Pump Recirculation Flow Deficiency - The licensee
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identified a design deficiency associated with the common recirculation
header limited capacity for the eight safety injection and containment
spray pumps operating simultaneously. The pumps required a total of about
1105 gpm recirculation flow and the licensee's calculations indicated that
a maximum capacity of about 600 gpm flow through the common recirculation
line. The licensee's evaluation of the design deficiency determined that
pressure safety injection (HPSI)quate to allow flow from the three highpumps
the common recirculation was ade
low pressure safety injection (LPSI) pumps (about 400 gpm total) operating
simultaneously (500 gpm recirculation flow total). One er all of the
three containment spray (CS) pumps (about 600 gpm total recirculation
flow) operating simultaneously with the HPSI/LPSI pumps could degrade the
recirculation flows for the pumps. The licensee modified the electrical
circuitry for the CS pumps such that the pumps will be started with the CS
valves open to eliminate the " choked flow" recirculation condition for the
CS pumps. The licensee also performed a special test (SP-SI/CS-3,
" Simultaneous Operation of LPSI/HPSI Pumps in Minimum Recirculation Mode")
on November 23, 1988, to ensure that adequate recirculation flow was
available for the SI pumps and the pumps exhibited normal conditions. The
test was deemed by the licensee to be acceptable.
The NRC inspectors reviewed selected emergency and abnormal procedures
(EOP-1, " Reactor Trip," Revision 2; E0P-2, " Electrical Emergency,"
Revision 3; E0P-3, " Loss of Coolant Accident," Revision 5; E0P-5,
" Uncontrolled Heat Extraction," Revision 3; and A0P-23, " Reset of
Engineered Safeguards," Revision 12) to ensure that the HPSI/LPSI pumps
and the CS pumps would not be operated on recirculation simultaneously,
which could result in degraded flow to the HPSI, LPSI, and CS pumps.
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procedural controls appeared to be acceptable with one minor exception:
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Step C.3.d of A0P-23 required that the CS valves (HCV-344 and
HCV-345) be closed, and then Step C.3.e required that the CS pumps
were to be shutdown. This sequence would place the CS pump on
recirculation with the operating HPSI and LPSI pumps.
Discussions with the licensee revealed that the CS pumps would be secured
quickly after the CS valves were closed; however, the procedure did not
specifically note this requirement to prevent operating the CS pumps in
parallel with the HPSI and LPSI pumps. The procedural weakness was
discussed with the licensee so the matter could be evaluated by(the
licensee and the necessary changes to the procedure considered
e.g., shut
off the CS pumps and then close the CS valves), as appropriate.
The modification to the CS pump initiation logic and the performance of
the special test to verify adequate recirculation flows for the HPSI and
LPSI pumps appeared to be adequate to resolve the identified design
deficiency.
The NRC inspectors have no further questions regarding the
HPSI, LPSI, and CS pumps recirculation flow deficiency at this time.
This open item is closed.
NOTE:
The licensee event report (LER 88-32-01) will remain open pending
NRC review of the licensee response to this matter as required by
NRC Bulletin 88-04, dated May 5, 1988.
(See paragraph 3.1)
During the running of Special Test SP-SI/CS-3, however, it was noted by
the licensee that the HPSI and LPSI pumps were extremely noisy.
Discussions and document reviews revealed that the high noise level was
attributed to potentially ungrouted or inadequately grouted motor / pump bed
plates. The licensee determined that no operability problem existed.
This matter is being further reviewed by the NRC resident inspector.
2.1? (Closed) Open Item (285/8836-05): Containment Spray (CS) Pump Initiation
Logic Changes - The licensee identified a problem related to the
electrical loads and the sequencing of the loads onto the essential buses
when being supplied by the normal offsite power sources. The bus loading
and load sequencing would result in the initiation of the offsite power
low signal (OPLS) and cause the operating engineered safeguards
feature (ESF) equipment to be removed (" stripped") from the essential bus,
the starting of the associated emergency diesel generator (EDG), and the
sequencing of the ESF loads onto the associated essential bus for a second
time.
The licensee evaluation resulted in the necessity to load the CS pumps
onto the essential buses after the initiation of the containment spray
actuation signal (CSAS), (the combination of th? safety injection
actuation signal (SIAS) and a containment pressure high signal (CPHSI),
and the alteration of the load sequencing of certain equipment onto the
essential buses (IA3 and 1A4) to prevent a degraded bus voltage and the
initiation of a OPLS. The changes to the load sequencing included the CS
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pumps, two containment fan coolers (VA-7C and VA-7D), and the load groups
(three, five, and six) to address all design scenarios considering normal
offsite or onsite power and a single failure.
The NRC inspectors reviewed selected items including:
FC-71, " Operators Log," Revision 45, page 31 of 44, to ensure
operator actions were specified when a 4160V low voltage was
encountered. The low voltage (OPLS) setpoints for TIA3 (3845 volts)
and TIA4 (3743 volts) were given as operator action points to declare
the bus inoperable and referenced the associated Technical
Specification (2.7).
OP-10-A17-13, Revision 12. "4160 Volt Bus IA3 Low Voltage," set at
90 percent of nominal. The actual OPLS setpoint of the bus low
voltage was approximately 92 percent.. This item should have been
changed as part of the design change package and the procedural
discrepancy was brought to the attention of the licensee for action.
OP-10-A18-17 Revision 9, "4160 Volt Bus 1A4 Low Voltage," set at
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90 percent of nominal.
The modification to the CS pump initiation circuit and associated load
sequencing appears to be acceptable. The NRC inspectors have no further
questions regarding this matter at this time.
This open item is closed.
3.
Followup on Licensee Event Reports (LERs).(92700)
The following LERs were reviewed to verify the specified corrective
actions had been completed and to ensure the corrective actions were
effective in preventing recurrence.
3.1 (0 pen)LER 88-032-01:
Safety injection / Containment Recirculation Piping
Design Deficiency - The licensee identified a design deficiency associated
with the common recirculation header for the eight SI and CS pumps
operating simultaneously.
The NRC inspectors reviewed the change to the CS pumps initiation logic
and special tests performed by the licensee to ensure adeouate SI and CS
pump recirculation flows to prevent pump degradation.
(See
paragraphs 2.11 and 2.12)
This LER will remain open pending the NRC review of the response to this
matter as reouired by NRC Bulletin 88-04, dated May 5,1988.
3.2 (Closed) LER 88-033:
Electrical Distribution System Design Deficiency -
The licensee identified a problem related to the electrical loads and the
sequencing of the loads onto the essential buses when being supplied by
the normal offsite power sources.
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The NRC inspectors reviewed the licensee evaluation associated with the
changes to the electrical distribution system and selected procedure
changes to implement the changos.
(See paragraph 2.12)
This design deficiency was apparently discovered during the. ongoing review
of the plant original design and as-built conditions and reported to the
NRC.
The NRC inspectors have no further questions regarding this matter at this
time.
This LER is considered closed..
4.
Compliance With Anticipated Transients Without Scram (ATWS) Rule (25020)
The purpose of this inspection was to verify that the licensee has
implemented the ATWS rule, 10 CFR 50.62, and the quality assurance program
as stated in the Generic Letter (GL) 85-06, " Quality Assurance Guidance
for ATWS Equipment That is Not Safety-Related." This inspection also
included verification that the modifications made conform to the
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licensee's commitments as endorsed in the Nuclear Reactor Regulation (NRR)
Safety Evaluation Report (SER).
This inspection consisted of a review of the following documents:
Modification Construction Package MR-FC-87-048, " Diverse Scram
System (DSS) Testing"
Installation Procedure No. MR-FC-87-048, E-1, " Diverse Scram System
Testing," Revision 1, dated October 29, 1988
Installation Procedure No MR-FC-87-048, T-1, " Diverse Scram System
. Testing," Revision 1, dated December 27, 1988
Calibration Procedure (CP)-120A, " Pressurizer Pressure Input to the
DSO Ct.annel A," Revision 4, dated November 29, 1988
CP-1208, " Pressurizer Pressure Input to the DSS Channel B,"
Revision 4, dated November 29, 1988
CP-120C, " Pressurizer Pressure Input to the DSS Channel C,"
Revision 4, dated November 29, 1988
CP-120D, " Pressurizer Pressure Input to the DSS Channel 0,"
Revision 4, dated November 29, 1988
Operating Instruction (01)-05S-1, " DSS Normal Operations,"
Revision 0, dated January 22, 1989
Operating Procedure (OP)-1, " Master Checklist for Start-Up or Trip
Recovery," Revision 36, dated January 25, 1989
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OP-10,~ " Annunciator Response Procedure," Revision 3, dated
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CP-DSS-1, " DSS Channel and Logic Checks"(New Procedure in Approval-
Cycle)"
In accordance with the SER, FCS, Unit 1, is comitted to install a DSS, a-
diverse circuitry to. initiate la turbine trip (DTT), and diverse circuitry
for. initiation of auxiliary feedwater (DAFW).
The NRC irispectors verified that the licensee' has implemented thel design
as approved by.NRR through the SER dated December 7, 1988, by document
reviews and installation walk-downs. The document review also included
review'of' procurement packages and installations drawings for ATWS
equipment.
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The documentation reviewed showed that the ATWS system hasbeen designed
according to. the'SER. .There was one open item in the SER concerning
isolation = devices that.NRR stated would need further review.
NRR will
close'this issue.
No violations or deviations were identified.
5. -
Exit Interview
The NRC inspectors met with the licensee personnel denoted in paragraph 1
on February 3, 1989. The NRC inspectors summarized the findings of the
inspection and discussed the violation identified in paragraph 2.8..
The
licensee did not identify as proprietary any of the information provided
to, or reviewed by, the NRC inspectors.
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