ML20235R060
| ML20235R060 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 09/18/1987 |
| From: | Dangelo A, Miller L, Myers C, Pereira D, Perez G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20235R007 | List: |
| References | |
| 50-312-87-20, NUDOCS 8710070780 | |
| Download: ML20235R060 (34) | |
See also: IR 05000312/1987020
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'U. S. NUCLEAR REGULATORY COMMISSION
REGION V
, Report No: 50-312/87-20
' Docket'No; '50-312 s
. License No. DPR-54
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Licensee: -Sacramento' Municipal. Utility District
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P. 0. Box 15830
Sacramento, California -95813
Facility Name: Rancho Seco Unit 1
- Inspection at: Herald, California (Rancho Seco Site)
Inspection'conductep May 30 - July 10, 1987
Inspectors: / !
Mb'.D'AnMM
A.-J o, Senior #Esident Inspector Cate Signed
> > Ju ~ > - V~l8~D
C. J. My s, F esident Inspector Date Signe'd
f %- 9llC/V7
G e ez, hf d t Inspector Date Signed
ou &
Difereira, Regiorfal Unspector
7/mf7
Date Signed
Accompanying Personnel: G. Johnston, M. Ely, Lawrence Livermore National
La atory
Approved By: / /
IL/ F. Miller, Chief, Reactor ProjectsSection II D&te Signed
Summary:
Inspection between May 30 and July 10, 1987 (Report 50-312/87-20)
Areas Inspected: This routine inspection by three Resident Inspectors and
by two Regional Inspectors, involved the areas of operational safety
verification, maintenance, surveillance, and followup items. During this
inspection, Inspection Procedures 25573, 30702, 30703, 35701, 37700, 37701,
42700,.55050, 55150, 61726, 62702, 62703, 71707, 71710, 72701, 92700, 92701,
92702,' 93702 and 94702 were used.
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Results: In the areas inspected, two violations were identified: failure to
use an appropriate heat treating process procedure and failure to properly
install certain cables, and associated electrical equipment.
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DETAILS
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1. Persons Contacted
a. Licensee Personnel
G. C. Andognini, Chief Executive Officer, Nuclear
- J. Firlit, AGM, Nuclear Fower Production
- W. Bibb, Deputy ' Restart. Implementation Manager
G. Coward, AGM, Technical and Administrative Services
- D. Keuter, Director, Nuclear Operations and Maintenance
- J. McColligan, Director, Nuclear Plant Support
D. Brock, Acting Nuclear Maintenance Manager
.B. Croley, Nuclear Technical Manager
G. Cranston, Nuclear Engineering Manager
-*W. Kemper, Nuclear Operations Manager
J. Shetler, Director, System Review and Test Program
T. Tucker, Nuclear Operations Superintendent
J. Grimes, Nuclear Mechanical Maintenance. Superintendent
L. Fossom, Deputy Implementation Manager
R. Colombo, Regulatory Compliance Superintendent
J. Field,-Plant Support Engineering Manager
S. Crunk, Technical Assistant, AGM
F. Kellie, Radiation Protection Superintendent
J.,Vinquist, Quality Assurance Manager
C. Stephenson, Senior Regulatory Compliance Engineer
R. Cherba, Quality Engineering Supervisor 1
T. Shewski, Quality Engineer
L. Conklin, Technical Assistant, AGM
D. Ross, Security
Other licensee employees contacted included technicians, operators,
mechanics, security and office personnel.
- Attended the Exit Meeting on July 14, 1987
2. Operational Safety Verification
The inspectors reviewed control room operations, including access
control, staffing, observation of decay heat removal system alignment, ,
and review of control room logs. Discussions with the shift supervisors j
and operators indicated understanding by these personnel of the reasons l
for annunciator indications, abnormal plant conditions and maintenance
work in progress. The inspectors also verified, by observation of valve
and switch position indications, that emergency systems were properly 4
aligned for the cold shutdown condition of the facility.
During this period, the licensee still relied on the operability of the
steam generators and auxiliary feedwater system for decay heat removal.
Tours of the auxiliary, reactor, and turbine buildings, including
exterior areas, were made to assess equipment conditions and plant
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conditions. Also, the tours were made to assess the effectiveness of
a . radiological controls and adherence to regulatory requirements. The.
inspectors also. observed plant housekeeping and cleanliness, looked for
potential . fire and safety hazards, and observed security and safeguards
practices.
Plant housekeeping appeared degraded in that tours made of the auxiliary
butiding roof revealed some trash such as Styrofoam coffee cups and food
containers, newspaper and cigarette butts on the floor. By the end of
this inspection, plant management had responded to the conditions noted,
had cleaned the' area, and, in addition, were having water jet cleaning of
the protected area of.the site performed.
No violations of regulatory requirements or deviations were identified.
'3. ESF System Walkdown
The "A" auxiliary feedwater system was walked down and found to be
operable. Required surveillance had been performed by the licensee and
found to meet requirements. Electrical bus lineups were found to be in
compliance with procedures. Valve positions were found to be proper and
in accordance with the current valve line-up procedure.
No violations of. regulatory requirements or deviations were identified.
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4. Monthly Maintenance 0bservation
TDI Turbocharger Temporary Support Inspection
The inspector observed portions of the preoperational startup testing of
'the new Transamerica DeLaval Incorporated (TDI) diesel generators under
STP-1009, " Diesel Generator Integrated System Phase 2 Testing." The
inspector noted the addition of a support for the turbocharger which was
not present during the previous testing of the diesel generators. The
inspector questioned whether the installation of the support was an
approved modification of the qualified final configuration of the diesel
. generators. In response to the inspector's concern, licensee personnel
identified the following controls established for the temporary support:
a. Turbocharger vibration on the "B" diesel generator during previous
testing had been determined to exceed the manufacturer's
. recommendations. In pursuing a solution to the problem with the
manufacturer, the licensee had designed and installed the temporary
support under a Drawing Change Notice (DCN).
b. The temporary support was shown to reduce turbocharger vibration to
within acceptable values during test runs of the diesel generators.
c. A permanent support was being designed to replace the temporary
support. This permanent support would be installed prior to
turnover of the diesel generators to operations. However in the
interim, preoperational testing of the diesel generators would
continue with the temporary support installed.
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. The inspector reviewed the. licensee's' engineering' analysis for the
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temporary turbocharger support and found that the analysis. appeared
adequate for the structural design of the support. However, the
inspector found limited analysis of the effect of the temporary support
on the turbocharger itself to ensure that its performance was not
affected by the support. The licensee acknowledged.that the
acceptability of the diesel generator test results was at risk pending an
-analysis of the acceptability of.the temporary support compared with the
permanent design support. No analysis on the effect of the support which
was not part of~the configuration of the engine which had been seismically
qualified had been performed. The licensee was conducting an evaluation
of what testing needed to be performed on the diesel generator when all
modifications were completed, including the turbocharger support, at the
end of this inspection. Plans which the licensee was considering
. included additional vibration measurements and disassembly of the
turbocharger after operation with the new support in place.
The inspector reviewed the installation of the temporary support and
found that the DCN contained an installation note that required the
support to remain unbolted from the turbocharger until the diesel
generator had stabilized at its " keep warm" temperature. The inspector
noted that STP-1009 had not been revised to include the support
installation note. After bringing this omission to the licensee's
attention, a caution tag'was subsequently hung on the electric keep warm
heater breaker requiring unbolting of the support before deenergizing the
heaters. The inspector concluded that this corrective action was
adequate.
The licensee was experiencing additional vibrations with the engines.
Excessive vibration apparently caused pipe wall and pipe support failures
in the lube oil return and engine jacket water systems. Although the
licensee's corrective action had not been finalized, during the
inspection period, the licensee had not determined the root cause of
piping and pipe support failures. The licensee had not established any
plan to pursue the failures to root cause other than to continue
measurement of engine vibrations and modify piping as required. No
documented comparison had been performed by the licensee of their engine
vibrations to other licensees with similar engines. The lack of root
cause determination for engine vibration will be followed as an open item
(50-312/87-20-01).
No violations or deviations were identified.
Review of Maintenance of Motor Operated Valves (MOVs)
a. During post maintenance testing of safety related valve SFV-25003,
excessive noise was observed by maintenance personnel. Subsequent
disassembly of the valve operator disclosed extensive gear damage.
Further investigation by the licensee identified that a spring
retaining washer required for that particular style operator
(Limitorque SMB-2) had not been installed during refurbishment. The
inspector observed that the maintenance instructions did not
include instructions to install the spring retaining washer. The
licensee considered the procedural discrepancy to have resulted due
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to inadequate procedures developed from insufficient vendor
descriptive data. The licensee determined that only one other safety i
related operator of this style was applied in the plant and also
found the missing washer to also sist on that operator.
The inspector observed portions of the troubleshooting activities
and damaged components and found that the licensee exercised careful
controls over the troubleshooting activities to identify the root
cause of the failure.
The inspector reviewed the vendor technical manual information for
the affected operator' style. Although difficult to clearly identify
as a distinct part due to the small scale of the assembly drawing,
the inspector found that the required spring retainer washer was
identified in the drawing part list and was called out in the
assembly drawing for the operator.
The licensee acknowledged the inspector's concern and indicated that
a previously planned review of the implementation of vendor
preventative maintenance manual data would be accelerated and
addressed as part of the licensee's new preventive maintenance =
program to be implemented by September, 1987. A review of the
preventative maintenance program is currently underway by the
inspector and will be documented in a subsequent inspection report.
b. During post-maintenance testing of refurbished motor operated valve
SFV-25003, the licensee observed cracks in the cast yoke of the
valve. The licensee subsequently modified the yoke to weld a
stiffener to the yoke in addition to repair welding the defect area.
A similar modification was made to the identical valve in the "B"
train.
The inspector observed portions of the repair activity. During
repair welding the licensee encountered unexpected distortion of the
yoke which required hot working to return the yoke to an acceptable
condition. The inspector found that the heat treating was not
controlled within the scope of the original work request nor was the
work request changed to include the expanded scope. Rather, work I
was controlled under verbal instructions from the welding engineer
without a procedure or other written instruction.
This is an apparent violation of the licensee's Quality Assurance
program and 10 CFR 50 Appendix B, Criterion IX, " Control of Special
Processes" (50-312/87-20-08).
c. During maintenance refurbishment of valve SFV-25003, the licensee
installed a manufacturer supplied conversion kit on the operator of I
the valve to reduce the inertia over thrust resulting due to the
fast acting response of the valve. The conversion unit adapted the
operator to install a spring pack above the stem nut to allow the
stem nut to move vertically after the stem had seated to absorb the
excess momentum in the operator.
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The 'inspec' tor observed that the conversion kit completely enclosed
the normally exposed rising valve stem, thereby eliminating the
normal indication of the open valve position. Furthermore, the
conversion kit included a spring pack compression indicator which
was labeled " stem movement". The inspector was concerned that
~1abeling was inaccurate and might be misinterpreted by operators
during manual' operation.
The inspector expressed his concern to. licensee management who
acknowledged the potential for confusion and agreed to review the
adequacy of'the local valve position indication resulting from the
addition of the SP conversion units on the valve operators.
d. The' inspector discussed the lack of a formal program document for
the licensee's program in response to NRC Information Bulletin (IB)
85-03 in several. meetings with licensee representatives. The
previous submittals to the NRC had referenced the licensee's
f' program" and summarized the status of its implementation. However,
no program document had been prepared. Due to the expanding and
contracting nature of the MOV refurbishment effort during the
restart effort, the inspector was concerned that the lack'of a
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controlled program defining the IB 85-03 commitments could result in
inadequate control of the program to insure that it is implemented
.as committed.
'The inspector noted, for example, that some portions of the program
originally submitted to the NRC were outdated and were no longer
being implemented as part of the MOV effort. Specifically, a torque
wrench was to be used to establish proper torque switch settings as
part of the existing MOV program submittals to the NRC, but one was
not being used due to changes in the practices used for valve
adjustment.
The licensee acknowledged the inspector's concerns and committed to
submit a formal complete program description to update their
commitments in response to IEB 85-03.
One violation of NRC requirements was identified.
5. Radiological Controls
The inspector observed controlled area refresher training conducted on
June 22, 1987. In general, the inspector found that the training
appeared to be appropriate and effective.
The inspector noted that approximately 75% of the multiple choice
examination questions in one copy of the exam administered after the ,
refresher training lecture had blue ink tick marks associated with
selected answers. The inspector noted that the examination booklet l
l contained general instructions prohibiting making any marks on the '
examination booklet. The inspector discussed his observation with the
instructor and subsequently with the Training department management and
expressed concern that lack of control of the test copies may compromise
the effectiveness of the training. The licensee responded to the
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inspector's concerns by reviewing all test copies and found no other
cases of answer markings. In addition to quarterly reissue of the
examination, the licensee committed to add a unique identification of-
each~ examination copy to be. recorded with each answer sheet to establish ~
traceability to deter future marking occurrences.
The inspector found the licensee's response to be. appropriate.
No violations or deviations were identified.
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.6. Facility Modifications
During the current outage, 57 facility modifications have been made to
the plant and will become operable for this restart. The installation of
the EFIC system (Emergency Feedwater Initiation and Control System) and
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the TDI (Transamerica DeLaval Industries) emergency diesel generators
E' were the two major modifications. selected for inspection.
The inspections were performed by Lawrence Livermore National Laboratory
(LLNL) personnel for the USNRC under supervision of the resident staff.
'LLNL's report is attached to this. inspection report. The NRC discussion-
of the LLNL findings is contained herewith.
Nuclear Welding General Inspection Procedure
-Inspection of Mechanical Installation of EFIC
Inspection of the mechanical and welding areas did not reveal any
violations of regulatory requirements. However, a concern was discovered
during thejinspection and pertained to the installation of the four EFIC
level tap nozzles on each steam generator.
During the inspector's observation of completed welds for the EFIC taps
on the steam generators, an apparent base metal reduction was discovered
near weld "I". The repair had been authorized, however, due to an error
in preparing the weld repair record form, the weld had not been repaired.
The EFIC system had not been turned over to the station at the time of
the inspection and therefore, the final inspections by the licensee had
not been completed.
The licensee committed at the exit meeting to reinspect the four EFIC
nozzles on each steam generator for any additional undercut on the EFIC
sensing lines which had not been repaired. In addition, the licensee
committed.to a review of all work request packages related to EFIC level'
taps on both stesm generators to identify any work activity which could
reduce base metal, such as grinding to prepare'for nondestructive
examination and ensure adequate measures existed to detect base l
, reduction.
No violations or deviations were identified.
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Inspection of Electrical Installations of EFIC and TDI Emergency
Diesel Generators
Electrical Cable Separation
Electrical cable separation within panels HISS (E), H4FWB, H2 DEA 2 and
H2 DEB 2 did not appear to meet separation criteria as described in Nuclear
Engineering Procedure (NEP) 5304.8c. The criteria which the licensee has
established was under review by the Office of Nuclear Reactor Regulation
(NRR). The NRR was reviewing the extent to which the licensee's
separation criteria was in conformance with IEEE Standard 384 (Electrical
Separation Criteria for Nuclear Power Plants).
The issue of electrical cable separation criteria will remain an open
item (50-312/87-20-02) pending NRR review of the licensee's separation
criteria and followup inspection.
No violations or deviations were identified.
Design, Design Changes and Modifications
Review of Design Drawings vs. As-Built Plant Configuration
A review of drawings associated with ECNs A-5415AA and A-5415AH
identified four drawing discrepancies. The four differences identified
appeared to have been approved, however, the design drawings did not
reflect this approved configuration. Both ECNs had not received final
review and approval, and, therefore, the facility modification had not
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The licensee's drawing control system will be reviewed in a subsequent
inspection for outstanding drawing change notices to be incorporated into
the base drawing. This will remain as an open item (50-312/87-20-03).
No-violations or deviations were identified.
Electric Cable Inspection of Completed Work - TDI Building
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The following is a summary of concerns raised as a result of the review
of completed TDI building work,
a. In manhole number 47 which contained safety-related electrical
raceway between the TDI building and the Nuclear Services Electrical
Building (NSEB), safety-related instrumentation cable was found
laying across safety-related power cable. The appropriateness of
this and the issue of electrical cable separation criteria
applicability to underground raceway was also under review by NRR, i
and will be tracked with a previous open item number
(50-312/87-20-02).
b. Mounting bolts for thermocouple TE55701 were found during the
inspection to be bottomed-out within their mounting channel. This ,
has caused the lockwasher under the bolt head to not be compressed )
against the base plate.
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Thermocouple TE55701 extension wires were also found to be
mislabeled in that the wire materials were not correctly shown.
These labels were used to correctly terminate the thermocouple to
field cable. The as found terminations were found correct with the
labels incorrect. Additional inspection by the licensee also
discovered cracked insulation on these wires. Following these
findings, the licensee generated NCR 56810 to document both NRC and
subsequent similar licensee findings.
Mislabeling of a HVAC damper actuator (HV-55713) model number was
identified on drawing M17.07-6 of ECN-3748.
These issues will remain as a single open item pending review of
licensee's corrective action program. (50-312/87-20-04)
c. TDI buiiding essential HVAC air handling unit (AH-DG-1B) motor
termination box contained electric power cable which was bent
tighter than the allowable bend radius. The discovered bend radius
was approximately 3/4" and the allowable radius for the 4 AWG cable,
as specified by licensee procedures MPEM 8304.8c was 1.6". Quality
control inspection of this connection had occurred on October 17,
1984. Apparently, the motor connection box was not of sufficient ,
volume to accommodate the larger cable with adequate cable radii.
The licensee documented this finding on NCR 56845, subsequent to
this finding.
An inspection was conducted of electrical cable as they exited
electrical conduit and entered electrical cable trays. Cable trays
A57El and L5782 both contained cables which entered the tray irom
conduit A57015 and L57046 where the cable crosses over the cable
tray side channel. To prevent electrical cable jacket deformation,
licensee procedure NPEM 5304.7.c required tray edge bumpers to be
installed.
, Contrary to procedure requirements, the tray edge bumpers were not
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installed on the tray conduit junctions of A57El with A57015 and
l L5782 with L57046. Quality control inspections of cable trays were
l performed on May 5, 1985, September 18, 1984 and September 19, 1984,
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which did not detect this condition. The licensee has documented
l the inspector's findings on NCR 56844, after this finding.
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Raceway color labeling was required by the licensee procedures in j
order to distinguish electrical cable separation groups. I
Specifically, procedure NPEM 5304.7c required the color markings be
installed, however, electrical pull box H7J2995 contained a
safety-related "A" channel cable and should have had a red label,
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but was found with a black (non-safety related) label.
The findings in this paragraph appeared to have been a violation of
the licensee's quality assurance program and 10 CFR Appendix B,
Criterion X, " Inspections." (50-312/87-20-05)
d. The inspector questioned the cable raceway system used by the
licensee in manhole number 47. The design of the raceway system
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used within this manhole consisted of a cantilevered arm which
attaches to a wall mounted bracket. The arm remained in place due
only to the deadweight load of the cable and arm'itself. There was
-no mechanical fastener for the arm such as a bolt.
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L ' The concern raised was the ability of _this raceway system to remain
L ' functional during a postulated seismic event with positive and
negative vertical ground acceleration. In response,.the licensee
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committed-to provide the existing analysis to support the design of
the' raceway system.
An additional concern was raised on the mechanical expansion anchor'
l- bolts used to secure wall brackets to the concrete. The bolts did
not have any markings to indicate a quality control inspection. The
licensee used a paint marking on all above ground bolt installations
to show the inspection has occurred. Also, two bolts appear to not
o be perpendicular.to the wall. This item will remain open pending
inspection of the QC documentation on the installation of the bolts
and a review of; installation criteria for acceptable ;
perpendicularity criteria. i
These questions will remain open pending further inspections.
.(50-312/87-20-06)
e. The inspector observed that the licensee had irtseveral
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authorized and used."Siltex" or "Fiberfrax" tape to meet required
cable separation criteria. The inspector requested the test reports
for qualification of the use of the Siltex or Fiberfrax tape as an
acceptable barrier. The-licensee committed to provide for
inspection the qualification records of the fire barrier.used.
This item will remain open pending review of the qualification
reports against the installed configuration of electrical cable
(50-312/87-20-07).
7.. Modification Testing / Monthly Surveillance Observation
Review of System Review and Test Program (SRTP)
The following test outlines were reviewed against the system functional
requirements as described in the System Status Reports (SSRs). No
significant deficiencies were identified.
STP-1030 Maximum Differential Pressure Test of Seal Return Valve
SFV-24004 and Seal Injection Valve SFV-23616
RT-HVS-011 Refueling Interval Inspection of Hardcast Tape and Drain
STP-1045 Letdown Valves Interlock Test
STP-970 Diesel Generator (G-886A) Synchronization Check Relay-
l Functional Test
STP-1074 Demonstration of Alternate Decay Heat Removal Methods
STP-1078 Hydrogen Recombiner Functional Test
STP-660 Rev.1 ICS Tuning at Power
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RT-HVS-013 Special Frequency Reactor Building Emergency Cooling
Performance Test
RT-HVS-012 Special Frequency Rx. Bldg. Normal Cooling Performance
Test
RT-HVS-014. Special Frequency Rx. Bldg. Cavity Cooling Performance
Test
STP-7928 "B" HPI Pump Lube Oil Modification
-STP-667 EFIC Hot Functional Test
STP-1063A/B Essential Filtration Unit Air / Aerosol Mixing Uniformity
Test Train "A" & "B"
STP-1057B "B" Component Cooling Water Pump P-4628 Performance
Monitoring Test
STP-199 NSEB Fire Zone 76 Carbon Dioxide System Concentration Test
STP-1033A- DHS Pump P-261A Performance
STP-1033B DHS Pump P-261B Performance
STP-1063A CR/TSC Essential Filtration Unit Air / Aerosol Mixing
Uniformity Test Train "A"
STP-792C Makeup Pump Lube Oil Modification Test
STP-1053 CCW Component Flow Verification
STP-790 Rev.1 RPS Module Removal Interlock Verification
STP-432 Rev.1 Post Accident Sampling System Gaseous Functional Test
STP-791 IDADS Calculated Points for EFIC
STP-7928 "B" HPI Pump Lube Oil Modification Test
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STP-10638 CR/TSC Essential Filtration Unit Air Aerosol Mixing 1
Uniformity Test Train "B" q
STP-1031B Nuclear Service Raw Water (NSRW) Component Flow _j
Verification Loop "B"
STP-783 MU Pump /MU Tank Outlet Valve Controls Verification
STP-1045 Letdown Valves Interlock Test
STP-1012 Emergency Diesel Generator (G-886B) Post Modification Test
STP-966 NSEB, Cross-Zoned Suppression (Fire Zones 81 & 82) Test
STP-974 HV-20596, HV-20569, HV-30801 differential pressure MOVATS
Test
STP-779A Rev.2 CR/TSC Essential Air Flow Transmitter Data Collection
Train "B"
STP-7798 Rev.1 CR/TSC Essential Air Flow Transmitter Data Collection
Train "B"
STP-1030 Maximum DP Test of Seal Bleedoff Isolation Valve SFV-24004
and Seal Supply Isolation Valve SFV-23616
STP-774 IAS Backup to TBV, CCW CIVs, FWS Cont. Valves; MFW, SFW, ;
AFW and ADV using CLBB and ADVs using App. R Bottles Test !
STP-780 Instrumentation Cross Correlation Test i
STP-1065 Rev.1 Flow Path Verification of the Waste Water System Piping
Mods
STP-1009B New Diesel Generator G-100B/GEB2 Engine Integrated System
Phase 2 Testing
- During the period, the inspector observed the performance of STP-1009A.
) Testing witnessed was conducted in accordance with the procedure and
recorded data was documented per the procedure. !
During the performance of this test, the licensee continued with their I
program for monitoring TDI diesel engine vibration. Several piping and l
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or pipe support failures occurred during the testing. These included two
failures of pipe supports (cracked welds) in the jacket water system.
Also, failures (pipe wall tear and a pipe support weld crack) occurred on
both engines in the lube oil return lines.
The licensee believed these failures were all related to engine vibration
and the licensee was preparing a repair. The inspector was concerned
with licensee efforts to make repairs in that there was not a plan by the
licensee to determine why the failures occurred, including their root
cause. This item will be tracked as open item 50-312/87-20-01 discussed i
earlier. in this report.
Review of Surveillance Performed on the Nuclear Service Raw Water
Heat Exchanger
On June 18, 1987, the licensee discovered complete blockage of the !
bearing cooling water supply and return lines for the bearing lube oil
cooler on the "B" reactor building spray pump. Previously, a similar
condition had been identified in the "A" train component but with partial
blockage.
The inspector observed portions of the maintenance activities involved in
investigating the extent of the fouling problem including disassembly and
visual inspection of the lube oil cooling water lines, and cleaning
activities involved with the nuclear service spray ponds. Samples of
potential micro-biologically induced corrosion attached to the interior
surface of the internally coated carbon steel piping were removed for
additional vendor analysis. The inspector found that the licensee's
corrective actions appeared to be thorough and appropriate.
The licensee determined that the event was reportable under 10 CFR 50.73
and initiated preparation of an LER. Followup of this item will be
conducted in review of the forthes ing LER.
No violations or deviations were identified.
Transamerica DeLaval Inc. (TDI) Diesel Generator Testing
The inspector observed portions of the preoperational testing of the TOI
diesel generators under STP-1009. The inspector noted that the diesel
governor oil was found to be contaminated with foreign particles a second
time. (Previously, the licensee had found the governors to have been
internally contaminated and removed and returned them to the manufacturer
for cleaning.) Reoccurrence of the contamination after reinstallation of
the governors caused the licensee to more closely inspect the control oil
reservoirs. Extensive flushing and visual inspection of the reservoirs
disclosed additional debris not previously noted.
The inspector questioned licensee personnel on the source of the debris.
The licensee considered that since the debris contained silica sand, that
it most likely was construction contamination which had not been cleaned
up during previous flushes of the system. The licensee responded to the
inspector's concerns by repeated flushes of the oil reservoirs to
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establish a baseline cleanliness for the system and by lockwiring the
filler caps for the reservoir to avoid tampering.
The inspector found that the licensee's actions appeared to be
appropriate to assess any further. repeat occurrences.
No violations or deviations were identified.
8. Review of Problem Statement Prioritization (0 pen)
Nuclear Service Cooling Water System (NSRW) Status Report Review
The inspector reviewed the licensee's System Status Report (SSR) for the
NSRW system for the acceptable prioritization of problem statements as
partLof an inspection of the licensee's Plant Performance and Management
Improvement Program (Safety Evaluation Report Item No. 2.3.2). The
inspector's criteria for acceptability as a post-restart item was whether
all regulatory requirements related to the item and system would be met
even if the item was not performed prior to restart.
The inspector noted that the NSRW system status report, Revision.1,
approved on 11/6/86, identified twenty-one problems. Of these problems,
four were considered priority 1, eleven were considered priority 2 and
six were considered priority 3. The inspector reviewed the priority 2
and priority 3 items (these items are considered post-restart by the
licensee) with the system engineer for the NSRW system and agreed with
the licensee's prioritization of them. In addition, it appeared that a
majority of the priority 2 and 3 items had already been completed.
The inspector brought to the licensee's attention at the exit meeting a
problem dealing with a resolution of an item identified in the SSR. The
item (QTS 26.0192) was classified as priority 2; it identified the
possible flaking of interior epoxy coating of the raw water piping. The
resolution defined for the problem was to monitor the pressure drop for
for the NSRW heat exchangers and if the pressure drop became excessive,
"... problem will be represented to PAG for possible upgrading to
Priority 1 status". The inspector's concern for this resolution was that
the monitoring of the pressure drop was considered priority 2, but, if
the results of the monitoring effort were poor then the problem would be
considered as a possible restart item. However, when the priority 2 item
was completed the plant would have already been restarted, and upgrading
the item to priority 1 at that time would be ineffective. The licensac
acknowledged this discrepancy, and noted that a "true-up" program had
been established to address similar inconsistencies systematically. The
inspector will continue to followup this issue and review the completed
work after the "true-up" program has resolved these types of issues.
No violations or deviations were identified.
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9. Inspection of Allegations
Characterization
Temporary power plant mechanics are inadequately trained for their work
assignments in most cases. (Allegation RV-A-87-10)
Implied Significance
Work activity performed on safety related equipment could possibly effect
the performance of equipment and not be detected by post maintenance
testing or surveillance. Such failures would then be prevented only by
carefully planned and conducted work activity by qualified crafts people.
Assessment of Safety' Significance
The individual is specifically commenting on the amount of supervision
crafts people received during the work activity. The alleger believed
there was inadequate supervision, leading to higher radiation exposure.
The concern centers on .a one-of-a-kind work activity which was being
performed for the first time.
The inspector reviewed three work requests covering work on motor
operated valves and installation of the Hydrogen Recombiners in
containment. In both cases, the work request /ECN covering the work
activity was properly documented and contained sufficient instructions to
perform the work. Craft foremen were observed by the inspector to be
carefully observing work practices of their respective crews. During
breaks or other periods of reduced work activity the crews were observed
to adhere to ALARA principles if they were in a high radiation areas. In
discussions with the crews they all appeared to have discussed their work
activity prior to commencing the job. In the case of MOV work all crafts
people had received training for their activity.
Staff Position
No evidence of inadequate training of temporary plant mechanics was
discovered. The allegation was not substantiated. (CLOSED) j
'
Action Required
None
Characterization l
The allegation (Allegation RV-A-87-16) concerns the following aspects
involved in welding:
a. Harassment by supervisors for raising quality / safety concerns
b. Inadequate welder training
c. Inadequate weld rod control
d. Supervisory pressure to weld without pre-heat
e. Inadequate welding tools
f. Inappropriate use of welder I.D. stamp
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Implied Significance
Safety'related welding activities must be performed in accordance with
committed codes and standards to insure that designed strengths are ,
obtained. Such inadequacies as discussed above could lead to failures of '
structures or plant systems.
Assessment of Safety Significance
Concern number 1 addressed the harassment due to the raising of a safety
or quality concern. For the work which occurred on safety related
equipment, Quality Control Inspectors were present to witness hold
points. The inspector questioned two QC personnel who inspect welding to
identify if they believed harassment had occurred or if they were told of
any harassment. Both stated that they are not aware of any harassment i
occurring. They also stated that they would report to Quality Management ,
any safety and or quality concerns raised to them. j
1
Concern number 2 addressed welder training onsite. All safety related
welding onsite was made in accordance with either ASME, ANSI or AWS
welding codes. All require welder training and performance tests. The
licensee's program for performance testing and training was in accordance
with committed code requirements.
Concern number 3 addressed weld rod control. The licensee subsequently
identified problems with weld rod control; this subject was discussed in,
Inspection Report 50-312/87-13. The problems were addressed by the
licensee's Quality Department, a report was made to licensee plant
management and a stop work was issued on onsite welding, The major issue
identified by Quality and the ANI concerned the traceability of weld
filler material. At the time of the inspection, onsite welding was in
progress with traceability maintained in accordance with requirements.
Concernnumber4addressedtgeuseofpre-heat. For most onsite welding,
a material temperature of 60 F was required. For all welding observed by
the inspector, the required temperatures were achieved without any
external heat source needed. In cases where a higher pre-heat
temperature was needed, QC was required to verify that the preheat
temperature was achieved. No instances were observed by the inspector
where pre-heat temperature was not maintained.
Concern number 5 centered on inadequate welding tools being used. The
inspector observed in process welding equipment onsite and welding
equipment in storage at the tool room. All equipment was in good working
,
'
order and delivering the required amount of current to the welding torch.
All hand-held welding devices were found to be operating properlys
Concern number 6 addressed the improper use of welder I.D. stamps. The
attached LLNL report in paragraph B.4 addresses a problem discovered by
the inspector of an incorrect I.D. apparently being used. The concern
here is similar. The traceability required is also achievable by
checking the " Weld Filler Material Withdrawal Slip" for welder I.D.
However, the proper I.D. stamp number must be used. This subject was
raised with licensee management and with the licensee welding engineer by
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the inspector. ' Additional direction w.d provided to welders for defrect
use of their I.D. number. N, f
,
No specific welds were heatified by the concerned individual Wnne' ' !i i/
incorrect welder I.D. numbers were used. J ...
Staff Position \ ,
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"pWocedures or records were
Noinadequateoranacceptableweldirgkeallegationwaspartially
discovered riuring this inspectich. . T. t
substantiated (Concerns 3 and 6?) The lf< ensee corrective action i \
discussed in inis paragraph and p3repa,ph 6 appears adequate. Th;s
allegation is' closed. .
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Acti,on Required s
None. ; i
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10. Followup on NRC Open Items '
,
3
Due to the ;1icensee's luge iGC ohen iden lis't, the licensee hai been
involved fri a program to providM aAore timely closure of NRC 'dentified "
4
open items. In this program a 'clor,ure /qckage for an item is assembled
and reviend by the licensing department and then verified as closed by
the quality assu'rance department. Af ter this procitss the package 'ir, i
available for the irdspector's review and inspection.
In previous inspection reports the'Hcensee's closure packages have been
discussed; however, it has. been apparent. that although there still exirts
a large number of open items, the raw of closure reportt3Mt have baen
available for inspection is very lcw 'In fut during this~ report period
there were times when no package; were avai'lable to inspect. The
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inspector discussed the need, duMng the exit meeting, to providr ,
information for open items on a timely tasis to facilitate the closure of
the open items. The inspecter will continw to inspect the licenseef s /
process for closing open items in a future inspection.
Enforcement Items
!
Violation 87-05-0} (CLOMD), " Voiding of NCR #S614(P
Inspection Report 87-06 identified a Notice ;,f Violation (Severity
Level IV) concerning licensee's measures which did not assure.that the
cause of the significant condition adverse to quality identified by
Nonconforming Report (NCR)-56140 dated 12/09/86 (emergency diesel
generator nonconforming condition) n s determined. In dadition, licensee
measures did not assure that corrective action to preclude repetition df 1
the significant condition adverse to quality identified by NCR-S6140 was j
taken. ,
I
The inspector's review of the licensee's responseidetermined the j
following: >
a. The licensee a$aitted that the violation occurred as stated.
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'f i The reason for the violation was a lack of communication between the
4 1 individual requesting the voiding of NCR-S6140 and the initiator.
f ' Under normal circumstances NCRs are voided only when no
3. i
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, ' nonconforming condition actually existed, and in this case, it was
y a? j initially thought that no nonconformance existed. However, later
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during a reevaluation subsequent to the voiding of NCR-56140,
Quality Assurance determined that the actual as-found condition was
, nonconforming.
I c. The corrective actions taken and the results achieved were that
Quality Assurance ~ Procedure QAP-17, " Nonconforming Material Control"
had been revised via Revision 5, dated March 16, 1987, to read "For
'
voided NCRs the QE Supervisor, or his designated representative,
will stamp, sign and date the NCR.. The originator or his supervisor
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must concur (sign and date) with the voiding. The Manager, Quality,
I can void an NCR without. concurrence of the originator."
(<t
In additien,sNCR-56140 was reopened by Revision 1 and processed in
accordance with QAP-17, Revision 5. The Governor Drive Coupling set
," screws were verified tight on 2/10/8'7 per work request #125701, and
NCR-56140 Revision 1 was closed.
+ The inspector concluded that the licensee's corrective action for this
violation was adequate.
, Violation 87-06-01 (CLOSED), " Failure to Issue NCR for Leaking Pipe"
Inspection Report 87-06 identified a Notice of Violation (Severity
Level IV) concerning that a Nonconforming Report (NCR) was not written
for non-isolable pipe leakage in the "A" train of the nuclear service raw
water (NSRW) system as identified in Work Request No. 110755 dated
2/2/86.
The inspector's review of the licensee's response determined the
following:
( a. The licensee admitted that the violation occurred as stated.
N b. The reason for the violation was the lack of adequate training in
'
Quality Assurance Procedure, QAP-17, " Nonconforming Haterial
Control" of the persons involved.
c. The corrective actions taken and the results achieved were that an
NCR-56480 was initiated on Feb. 27, 1987 to address this
non-isolable pipe leakage, and Revision 5 of QAP-17 was made
concerning this item on March 16, 1987. In addition, as part of the
revision process, plant personnel identified as requiring training
have been trained on QAP-17.
s
1b The inspector reviewed the training records and briefing notes for
! , , Revision 5 of QAP-17 which indicated satisfactory coverage of QAP-17
? I topics and attendance. NCR-56480's disposition was to replace the spool
oiece and work request #110755 fabricated and installed the spool piece
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on 3/9/87. Based on the licensee's corrective actions, the inspector
considers Violation 87-06-01 closed.
Violation 87-06-02 (CLOSED), " Failure to Post Abnormal Tag for Repair"
Inspection Report 87-06 identified a Notice of Violation (Severity
Level IV) concerning the licensee's failure to prepare an Abnormal Tag
for the temporary modification of the "A" train nuclear service raw water ,
(NSRW) pressure boundary as identified in Work Request No. 110755 dated
2/2/86. The "A" train NSRW as returned to service with a rubber patch
clamped onto the exterior of a defective pipe section to control
non-isolable leakage. This condition existed for over one year.
The inspector's review of the licensee's response determined the
following:
a. The licensee admits that the violation occurred as stated.
b. The reason for the violation was a lack of adequate training in
AP-26, " Abnormal Tag Procedure" of the person involved.
c. The corrective actions taken and the results achieved were that the
person responsible for failing to write the Abnormal Tag was
retrained in the requirements of AP-26 by June 2, 1987, and training
in the requirements of AP-26 and QAP-17 were given to Maintenance
Department personnel involved in the performance of corrective
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maintenance.
During this training period, personnel were asked to identify other
cases of failing to issue an Abnormal Tag. No cases were identified
during this session. This action was completed May 2, 1987.
Based on the above licensee's corrective actions, the inspector considers
Violation 87-06-02 closed.
Violation 87-06-03 (CLOSED), " Failure to Provide Immediate NRC
Notification for LER 85-13 and 85-20"
Inspection Report 87-06 identified a Notice of Violation (Severity
Level IV) in which the licensee did not notify the NRC Operations Center
as described below of the following events:
a. Actuation of the "A" diesel generator on June 22, 1985. (Required
under 50.72(b)(2)(ii) and later identified in LER 85-13.
b. Discovery on October 7, 1985 of a failure of the essential control
room HVAC system to function per design requirements and Technical
Specification requirements. (Required under 50.72(b)(2)(iii) and
later identified in LER 85-20.)
c. Discovery on January 7,1986 that insufficient voltage would be
available to the essential control room HVAC power supplies
following a design basis LOCA. (Required within four hours under
50.72(b)(2)(iii) and in LER 86-23.)
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The inspector's review of the licensee's response determined the
following:
L a. The licensee admitted that the violation occurred as . stated.
1
b. The reason for the violation was that' administrative procedures.
established to control the reporting of plant occurrences did not-
, adequately define the responsibilities and process for evaluation
and reporting of events in accordance with the requirements of-
'10 CFR 50.72.,
c. The corrective actions taken and the results achieved we're that
Administrative Procedure AP-22, " Occurrence Description Reports
(0DRs) Reporting and Resolution" has been revised to amplify the
reporting requirements of 10 CFR 50.72. A'-22, Revision 12, was j
issued.on January 16, 1987.' Personnel identifying an unsafe,
atypical, or off-normal condition are now required'to promptly
.
notify the Shift Supervisor and to provide an ODR to the Shift'
Supervisor with one hour. The Shift Supervisor has been clearly
designated as having the responsibility for evaluating 10 CFR 50.72 ,
deportability and initiating 10 CFR 50.72 reports. As part of
Senior License training and requalification training, the Shift
Supervisor received instruction on the requirements of 10 CFR 50.72.
Since revisions to AP-22 have been implemented, there had been no
additional deficiencies in the reporting of 10 CFR 50.72 occurrences.
The inspector reviewed AP-22, Revision 12, which provided instructions to
ensure that an unsafe,. atypical or off-normal condition is promptly
verbally notified to Shift Supervisor (Para. 4.1.1), and an ODR delivered
to the Shift Supervisor with I hour of initiating the ODR-(Para. 4.1.3).
The procedure phone notifications shall be made by the Shift Supervisor
in accordance with paragraph 4.3 of AP-22. Para. 4.3.2,-states that if
an NRC telephone notification is required, the Shift Supervisor shall
make the notification to the NRC Emergency Operations Center within the
time limit specified in 10 CFR 50.72. Based on the above licensee's
corrective actions, the inspector considers Violation 87-06-03 closed.
Violation 87-14-01 (CLOSED), " Failure to Issue NCR on Capscrew Heads
_
Broken"
Inspection Report 87-14 identified a Notice of Violation (Severity
Level IV) concerning broken capscrews for the bearing positioner on the
RCP at the time of the Inservice Inspection of Reactor Coolant Pump (RCP)
P-210B on April'15, 1985. This nonconforming condition was not reported
on an NCR'and dispositioned accordingly. .As a result the required
management review and approval were not completed.
The inspector's review of the licensee's response determined the
following:
a. The licensee admitted that the violation occurred as stated.
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L. b. The reason'for the violation was inadequate training of the
personnel who were responsible for implementing Quality Assurance
Procedure, QAP-14, " Nonconforming Material Control".
c. The corrective actions-taken and the results achieved were that
plant personnel who were identified as requiring additional training
in the Nonconformance Report (NCR) process have been trained in
QAP-17, Revision 5, dated March 16, 1987. The nonconforming Reactor
Coolant Pump capscrews were replaced after management review
. determined that replacement was the. prudent corrective action. This
corrective action.was reviewed by the Plant Review Committee (PRC),
t
MSRC, and upper plant management, including the. Quality Department,
!. through the Occurrence Description Report (0DR) response review
L- process. These capscrews apparently provide. support of the RCP
bearing assembly during assembly. The licensee contacted the pump
manufacturer, Bingham-Willamette Company, and determined that the
failure of the capscrews had occurred due to relaxation of the mild
,
fastening. torque.and cycle fatigue. The licensee requested and
received a change in the vendor technical manual torque value and
, installed new capscrews with the higher torque values.
Based on the above corrective actions, the inspector considers Violation
87-14-01 closed.
Unresolved Item (URI)
>
URI 85-31-01 (CLOSED), " Procedures for USAR Revisions"
This item identified that in an amendment to the licensee's Updated
Safety Analysis Report (USAR) the licensee inadvertently omitted
revisions to one section of the USAR but the revision was made to another
section of the USAR.
The inspector reviewed this item in report 87-11 and reported that the
licensee's corrective actions appeared to be appropriate; however, the
item remained open pending a review of an issued Nuclear Directive
prescribing USAR revisions. The procedure, LDAP-0003 Revision 0,
effective date, 7/3/87, "USAR Revision Control", appeared procedurally to
control the revision process, therefore this item is considered closed.
1
Licensee Event Reports (LERs)
LER 85-19-L0 (CLOSED), " Reactor Trip on High RCS Pressure" ,
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This.LER dealt with the October 2, 1985 reactor trip and subsequent
cooldown. This event occurred with the reactor at approximately 15%
power in preparation for a turbine overspeed trip. The licensee provided
a brief description of the event in the LER and provided a reference to a
letter submitted to the NRC (October 18, 1985) which contained the
licensee's corrective actions.
I
The event and subsequent followup has been addressed in NRC inspection
reports. In addition, the NRC immediately issued a Confirmatory Action
Letter (CAL) to the licensee, on the day of the event, which requested an
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explanation of the event and corrective actions prior to restarting the
plant. The licensee responded to the CAL and subsequently restarted in
. late 1985.
The inspector reviewed the action list generated from'the event. It
appeared that appropriate actions'have been identified and corrective
actions have been implemented. In addition, due to the December 26, 1985
event, the licensee has developed a more far reaching improvement program
which is being implemented to date and for certain items will be
completed prior to restart. Therefore, this LER is closed.
LER 87-11-LO, 87-11-L1-(CLOSED), " Snubbers Found Out-of-Tolerance
per New Temperature Compensated Acceptance Criteria"
This LER reported that previous testing done on hydraulic snubbers used
inadequate acceptance criteria which allowed snubbers to be declared
operable when it was later demonstrated that they were not operable based
on the corrected acceptance criteria. The original LER was inspected and
documented in NRC report 87-12. The revision to this.LER was submitted-
to the NRC on May 14,:1987.
The inspector reviewed the LERs and found that the LER reflected the
appropriate reporting requirements and descriptions. The root cause of
the failure appeared to be appropriately determined and corrective .
actions should correct the cause of the event. The inspector reviewed
the procedure changes initiated by the corrective actions and had no
further-questions. Based on the above review and previous inspection,
this LER is closed. LER 87-11-LO, 87-11-L1 is closed.
LER 87-15-LO (OPEN), "CO, Deluge System Deactivated"
Licensee Event Report 87-15 described several carbon dioxide protected
fire zones CO, deluge systems left deactivated and the zones unattended.
These numerout occurrences happened because personnel who, for safety
reasons, disabled fire protection C03 systems in the plant, and then
failed to reactivate the deluge system for over an hour after the safety
concern for personnel was no longer present.
The licensee's corrective actions were reviewed by the inspector who
determined the following:
a. A letter was transmitted to all site employees on tag removal and
zone reactivation to raise the awareness'of those requirements,
b. An item was placed in the plant daily newsletter " WATTS HAPPENING"
on March 10, 1987 to reinforce the requirement for tag removal and
zone reactivation.
'
c Safety meetings added on agenda item concerning tag removal and zone
reactivation.
d. Administrative Procedure (AP) 30, " Entrance into Carbon Dioxide Fire
Protected Zones" has been revised to explicitly state if no other
identification tags are present, remove your identification tag.
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Procedure temporary change No. 87-0392 added paragraph 3.3.5.3 to
state: " Remove your identification tag" and was approved on
April 6, 1987.
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e. In addition, corrective action was instituted to provide an hourly i
verification that personnel are, in fact, in the deactivated CO 1
Zone when the system is deactivated. Thisverificationwillrebain
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in place until system design or the entry procedure is revised to
provide improved control of system and personnel status or until it
has been demonstrated to the Plant Review Committee that the
corrective measures have been effective. Per conversation with
Regulatory Compliance by the inspector revealed that there has been
a marked decrease in the number of Occurrence Description Reports
written for CO2 related events.
This LER will remain open based on findings discussed in inspection
report 50-312/87-15 and not yet resolved.
LER 87-19-L0 (CLOSED), " Failure to Post Fire Watch"
Licensee Event Report 87-19 described the failure to post a continuous
fire watch posting in Fire Zone 78 during a welding job, after removing
the fire detection system from service. After removing the detection
system from service the Control Room should have taken compensatory
measures, and established a continuous fire watch as required by
Technical Specification 3.14.6. The continuous fire watch posting was
required due to the breaches between the areas comprising the fire zone.
The Control Room did not upgrade the hourly fire watch patrol to a
The licensee's corrective actions were reviewed by the inspector who
determined the following:
a. The Operations Department issued Special Order 87-15 to its staff.
Due to the complex nature of the fire protection components in the
plant, special order 87-15 requires that the Shift Supervisor must
rely on the Fire Protection Coordinator and his group to provide an
analysis of each situation to justify why a continuous fire watch
should not be used. AP.60, " Control of Fire Protection Limiting
Conditions for Operation", step 4.3.2.2.5.2, states, " Post a
continuous fire watch until the Nuclear Operations Fire Protection
Coordinator (N0FPC) can provide further analysis."
b. The corrective action regarding the missed continuous fire watch
posting in Fire Zone 78 was accomplished when detection was restored
on February 13, 1987, at 2:30 a.m.
,
c. This LER described an event which has recurred at Rancho Seco, and
the writing of the special order is an outgrowth of the review of
the licensee's fire watch policies.
,
Based on the above licensee's corrective actions, the inspector considers
Licensee Event Report 87-19-L0 closed.
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, 11. ' Unresolved Items
Unresolved items are. matters'about which more information is required to
o determine whether they are acceptable or may involve violations or_ -
? . deviations.
- 12. Significant Meetings
On June. 18, 1987, the. inspector attended the bimonthly meeting of
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, ' licensee's board of supervisors at.the licensee's headquarters in-
- Sacramento. The licensee's new Chief Executive Officer, Nuclear,
- presented a revised-schedule for plant restart in January,1988, with an ;)
. extensive power ascension test program to extend.into June, 1988. In
. addition, the licensee announced the selection of Karl Meyer as the
Nuclear Licensing Manager.
)
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A meeting was held at Rancho Seco on June 7, 1987 to discuss the status j
of the system' review and test program (SRTP). The attendees in the j
meeting. included J. Martin,.L. Miller,.and the resident staff of the NRC' '
and W. Bibb and G. Blackburn of SMUD.
The licensee presented.an overview of the SRTP which included in part the
organization chart, the system status report flow chart, system test
matrix. outline, and an outline for the hot shutdown and power ascension
testing.
13. Exit Meetina
The. inspector met with licensee representatives (noted in' Paragraph 1) at
various times during..the report period and formally on July 14, 1987.
The scope and findings of the inspection activities described in this
report were summarized at the meeting. Licensee representatives 1
acknowledged the inspector's findings and violations identified. l
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Enclosure C
Review Conducted by Lawrence Livermore National Laboratory Personnel
of Major Facility Modifications at Rancho Seco
The inspections performed and documented in this paragraph were performed by
Lawrence Livermore National Laboratory (LLNL) personnel for the USNRC, and
under the supervision of the Senior Resident Inspector. ,
I
The following documents, procedures, and standards were reviewed as background 1
for. inspection of recent construction involving the major plant modifications. {
i
- NRC Inspection Report #50-312/85-01
Re: NSEB Construction Review
- NRC Inspection Report #50-312/85-19 ,
Re: High-Point Vent Review I
- SMUD Procedure, NEP 4109
" Rancho Seco Configuration Control"
- SMUD Construction Specification, NEP C5307 series
Re: Installation of piping, welds, and supports
- SMUD Construction Specification, NEP C5303
Re: Expansion Anchors in Concrete
- SMUD Procedure, NEP 6405
" Pipe Supports"
- Rancho Seco Welding Manual
- Rancho Seco Nondestructive Examination Manual
- SMUD Procedure, NEP 6118
" Field Problem Report"
- ECN #A-5415A
Re: Level Taps on Steam Generator
- ECN #A-5415AB
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ECN #A-5415AC
Re: MOVs on ADVs isolation of steam generator loop
- ECN #A-5415AD
Re: MOVs on Turbine Bypass isolation
A. Items Inspected
1. ECN #A-5415A Re: EFIC Level Taps on Steam Generator
The inspector used the as-built drawings in the field to verify
hardware condition and location for the four EFIC nozzles on Steam
Generator E-205-A. No hardware problems were found, dnd the
as-built drawings reflected the field condition adequately.
2. ECN #A-5415AC Re: MOVs on two ADVs
(
The structural member upgrades, pipe support modifications, and the '
piping welds for installation of MOVA on two ADVs (in Areas 6 & 7
between the reactor building and the fuel storage building) were
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inspected. No field problems were encountered in comparing ,
- Drawing C-612, sheets 1 & 2 with the hardware. '
3. ECN #A-5415AC Re: MOVs on Turbine Bypass isolation !
Structural member upgrades, pipe support modifications, and piping !
welds were field reviewed by the inspector for the installation of I
l four MOVs on the main steam turbine bypass lines (in Areas 6 & 7 J
!
between the reactor building and the fuel storage building and the j
l fuel storage building). Drawing C-611, sheets 1, 4, & 6 adequately l
reflected the as-built condition of the hardware.
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! B. Followup Documentation Review
As part of the review of ECN #A-5415A involving the EFIC Nozzles on the
! steam generators, six work request (WR) packages were selected by the l
1 inspector: WR 125164, WR 125011, WR 125165, WR 116569, WR 125166, and I
WR 125167. These work request packages were checked for adequate {
,
documentation of QC records, welder identification, material
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traceability, valve serial number, weld procedure, and weld {
identification. '
The following are findings made by the inspector:
l 1. For WR 125165 (line 20601 - 3/4" - CA, steam generator E-205-A) of
i
ECN #A-5415A, there was no weld filler material issue (FMI) slip ,
l attached to the work request package, for the repair of weld "I" at (
l the steam generator shell. The FMI slip is routinely used to record
l the work request number, date, weld numbers, welding procedure,
I filler material size and type, heat number or lot number, and welder ?
l identification. The weld repair record form in the work request
i package included instructions to the welder for which welding
i procedure and filler material to use and the location of the weld;
! but no record of heat number traceability was shown for the weld I
l repair. l
l i
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This work request package (WR 125165) had been reviewed by the
licensee's " Filler Material Task Force" several weeks prior to
review by the inspector. The lack of heat number traceability in
the work request package for the weld repair was not detected by the l
! licensee during it's in-house review by the task force. The I
licensee decided to initiate an NCR for this discrepancy. The weld
inspection check list (ICL) for weld "I" indicated that it was
nuclear class 1 with ASME section III, subsection NB as it's
I applicable code. Also, the work request cover sheet lists the
welding of the EFIC nozzle to the steam generator shell as an ASME i
"Section XI work package". l
2. It should be noted that performance traceability of weld "I" was
possible in the field, since the welder identification (SM 319) was
stamped nearby the repaired weld on the pipe. In reconstructing
what had occurred, SMUD produced a copy of the FMI slip for Welder
SM 319, dated 12/22/86 for weld "A" on WR 116569. Weld "A" was only
3 inched away from weld "I" on the same line (20601) but a different
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work request package. When the licensee contacted Welder SM 319 on
6/10/87, he explained that he had noticed some base metal reduction
on weld "I" on 12/22/86.
Welder SM 319 concluded that the base metal reduction had occurred
during preparation of weld "I" for a magnetic particle test (MT) on
12/21/86 when some grinding at the toe of the weld occurred. The MT
report on 12/21/86 showed weld "I" as acceptable, with no mention of
base metal reduction. On 12/22/86, the plant welding engineer
authorized the repair of weld "I" by initiating a weld repair record
form, per information from Welder SM 319. At that time, WR 116569
'
was entered at the top of the weld repair record form in error,
since the nearby weld "A: was what Welder SM 319 had been working
on.
Later, on 1/12/87, the Field Engineer corrected the weld repair
record form to show WR 125165 for weld "I". Also, Welder SM 319
explained that he originally described to the plant welding engineer
on 12/22/86 that weld "I" had 1/16-inch undercut all-around, instead
of the base metal reduction which he later described on 6/10/87. It
turned out that the required filler material for weld "A" and for
repair of weld "I" was the same type; but weld "I" had no record of
filler material issue.
3. In reviewing several work request packages, the inspector noted that
all of the records showed the proper weld identification (line
number plus alpha designator) except for the magnetic particle test
(MT) reports. The design drawing, weld inspection check list, weld
repair record form, and liquid penetrant test (PT) report typically
showed the proper weld identification. The MT reports should show
the same description of weld identification for clarity and
consistency in records.
4. The weld filler material issue (FMI) log on 1/5/87 and 1/6/87 showed
the welder somctimes as SM 314 and other times as SM 319. The
correct identification is SM 314 per the Master List of Welder
Qualifications.
Followup Actions by Licenses
At a pre-exit meeting on 6/12/87, the Plant Welding Engineer and the QC
Supervisor suggested that an improvement to the nondestructive
examination (NDE) manual would be to: .
.
a. Require visual weld inspection (VT) on all NDE that required
grinding; and
b. Require that any post-weld heat treatment (PWHT) occur prior to
the " final" visual weld inspection (VT).
At the exit meeting on 6/12/87, the licensee committed to the following:
a. District will perform a re-inspection of the four EFIC Level
Tap nozzle welds on each steam generator; and
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. . b '. District will perform an additional. documentation review on the
i: . corresponding work request packages.
Electrical, Instrumentation and Control EI&C of.EFIC and TDI
Four EFIC Engineering Change Notice (ECN) packages were selected for-
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audit,- These were:
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'ECN A-5415B -Installation.of.new once throughisteam generator
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_ level transmitters
ECN A-5415E- . Electrical.and fiberoptic cable to the EFIC local
control panels
ECN A-5415AA Power and control to valve motor operators HV-20581'
and HV-20582
ECN A-5415AH ' Installation of the Control Room EFIC panel
These packages _were selected to. provide a representative sample of EFIC
construction work activity. These packages include power, instrument,
and control' circuits. They also provide a cross section of the EFIC work
in the Control Room, Containment, Nuclear Services Electrical Building-
(NSEB), and the Tank Farm.
'
A number of. Design Change Notices (DCNs) were selected from each package.
~
These were_ walked down to confirm that the'as-built configuration
conformed with the design.as defined by the DCNs, and Rancho Seco
construction specifications and procedures.- The specific items inspected
included:
Power cable routing to motor operated valves HV-20581 and HV-20582
including breakers.and terminations in load centers SOC 2, 50D2, and
several field terminal boxes.
Instrument cable routing between level transmitters LT-20505A, and
LT-205078 including terminations at transmitters, containment
penetrations, and field terminal boxes. _Also examined transmitter
installation to determine compliance with equipment qualification
requirements.
Selected terminations and channel separation in the HISS (E) panel.
Also inspected device installation and human factors enhancements.
Selected terminations and separation in the EFIC field panels H4E181
and H4FWB. 'Also examined fiberoptic cable installation in the H4FWB
panel.
Witness of electrical construction activities in panels S1B2 and
,
S102. This work in progress is covered by Work Request 120609.
. The EFIC inspection included review of several Work Requests which
'
implemented these ECNs. Work Requests include detailed work
instructions, material records, and inspection records.
Inspection of TDI construction included:
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Diesel engine. control panels H2 DEA 2 and H2DEAB2
-Safety related wiring on the "A" side diesel skid
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l "A" side raceway installation
Portions of the electrical' distribution system associated with the
TDI including both 480 volt and 4160 volt circuits
Portions of the TDI building essential Heating, Ventilating and Air
Conditioning (HVAC) system
The "A" side motor control system
The "A" side engine radiator fans
The "A" side fuel oil system pumps and level instrumentation >
Findings
Except for one issue, the EFIC items inspected conform with design.
The area of exception is wiring separation inside of control panels.
Three of the EFIC panels inspected had the potential for wiring
separation problems. Two of these contained separation problems
that were not detected by the licensee Quality Control (QC)
' inspections. ,
!
ECN A-5415 AH, drawing change notice OE, to drawing N25.08-46, ;
Revision 3 requires: "The wiring of individual channels (separation !
groups) must be bundled together and separated from other channels
by a minimum of 6-inch air space. Where conduit or Siltemp is used
a minimum of 1 inch air space is required." Contrary to this
requirement, LLNL's inspection identified fifteen cases in control
room panel HISS (E) where vendor installed internal panel wiring of
different channels was not separated by 6 inches and was not
protected by Siltemp or conduit. The vendor installed wiring was !
inspected and accepted by the district and the inspection documented. ;
in a Source Inspection Data Report dated November 26, 1986. !
!
Nuclear Engineering Procedures (NEP) Manual section 5304.8C, I
" Electrical Cable Installation and Termination" requires that safety ;
related and nonsafety-related cables, or different channels of )
safety related cables be separated by 6 inches of air space or a
metal barrier. It also allows an alternate barrier consisting of
Fiberfrax sleeving or tape covered by Scotch 69 tape where 6 inch
separation does not exist. Field cables entering the HISS (E) panel i
are protected in'this manner where the 6 inch separation criterion
cannot be met. In a number of cases, however, LLNL's inspection
found the Fiberfrax installation stopped short of the quick
disconnect connectors used as field wiring terminations.
Consequently, about 1/4 to 1/2 inch of cable is left unprotected !
where the cable enters the connector. Additionally, the cable entry
hole in the connector backshell is not protected by fire retardant
material. This installation of field cables was installed,
inspected, and accepted by Work Request 124925 which was closed on j
March 21, 1987.
I
A similar situation exists in panel H4FWB. Non-safety field cables
in this panel were wrapped with Fiberfrax tape to the point where ,
the cables pass through a hole in a terminal box which separates :
non-safety terminal blocks from safety related terminal blocks. .
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LLNL's inspection found that the tape did not extend to the point
where 6 inch separation or a metal barri6r exists between safety
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related and nonsafety-related cables. This configuration was
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installed, inspected, and accepted by Work Request 111991, closed
out on February 2, 1987. Subsequent to the above inspection
finding, inspected the similar panels H4FWA, H4FWC, and H4 FWD and
found problems of the same nature in these panels as well.
The district issued Nonconforming Reports 56754, and S6847 to document
and resolve the above LLNL inspection findings.
Cable separation problems inside of panels.are not confined to the EFIC
modification. In the TDI system, diesel generator engine control panels
H2 DEA 2 and H2 DEB 2 both contain safety and nonsafety vendor installed
wiring separated by only 4 inches of air space. Also, they contain
nonsafety-related wiring installed bundled with safety related vendor
installed wiring. This LLNL inspection finding was documented by
Nonconforming Report S6811, dated July 3, 1987. The installed wiring was
installed and inspected under Work Requests 115274 and 115276, both of
which were closed out on March 27, 1987. Neither work request included l
requirements or criteria for the inspection of wiring separation.
Except for the case of internal panel wiring separation, the inspected
areas of EFIC El&C construction exhibited acceptable workmanship and
conformed with design. The minor differences between drawings and the
as-built condition were:
ECN A-5415AA, DCN 0, E-342, Sheet 88, Rev. O. The terminations
identified on this drawing as "***" were made in the field using a
different combination of Raychem tubing than specified on the
drawing. This deviation was approved by FPR-6 to ECN A-5415AA,
however, the drawing was not revised to reflect this change.
ECN A-5415AH: DCN 0, I-1485, Sheet 1, Rev. 1., DCN0, I-1486,
Sheet 1, Rev. 1; DCN 0, I-1487, Sheet 1, Rev. 2; DCN 0, I-1488,
Sheet 1, Rev. 2. These drawings indicate that the OTSG water level
and pressure indicators use soldered terminal connections. The 3
installed units use screw-compression terminal connectors. !
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ECN A-5415AH: DCN 0, I-1485, Sheet 1, Rev. 1, DCN 0, I-1486, l
Sheet 1, Rev. 1. These drawings identify the control room EFIC 1
panel as numerical displays as three digit displays. The installed l
displays are four digit. l
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ECN A-5415AH, DCN 0, Sheet 3, E-402, Sheet 19A, Rev. O. The edges i
of the "EFIC Initiate / Test Matrix" label are not painted in I
accordance with the drawing.
None of the above deviations from design affect the functionality of EFIC
equipment.
LLNL's inspection of TDI EI&C construction indicated that the TDI I
construction exhibited more deviations from requirements than did the l
EFIC system. In one case a condition was noted which deviates from the i
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provisions for cable separation by voltage level described.in the Updated :
Safety Analysis Report (USAR).
Manhole'47, part of.the channel A underground raceway between the
TDI building and the NSEB, contains safety related and non-safety
related instrumentation cable intermingled with safety related power
cable. In one' case safety related instrumentation cable are laying
across 350 MCM, 480 VAC conductors. This is contrary to the
criteria for cable separation by voltage level described in
Rancho Seco USAR section 8.2.2.11.H.5. This section states " Power
and Control circuits are not mixed with instrumentation circuits in
any raceway in any system." LER 87-26 reported a similar but
separate occurrence of mixing power and instrumentation cables.
licensee engineering is taking action to address this finding.
(' In addition'to the above mentioned separation problems in the manhole and
diesel engine control panels, LLNL's inspection observed several
deviations from design documents or construction specifications in the
TDI building. Approximately one-half of the items inspected contained-
some deviation from requirements. The specific items noted are as
l follows. None of these were previously identified and documented by
L licensee inspections.
Three of the four mounting bolts attaching safety related
thermocouple TE55701 to it's Unistrut support are not snugged up to
l the thermocouple support base plate. The lock-washers were not
compressed and about 1/8 inch gap existed between the bolt heads and
the base plate. The bolts appeared to have been torqued, however,
the bolts bottomed'out against the base of the Unistrut to which
, they were attached before they seated against the thermocouple
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support's base. plate. Subsequent to LLNL's inspection, the licensee
documented these findings in NCR 56810.
The thermocouple extension wires for essential HVAC room temperature
thermocouple T#55701 were labeled such that the connections in the
thermocouple connection box appear to be reversed from the
connection drawing. Careful examination, however, showed that the
extension wires are properly connected,.but the wire number labels
are reversed. Licensee inspection noted also that the insulation of
the thermocouple extension wire was cracked. This wiring was
installed and inspected under Work Request 82910. The inspection
was conducted on October 4, 1984 as documented by the termination
cards. Licensee has documented LLNL's inspection findings in
NCR S6810.
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Essential HVAC damper actuator HV-55713 is shown on ECN-3748, DCN-0A
to drawing M17.07-6 as model number MNQ-12-8-1600, the installed
model number is MNQ-12-4-1600. Licensee documented this finding in
NCR 56813.
Essential HVAC air handling unit AH-DG-1B termination box contains
4 AWG triplex cable in which individual cables are bent tighter than
the allowable bend radius. The worst observed bend radius was about
3/4", the allowable specified by NPEM 8304.8C section 5.1.27 is 1.6
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inches. Installation and QC inspection of this item occurred under
Work Request 82910. QC inspection occurred on October 17, 1984 as
documented on the termination card. Licensee inspection determined i
that a similar condition exists in the terminal box of opposite j
train's air handling unit. Licensee has documented these inspection !
findings in NCR 56845. l
Two cases were noted in which tray edge bumpers were not installed
as required by Licensee procedure NPEM 5304.7.C section 5.2(12) to
protect cable entering over the edge of the tray. These cases are
in tray A57El where cable enters from conduit A57015, and tray L5782
where cable enters from conduit L57046. In the former case
deformation of the cable jacket has resulted from contact with the I
tray edge. In the later case the cable does not actually rest on ;
the tray lip because the cable hangs by rope from the tray above. l
This configuration is not addressed by licensee procedures, and it I
appears that the cable would rest on the tray edge if the rope were !
removed. These items were respectively installed and QC inspected j
under Work Requests 84414 and 82910. Inspections occurred on j
May 5, 1985, September 18, 1984, and September 19, 1984 as j
documented by the pull cards. There remains some question whether
the construction specifications required the installation of bumpers
at the time of this installation. Nevertheless, licensee has
prepared NCR S6844 to track resolution of these items. The above
two items, dealing with the bend radius of cables in the HVAC air
handling unit and the lack of tray edge bumpers noted above are j
considered together as an apparent violation. (87-20-02). 1
Pull box H7J2995 is labeled as " black", i.e. , non-safety, when, in 3
fact, it is part of the channel "A" raceway system. NPEM 5304.7C )
requires installation of appropriate separation group color
markings. Licensee has prepared NCR S6843 to document this finding.
Terminal box H7J2968 has a cable labeled as 1G1DIEA2B installed in
the location where DCN OF sheet 2, ECN A3748, drawing #344 sheet 17 {
shows cable 1G1DGA2V drawing. NCR S6846 was prepared to document
^
this finding.
In motor control center S2A4 cubicle 4, one of the wires to the i
starter contactor was terminated with a lug on which the barrel was l
bent backwards from the ring by about 90 degrees. This panel was
inspected and the condition accepted by licensee on
February 6-9, 1984, and on February 14, 1984 as documented by the
Source Inspection Data Report and the Receipt Inspection Data Report
for this motor control center. Licensee documented this finding in
NCR 56809.
Safety related terminal boxes H7J2967 and H7J2968 on the "A" TDI l
diesel skid contain several terminal lugs which do not have the i
required insulation crimp. Licensee has documented this inspection
finding in NCR 56846.
During walkdowns in the TDI building three " automotive type" electrical
"T" connectors were found adrif t in the HVAC mezzanine of the TDI
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building. This type of connector is not approved for plant use by the
licensee as it does not provide a reliable permanent connection. Even
temporary use damages conductor insulation. Inspection of nearby
terminal boxes failed to identify any evidence of their use. It is
possible, however, that the previously noted insulation damage of the
TE55701 extension wire may have been caused by this type of connector.
Licensee is evaluating the need for a program to determine if these kinds
of connectors are used at the plant, report on corrective action taken to
repair damage caused by these types of connectors, and establish controls
to prevent their use in the future.
The EFIC and TDI inspection also identified several items of potential !
concern. Licensee was, however, unable to produce documentation during ,
the inspection to resolve these items or confirm that they represent {
deviations from their design, safety analysis, or commitments to NRC.
These open items are:
The cable within manhole 47 is supported on cable racks held in ,
place only by the weight of the cable. In addition, the supports
for the racks are anchored with concrete expansion anchors which are )
l
not marked with torque stripes, and which, in some cases, do not
appear to be installed perpendicular to the concrete wall. During
the inspection the licensee was unable to provide documentation
which established the seismic qualification of this configuration. ;
If indeed this is not a seismically qualified installation this
would be contrary to USAR Appendix 5B section 58.2 which indicates
" seismic class I structures and components are... designed to
withstand appropriate seismic loads and other applicable loads i
without loss of function." Licensee Engineering is investigating {
this issue. J
Attachment bolts for Unistrut used as supports for safety related ,
conduit L57065, and A57081 appear to have bottomed out against the
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embedded Unistrut to which they attach. While the lockwashers are
compressed in these cases, that the bolts are bottomed out raises i
the question whether torquing generated the intended preload. !
ECN A-3748, DCN 5 to drawing C906 sheet 2 revision 1 identifies the I
torque requirements for these bolts. Followup inspection by
licensee determined that a number of similar cases exist in the TDI
building. Licensee is issuing an Engineering Action Request te
evaluate this potential problem.
Separation of redundant cables and separation of safety and
non-safety wiring within panels is sometimes provided using two
layers of Siltex or Fiberfrax tape or a single layer of Siltex or l
Fiberfrax tubing. These installations are in accordance with
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construction specification NPEM 5304.8C. Licensee, however, was not
able to produce test reports for the Fiberfrax to demonstrate that
this installation provides an adequate barrier. A portion of a test
report was available to demonstrate the capability of the Siltex
tape. Information was not, however, available to confirm that the
single layer of Siltex tubing provides protection equivalent to that
of two layers of tape.
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The inspection also' identified several items which are in accordance.with i
the design, and in compliance with site procedures and specifications yet-
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... i are. a. concern to the inspector and should be addressed by the licensee. i
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The following items will be tracked as one open item. (87-20-03). ;
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The channel A cable ducts from the NSEB to the Auxiliary Building
contain channel.C flex conduits laying on the A channel cable. !
Similarly, the' channel B ducts contain D channel flex conduits. !
This condition was previously identified by licensee to.the NRC i
Division of Nuclear Reactor Regulation (NRR). NRR is reviewing- j
SMUD's justification for this deviation.from the physical separation .i
guidance of Regulatory Guide 1.75. '
Procedure NPEM-5204.13, Revision 0, " Preparation of Schematic. l
Diagrams" requires labeling of electrical schematic drawings as ;l
" Nuclear Safety Related" if the drawing includes safety related !
circuits. Procedure NPEM-5205.1,~ Revision 0, " Instrumentation Loop
Diagrams", contains no such requirement. .The inconsiste.ncy.between '
the requirements for electrical:and instrumentation schematic
drawings may-cause confusion.in determining if instrumentation
circuits are safety related. D
-The human engineering review of the H1SS(E) panel was. performed on
the equivalent' simulator panel and on the de-energized control room
panel. While there are differences between the simulator panel and
the control room panel, there is no plan to review the energized
control room H1SS(E) panel against human factors design criteria.
Therefore, the HISS (E) panel will not be completely reviewed with
respect to those human engineering criteria which can only be-
evaluated using the energized panel.
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' Construction specification NPEM 5304.8C does not contain
requirements for separation of redundant' channels except within
equipment enclosures, or for separation of different voltage
classes. Therefore, QC inspectors have no clear basis for
inspection of cable separation in areas such as manholes or the
training of cables between raceway. Licensee procedures, in
general, do not acknowledge the licensee's commitment to meet the
intent of Regulatory Guide 1.75 for.TDI and EFIC construction.
Therefore, this attribute of the system design is not being
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inspected by QC.
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