ML20128D249
| ML20128D249 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/29/1993 |
| From: | Collins S, Mullikin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20128D229 | List: |
| References | |
| 50-298-93-01, 50-298-93-1, NUDOCS 9302100093 | |
| Download: ML20128D249 (28) | |
See also: IR 05000298/1993001
Text
- . -. - . . _ - .. . - . . - .__- -_ -. --
.
'
j,*
['
Anoendig
U.S. NUCLEAR REGULATORY COMISSION-
,
REGION IV
L
l NRC Inspection Report: 50-298/93-01
- -
Operating License: DPR-46
,
Licensee: Nebraska Public Power District- (NPPD)
, P.O. Box 499
Columbus, Nebraska 68702-0499
j Facility llame: Cooper Nuclear Station
.
Inspection At: Brownville, Nebraska
l Inspection Conducted: January 4-8, 1993
l Acting Team Leader: 7 1Z M /-2 f- f3
R. P. Mullikin, Acting. Tea 6 Leader -Date
Technical Support Section-
Division of Reactor Projects
i
leam Members: T. F. Westerman, Chief, Engineering-Section
Division of Reactor Safety-
j
,
R. K. Frahe, Jr., Vendor Inspector (Rotational Assignment)
4
Engineering'Section, Division of Reactor Safety
l P. A. Goldoerg, Reactor Inspector, Engineering Section
Division of Reactor Safety
o
- M.'F. Runyan, Reactor Inspector, Engineering Section
Division of Reactor safety.
l R.-B. Vickrey, Reactor Inspector, Engineering Section-
,
Division of Reactor Safety.
4
Accompanying Personnel: H. Rood, Project Manager, Office _.of Nuclear Reactor
! - Regulation
Approved: IN% i ~4 *t- 9~5
Date
y Samuel J. Collins, Director, Division of Reactor
Safety
L
r
!
>
1
'
93021000N 203
PDR ADO * JOO298
G PDR
I '*
_. _ . - . _ _ . . . _ . _ . . , - _ . . - - _ ,
_ _ _ _ _ - _ _ _ - - _ _ _ ____-_- _ ____ - - -
,
,. <
.
TABLE OF CONTENT 4
EASA -
EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . iii
DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I
1. ENGINEERING AND TECilNICAL SUPPORT ACTIVITIES . .-. . . . . . . 1
1.1 Design Chenges_and Modifications ............. -
I
1.1.1 Permanent Design Changes and Modifications .-. . . I
1.1.2 Temporary Modifications .............
-
6
1.1.3 Conclusions ...,............... 8-
1.2 Offsite Support Staff . . . . . . . . . . . . . . . . . . . -
8
1.2.1- Nuclear Power Group . . . . . . . . . . . . ._. . . 8
1.2.2 Nuclear Engineering and Construction Divisinn . . . 9
1.2.3 Nuclear Configuration Management Department . . . . 9
1.2.4 Nuclear Engineering Department .......... 10
1.2.5 Nuclear Projects and Construction Department- ... 13
'
1.2.6 Nuclear Fuels Department ............. 13
1.2.7 Cooper Nuclear Statf an Engineering Department . . . 13
1.2.8 Configuration Management ............. 15
1.2.9 Design Basis Program ............... 16
1.2.10 Engineering Disposition of Nonconformance Reports . 16
1.2.11 Audits and Assessments .............. 19
1.2.12 -Engineering Initiatives . . . . . . . . . . . . . .- 19
1.2.13 General Observations-Related to Engineering . . . . 19
1.2.14 Conclusions . . . . . . ............. 19
ATTACHKENT 1 - EXIT MEETING AND ATTENDEE 5
ATTACHMENT 2 - ENGINEERING STAFF GROWTH
ATTACHMENT 3 - ENGINEERING INITIATIVES
11
l
l
c- _ - - - - - - - - _ _ - - - - - -- _ . . _ ___ - - - - - - - - - - _ _ _ _ _ _ _ - - _
.__ _-_____- ____ _ __ _ _ __
-.
. c
,
.
EXECUTIVE SUMMARY
A team of NRC staff members conducted an inspection of engineering and
technical support at the Cooper Nuclear Station. The inspection was conducted
from January 4-3, 1993.
The NRC team utilized the guidance provided in NRC Inspection Procedures
37700, " Design Changes and Modifications," 37702, " Design Changes and
Modifications Program," and 40703, "Offsite Support Staff."
The inspection team reviewed the engineering organization for organizational
structure and interfaces, manpower and work backlogs, scheduling and
prioritization of work activities, support of plant operations, and
qualifications of personnel. The quality of engineering performance was
evaluated by reviewing completed station modification, design change work
packages, temporary modifications, and engineering disposition of
nonconformance reports. -The quality assurance audits and assessments of the
engineering and technical support organization, and the actions taken with
respect to the assessments and audit findings were reviewed.
The inspection team observed the following:
- The engineering organization (onsite and corporate) had grown from
approximately 27 engineers in 1980 to 154 engineers by 1993.
- In general, all design modification activities are performed
.within the licensee's organization with limited outside contractor
augmentation.
- The overall process to control projects and design modification
activities appeared to be very effective. The backlog of work was
sina11.
- Procedures for design changes and modifications were found to be
,
comprehensive and well written.
- The plant modification packages were found to be well written with good
safety evaluations.
- A great deal of engineering effort was incorporated into the
modification process. Conservative engineering practices were observed.
- The temporary modification process was found to be good. Particular
strengths were noted in the weekly audit performed by senior licensed
operators and the use and control of permanent temporary modification
tags. The use of temporary modifications was limited and were not left
in place over 6 months.
- The interface between corporate-engineering and site engineering-
appeared effective.
iii
_ _ _ _ _ _ _ _ - _ _ _ . ________ _-- . _ _ . - - _ _ _ _ - - _ . _ - - _ _ _ - _
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
_ _
_ _
,. :
- There was a very stable engineering staff with a low turnover
rate.
- Engineering personnel were qualified, trained, and their
responsibilities defined. Good morale was observed.
- Staff levels appeared consistent with the workload.
- Of particular note, with regard to qualification, was the emphasis
on shift technical advisor certification for system engineers.
- The development of the program engineering department was seen as
an enhancement to the Cooper Nuclear Station engineering
department (onsite engineering.)
- Engineering appeared to have good credibility and working
relationships within the licensee's organization.
'
- Configuration management was found to be effective.
In response to team observations, the licensee made the following comitments:
1. The licensee committed to annotate control room drawings and procedures
to indicate that an open temporary modifications affects those
documents. Inspection Followup Item 298/9301-01 was opened.
2. The licensee comitted to review the open temporary design changes
to determine if any could be rettaved or made permanent plant
reodifications. Inspection Followup Item 298/9301-02 was opened.
>
tv
L- _
.
. _ _ _ _ _ _ _ _ _ _____ _ _ . _ _ _ __ _ _ _ _ _ _ _ _ _ _ _
_ - _ _ - _ _ _ - _ - .. .. . . . .
._
. .
,
.
DETAILS
1 ENGINEERING AND TECHNICAL SUPPORT ACTIVITIES
1.1 Quian Ch1D90_IDitiodifications (37700 and 37702)
1.1.1 Permanent Design Changes and Modifications
The team examined six design modifications to verify that the design
modifications were in conformance with the requirements of the Technical
Specifications,10 CFR Part E0.59, the updated safety analysis report (USAR),
and applicable codes and standards. The design change packages reviewed were:
- DC 89-252B, " Trim Hodification to RHR-H0V-H034A"
- DC 90-174B, " Service Water Pump Bearing and Shaft Hodifications*
- DC 90-283, "CS-HOV-H026A&B and SW-HOV-H089A&B Rerate Due to Insufficient
Wall Thickness"
- DC 90-292, "No Break Power Panel Static Inverters"
- DC 90-302, "ASCO Transfer Switch Hodifications"
The team reviewed the licensee's process associated with plant modifications.
The governing procedure for all permanent modifications to structures,
systems, components or equipment located at Cooper Nuclear Station (CNS) was
CNS Operations Manual Engineering Procedure 3.4, " Station Modifications,"
Revision 14. This procedure described the station modification process and
provided an overview on the use of the subset of procedures required to
process a design change. In addition, the team reviewed Engineering
Procedure 3.4.3, " Design Change," Revision 3, and Instruction Number NED-24,
"DC Writer's Guide," Revision 5. Procedure 3.4.3 specified the method for
preparing a permanent design change at CNS, and the instruction provided
information concerning the format and items which should be addressed when
preparing a design change. The team found the procedures to be comprehensive
and well written.
The team found that the development of a design modification package and the
assignment to a design engineer formally began after an engineering work
request was approved. Prior to the preparation of the package, a scoping
meeting was held at the site with participants from design engineering and
various site eganizations. As the package was developed, a conceptual design
meeting was held with the site organizations to review the design criteria,
proposed design fix, and perform walkdowns. The design change package
prepared by the design engineer consisted of a description of the change,
design analysis, safety evaluation, installation procedure, and acceptance
1
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ -
. <
l
.
testing. Once the package was complete, a General Office disciplinary review
and CNS site review were performed. In addition, selective reviews would be
performed as determined necessary. The package was approved by engineering
and the station operations review committee prior to implementation.
1.1.1.1 Design Change DC 89-2528
The team reviewed DC 89-252B, " Trim Modification to RHR-H0V-M034A," which
installed anticavitation trim into essential Valve RHR-H0V-H034A. The valve
is an 18-inch globe valve manufactured by Anchcr/ Darling Valve Company. This
valve is the res11ual heat removal Loop A suppression pool cooling throttle
valve. During the 1990 refueling vutage, the same trim was installed in the
Loop B equivalent valve and was found to be performing in a satisfactory
manner.
The plant previously had determined that the wall thickness of
Valve RHR-H0V-M034A was being gradually reduced due to cavitation / erosion
taking place. The modification was developed to arevent further
cavitation / erosion of the valve walls to insure tlat the minimum allowable
wall thickness was not violated.
The valve tria design change was completed in October 1991. The modification
consisted of replacing the original disc and seat ring, which required
complete disassembly of the valve. The original seat ring was machined out
and the new design welded in place. Welding and nondestructive examination
were performed in accordance with the ASME Boiler and Pressure Vessel Code,
Section III, 1986 Edition. The new trim utilized stellite as the hardfacing
material. Acceptance testing was performed in November 1991, which consisted
of a residual heat removai Loop A flow test and an inservice leak test.
The team reviewed the evaluation performed in accordance with the provisions
of 10 CFR Part 50.59, as well as the ALARA review and the impact on the USAR
and Technical Specifications. The team found that the safety evaluation was
complete and weli written. The team noted that a great deal of engineering
effort had been incorporated into the modification and conservative
engineering practices had been utilized.
1.1.1.2 Design Change DC 90-1748
The team reviewed DC 90-174B, " Service Water Pump Bearing and Shaft
Modification," which was prepared to modify the bronze bearings, shaft
sleeves, and shafts in the four Byron Jackson service water pumps. This
modification consisted of replacing the original bronze bearings with rubber
bearings housed in stainless steel casings. This bearing replacement was done
since the rubber bearings were designed to allow debris up to 1/8 inch in
diameter to be flushed through the bearing without affecting its performance.
Since the rubber bearings were larger in annular area than the bronze bearings
f9r strength purposes, it was necessarv to remove the stainless steel shaft
sleeves at the bearing locations and hard face the shafts to minimize shaft
wear.
2
.
,,.-- _a - - m.,__.
- .. .- . . .-- - .- - .
'
. .
.
This modification was developed due to the difficulty of operation and the
high raintenance costs associated with maintaining the original service water
pump essential gland water injection system. The original dand water system
used cyclone separators to remove suspended solids. Because the components of ,
the essential gland water injection system were not no.mally operated, !
1
accumulation of particles in the components led to increased maintenance and
cleanup requirements. By using the rubber bearings which allowed debris to be
flushed through, this modification would allow the essential gland water
system to be simpitfied to delete the cyclone separators and allow the direct
use of the Missouri River water. This design change package only covered the
service water pumps' bearing and shaft modifications. The essential gland
seal modification was the subject of another design change package.
The rubber bearings and hardfaced shaft were initially installed on Service
Water Pump B and then tested for ar.ceptable performatice prior to modifying
Service Water Pumps A, C, and D. The licensee stated that, prior to the 1993
refueling outage, the remaining three service water pumps will have the
modification completed. The team performed an inspection of the physical
installation of the modification and found that the modification was installed
in accordance with the approved design.
The team found that a considerable effort had been made to address issues of
safety significance created by this modification. The team considered the
10 CFR Part 50.59 safety evaluation to be thorough-and well written. The team
also noted that the design change package reflected conservative engineering
practices.
1.1.1.3 Design Change DC 90-283
The team reviewed DC 90-283, "CS-MOV-M026A&B and SW-MOV-M089A&B Rerate Due to
Insufficialit hil Thickness." This design change rerated the core spray and
- service water valves from the original purchased ratings to ratings which
l corresponded to the design temperatures and pressures of the valves. This
- modification was a paper change only ard did not physically modify the valves
or-their systems.
The licensee had previously determined that the valves were below their
minimum wall thickness based on the ANSI Class they were purchased to. In
response to NRC Information Notice 89-01 concerning erosion in valves used for
throttling, the licensee had performed ultrasonic testing inspections and
found that the two service water valves and one of the core spray valves had
wall thicknesses below the minimum required for their ANSI Class. As a
result, nonconformance reports (NCRs) were generated to document the wall
thicknesses. The root causes for the insufficient wall thicknesses were
documented in the NCRs. The core spray valve was determined to have been
l originally supplied with a reduced wall thickness and was not subject to
erosion. However, the service water valves were determined to be susceptible
to erosion. .To prevent further erosion of the service water valves, the
licensee had applied an epoxy coating to the internals of the valves which was
documented in Special Test Procedure STP 89-173, " Application of Abrasion
l Control Putty on SW-MOV-M089A/M0898." The licensee is monitoring the epoxy
i
coating in accordance with Preventive Maintenance Procedure 06821.
3
. -. . .. . -. ,- .. .
-
. i
This design change rerated the core spray valves (CS-MOV-M26A&B) from ANSI
Class 300 to ANSI Class 250, and rerated the service water valves
(SW-MOV-M089A&B) from ANSI Class 300 to ANSI Class 180. This was accomplished
in accordance with ANSI B16.34, 1977 Edition. The design pressures and
temperatures of the valves were used to determine the intermediate ANSI Class
ratings and minimum wall thicknesses. The minimum ANSI Class ratings were
documented in Calculation NEDC 89-1908.
The team found that the safety evaluation and checklists were thorough and
well written. The team also noted that the assertions and assumptions were
documented and reflected conservative engineering practices.
1.1.1.4 Design Change DC 90-292
The team reviewed DC 90-292 which modified the 120/240-Vac power supply to the
no break power panel (NBPP) by replacing the Static Inverter IA, static
switch, and manual bypass switch. This design change also modified the power
supply to four reactor core isolation cooling system control components by
installing a separate static inverter to provide essential 120-Vac power.
CNS initiated this modification because they felt the existing inverters and
associated components, which were original plant equipment, had an increased
likelihood of failure as the equipment continued to age. Failure of a
component in the inverter would likely cause a plant trip because a
significant number of instruments and controls derive their power from the
NBPP. The design change would minimize the likelihood of static inverter
failure and hence provide a more reliable power source to the NBPP.
The design basis for the inverter and NBPP was to provide a uninterruptible
power source for nonessential components that are critical for plant operation
and was unaffected by this modification. Performance of this design change
necessitated a change to the USAR to reflect the NBPP in its final
,
l
l configuration, but did not require a change to the Technical Specifications.
The modification was reviewed and approved by the appropriate technical
disciplines prior to installation. The team reviewed the safety evaluation
' and 10 CFR Part 50.59 reportability analysis and considered them to have been
very detailed and well researched. All of the additional detailed evaluations
were made as part of the design change package, and the listing of affected
.
document changes including drawings, procedures, vendor manuals, calculations,
and the USAR appeared complete. A license change request form was completed
'
to assure a description of the change would be included in the annual report
'
because a USAR change is reportable under 10 CFR Part 50.59(b).
- Installation and acceptance testing of this design change was completed during
the 1991 refueling outage. The team walked down portions of the physical
osta11ation of the modification and found no problems. The team viewed that
..iis modification thoroughly addressed safety considerations and was well
prepared and installed,
4
l l
<
. .
,
1.1.1.5 Design Change DC 90-302
The team reviewed DC 90-302 which removed the automatic transfer function from
the ASCO Model 935-307 transfer switches by physically removing the
unnecessary circuitry and components. The modification also added isolation
fuses in each of the transfer switches to allow local (manual) de-energizing
of the control circuit and provide electrical isolation of the nonessential
control power from the essential busses. Transformers were also installed to
lower the control circuitry voltage to 120 Vac to reduce potential safety
hazards during maintenance activities.
This design change was initiated to close out Temporary Design Change
TDC 90-300 which electrically removed the automatic transfer function from
Transfer Switch EE-MCC-X(IA) and exchanged the normal and emergency feedere to
provide a more appropriately distributed diesel loading. This design chm;;c
removed the disconnected components from that switch and performed a similar
modification on several other ASCO transfer switches. The control logic in
these transfer switches had been defeated procedurally by de-energizing the "
emergency feeders to the switches.
The design basis for the transfer switches, which was to provide a continuous
source of power, was maintained. This design change required a minor
modification to two figures in the USAR, but did not require any changes in
the Technical Specifications. The modification was reviewed and approved by
the appropriate technical areas prior to insti.llation. The safety evaluation
and 10 CFR Part 50.59 reportability analysis were completed and analyzed. All
of the additional detailed evaluations were ?ade as part of the design change
package, and the listing of affected document changes including drawings,
procedures, vendor manuals, calculations, and the USAR appeared complete. A
license change request form was completed to assure a description of the
change would be included in the annual report because a USAR change is
reportable under 10 CFR Part 50.59(b).
Installation and acceptance testing of the modification was completed during
the 1991 refueling outage. The team inspected the physical installation of
the modification for several of the transfer switches and found that they were
installed in accordance with the approved design change. The team concluded
that this design change package was clearly documented and implemented with a
great deal of engineering consideration of safety concerns.
1.1.1.6 Design Change DC 92-141B
The team reviewed DC 92-141B which modified the control power circuit for the
low pressure coolant injection (LPCI) and reactor recirculation pump discharge
valves. The contro~l power supplied to the starters was from a 125-Vdc source,
and this modification changed the source to the 250-Vdc motive power which was
already in each starter. The modification involved replacement of the starter
contact coils and control circuit fuses with components rated at a nominal
250-Vdc. It also included removal of starting resistors and associated
relays, installation of contactor coil surge suppression resistors and relays,
and miscellaneous wiring changes.
5
l
l
_ _ _ _ _ __ _ _____________ __________._______.________.__ _____ ______ _ ___.___ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _
' .
.
_
.
This design change permitted control of LPCI injection and reactor
recirculation discharge valves to be independent of the 125-Vdc system. The
need for this modification was discovered during CNS's design basis review
efforts, when a failure moda was identi?ied in the 125-Vdc system which would
be more limiting than the previously analyzed LPCI injection valve failure.
Completion of the modification resolved the concerns related to the LPCI
single failure vulnerabilities and returned CNS to its original @ sign bases.
The loss of one division of 250-Vdc power is similar to the loss of one LPCI
valve and had already been aalyzed and documented in the USAR, so the
determination was made that this design change did not require a change to the
USAR or Technical Specifications. The modification was reviewed and approved
by several technical areas including systems, operations, maintenance,
radiation )rotectirtn, and quality assurance prior to installation. The team
reviewed tie safety evaluation and 10 CFR Part 50.59 reportability analysis
and considered then to have been quite thorough and well written. Additional
detailed evaluations were made as part of the design change package to
determine effects on ALARA, fire protection / Appendix R, inservice
inspection / testing, human factors, environmental qualification, Regulatory
Guide 1.97, emergency operating procedules, motor operated valves, and others.
The packaga also included a detailed listing of affected document changes
including drawings, procedures, vendor manuals, and calculations.
Installation and acceptance testing of the modification was completed in
September 1992, while the plant was in cold shutdown. Five changes were made
to the original modification package and were documented by on-the-spot change
sheets as detailed in Engineering Procedure 3.4.10. " Revisions, Amendments,
and On-The-Spot Changes," Revision 2. The team concluded that this design
change package was well prepared and reviewed by CNS to identify and address
all potential issues of safety significance created by the modification.
1.1.2 Temporary Modifications
fhe team reviewed six ,,lant temporary modifications. The temporary
modifications reviewed were:
- PTM-92-05 Leak Repair Clamp on HPCI Steam Line Dripleg
Drain Line to Main Cendenser
- PTM-92-06 Pipe Cap on B-5 Heater Inlet Drain Valve MC-V-524
- PTM-92-09 Lifted Leads on MS-LS-101 and 102 Due to Grounds
- PTM-92-12 Pipe Cap on Drain Valve DGD0-V-44
- PTM-92-14 Mechanical Jumper on Service Water Radiation Monitor
- PTM-92-15 Reactor Recirculation Pump Motor Bearing Vibration
Setpoint Change
6
4
. .
,
4
The team reviewed Conduct of Operations Procedure 2.0.7, " Plant Temporary
Modifications Control," Revision 14, and Engineering Procedure 3.4.4,
" Temporary Design Change." Revision 4. The team concluded that the procedures
properly provided for the control of temporary modifications to safety-related
plant systems as required by the Technical Specifications and
10 CFR Part 50.59. These procedures applied to both safety-related and
nonsafety-related temporary modifications. The licensee attempted to limit
the number of temporary modifications by utilizing the permanent design change
process for major scope temporary modifications. There were relatively few
temporary modifications installed during the past 2 years (17 in 1992 and 22
in 1991). However, the licensee utilizes temporary design changes, which are
temporary modifications installed greater than 6 months. The use of temporary
design changes had the benefit of drawing and procedure updates. There were
19 temporary design changes still open during the time of the inspection, of
which 3 were installed in 1986. The team reviewed these temporary design
changes and found, although minor in significance, that several could possibly
be removed or made permanent plant modifications. The licensee committed to
review the open temporary design changes to determine which could be removed
or made permanent. The review of the licensee's review of temporary design
changes is an inspection followup item. (50-298/9301-01)
The team reviewed the six temt irary modifications and found the
10 CFR Part 50.59 evaluations to be good. The modifications had proper review
and approval by the licensee. The team noted that control room drawings
affected by temporary modifications were not procedurally required to be
annotated in any manner to indicate an outstanding temporary modification.
This created the possibility of an operator making a decision using a drawing
that was not indicative of the actual plant configuration. However, the
temporary modifications installed that afi'ected control room drawings would
have very little safety significance. In addition, there were a small number
of open temporary modifications and the shift supervisors interviewed were
aware of the temporary modifications. The licensee committed to enhance the
temporary modification process by requiring that control room drawings end
! procedures be annotated to reflect that a temporary modification affected that
!
document. The team's review of the licensee's enhancements to the temporary
modification program is an inspection followup item. (50-298/9301-02)
The licensee performs various audits of the teroorary modification process.
The operations supervisor or designee performed a monthly administrative audit
of the control room temporary modification log. This audit mainly determined
.
'
i which modifiestions were no longer needed and which ones were to be made
temporary design changes. A weekly audit was performed by the shift
,
supervisor or a . senior reactor operator. This audit was a physical walkdown
I
to verify that the installation is as described by the temporary modification
package. The team accompanied the shift supervisor on his weekly audit and
found no discrepancies with the field installation. The team noted two
strengths from this audit. The verification was very detailed and would
,
provide excellent assurance of proper plcnt configuration of the temporary
modification. Secondly, the licensee utilizes permanent plastic temporary
modification tags which were visibly attached near the modification. These
tags were noted by number in the control room log and must be returned to the
control room efter the temporary modification is removed. The tags were
i
7
- . .- . . -
.
' .
.
audited to verify that all tags are accounted for. This provided for added
assurance that closed temporary modifications are actually removed.
The team also rufewed the results of recent quality assurance surveillances
of the temporary modification and temporary design change processes, and found
that a good review was parformed by the-licensee. The team also reviewed a
computer printout describing all the safety-related maintenance work requests
generated in 1992. The team found no instances where a maintenance work-
request was used in lieu of a temporary modification, temporary design change,
or a perwanent plant design change.
1.1.3 Conclusiong
- Procedures for design changes and modifications were found to be
comprehensive and well written.
- The plant modification packages were found to be well written with good
safety evaluations.
- A great deal of engineering effort was apparently incorporated into the
modification process. Conservative engineering practices were apparent.
- The temporary modification process was found to be good. Particular
strengths ware noted in the weekly audit performed by senior licensed
operators ar.d the use and control of permanent temporary modification
-
tags.
- Enhancements to the temporary modification and temporary design
j changes process were noted. The licensee committed to annotate
, control room drawings and procedures to indicate the existence of
'
an open temporary modification. Also committed to was the review
of open tusporary design changes-to determine if any could be
removed or made permanent plant modifications.
1.2 Offsite Suonort Staff (40703)
1.2.1 Nuclear Power Group (NPG)
The engineering and technical support functions for CNS were an integral part
of NPG,-located both in the NPPD General Office in Columbus, Nebraska and
onsite. All functions of NPG reported via the NPG Manager, located in the
NPPD General Office, to the Vice President - Production. Reporting to the NPG
Manager were the CNS Site Manager - Nuclear Operations Division, the Division
Manager of Nuclear Engineering & Construction, the Division Manager of Quality
Assurance, and the Division' Manager of Nuclear Support. The specific
responsibilities of each manager were outlined in functional job descriptions.
NPG Directive 1.2 stated that within NPG a single integrated organization is
provided for the safe and efficient operation, maintenance, modification, and
support of CNS. The inspection team observed that the-staff within NPG has
evolved from around 280 in 1985 to a present allocated staff of 569 with
8
-- ~. - _ _ _ . - _ . _.
.
' ,
.
around 10 vacancies. This number did not include a contract security force of
around 85, general office personnel of around 100, and outside contractors.
The use of contractors was observed by the inspection team to be generally
limited (indicated presently to be 13). Engineering (onsite and corporate)
had evolved from approximately 27 engineers in 1980 to 154 engineers by 1993.
Attachment 2 is a giaphical chart of engineering staff growth from 13 NPPD
deneral Office and 14 CNS engineers in 1980 to 97 NPPD General Office and 57
CNS engineers in 1993. The overall organization appeared stable with a very
low turnover rate.
Engineering functions within NPG were principally performed by the nuclear
engineering and construction (NE&C) division in the General Offices of NPPD
and onsite by the CNS engineering group. The NE&C division has established a
site engineering group to improve their interface and support of the site.
The CNS engineering group reported to the CNS Site Manager.
The hierarchy of procedures observed by the inspectors consisted of NPG
directives (establishing corporate policy and direction), CNS policy
. directives (site), NE&C division procedures, and HE&C division guidelines.
1.2.2 NE&C Division
The NE&C division consisted of the nuclear configuration management
department, the nuclear enginewring department, the nuclear projects and
construction department, and the nuclear fuels department. The inspection
'
team found that there were 102 engineers and technicians allocated to this
division with five vacancies.
-
The inspection team found that major activities and planning were set forth in
, a 5-year business plan that is updated annually. Goals and objectives are
also assigned to each division. The implementation of the goals and
objectives are reviewed quarterly and a status issued by each division. The
inspectors observed that the goals and objectives established for the NE&C
,
division provided for continued enhancements to the NESC division. Goals
'
established included the completion of all modification packages and
procurement 30 days prior to a planned refueling outage and 90 days prior to
work that does not require an outage. An outage scope list was established by
the work planning and ianagement committee (composed of corporate and site
managers) for each refueling cycle. The outage scope list established the
engineering activities to be accomplished during each cycle. Licensee
management indicated that the outage scope list for the next fuel cycle was
near completion. The inspectors found that a monthly NED status of activities
l for NED was maintained. There was also a NE&C division programs book for
!
setting priorities and tracking engineering initiatives and programs. Project
managers issued a monthly report. The overall scheduling and prioritization
of work was computer tracked by the nuclear item tracking (NAIT) or the CNS
action and comitment tracking (ACT) systems. Licensee management has '
indicated that changing the CNS fuel cycle to 18 months over the last two - 4
cycles has provided engineering with more time to plan and complete their
engineering activities. Work backlogs have essentially been eliminated.
9
l
. - .
4 .
,.
.
1.2.3 Nuclear Configuration Management Department !
!
The nuclear configuration manaDement department had three major groups. These
consisted of the configuration management group, the PRA and engineering
review group, and the design basis group. There were 12 personnel allocated
with no vacancies.
The configuration management group was a special group that had been verifying
the as-built configuration of the 245 safety-related drawings utilized by the
,
o>erators in the control room and the 1300 related drawings. Supervision in
t11s group indicated that approximately 98.2 percent of control room drawings
-
have been upgraded. ,
The PRA engineering review group had responsibility for the. site PRA related
,
activities and engineering reviews. The status of the individual plant
examination (IPE) program was reviewed by the team. The program was also
~
l
supported by NED personnel, the CNS operation support group, and CNS system
engineers. Science Applications international Corporation (SAIC) has been the
'
primary contractor. Engineering & Research, Inc.-(ERIN) has been contracted
and was providing prog am advice and third party review. The licensee stated
'
that preliminary results indicate that there was-no significant vulnerability.
The core damage frequency for CNS is 7.9/ E-5. The licensee indicated that
the core damage frequency was dominated by station blackout (35 percent) and .
transient induced LOCA (30 percent). Their IPE (PRA Level I and II) submittal
in response to Generic Letter OL 88-20, " Individual Plant Examination of
Severe Accident Vulnerabilities," was scheduled for March 1993. This group
will continue to have the responsiFility to maintain the PRA. The utilization
of PRA was indicated to have been limited due to the concentration of effort
to complete the IPE. The future uttiization of PRA within the licensee's
, activities was under development. The IPEEE (PRA Level III) was being
. developed by the project management group and was sr.heduled for submittal in
i June 1994.
The design basis group has responsibility for the design basis program. This
'
program was observed by the team to have been through a significant
evolutionary process. The NPPD first pilot program began in 1987. There were
,
several starts and stops in the program. All work was stopped as late as
September 1992 following issuance of the August 10, 1992, NRC Polic' Statement
which was issued in conjunction with SECY 91-364 (November 12, 199.j and
. SECY 92-193 (May 26, 1992). The scope of the program was to complete 21
system design criteria documents and five topical design criteria documents.
. The diesel generator design criteria document (one of three documents
completed) was reviewed by the team and appeared to be of good quality. The
.
scheduled completion was in 1997. The licensee planned to complete the
'
program in house and maintain the expertise within NPPD. The licensee has
indicated that the plant design appears to be " robust." The identification of
the recent control wiring routing in a non-Class IE area which could have
rendered both emergency diesel generators inoperable was observed by the team
to be indicative of the present level of review.
10
_ _
>
. .
l
1.2.4 Nuclear Er,gineering Department (NED)
The NED was responsible for the dest n of the plant, maintaining the plant
designbasis,andprovidingengineer$ngsupporttotheplantonanas-needed i
basis. The NED organization included instrumentation and control (!&C),
electrical, civil / structural, mechanical, aid technical support disciplines
based in the NPPD General Office, as well as a site engineering group based at
the CNS site. The technical support group focused on regulatory issues so
that the other engineering disciplines could concentrate on plant
modifications. NED was composed of 69 people including 46 degreed engineers,
21 technicians, and 2 secretaries.
The first four groups identified above performed all design modification
activities for NPPD in each of the identified engineering disciplines. The
inspectors found that the licensee performed all design work with in-house
personnel, with limited utilization of outside contractors t3 provide staff
augmentation when necessary. When assigned the lead, a design package would
be completely developed by the assigned engineer with the assistance of the
other discipline groups as required. This same engineer would go to the site
for full coverage of the modification during installation as part of the site
engineering group. An engineer may be assigned to work as a part of a
projects group, but is technically responsible for his work to the supervisor
for his discipline. The inspectors observed that there were some safety-
related software programs used by the civil / structural groups in performing
Class I seismic analysis and some within the electrical engineering group. In
discussions with the supervisors in NED, the inspectors found that the
licensee has been able to hire quality angineers with high grade point
averages. Most successful hires have been from engineers coming back to the
area. In accordance with the NPPD training program description, technical
training was provided as )osition required, task required, and optional. In <
addition, most engineers 1 ave Ld system training. Additional general
training was recently completed for all engineers in response to a third party
audit. All engineers were encouraged to obtain their professional engineer's
license. A monthly salary bonus is provided to all engineers with a
professional engineer's license.
The technical support group performed special studies, responded to some NRC
requests such as generic letters, was responsible for fire protection long-
term compitance, performed engineering reviews, coordinated procurement for
modifications, and was responsible for the development of engineering
directives, policies, and procedures.
The site engineering group provided the onsite interface with NED and
coordinated all onsite NED engineering activities (particularly during
refueling or other outages). There was also a dedicated group of onsite
draftsmen that assisted with walkdowns and drawing upgradas.
The team interviewed 20 NED engineers assigned to the I&C, electrical,
11
i
, _ - . _ . _ _ . .
'
. .
,
civil / structural, mechanical, and technical support disciplines to evaluate
the functionality of the NED staff. The engineers were knowledgeable of their
res)onsibilities and interfaces as well as the availability and use of
tec1nical information. Those interviered appeared very pleased with the
current operation of NED and many emphasized that recent improvements were
effective, including the addition of the NED site manager and the increased
availability of computers.
The team determined that staffing levels of each discipline appeared to be
consistent with the workload. Between outages, the design engineers worked
little or no overtime. During outager., however, the design engineers
typically worked six 12-hour days, e ,secially if they were the responsible
engineer for any of the modifications being installed. NED developed the !
large majority of their own modifications, and the small portion delegated to
contractors were reviewed by NED engineers. l
4
The team found that the NED staff's average nuc1'.c experience was 6 years,
and 19 of the NED engineers had obtained their pwfessional certification.
The staff also had available tuition assistance for job-related education, but !
few engineers participated due to the lack of local opportunities. The NED l
had a low attrition rate which could indicate that the engineers had a great
deal of job satisfaction.
The team noted that the design engineers were respensible for their assigned
modifications from conception to completion. The design engtheers stated that
they were not involved in day-to-day site activities and that those activities
were the responsibility of the plant systems engineers. There appeared to be
little if any duplication of effort within the NED or with plant systems
engineering.
The team observed that interfaces between the NED and site personnel appeared
effective despite the fact that the NED design engineers were stationed at the
General Office. The NED design engineers stated that they made frequent
visits to the site as necessary when preparing their design change packages,
perhaps once or twice a month, and were predominately at the site during
outages when their :nodifications were being installed. Management encouraged
the design engineers to spend time at the site, and made available two
airplanes and numerous company cars for transportation. The scoping and
conce)tual design meetings in the early stages of the design change process
also 1elped to improve communications with plant systems engineering,
operations, maintenance, and other site groups. The addition of the NED site
manager and his staff also greatly increased the effectiveness of
communications between the NED and site personnel.
The training program for NED employees appeared to be effective. Subject
matter included root cause analysis,10 TR Part 50.59 reviews, a boiling
water reactor system overview, industry codes and standards, the CNS Technical
Specifications, the USAR, and numerous others. Some of the engineers lacked
training in operability determinations and regulatory requirements, but were
not directly responsible for these activities and were cognizant of how to
find the requirements if necessary.
12
-
>
. .
,
The team discovered that there were several tools available to the design
engineers. Most of the engineers had personal computers on their desks with
word processing and other capabilities. They also had access to a main frame
computer which provided information on component data, action item tracking,
nonconformance report (NCR) tracking, drawing numbers, purchase orders, spare
parts inventory, microfilm record addresses and others. There were
controlled sets of vendor manuals, applicable procedures, Technical
Spacifications, and the USAR readily available. There were only a few design
basis documents issued to date, but the engineers felt they would be a useful
tool in the future.
The team concluded that employee morale was high and management support was
strong. NED personnel were qualified, trained, and their responsibilities
defined. The team considered the staffing levels to be consistent with the
work load and interfaces between NED and site personnel to be sufficient.
Overall, the team viewed NPPD as having an effective design engineering
program.
1.2.5 Nuclear Projects and Construction Department
Within this department was the nuclear projects group and construction group.
There were four project managers and a construction manager. A total 14
allocated positions, including two secretaries formed this department. The
team found that the )rojects group operated as a matrix organization with a
project manager in c1arge. Each project manager normally had several projects
assigned. Normally, the more major projects assigned can draw from all of NPG
for personnel to perform within the project. The design, procurement,
installation and testing of the hard vent was one of the projects assigned to
this group.
The construction group was responsible for providing onsite craft personnel to
accomplish both outage and nonoutage activities.
1.2.6 Nuclear Fuels Department
Within this department were three allocated positions, all associated with
'
nuclear fuels procurement. The manager position was vacant. The team fou~i
'
that all engineering and reload analysis was performed by General Electric
(GE). This department provided oversight of GE. As a part of this oversight,
staff members participated in an annual audit of GE in the fuels area.
1.2.7 CNS Engineering Department
The CNS engineering department was the )rimry onsite engineering resource,
consisting of three departments. The t1ree departments were operations,
plant, and programs engineering, with each reporting to a department
supervisor. Also, three senior engineers reported to an assistant engineering
manager. The supervisors and the assistant manager reported to the
engineering manager, who reported to the plant manager. Additionally, smaller
staffs of engineers were assigned to the maintenance department, the technical
staff department, the quality assurance department, and the onsite design
engineering department.
13
= _ _ _
-
- , - -
- - . . - _ _ _ _ . _ . . _ ___ . . - _ _ -
.
.o .
.
The system engineers were divided into two departments. The operations
engineering department consisted of 10 budgeted engineers with no vacancies.
This department was responsible for power production systems including the
reactor, turbine, and associated support systems. The engineers were divided
into three groups; reactor, operations, and performance engineering. Each
group reported via a lead engineer to the operations department supervisor.
The plant engineering department consisted of 13 budgeted engineers with no
vacancies. This department was responsible for safety systems incit ding
reactor protection, containment isolation, em^rgency power, and emergency core
cooling. The engineers were divided into two groups; electrical /I&C and
mechanical engineering. Each group reported via a lead engineer to the plant
engineering supervisor.
The main functions of the system engineers were:
- Observing surveillance tests
- Monitoring equipment performance
- Reviewing maintenance work req 1tsts ato postmaintenance testing
- Generating maintenance work request special instructions
- Technical evaluation of temporary modifications
Of the 70 systems in the plant, all were assigned to system engineers in the
system engineering departments.
The team found that systems engineers were given extensive systems training,
and received a detailed system checkout on their assigned systems as a part of
their qualification arocess. The team also found th'it, in addition to other
rsquired training, w11ch included root cause and safety evaluation tra,ining,
system engineers were allnwed the opportunity to attend other scheduled
courses such as maintenance training and career enhancement courses. Of
particular note was the licensee's efforts to strengthen system engineering
with shift technical advisor certification of system engineers. The team
observed that the system engineering staff had experienced a very low turnover
- rate. The licensee appeared to have a goed development program for system
engineers.
'
The licensee had developed an extensive trending program. This program was
under the administration of the assistant engineering manager. The program
tracked over 2000 individual tr# nds. System re) orts were updated quarterly
and a quarterly adverse trend rsport was publisied. The system was readily
j available to individual system engineers through the local area network
l system.
The programs engineering department consisted of eight budgeted engineers,
with one vacancy. This department was responsible for the tuhnical direction
of 21 progr!.ms. The department was composed of mechanical and electrical
l engineers who were respon;ible for the various programs under engineering such
l as ASME, Ap)endix J, environmental qualification, procurement, etc. This
department 1ad been initiated to focus engineering programs into one
department. Since its inception, a large portion of the resources in this
14
.- .__ -
.
- .
,
'
department have been devoted to upgrading these programs. The department
responsibilities included: direct technical su> port of maintenance, project
management / program development, independent tecinical overview / review, and
specialized technical expertise.
The CNS engineering action items were tracked, scheduled, and prioritized
through a computerized database (ACT). Regulatory compliance actions items
were also independently tracked by a separate database (NAIT). The status of
critical items was discussed daily in manager and supervisor staff meetings.
These mechanisms were used to ensure that action items were identified and
completed in a timely manner and that management was aware of action item
status and emerging issues. 1he team concluded that these mechanisms had been
effective in reducing previous backlogs and maintaining subsequent bachlogs at
a reasonably low lavnl.
The taam concluded that the CNS engineering department training program was a
strength in development of an effective engineering program. Of particular
note was the licensee's emphasis on shift technical advisor certification of
system engineers. The team viewed the engineering department's coordination
of related activities as superior. The licensee's initiative in the
development of the program engineering department wa: seen as an enhancement
to the CNS engineering departoeat. The team found the engineering department
engineers had developed very good credibility and working relationships
throughout the licensee's organization. The team found a strong sense of
ownership of systems by the system engineers.
1.2.8 Configuration Management
The team reviewed the NPPD 1rocess for controlling CNS design changes to
determine the adequacy of t1eir configuration management program. The nuclear
power group directive tiefining and describing the program was documented in
" Configuration Management," Operations Section 4.4, Revision 1. CNS
criginated drawing changes were documented on the drawing change notice (DCN)
form and forwarded to the record; administration section (RAS) as described in
Engineering Procedure 3.7, " Drawing Change Notico," Revision 7. RAS verified
duplication of the change, assigned DCN control numbers, issu'ed DCN
transmittals to all drawing custodians in accordance with the distribution
list, and tracked the status of the DCNs to assure timeliness of
incorporation. The team noted that NPPD seemed to incorporate design changes
in a tirrely manner with a minimal backlog.
The team reviewed the controlled drawings that )rovide the control room
operators the necessary information to assure tie plant raaintains safe
operation. Control room drawings (CRDs) were maintained by the CRD custodian
in accordance with Operations Instruction Number 3 " Control Room Orawings,"
Revition 3. When a modification was approved which would affect a CRD, the
CRD custodian stamped the affected drawing to inform the operators that the
drawing was subject to revision. When the modification was installed, the CRD
custodian would receive a status report wi'.h a marked-up interim drawing
attached as detailed in Engineering Procedure 3.4.11, " Status Reports,"
Revision 3. The CRD custodian then filed the interim drawings and highlighted
the stamp on the affected drawing to signify the design change installation
15
l
. - ,
-
-
)
> .
.
.
and the availability of interim drawings. Once the DCN was issued, the CRD
custodian stamped the affected drawings to reference the DCN number, filed the
DCN, and discarded the interim drawings. U)on receipt of the revised drawing
transmittals, the CRD custodian discarded tie old drawing revision and
replaced it with the new revision. Tne CRD custodian also performed a
quarterly audit of the CRD drawings and drawing files to assure accuracy. The
team spot checked several drawings to verify proper revision and status and
found no discrepancies. The team concluded that control room drawings were
controlled.
The team also reviewed the program for ensuring the accuracy of vendor
manuals. The engineering programs department (EPD) maintained the vendor
manual control system as described in Engineering Procedure 3.11 " Vendor
Contact for Verification of Manuals," Revision 4. The EPD contacted
applicable vendors yearly for essential components, and every two years for
nonessential components, to assure the CNS controlled vendor manuals were the
latest revision pertinent to equipment installed at CNS. CNS utilized a
vendor manual change form to assure manuals were evaluated and updated as
necessary based on information originating either within or outside NPPD. The
program requirements for this activity were contained in Engineering
Procedures 3.11.1, " Generating and Dispositioning Vendor Manual Change Forms,"
Revision 2, and 3.11.2 " Vendor Manual Update Control," Revision 1. The team
viewed CNS's program for updating vendor manuals as very good, but did not
verify the implementation of the program,
1.2.9 Design Basis Program
CNS's " Design Basis Program Description," dated November 30, 1992, documented
the )regram and assigned responsibility for its develo) ment and implementation
to tie nuclear configuration management department. T1e purpose of the
program was to document the design basis in design criteria documents (DCDs)
for the identified safety systems, to ensure that the physical plant and
design documents avset the design basis. CNS identified several uses for the
DCDs including serving as a reference to support operability and reportability
evaluations, safety reviews, Technical Specification reviews, and
design /setpoint reviews.
The current procedure to control the process for preparatlon, review, and
maintenance of the DCDs was Procedure NECDP-03 " Design Criteria Document
Production," Revision 1. Design basis teams were developed with
representatives from NED, licensing, operations, and system engineering to
coordinate the review and approval of the DCDs within their departments. The
procedure included an extensive listing of design documents which should be
reviewed for applicebility when preparing DCDs. A logic diagram would then be
produced which identified the system safety objective, design criteria, design
requirements, and critical components to serve as a template for the system
design basis development.
Individuals which identified design basis information that deviated from, or
should be included in the DCD, were instructed to complete a change request'
form to assure incorporation into the DCDs. Procedure NECDP-09, " Evaluation
of Open Items Identified during DCD Production," Revision 2, provided controls
16
.
. . .
4
for processing and resolving open items and discrepr.ncies identified during
the production and maintenance of DCDs. The team concluded that the use of
DCDs is clearly defined and measures have been implemented to maintain the
documents current to accurately reflect plant modifications.
1.2.10 Engineering Disposition of Nonconformance Reports (NCRs)
in addition to developing design modification packages. HED provided support
to the plant by evaluating a small percentage of NCRs that are specifically
directed to them for disposition. The team reviewed the following NCRs that
fell into this category:
- 91-045 Loss of Voltage Relay Setting
- 91-055 Diesel Generator Fan Coil Unit Supports not Seismically
Qualified
- 91-075 Inadvertent Isolation of Reactor Water Cleanup System
- 91-077 Loss of 161 kV Power Due to Operator Error
- 91-099 Diesel Generator Manual Start Problems
- 91-128 Potential Overheat of Service Water Pumps Due To Loss Of
Nonessential HVAC
Each of the NCRs that were reviewed showed extensive effort and conservative
judgement. Of special note, a root cause analysis was performed for each NCR
and was used in conjunction with a study of previous similar NCRs to develop
an effective preventive action plan. The team determined that the actions
taken for each of the NCRs and the level of documentation provided in the
finished document were appropriate. The team interviewed several of the
engineers responsible for the disposition of the reviewed NCRs and determined
that a high degree of professionalism and expertise was evident. Overall, the
effort provided by the NED engineers to address the reviewed NCRs was
exemplary.
To assess the quality of support provided by the site engineering group for
plant operations, the team selected for review several NCRs and deficiency
,
reports that had been directed to site engineering for disposition. The NCRs
'
that were selected are listed below:
- 91-050 Service Water Pump low Discharge Pressure
o 91-074 Indications on Reactor Pressure Vessel Head Studs
17
l
l
,
,,
. . . _ .
.
- .
,
.
- 91-093 Use of a Nonessential Mechanical Seal
i * 91-103 Safety Relief Valve Lift Pressures Out of Tolerance
- 91-118 Motor Operator loosened and Rotated on a Yalve
- 92-016 Temperature Switch Setpoint Out Of Tolerance
- 92-070 Olesel Generator Fuel Oil flash Puint less Than Minimum
Specified ,
1
- 92-102 Suppression Chamber Air Temperature Detector failure
Similar to the NCRs performed by NED, the above NCRs dispositioned by the site
engineering group showed extensive effort and conservative judgement. A
detailed and well-conceived root cause analysis was provided for each NCR
along with a review of previous NCRs to detect potential repetitive problems.
- In all cases, the immediate corrective and long-term preventive actions
, a)peared appropriate. .To gather additional information on some of the NCRs, ,
'
tie team interviewed the responsible engineers. Two observations were
discussed with the licensee. NCR 92-102 documented a failure of a suppression
chamber air temperature detector due to a power supply failure. The detectors
and associated power supplies were manufactured by Honeywell. The power
supply failed due to a failed diode and capacitor, apparently an.end-of-life
consideration. At least one previous Honeywell power supply failure was due
to a similar cause. The licensee had considered the failure rate to be too
. sporadic to justify generic action (i.e., replacement), but the team was
sensitive to the potential for an increasing failure rate due to a shared end-
of-life (20 years since installation) condition. The team did not expect the
licensee to take imediate action on this observation but to be aware of the
potential implications. The other observation concerned NCR 91-103 which
documented the fact that all eight of the main steam safety relief valves had
been found to lift at pressures ranging from 1.7 to 13.3 percent above the
,
Technical Specification setpoint. The allowed toler;nce was only
+/- 1 percent. The licensee detennined the root cause to be corrosion bonding
and labyrinth seal friction. An analysis had been perforo d showing that the
as-found lift pressures would not have compromised the safety of the plant.
Corrective action on this problem was being held in abeyance pending future
initiatives anticipated from the boiling water reactors owners group. The
team did not identify a concern with the proposed corrective action plan, but
did note that the licensee had considered the reporting of this event to be
voluntary.
The team reviewed the following deficiency reports that had been directed to
the site engineering group for dispositioning:
- 92-025 Small Boric Piping Does Not Meet Code Allowable Stress
18
_ - . -
--- . -. .- - . - - _ _ . - . - . _ - - - - - - -
s . o .
,
- 92-028 Service Water d/p indicator failure
- 92-029 Battery Cell Voltage Low ,
- 92-035 U-Bolt Support Loose i
l
The team considered the corrective and preventive actions taken for the above
deficiencies to be proper. Additionally, none of the reviewed deficiency
reports appeared to be of sufficient severity to be more appropriately
categorized as an NCR. The deficiency report process was implemented in
October 1992,
i
1.2.11 Audits and Assessments l
The overviev function of engineering activities was the responsibility of the
'
Division Manager of Quality Assurance. This organization performed audits, ;
surveillances, safety system functional inspections (SSrls), and self
- assessments.
!
i The inspector observed that an audit of design control was performed annually.
The inspector reviewed the findings of the audits performed in 1991 and 1992.
i The 1992 audit report iridicated that 236 surveillances and in-line reviews
- were performed. The report indicated that 69 were )erformance bued. Only
'
one finding was issued. The 1992 audit concluded t1at the functional area of
-
design control was being effectively implemented. The licensee indicated that
job task training was being given to the auditors. More performance Leased
i audits and surveillances were being planned and conducted. One SSFI was
performed of the core spray system. The use of SSF!s has been limited. In
accordance with the NPG self-assessment program, c self assessment of
electrical distribution system was performed in 1991 which concluded'that tha
- electrical distribution system and related support systems at CNS met its
-
intended design functions. More self-assessment activities were being
- planned. A self assessment of shutdown risk activities was just being
completed. A self assessment of the service water system was being scheduled.
The licensee was encouraged by the team to consider conducting a self
.ssessment of engineering. .
i 1.2.12 Engineering Initiatives
The licensee provided the team with a list of 35 NPG engineering initiatives ,
(Attachment 3). The team observed that the initiatives were being implemented ,
in a number of different projects throughout the organization.-
1.2.13 General Observations Related to Engineering
The team observed that the licensee has established a program in response'to
the NRC program for systematic asse:ssment of licensee performance (SALP). The
program provided for a quarterly status of the enhancement and actions taken
in response to the SALP identified issues. The program document stated that
nuclear safety is of paramount importance, that rising standards of
performance are necessary to achieve and maintain excellence of performance,
19
- -. - --, -. . . - - , - - . . . . - - , -
- - . _ _ . - - - . -
,
_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ .___ _ _ _ _ _ .. _ ._ _-
.=,
.
.
and the importance to develop and sustain a self critical, questioning
attitude among all employees. The team reviewed the engineering and technical
'
, support section of the quarterly status report and found that each of the
l issues identified in the previous NRC SALP report have or were being responded
l to.
1.2.14 Conclusions l
- There was a very stable engineering staff (onsite and corporate) with a
low tiirnover rate, which had evolved from approximately 27 engineers in
1980 to 154 engineers by 1993. q
- In general all design modification activities are performed within
the itcensco's organization with limited outside contractor
augmentation.
l
- -The overall process to control projects and design modification
activities appears to be very effective. The backlog cf work is i
small.
- The design basis program has been through a significant evolution.
The recent identification of design basis issues was indicative of
the effectiveness of the program.
- The individual clant examination for PRA Levels I and II was t
. scheduled to be'sutaitted in March 1993. The licensee indicated
that no significant vulnerabilities have been identified.
- The interface between corporate engineering'and site engineering
appeared effective. .
- Employee morale was high and management support was strong.
!
'
- Engineering personnel were qualified, trained,- ar.d their
responsibilities defined.
- Staff levels appeared consistent with the workload.
- Of particular note, with regard to qualification, was the emphasis ,
en shift technical advisor certification-for system engineers.
- iLe development of the program engineering-department was seen as -
u enhancement to the CNS engineering department.
- .12gineering had good credibility and working relationships within -
m e licensee's organization.
- fanfiguration management was found to be effective.
.
20-
__
_ , - - - _ .,m . . - ,m , , , _ . , , . . . . . _ , - - , ,m.- ._
- . . m.__.... . - - , , ~ . , ~ , . - , , . . . - _ . . - . - - - , _ . - . . - . ~ . . . _ . . - , ~ , , - . - . _ , . - .
'
, a ,
,
'
ATTACHMENT 1
1 PERSONS CONTACTED
1.1 Licensee Personnel
R. Augspurger, Lead Civil / Structural Civil Engineer
A. Boesch, I&C Engineering Supervisor l
M. Borce, Design Basis Superviser
J. Branch, PRA and Engineering Review Supervisor
- L. Bray, Regulatory Compliance Specialist !
D. Busan, Lead Electrical Engineer
R. Bussard, I&C Electrical Engineer
K. Curry, Mechanical Engineer
J. Dewyke, Mechanical Engineer
K. Done, Mechanical Engineering Supervisor
T. Duren, Civil / Structural Engineer
- J. Dutton, Nuclear Training Manager
W. Fisher, Electrical Engineering Supervisor
- J. Flaherty, Cooper NJClear Station Engineering Manager
- R. Foust, Assistant Engineering Mar,ager
- S. Freborg, Plant Engineering Supervisor
- R. Gardner, Plant Manager
M. H111strom. Technical Support Supervisor
- G. Horn, Nuclear Power Group Manager
M. Kennedy, Electrical Engineer
S. Kochanowicz, Mechanical Engineer
L. Kohles, Nuclear Projects and Construction Manager
E. Lanning, Acting Nuclear Fuels Manager
- E. Hace, Senior Manager, Site Support
M. Mager, Senior Mechanical Engineer
- G. McClure, Nuclear Engineering Department Manager
T. McC1'Jre, Hechanical Engineer
D. McManaman, Electrical Engineer
B. McMillen, Lead I&C Electrical Engineer
- J. Neacham, Site Manager
- C. Moeller, Technical Staff Manager
J. Murphy, Acting Nuclear Projects and Construction Manager
- S. Peterson, Senior Manager of Operations
R. Rexroad, Electrical Engineer
- D. Robinson, Quality Assurance Manager
M. Siedlik, Civil / Structural Supervisor
- G. Smith, Licensing Manager
- M. Spencer, Engineering Programs Supervisor
- W. Swantz, Project Manager
S. Thompson, Lead Electrical Enginaer
G. Tillotson, Civil / Structural Engineer
J. Ullmann, Configuration Management Supervisor
K. Walden, Nuclear Configuration Management Department Manager
- R. Wenzl, Muclear Engineering Department Site Manager
A. Wiese, lead Mechanical Engineer
1
-
. ' <* ,
,
4
- R. Wilbur, Nuclear Engineering and Construction Division Manager
B. Wilcox, Lead Mechanical Engineer
J. Wright, Mechanical Engineer
B. Vants, IAC Electrical Engineer
1.2 NRC Periknn.tl
- S. Collins, Division Director, Division of Reactor Safety, Region IV I
- J. Gagliardo, Chief, Reactor Projects Section C, Region IV
- W. Walker, Resident Inspector
in addition to the personnel listed above, the inspectors enntacted other
personnel during this inspection period.
- Denotes personnel that attended the exit meeting.
4 # Denotes personnel that attended the exit meeting via telephone-from Corporate
Headquarters ,
l
2 EXII REETING l
i
An exit meeting was conducted on January 8,1993. - During this meeting, the
team reviewed the scope and~ findings of the report. The licensoe did not
identify, as proprietary, any information provided to or reviewed by the
inspector.
,
.
>
2
- =
v ,r ,-
-m,*7 , ,- -- y- em g7 9, &r--w4Wgg-qg.-wp-6,,,,,gyy a wy _y=,,4 y--w w.m ,---.+.w,gw=, 4. ggygyt-'cg- 'v.. g -e.+eagw r vwes ev'
, l-; 1l{ - 7 ) I
,
= .
'
it5o$G m
1 1
1
9 m.s
.
,
,
a
2
0 o
s
o " c
c
2
0
.-
0
_
6
0
-
a
n
.
8
0 e l __
-
- T - - -
-
-
-
,-
.
8 :
-
1 . (
S
e
e
_W '_#
-
E
r
8
2
- va -
f -
/
,
8
3
% a
- n
d
C
G
e 8:
u
i
va
E
gnN
4
n
o t
r er
d
A
i u .
O
m
o 8:
p
p nc _
_
e :
5 l
i
c el e
.
t
.k,
e
r a
ir
O
8
6
s -
s
v gPn
C
s.
N o
N
S ,D
E Sw
t
8
7
'
C
o
ae
fr
f
n
e g
5
u
r Gr G
ro
'
8
8
- a
o
wpu
t
i
o
n
A.
. t
T 8 :
9
p'h
o
t
a
l
9 : , .t
0
P
r
o
/
j
e
_
ct
_
_
9 : - s
1
a
- # d
n _
9 :
2
- j )s
F
u
e
l
w /
'
9 : '
3
_
e
- l .
y ; . !; ! ' ;; ; ! . *, !
I I, .
O ek 4
ATTACHMENT 3
1
NPG ENGINIiERING INITIATIVES
'
- Electrical Calculation Updates
- Dapper Program Voltage Drop Study
- Instrumentation Setpoints Program
- Appendix "R" Cable Sepamtion Database
- Pire Protection Program
- Individual Plant Evaluation
- Source Term Resolution
- Steady State Core hWel ,
- Spare Parts Bar Coding Program
- Diesel Generator Owner's Group
- Motor Operated Valve Program
- Erosion / Corrosion Program Upgrade
- Procurement Engineering
- As Building Program
- Design Change Writer's Guide
- Technical Support Training
- Vibration Monitoring
- Vendor Mt.nual Program Upgrade
- Equipment Spare Pstts List
- ISI Program Upgrade
- Turbine Six Year Plan
- Core Plate Plug Lifetime Evaluation
- Check Valve Program Development
- Trending Program Development
- Non-Essential Instrument Air Project
- GL 89-13, Service Water System
- IST Program Upgrade
- Fuse Control Program
- CGI Dedication Program Upgrade
- DG Spare Parts Upgrade
- Snubber Program Upgrade
- ILRT/LLRT Program Upgrade
- SNM Inventory Program Upgrade
- Design Basis Document Program
!
l
l arnt.epsi
, -. , , -. , -. ,