ML20133C113
| ML20133C113 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/31/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20133C088 | List: |
| References | |
| 50-482-96-21, NUDOCS 9701070037 | |
| Download: ML20133C113 (55) | |
See also: IR 05000482/1996021
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ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.: 50-48.2
License No.: NPF-42
Report No.: 50-482/96-21
Licensee: Wolf Creek Nuclear Operating Corporation
Facility: Wolf Creek Generating Station
Location: 1550 Oxen Lane, NE
Burlington, Kansas
Dates: October 7-11 and 21-25,1996
Team Leader: J. Tedrow, Senior Resident inspector
Inspectors: R. Azua, Project Engineer
P. Campbell, Mechanical Engineer
M. Fallin, Consultant, Scientech, Inc.
P. Goldberg, Reactor inspector
F. Ringwald, Senior Resident inspector
J. Stone, Project Manager
Approved By: C. VanDenburgh, Chief, Engineering Branch
Division of Reactor Safety
Attachment: Supplemental Information
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9701070037 961231
l PDR ADOCK 05000482
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TABLE OF CONTENTS ,
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EXEC UTIVE S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv I
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R e p o rt D e t a il s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 1 j
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111. En g i n e e ri n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l
El- Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
E1.1 G eneral Comme nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i
E1.2 Permanent Plant Modification Review . . . . . . . . . . . . . . . . . . . . . 1
E1.3 Temporary Plant Modification Review . . . . . . . . . . . . . . . . . . . . . 4 {
E1.4 Review of Engineering Calculations . . . . . . . . . . . . . . . . . . . . . . 5
E1.5 Review of Performance Improvement Requests . . . . . . . . . . . . . . 6
E1.6 Work Pac k age Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
E2 Engineering Support of Facilities and Equipment .................. 7
E2.1 General Com ments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
E2.2 Review of Facility and Equipment Conformance to the Final
Safety Analysis Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
E2.3 10 CFR 50.59 lmplementation . . . . . . . . . . . . . . . . . . . . . . . . . 10
E2.4 Unsupported Operability Determination . . . . . . . . . . . . . . . . . . 19 ;
E2.5 System Walkdowns (37550) . . . . . . . . . . . . . . . . . . . . . . . . . . 20 l
E2.6 Engineering Work Backlog . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
E2.7 Surveillance Te sting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
E2.8 Industry Event Assessment and Lessons Learned ........... 27
E3 Engineering Procedures and Documentation .................... 28
E3.1 Review of Design Basis Documents .................... 28
E5 Engineering Staff Training and Qualification .................... 29
E5.1 System Engineering Staff Training and Qualification . . . . . . . . . . 29
E6 Engineering Organization and Administration . . . . . . . . . . . . . . . . . . . . 30
E6.1 System Engine ering . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
E6.2 Desig n Enginee ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . . 33
E8 Miscellaneous Engineering Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
E8.1 (Closed) Inspection Followup Item 50-482/9504-03: Use of
gear operator stop nut for actuator braking . . . . . . . . . . . . . . . . 33
E8.2 (Closed) Licensee Event Report 50-482/96001: Loss of
circulating water due to icing on traveling screens . . . . . . . . . . . 34
E8.3 (Closed) Licensec Event Report 50-482/96002: Loss of
essential service water train due to !cing on trash racks . . . . . . . 34
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V. M a n a g e m e nt M e e tin g s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
X1 Exit M eeting Sum m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
ATTACHMENT: Supplemental Information
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EXECUTIVE SUMMARY
Wolf Creek Generating Station
NRC Inspection Report 50-482/96-21
This team inspection evaluated the effectiveness of the Wolf Creek system and design !
engineering organizations to respond to routine and reactive site activities which included I
the identification and resolution of technical problems. The performance of safety and
operability evaluations, and self-assessment activities were also included in this inspection.
Enaineerina
- The inspection team found that modification packages included appropriate safety l
evaluations, and appropriately specified post-modification testing. In addition, !
associated drawings and procedures were generally updated as required, and the
engineering calculations were satisfactory. However, the inspection identified a I
design control violation regarding the use of outdated calculations for capping
containment air cooler tubes. In addition, the team considered the licensee's
control of the design basis information to support the safety function of the
emergency core cooling system to properly operate following a postulated internal l
missile generation and impact to be poor (Sections E1.2 a.'d E1.4).
- The inspection team determined that the administrative procedures that the licensee
had developed for the review and evaluation of changes in accordance with
10 CFR 50.59 were appropriate. However, the team found numerous discrepancies
between the Updated Safety Analysis Report and the actual plant conditions and
identified problems in the licensee's implementation of the 10 CFR 50.59 review
process. The team identified one apparent violation involving four examples, which
were indicative of a programmatic breakdown in the control of this activity. These
examples involved. (1) the Operation nf the essential service water self-cleaning
strainer backwash setpoint differently than described in the Updated Safety
Analysis Report, (2) the performance of inservice inspection and testing of the ,
reactor coolant pump flywheel examination differently than described in the l
Technical Specifications, (3) the performance of underground pressure testing of I
essential service water piping differently than described by the Updated Safety
Analysis Report, and (4) the performance of a safety evaluation regarding changing
the main turbine overspeed protection test frequency without performing sufficient
evaluation to conclude that an unreviewed safety question was not involved
(Sections E2.2, E.2.3, and E2.7).
- Although the licensee's cc,rrective action for a 1993 quality assurance audit required
the performance of a 10 CFR 50.59 screening of Technical Specification
clarifications, the screening did not identify potential conflicts between the
Technical Specifications and the clarifications. Specifically, the licensee screenings i
of nine Technical Specification clarifications, which were performed to resolve the i
concerns of the quality assurance audit, failed to determine that these clarifications
involved unauthorized changes to the Technical Specification requirements. In
addition, a followup quality assurance audit failed to recognize that the conditions
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found during the original audit were not corrected. This f ailure was identified as an
apparent violation involving inadequate corrective action. The inspectors also noted
that the screenings of the Technical Specification clarifications were subsequently
reviewed by the Plant Safety Review Committee, and they also failed to identify the
issues involving Technical Specification compliance (Section E2.3).
- Based on the number of findings in the 10 CFR 50.59 area and the recent
indications of improper screenings for Updated Safety Analysis Report change
requests, the team concluded that training did not appear to have been effective in
avoiding continuing deficiencies (Section E2.3).
- The team identified that a shift supervisor violated the licensee's administrative
procedures regarding operability determinations when he relied, in part, on an
out-of-date calculation. Previous examples identified by NRC inspectors indicated a
declining trend in the performance of on-shift operability determinations
(Section E2.4).
- The team found that housekeeping was generally very good and noted that the
material condition of system components had little evidence of boric acid leakage
and few deficiencies. A very good threshold for deficiency identification had been
established. However, the inspection team identified that system walkdowns by
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the safety injection system engineers did not include all plant areas where system
components were located (Section E2.5).
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- The team considered temporary shielding controls to be weak because they did not
require an engineering review of erected temporary shielding and periodic
inspections of installed temporary shielding. Ir- Jdition, the residual heat removal
system engineer was not knowledgeable of the condition of temporary shielding,
even though it had been installed for several years (Section E2.5).
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- The licensee managed the engineering open item workload appropriately, but the
licensee did not have a formal program to control the backlog. The inspectors were
concerned that the program had a high threshold for backlog criteria, and f ailed to
trend the impact on engineering personnel workload (Section E2.6).
- In general, the inspection team found that surveillance tests for the systems
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selected had been accomplished in accordance with Technical Specification
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requirements and were performed at the correct periodicity. However, the team
identified one violation associated with an inadequate procedure to verify {
emergency core cooling throttle valve mechanical position stops (Section E2.7). '
- Uncontrolled and out-of-date design basis notebooks hindered the licensee's control
of design basis information. The licensee's control of design basis information was
found to be weak, in that, it did not provide a central location for the design basis
information. In general, licensee personnel had difficulty retrieving some design
basis information (Section E3.1).
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Although system engineering knowledge was excellent, it appeared to be the result
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of the personalinitiative taken by system engineers and their immediate supervisors, '
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and not due to any specific management guidance or administrative requirement.
j Training guidance was found to be very general and did not provide a minimum
standard for system engineer training or knowledge. Overall, licensee management
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communication of system engineering expectations has improved; however, the
weaknesses identified in the previous NRC engineering inspection in May 1995, had
not been corrected (Sections E5.1 and E6.1).
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Report Details
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E1- Conduct of Engineering
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E1.1 General Comments (37550)
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Using Inspection Procedure 37550, the team reviewed three safety-related systems
, to verify the licensee's ability to maintain these systems in an operable status. The
l three systems reviewed were: (1) essential service water, (2) residual decay heat
- removal, and (3) safety injection. The team reviewed the adequacy of the
- licensee's plant modification processes (permanent and temporary), engineering
! calculations, performance improvement requests, and documentation of work
performed on system components.
E1.2 Permanent Plant Modification Review
a. Inspection Scooe
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i The team reviewed several safety-related plant modification records listed in the
attachment to verify conformance with applicable installation and testing
i requirements as prescribed by procedures. Specific attributes reviewed and/or
i verified by the team included: (1) 10 CFR 50.59 safety evaluations, (2) post-
i modification testing requirements, (3) safety-related drawing updates, (4) Updated
j Safety Analysis Report updates, (5) training requirements, and (6) field installation.
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b. Observations and Findinos
in general, the team found the modification packages reviewed included appropriate
safety evaluations. The specified post-modification testing in the modification j
packages was appropriate and associated drawings and procedures were genera!!y )
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updated as required.
Outdated Calculations Used for Cabbino Containment Air Cooler Tubes
The essential service water system supplies the containment air coolers under
accident conditions. The system contains four coolers total, with two coolers for
each of two safety-related trains of essential service water. Each cooler has
12 coils with 32 circuits of 6 multiple passes, totaling 2304 tubes per cooler.
' The team reviewed Configuration Change Package CCP-07111, Revision 0, which
was initiated on October 17,1996, to address a leaking tube which had developed
in one of the 12 cooler coils in Containment Air Cooler SGN-01C, one of the two in
the A train of essential service water. The package was issued to assess the effect l
of plugging (or capping) the tube and continuing to use the cooler. A 7-day action
statement was entered on October 17,1996, and an engineering review was
initiated. The assessment for this change concluded that up to 64 tubes could be
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plugged based on Calculations SA-90-030, CWR-02424-90, and GN-MW-005. The
team noted that these calculations used a flow rate of 2000 gpm through each .
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cooler instead of more recent calculations which were based on a flow rate of
1000 gpm through each cooler.
Change Package CCP-07111, Revision 1, was issued, and approved by the Plant !
Safety Review Comn ntee, on October 18,1996, because cooler SGN-01C ,
continued to have leakage problems. Plans were to install a blind flange on the
supply header flange and on the return header flange to the leaking coil. The one
l affected coil was to be abandoned in place until it could be replaced. The change
l package stated that the removal of one coil bundle,32 circuits, would reduce total
- flow through the containment cooler pair by a maximum of 2 percent and
referenced Calculation GN-MW-005, Revision O. The change package also stated -
that the removal of one coil bundle would reduce the heat transfer capacity
Contair..nent Coolers SGN01 A and SGN01C, by approximately 1/24, which was
previously analyzed under Calculation SA-90-030. Calculations SA-90-030, dated
April 23,1990, and GN-MW-005, dated April 25,1990, used a flow rate of
2000 gpm per cooler (4000 gpm per pair of coolers).
Change Package CCP-07111, Revision 2, was issued, and approved by the Plant
Safety Review Committee, on October 20,1996, when a second coil on cooler
SGN-01C developed a leak. The package stated that one objective was to allow up
to three cooling coils to be blanked off if needed. The package stated that the
removal of one coil bundle,32 circuits, will reduce total flow through the
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containment cooler pair by a maximum of 2 percent for a total of 6 percent with
three coils removed and again referenced Calculation GN-MW OOS, Revision O. The
change package also stated that a sensitivity study was performed to determine the
effect of degraded performance of containment coolers on the containment pressure
and temperature response fc!!owing a postulated main steam line break accident. c
The change pukege referenced Calculation SA-90-025, dated April 9,1990, which
stau ubed 2000 gpm flow through each cooler, for this sensitivity study.
Subsequent to these calculations, the licensee had identified that the essential
service water system total flow had degraded due to erosion and corrosion in the
system and was concerned that the analyzed flow rate to the containment air
coolers, along with other cooling loads, may not be assured. Calculation
SA-90-057, dated November 1990, determined the containment peak temperature
and pressure that.would resul+. if the capacity of the containment air coolers were
assumed to be only 45 percent of the original capacity due to a reduction in the
flow rate through each cooler from 2000 to 1000 gpm, or 4000 gpm per train to
2000 gpm per train. The calculation supported Technical Specification Amendment
50, issued November 4,1991, which changed the required minimum flow rate
specified in Technical Specification 4.6.2.3.b from 4000 gpm per cooler group to
2000 gpm per cooler group. Calculation SA-90-057 concluded that sufficient heat
removal capability existed with the lower flow rate.
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The licensee's most recent flow balancing of the essential service water system
was conducted in the 1994 refueling outage and set the measured flows, by
throttling valves to the desired position, as follows:
Train A: Cooler SGN01 A 1022 gpm
Cooler SGN01C 1034 gpm
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Train 8: Cooler SGN018 1150 gpm l
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Cooler SGN01D 1440 gpm
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The team determined that Calculations GN-MW-005, SA-90-025, and SA-90-30 did
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not reflect the current operation of the coolers (i.e.,1000 gpm current flow versus I'
2000 gpm flow) and predated the calculation for 1000 gpm and the subsequent
. Technical Specification change. Both Revisions 1 and 2 of Change Package
CCP-07111 included an unreviewed safety question determination concluding that
the removal of three coils from service did not constitute an unreviewed safety
question. The conclusion was based on the outdated calculations discussed above.
None of the referenced calculations based on a 2000 gpm flow rate for each cooler
3 were denoted as either out-of-date or as not reflecting the current configuration of
- the equipment. However, the essential service water system engineer, who
- coordinated the efforts, was aware that the flow rate had been reduced to j
approximately 1000 gpm per cooler subsequent to the Technical Specification l
amendment.
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Performance Improvement Request PIR-962669 was initiated on October 20,1996,
based on questions from the Plant Safety Review Committee on the 10 CFR 50.59
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safety determination associated with Change Package CCP-07111, Revision 2. In
this improvement request, the difference in the margins between the capacity of the
coole:3 with 1000 gpm versus 2000 gpm was explained, and the impact of
blocking three coils was addressed. The improvement request concluded that the
containment peak pressure would not be exceeded based on Calculation SA-90-057
results. As of October 25,1996, Change Package CCP-07111, Revision 2, had not
been revised to reference the design information that reflected current operation of
the coolers with a flow rate of 1000 gpm each (or 2000 gpm flow rate per a group
of two coolers). However, the team considered the information provided in the
improvement requests addressed the current operability conclusion of the cooiers
with the blocked coils (2 of 12 in the C cooler).
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10 CFR 50, Appendix B, Criterion 111, requires, in part, that rneasures be established l
to assure that regulatory requirements and the design basis are correctly translated
into specifications, drawings, procedures, and instructions. These measures shall
include provisions to assure that appropriate quality standards are specified and
included in design documents. The suitability of continued use of Containment Air
Cooler SGN-01C with 2 of 12 coils blocked from essential service water flow,
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assessed in Change Package CCP-07111, was determined based on calculations
that did not reflect the current operating configuration of the equipment (i.e., the
reduction in flow requirements from 4000 gpm per cooler group to 2000 gpm per
cooler group), which is considered to be a violation of 10 CFR 50, Appendix B,
Criterion lli (50-482/96021-01).
Licensee management stated that they considered references to outdated
calculations and information to be acceptable as long as current data was utilized in
present calculations. The team recognized that the licensee could have used the
calculations based on 2000 gpm flow per cooler as a comparison analysis for
1000 gpm flow per cooler if the engineering analysis had stated such,
c. Conclusions
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In general, the team found the modification packages reviewed included appropriate
safety evaluations. The specified post-modification testing in the modification
packages was appropriate and associated drawings and procedures were generally ,
updated as required. The team identified one violation regarding the use of !
outdated calculations for capping containment air cooler tubes. '
E1.3 Temocrarv Plant Modification Review
a. Inspection Scope
The team reviewed a number of the licensee's active safety-related temporary
modifications listed in the Attachment. This effort was performed to verify that
these modifications were in conformance with plant procedures. In addition, ;
nonsafety-related temporary modifications were also reviewed to determine if they
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were appropriately categorized, and if 10 CFR 50.59 evaluations were appropriately
performed,
b. Observations and Findinas
The team identified that the licensee had only 14 temporary modifications installed
in the plant. Of these modifications, five were identified as safety related. The
team reviewed these temporary modifications against the requirements of
Administrative Procedure AP 211-001," Temporary Modifications," Reviainn 1, and
did not note any discrepancies. Affected procedures and drawings were also
reviewed to determine if appropriate changes were annotated. No problems were
noted.
The licensee had assigned an engineering supervisor to monitor temporary ;
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modifications in the plant. The licensee maintained a computerized log of these
modifications, with assigned durations. The team interviewed the engineering
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supervisor and found him to be cognizant of the temporary modifications installed in
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the plant. The team noted that this effort was designed to identify those temporary
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plant modifications that could be easily removed or corrected, and to make sure that
long term corrective actions were applied to the remaining temporary modifications
in a reasonable time,
c. Conclusions
The licensee efforts in reducing the number of temporary modifications in the plant
have been very successful.
E1.4 Review of Enaineerina Calculations
( a. Insoection Scope
The team reviewed the adequacy of several design engineering calculations listed in
the Attachment associated with the three subject systems to determine whether the
calculation assumptions were technically reasonable Lnd properly supported.
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b. Observations and Findinas
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The team found that the licensee's calculations were satisfactory. The calculations
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revbwed provided sufficient information and assumptions to reach the conclusion
stated. ".; *aam found some minor mistakes in the calculations regarding the
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correct atmospheric pressure for the elevation of the plant, and conversion of pump
horsepower to heat transferred to the coolant system, which did not adversely
affect the calculation's conclusion. Licensee personnel were informed of these
mistakes for correction.
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Inadeauate Suocort of Desian Basis
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The team reviewed Calculation IMS-01, " Missiles," Revision 0, to verify a statement
in the Updated Safety Analysis Report, Section 6 31.1, regarding the design bases
for the emergency core cooling system. The Updated Safety Analysis Report
contained general information that stated the system was designed to withstand the
effect of generated missiles. The calculation also contained an unlisted attachment
which listed the summary of rotating equipment in safety-related areas, by room
number. This attachment utilized Resolutions (1) and (2) which stated that room
- coolers and pumps were not considered to be credible missile sources based on
"The Internal Missile Hazards Analysis Program Overview," items B.4.C and B.4.A.
The team requested these documents for review, but the licensee was unable to
locate or retrieve them during the inspection. No other documentation was
available to justify these assumptions. The team was, therefore, unable to
determine if the design of the system was adequate to support the system's safety
, function under postulated generated missiles.
On November 8,1996, the licensee obtained the missing information from the
architect-engineer. These documents were provided to the team on November 12,
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1996. The documents were hand-written and contained justification for omitting
the pumps as credible missile sources due to the thickness of the pump casings.
The licensee stated that they disagreed with the inspection team's finding, in that,
the missing information was not part of the design bases of the plant and,
therefore, need not be readily available. The team noted that the missing
, information was an element of the licensing basis for the emergency core cooling
system as described in the Updated Safety Analysis Report, Section 6.3.1.1,in
Safety Design Basis Two. Since the design basis includes information identifying ,
the specific safety functions of the system and supporting analysis for reference l
bounds for the system design, the team considered the plant design basis to be
inadequately supported without this documentation. Due to the difficulty the
licensee experienced with retrieving this information, the team considered the
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licensee's control of this design information to be poor.
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c. Conclusions
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in general, the calculations were found to be satisfactory. The control of the design I
basis information to support the safety function of the emergency core cooling
system to properly operate following a postulated internal missile generation and
impact was considered to be poor.
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E1.5 Review of Performance Imorovement Reauests
a. Inspection Scope
l The licensee issued performance improvement requests as a means to identify
problems with components and systerns and to place these problems in their
corrective action system for resolution. The team reviewed performance
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improvement requests listed in the Attachment associated with the three subject
systems to determine the adequacy of the resolution, whether the systems'
operability was properly determined, and that the proposed corrective actions were
adequate to preclude recurrence.
b. Observations and Findinas
1 The team found that the performance improvement requests had resolutions with
. proper engineering justification and that the proposed corrective actions were
adequate to preclude recurrence.
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4 E1.6 Work Packaae Review
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a. inspection Scope
The team reviewed work packages listed in the Attachment associated with the
three subject systems, and work history printouts, to determine if repetitive
problems existed and to determine the present material condition of the system.
This information was compared with the results of the system walkdowns.
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b. Observations and Findinas
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- The team found that the work packages were performed in accordance with their
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- No recurrent problems were noted. The team's walkdown results indicated that the
[ licensee was maintainm9 the systems in good condition and a very low threshold
for deficiency identification had been established.
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E2 Engineering Support of Facilities and Equipment
E2.1 General Comments (37550)
To ascertain engineering support of plant activities, the team walked down the
selected systems with the system engineer, reviewed the system description
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provided in the Updated Safety Analysis Report, compared the Updated Safety
Analysis Report description with design basis information, evaluated the engineering
work backlog, compared surveillance testing records and test . >cedures with
design basis information and Technical Specifications, and re- swed the engineering
disposition of selected industry events for lessons learned.
E2.2 Review of Facility and Ecuioment Conformance to the Final Safety Analvsis Reoort
Descriotion
a. Insoection Scoce
A recent discovery of a licensee operating its facility in a manner contrary to the l
Safety Analysis Report description highlighted the need for a special focused review
that compares plant practices, procedures and/or parameters to the Safety Analysis
Report description. While performing the inspections discussed in this inspection
report, the inspectors reviewed the applicable sections of the Final Safety Analysis
Report that related to the selected inspection areas,
b. Observations and Findinas
The team found that the Final Safety Analysis Report was generally consistent with
the actual plant configuration. The team noted several discrepancies in the
descriptions as noted below:
Imorocer Chanae to Essential Service Water Self-Cleanina Strainer Backwash
Setooint
The team reviewed Section 9.2.1, " Station Service Water System," and
Table 9.2-5, " Essential Service Water System Component Data," of the Wolf Creek
Updated Safety Analysis Report. The team noted that Table 9.2-5 for the essential
service water system self-cleaning strainers listed a strainer capacity of 15,000 gpm
with a maximum differential pressure of 3.0 psi. The team asked the licensee to
verify the capacity at this differential pressure. The licensee stated that the signal i
to start the self-cleaning strainers was 5.0 * 0.5 psi not the 3.0 psi stated in
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Table 9.2-5. During the first week of the inspection, the licensee was not able to
determine the reason for the difference in the maximum strainer differential
pressure.
i
During the inspection, the licensee contacted the strainer vendor to determine if a
maximum strainer differential pressure of 5.5 psi was acceptable. The licensee
stated that setting the maximum differential pressure at 6.0 psi would not cause
any physical damage to the strainer. However, it might detract from the strainers
ability to self clean upon initiation of the backwash cycle. The licensee stated that
the vendor indicated that a pressure drop of 1.0 psi, clean, across the strainer was
4
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based on laboratory tests and did not account for the pressure drop across the inlet
and outlet connections, or specific piping connections, in addition, the vendor i
recommended a strainer backwash initiation at a pressure drop 2.0 psi greater than
the clean pressure drop.
The team reviewed vendor data on the strainers. One chart plotted pressure loss l
versus flow. The team noted that for a clean strainer there was a pressure drop of
1.0 psi at a flow of 15,000 gpm. The team reviewed another plot of pressure loss
versus percent of strainer clogged. The team noted that, with a differential
pressure of 5.0 psi, the plot indicated that the strainer surface was 95 percent
clogged in additbn, the team reviewed the licensee's data on strainer differential
pressure and system flow. The team found that, since 1994, the normal differential ;
pressure across the strainers has been approximately 3.0 to 3.5 psi and the system
flow was approximately 15,000 gpm. In addition, the team reviewed startup test
data from 1984 which listed a strainer differential pressure less than 1 psi at a flow
'
over 15000 psi. The licensee could not explain what caused the pressure to
increase from less than 1.0 psi in 1984 to more than 3.0 psi in 1994. ;
The team considered the Updated Safety Analysis Report setpoint discrepancy to be
important since a change in strainer differential pressure could directly affect system
flow rates. Based on reviewing the licensee's recent test data, which showed
system flow greater than the design flow rate of 15,000 gpm, the team concluded
that there were no operability concerns on account of the discrepancies.
10 CFR 50.59(a)(1) allows the holder of a license to make changes to the facility I
and procedures as described in the final safety analysis report without prior
Commission approval unless the proposed change involves a change in the
Technical Specifications or an unreviewed safety question.
The team reviewed setpoint Change Request EF-84-01, dated March 13,1984.
This document requested a setpoint change for the self cleaning strainer pressure l
. instruments to change the setpoint to 5.5 psid. The cover sheet was annotated
with an "N/A" following questions concerning if any Updated Safety Analysis
Report section or limit was affected by the change. The modification had a
10 CFR 50.59 screening, but no safety analysis. The team found that the screening
stated that the change described in the primary document did not involve a change
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to the Updated Safety Analysis Report. However, the strainer table was a part of
the Updated Safety Analysis Report and included the 3.0 psi maximum differential j
pressure for a dirty strainer. The team considered the licensee's failure to perform a
'
safety evaluation to be the first example of an apparent violation of 10 CFR 50.59
(50-482/96021-02). l
Emeraency Core Coolina System Water Hammer
i
The team noted that Updated Safety Analysis Report, Section 6.3.2.2, stated that
all emergency core cooling system discharge piping is water solid during plant
operation and, therefore, water hammer in the injection line is precluded. The team ;
questioned this statement since solid pipe operation alone will not always preclude l
waterhammer events depending upon the piping configuration and flow ;
characteristics. The licensee responded by acknowledging that this statement was I
not appropriate. The licensee initiated Plant Improvement Request 96-2675 and
stated that the Updated Safety Analysis Report would be revised to clarify the l
water hammer statement. The licensee provided applicable sections of the safety
evaluation report which discussed now the residual heat removal system design
features and Mper venting and filling procedures prevented water hammer. The
team concluded that no operability concern existed.
Containment Pressure Used in Pumo Net Positive Suction Head Calculations
in Section 6.3.2.2 of the Updated Safety Analysis Report discussion about i
net-positive suction head, the statement is made that the calculation of available
net-positive suction head in the recirculation mode assumes that the vapor pressure j
of the liquid in the sump is equal to the containment ambient pressure. This is the
case only when containment ambient pressure is atmospheric in accordance with
Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and
Containment Heat Removal System Pumps." The actual net-positive suction head
calculations use atmospheric ambient conditions. The team considered the Updated
Safety Analysis Report statement to be misleading. Licensee personnel
acknowledged the inspector's comment and initiated an Updated Safety Aaalysis l
Report change to clarify the wording.
Incorrect Capacity of Essential Service Water Pomo Prelube Storaae Tank
The team reviewed Section 9.2.1.2.2.2 0f the Updated Safety Analysis Report,
which stated that the essential service water prelube storage tank size was based
on supplying a minimum of 6 gpm water for 5 minutes to the essential service
water pump bearings without any makeup from the essential service water line.
The team asked the licensee how they verify this statement.
The licensee verified that the tank would hold enough water to supply 30 gallons of
water without any mekeup. However, the licensee determined that the maximum
flow to the bearings would only be 1.0 to 1.5 gpm due to the size of the piping
from the prelube tank to the pump bearings. The licensee stated that there was no
operability concern since the pump vendor had installed bronze bearings in the
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pump because of the possibility for pump start without prelubrication. Therefore,
the tank was not needed for pump operability requirements. In addition, the team
determined that the licensee did not knnw the necessary flow rate of water to
properly lubricate the bearings as recommended by the pump vendor to reduce
wear. The team noted that Table 9.2-5 listed the capacity of the prelube tank to be
43 gpm. The team determined that 43 gallons was the volume of the tank with a
usable volume of 35 gallons. The licensee prepared Plant improvement Request
96-2617, dated October 16,1996, to resolve these discrepancies and correct the
Updated Safety Analysis Report.
c. Conclusions
Although there were numerous discrepancies between the Updated Safety Analysis
Report and the actual plant conditions, the inspection team determined that the <
discrepancies did not present an operability concern. The inspection team identified
one apparent violation regarding operation of the essential service water self-
cleaning strainer backwash setpoint differently than described in the Updated Safety
Analysis Report. In addition, the team noted that the licensee had difficulty in
retrieving design information.
E2.3 10 CFR 60.59 Imolementation (37001)
a. Insoection Scooe
The team reviewed the licensee's program guidance, training program information, a
sample of 50.59 screenings and associated unreviewed safety question
determinations, a sample of 50.59 screenings that did not require an unreviewed
safety question determination, and interviewed a number of individuals who perform
( 50.59 screenings and prepare unreviewed safety question determinations. In
addition, a sample of Updated Safety Analysis Report changes were reviewed,
b. Observations and Findinas
The licensee's safety evaluation process for changes to the facility is controlled by
Procedure AP 26A-003," Screening and Evaluating Changes, Tests, and
Experiments," Revision 1. This procedure was recently revised in February 1996.
The procedure delineated the licensee's methods, training requirements, and
responsibilities to determine and document whether facility changes can be made
without prior NRC approval. The process used to determine if an unreviewed safety
question exists is a two step process. l
- The first step was a screening process that made a determination as to
i whether or not the proposed change was a change to the facility as
j described in the Technical Specifications, Updated Safety Analysis Report,
nonradioactive liquid or gaseous discharges, nonradiological solid waste,
thermal discharges, security plan, safeguards contingency plan, security
. guard training plan, radiological emergency plan, an NRC or INPO
,
commitment, and physical changes within the site boundaries. If the answer
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to all questions was negative, then a change to the Updated Safety Analysis
Report was deemed not to exist and the change could proceed without an
unreviewed safety question determination prepared. An affirmative answer
to any of the questions required further evaluation. Or'/ if the screening
determined that it was a change to the Updated Safety Analysis Report, was
an unreviewed safety question determination required.
The second step involved documentation of an unreviewed safety question
determination on Form APF 26A-003-03,"10 CFR 50.59 Unreviewed Safety
Question Determination," by answering a series of questions and recording
the basis for each answer. If the answer to all questions was "no," then an
unreviewed safety question did not exist and the change could be
implemented without prior approval of the NRC. If the answer to any
question was "yes," then NRC approval was required prior to implementing
the proposed change. Procedure AP 26B-003," Revisions to the Updated
Safety Analysis Report," provided instructions for issuing changes to the
Updated Safety Analysis Report.
The team determined that these procedures provided appropriate guidance for the
development and approval of reviews and approvals under 10 CFR 50.59.
The licensee developed a training program for personnel that performed 50.59
screenings and prepared unreviewed safety question determinatinns. The team's
review of the training program determined that the program covered all the essential j
aspects of the 50.59 screenings and unreviewed safety question determinations. In !
addition, there was a requirement that by the end of calendar year 1996, personnel j
performing 50.59 screenings and preparing unreviewed safety question i
determinations must have taken the training. The need for requalification training
will be determined by significant changes to Procedure AP 26A-003, an increasing
trend in the number of Plant improvement Requests indicating deficiencies in I
cornpleted screenings or unreviewed safety question determinations, self- i
'
assessment results and quality assurance audit results.
The team evaluated the implementation of the 50.59 program by reviewing a
sample of completed 50.59 screenings and determinations as contained in the Wolf
Creek Generating Station Annual Safety Evaluation Report for 1995, a listing of the
changes approved since January 1,1996, and interviewing a number of personnel
involved in the preparation of 50.59 screenings and determinations. Several
deficiencies were identified as delineated below:
Inadeauate Justification of Chanae to Turbine Overspeed Protection
Unreviewed Safety Question Determination 59 96-0067 and associated Updated
Safety Analysis Report Change Request 96-044 evaluated and changed the
surveillance frequency for the four high pressure turbine stop valve, six low
pressure turbine reheat stop valves and six low pressure reheat intercept valves
from once per seven days to once per 92 days. This change was based on NRC's
Generic Letter 93-05, "Line-Item Technical Specifications improvements to Reduce
11
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Surveillance Requirements for Testing During Power Operations." The guidance
provided in the generic letter for changing the turbine valve surveillance fraquency,
requested that licensees include a statement in their amendment request that the
proposed change is compatible with plant operating experience and a statement that
the turbine manufacturer concurred with the proposed change. However, the
inspection team noted that the unreviewed safety question determination did not
address the licensee's experience with the testing of these valves and did not
contain any information as to the acceptability, by the turbine vendor, of the
decreased surveillance frequency of the turbine valves. Based upon interview.s with
licensee personnel, the team determined that the licensee had not fully considered
these factors and that the turbine vendor had not been contacted.
10 CFR 50.59 (b)(1) requires that records of changes include a written safety I
evaluation which provides the bases for the determination that the change, test, or I
experiment does not involve an unreviewed safety question. Even though the ,
turbine test frequency change did not involve a licen:,e amendment, the licensee !
should have been aware of the specific information the NRC deemed appropriate to
include in this unreviewed safety question determination based on the generic letter.
Therefore, the team determined that the basis included with this change did not
provide adequate information to come to the conclusion that an unreviewed safety j
question did not exist. The team considered the failure to fully evaluate that the
change did not involve an unreviewed safety question to be the second example of ;
an apparent violation of 10 CFR 50.59 (50-482/96021-02). '
The licensee subsequently informed the team that the information needed to justify
the change did not involve an unreviewed safety question was available and the
determination would be revised to include it.
Inadeauate Screeninas of Technical Specification Clarifications
The team reviewed several proposed Updated Safety Analysis Report changes,
including three that would have incorporated Technical Specification clarifications
into the Updated Safety Analysis Report. These clarifications had been screened
and determined to neither change the Updated Safety Analysis Report nor the
Technical Specifications and had been issued for review and approval.
Change Request 96-094 was written is add existing Technical Specification
Clarification 009-85 for a Technical Specification that had been relocated to
Chapter 16 of the Updated Safety Analysis Report. The clarification allowed closing
the breaker and operation of a second centrifugal charging pump while swapping
pumps when in operating Modes 4,5, or 6. The team reviewed current Technical
Specifications 3/4.5.3 (applicable in Mode 4) and 3/4.5.4 (applicable in Modes 5
and 6) and determined that both allowed only one charging pump to be operable.
On October 2,1995, a change to Technical Specification 3/4.5.4 (Amendment 89)
was made that added a 4-hour action period to disable one pump.
The. team determined that the licensee had changed Operating Procedure
SYS BG-201, " Shifting Charging Pumps," in 1985 to incorporate the Technical
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Specification clan'.. cation. The clarification received a further screening in March
1994 as a result of a quality assurance finding. The team was informed that the
, operating procedure had been previously used, and the 4-hour action period
'
exceeded, on March 22 and 26,1996. In addition, during two occasions on
October 24,1994, while the plant was in Mode 5, both charging pumps were
operable. The team considered the initial screening done for the operating
procedure and the subsequent screening done for this clarification in 1994 to be
inadequate as they changed a Technical Specification requirement and resulted in
operation of a second charging pump while in Mode 5, contrary to Technical
Specification 3.5.4. Failure to perform the required actions of Technical
Specification 3.5.4 is considered to be an apparent violation of the Technical
Specification (50-482/96021-03).
The licensee subsequently voided this proposed change request and the Technical
Specification clarification. A revision to the operating procedure was also initiated
to prohibit this action.
Following the identification of the team's concerns about Technical Specification
clarifications, the licensee formed an internal investigatior, wam to review and
determine the adequacy of all 45 active clarifications and whether or not
compliance with Technical Specification requirements was being achieved. As a
,
result of that continuing review, the licensee identified two additional clarifications j
which were improperly screened and that resulted in Technical Specification
non-compliance as follows:
Technical Specification Clarification 004-86 allowed cold-leg accumulators to
be considered operable upon receipt of level and pressure alarms if
accumulator level and pressure was within prescribed limits. This
clarification involved a change to Technical Specification Surveillance
Requirements 4.5.1 and 4.0.3, which required the accumulators be
considered inoperable upon receipt of alarms.
The licensee determined that from September 25 to October 2,1996, the
associated level alarm was energized and the Technical Specification action
statement was not met because of the failure of one levelindication channel
on Cold Leg Accumulator B. The Technical Specification action statement
required restoration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
followed by reactor coolant system depressurization below 1000 psig within
the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The team noted, however, that the alarm function did not
affect the ability of the system to perform its safety function.
J
- Technical Specification Clarification 005-94 allowed hot restart testing of an
emergency diesel generator to be per'ormed any time before or after the
24-hour load test, as long as the hot restart test was performed within
5 minutes of a 2-hour diesel run. This clarification involved a change to
Technical Specification 4.8.1.1.2 g,7, which specified that a hot restart test
be performed within 5 minutes following tne 24-hour test. There was a
footnote to the Technical Specification that allowed the hot restart test to be
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done following a warmup run if it failed the hot restart test following the load J
test. This clarification allowed the complete decoupling (i.e., allowing the l
hot restart test to be performed anytime after engine warmup and not !
requiring a failure of the hot restart test following the load test) of the Joad
test and the hot restart test. This Technical Specification was changed by
the NRC with Amendment 101, issued on August 8,1996, and allows the i
decoupling of these two requirements. This amendment was implemented l
by the licensee on November 7,1990.
The licensee determined that prior to issuance of this amendment, hot restart
testing of the diesels was not performed in accordance with the Technical I
Specifications. Specifically, during Refueling Outage 7, Emergency Diesel l
Generator A was load tested on September 17,1994, and the hot restart
{
test was not performed until October 15,1994. Emergency Diesel j
Generator B was load tested on September 16,1994, and the Nt restart '
test was not performed until October 17,1994. I
!
The licensee also determined that during Refueling Outage 8, Emergency l
Diesel Generator A was load tested on February 6,1996, and the hot restart
test was not performed until March 26,1996. Emergency Diesel Generator
B was load tested on March 16,1996, and the hot restart test was not
performed until March 23,1996. Again, since the licensee's hot restart test I
method was allowed by the Technical Specifications under certain
conditions, the team considered the consequences of these violations to be
nnnor,
in addition, the team evaluated the licensee's review of all the clarifications and l
identified the following clarifications that provided guidance contrary to Technical
Specification requirements and could have resulted in non-compliance due to
inadequate screenings:
- Technical Specification Clarification 010-85 allowed daiiy containment
closecut inspections following multiple containment entries in one day. This
clarification involved a change to Technical Specifications 3.5.3 and 4.5.2
which specify a containment visual inspection for loose debris be performed
following each containment entry.
- Technical Specification Clarification 026-85 allowed increasing power while
the quadrant power tilt ratio exceeded a prescribed limit. This clarification
involved a change to Technical Specification 3.2.4.a.4 which prohibited
increasing power with the quadrant power tilt ratio greater than the
prescribed limit.
- Technical Specification Clarification 033-85 allowed containment i
penetrations to be considered operable if dedicated operators were assigned
to close inoperable containment isolation valves. This clarification involved a
change to Technical Specification 3.6.1.1 which specified that all
containment penetrations be operable by automatic isolation valves.
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- Technical Specification Clarification 001-94 allows the reactor coolant
system to be cooled down, an activity which involves a positive reactivity
change, with one source range channel of nuclear instrumentation
inoperable. This clarification involved a change to Technical Specification 3.3.1, Table 3.3-1, Functional Unit 6.b, " Source Range Shutdown,"
Action 5, which specified that with one source range channelinoperable, all
operations involving positive reactivity changes be suspended.
Technical Specification Clarification 004-94 deleted emergency diesel
generator testing of the redundant dieselif the inoperable diesel was
rendered inoperable by a support system failure. This clarification involved a
change to Technical Specification 3.8.1.1 which specified that the redundant
emergency diesel generator be tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one emergency
diesel generator was inoperable for any reason except for preplanned
preventive maintenance, testing, or maintenance to correct a deficiency
which, if left uncorrected, would not affect the operability of the diesel
generator. This clarification extended this footnote to include inoperable
support systems on one diesel as a condition that would not require a start
test of the other diesel. This Technical Specification was changed by the
NRC with Amendment 101, issued on August 8,1996, and was
implemented by the licensee on November 7,1996.
- Technical Specification Clarification 002-96 allows one of the two required
source range neutron flux monitors to be considered operable when in the
refueling condition when powered from a nonsafety-related power supply.
This clarification involved a change to Technical Specification 3.9.2, which
specifies that two source range neutron flux monitors to be OPERABLE in the
refueling condition (Mode 6). Although Technical Specification 3.9.2 does
not specify the power source requirement, the definition of OPERABILITY
does include a requirement for electric power, which refeis to the normal
safety-related power supply.
The licensee provided the result of an audit done of the existing clarifications by
their quality assurance group in February 1993. This audit identified the following
potential consequences that could result in the use of Technical Specification
clarifications:
- Failure to comply with Regulatory, Technical Specification, or other
applicable requirements;
- Poor performance ratings, concerns, or more severe actions from the fJRC
for a potentially inadequate or incorrect Technical Specification clarification
program;
e inappropriate actions being taken by operators;
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Potentially non-conservative actions which could require NRC approval prior
to implementation; and/or
- Overly conservative actions for plant shutdown without consideration of
other risks involved. l
A-s a result of that audit, the licensee reviewed membership on the Technical
Specification clarification committee for appropriateness; reviewed guidance for
preparation of clarifications; and performed a 10 CFR 50.59 review (screenings) of
all current clarifications. In addition, the screenings of the clarifications were
reviewed and approved by the Plant Safety Review Committee. These activities
resulted in voiding eleven clarifications, revision of six clarifications, and one ,
clarification was considered for a Technical Specification amendmer>+ The l
remaining clarifications were deemed by the licensee to meet requim.ients. This
action was completed in March 1994. The quality assurance group performed a
follow up audit to evaluate the effectiveness of the corrective actions which
concluded that the corrective actions were adequate to resolve the concern. This !
audit and review of the completed corrective actions failed to identify additional
potential conflicts between the clarifications and Technical Specifications.
10 CFR 50, Appendix B, Criterion XVI, requires in part, that measures be
established to assure that conditions adverse to quality are promptly identified and
corrected. The team determined that the licensee's corrective actions, done
following the quality assurance finding, were inadequate and failed to identify the
conflicting statements in the clarifications with the Technical Specifications. Based
upon the numerous deficiencies in this area, the team concluded that a
programmatic breakdown in the licensee's 10 CFR 50.59 screening program had i
'
occurred. This breakdown included the licensee's quality assurance group which
initially identified potential concerns with the clarifications, but did not properly
assess the adequacy of the licensee's corrective action, and the Plant Nuclear
Safety Review Committee which reviewed the clarification screenings and also
failed to note that changes to the Technical Specifications were involved. The
failure to perform adequate corrective action for the identified clarification
deficiencies is contrary to the requirements of 10 CFR 50, Appendix B,
Criterion XVI, and is considered to be an apparent violation (50-482/96021-04).
At the time of the exit meeting on November 8,1996, the licensee had reviewed
5 the clarifications and determined that occasions had occurred in which the
Technical Specifications were violated and planned to submit five licensee event
reports on these items, i
Imoroper Chanae to Reactor Coolant Pomo FivwheelInsoection Freauency
!
The team reviewed Updated Safety Analysis Report Change Request 95-003,
" Screening for Licensing Basis Changes," approved January 11,1996, regarding a
change in the examination schedule for the reactor coolant pump flywheels.
Specifically, the description of the proposed change stated that Regulatory
Guide 1.14, Revision 1, required a 10-year reactor coolant pump motor flywheel
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examination coinciding with the inservice inspection orogram interval. This change
clarified the intended examination schedule by revising Chapters 3A and 5.4.1 of
the Updated Safety Analysis Report to include an exception to the Regulatory Guide
examination schedule. The examination schedule was changed to 12 years to
accommodate the "D" reactor coolant pump flywheel which had not been inspected
per the previously established schedule. The response to Screening Question 2 on
whether the change results in a revision to the Operating License, including the
Technical Specifications, was marked "No."
Technical Specification 4.4.10, which was applicable January 9,1995, stated that
each reactor coolant pump flywheel shall be inspected in accordance with the
recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14,
Revision 1, August 1975. This Technical Specification was subsequently
'
superseded by Technical Specification 6.8.5.b in License Amendment 89, issued
October 2,1995, which contained the same statement. Regulatory Guide 1.14,
" Reactor Coolant Pump Flywheel Integrity," Revision 1,1975, Paragraph C.4.b.(2)
states that a surface examination of all exposed surfaces and complete ultrasonic
volumetric examination of the flywheel be performed at approximately 10-year
intervals, during the plant shutdown coinciding with the inservice inspection
schedule as required by Section XI of the ASME Code.
The interval for inservice inspection is based on 120 months pursuant to 10 CFR
50.55a(g)(4), with the initial interval beginning on the date of commercial operation.
Commercial operation for the Wolf Creek plant commenced September 3,1985. >
Provisions in Paragraph IWA-2400(c) allowed that each inspection interval may be
decreased or extended by as much as 1 year. The provisions of Paragraph C 4 b of
Regulatory Guide 1.14 specified that the surface and ultrasonic examination of the
flywheel be performed ". . . at approximately 10-year intervals." Therefore, using
the code provisions for the inservice inspection interval, the surface examination of
all of the reactor coolant pump flywheels should have been completed by
September 3,1996. The licensee confirmed on October 25,1996, that the surface
and ultrasonic examination of the "D" reactor coolant pump flywheel has not yet
been performed and is currently scheduled for the Fall 1997 refueling outage during
reactor coolant pump maintenance.
Section 50.59, " Changes, Tests, and Experiments," allows licensees to make
changes to licensed f acilities or to perform tests and experiments at licensed
facilities when these changes, tests, and experiments (1) do not change the
parameters specified in the f acility operating license, including Technical
Specifications, or (2) present an unreviewed safety question. If the changes, tests,
or experiments change the operating license, including Technical Specifications, or
present an unreviewed safety question, NRC approvalis required prior to
implementing the change or performing the tests or experiments. By reference in
the Technical Specifications, any exceptions to the reactor coolant pump motor
flywheelinspection program delineated in paragraph C.4.b of Regulatory
Guide 1.14, must be approved by the NRC.
17
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The team considered r, change to the examination schedule would result in a
-
change to the Tec,hn! cal Specifications by reference in paragraph C.4.b of
Regulatory Guide 1.14. Therefore, the proposed change to the examination would
require NRC approval prior to implementing the change. Failing to properly perform
the screening for the proposed change to the surface examination schedcle for the
reactor coolant pump flywheels to identify a change to the Technical Specification
is contrary to 10 CFR 50.59 and is considered to the third example of the apparent
violation discussed in Section E2.2 of this report 6 0-482/96021-02).
.
'
Af ter being informed of this discrepancy, the licens. e performed an operability
determination for the "D' eactor coolant pump whic., concluded that the pump was
capable of performing its safety related design function. This determination was
3
based upon satisf actory examination results c i f.he flywheel keyways and bore
i
which were last performed during Refueling Ou'tage 7. In addition, nuclear industry
- experience has indicated that a decrease in inspection requirements is appropriate in
!
some cases. Based upon this inf ormation sod consultation with the Office of
Nuclear Reactor Regulation, the taan< cmAded that continued operation of the
pump until the examination could be performed was not a safety concern.
c. Conclusions
Numerous problems were identified with the Fcensee's implementation of the 50.59
review process, which were indicatise of a programmatic breakdown. Further
. evidence of a continuing breakdown in the review process was evident by the
exister,ce of changes made since 1994in which the licensee did not recognize
changes to the Technical Specifications (reactor coolant pump flywheelissue) or
other NRC approved programs (esrential service water system buried pipe testing
discussed in Section E2.7).
1
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The team determined that the program procedures the licensee has developed for
'
the review and evaluation of changes in accordance with 10 CFR 50.59 were
appropriate. Based on the number of findings in the 50.59 area, and the recent
indications of improper screenings for Updated Safety Analysis Report change
requests, the team concluded that trairting did not appear to have been effective in
avoiding continuing deficiencies.
>
The licensee's corrective action for a quality assurance audit, initiated in 1993,
identified potential problems with the use of Technical Specification clarifications,
did not identify potential conflicts between the Technical Specifications and the
clarifications. The followup audit by quality assurance failed to recognize that the
conditions found during the original audit finding were not corrected. This was
considered to be an apparent violation involving inadequate corrective action. In
addition, the review of the clarifications by the Plant Safety Review Committee, and
their failure to identify continuing iss.ues involving Technical Specification
compliance, calls into question the performance of that group.
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E2.4 Unsucoorted Operability Determination l
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a. Inspection Scope (37550)
The team reviewed one operability determination made during the inspection by a
shift supervisor associated with team observations.
b. Observations and Findinas
On October 22,1996, the team noted that the shift supervisor reviewed an informal
listing of inspection issues raised by the team. Item 133 noted that several
different documents, Technical Specification requirements, Updated Safety Analysis
Report sections, and a calculation identified conflicting essential service water flows
through the containment air coolers. Item 133 also identified two questions
regarding the correct number for essential service water flow through the
containment air coolers and, the correct number for heat removal rate of a single
containment air cooler. The shift supervisor reviewed this listing, then logged the
following entry into the Shift Supervisor Log: "1410 Reviewed items 130-134 on !
Engineering and Technical Services NRC Inspection list - No operability /reportability l
issues noted."
The team asked the shift supervisor what the basis was for the log intry identifying
no operability issues for item 133. The shift supervisor stated his baus was
Calculation GN-MW-OO5, Revision 2, which used 4000 gpm flowrate per cooler
group, and that the assumption had been made that, ' ...the engineers knew what
they were doing." The team noted that the flow information used by this
calculation had been superseded, and that the present containment cooler flow was
2000 gpm flowrate per cooler group. The team questioned the engineer regarding
how the list had been presented to the shift supervisor. The engineer stated that
the list had been handed to the shift supervisor, and that there had been no
substantive discussion regarding item 133.
Administrative Procedure ADM O2-024," Technical Specification Operability,"
Revision 3, step 5.3.2, required the shift supervisor to perform a number of actions
associated with the operability determination to ensure sufficient scope of review.
This step required the shift supervisor to determine the requirement or commitment
established for the equipment, and why the requirement or commitment may not be
met. In cases where the operability determination was not straightforward,
Procedure ADM O2-024 also required the shift supervisor to use the information
available to make the determination, and start the actions stated in
Procedure AP 28-001," Evaluation of Nor. conforming Conditions of Installed Plant
Equipment," Revision 4, to obtain sufficient information to completely answer all
questions.
l
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The team determined that the operability evaluation performed by the shift
supervisor failed to include all the required actions, in that, the shift supervisor did
not properly identify the minimum acceptable flow rate for the containment air
cooler given the conflicting statements of containment air cooler flow in the
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Updated Safety Analysis Report and other documents, and compare the actual
cooler flow with the minimum flow requirement as stated in Technical
Specification 4.6.2.3.b. This is a violation of 10 CFR 50, Appendix B, Criterion V
(50-482/96021-05).
The inspection team noted that NRC Inspection Reports 50-482/96-012,
'
50-482/96-11,and 50-482/96-09,had previously identified several examples where
the NRC had identified poorly supported operability determination.',. The team
determined that while the previous examples of poorly supported orarability
evaluations were not identified as violations of requirements, they indicated a
declining trend in performance. The violation identified in this paragraph was
determined to be more significant than the previous examples, in that, the shift
supervisor stated that the operability determination was, at least in part, based on
'
an out-dated calculation and an unsupported reliance on engineering,
c. Conclusions
The team concluded that the shift supervisor violated 10 CFR 50, Appendix B,
Criterion V, when an operability determination failed to comply with the licensee's
procedure on operability determinations, and relied, at least, in part, on an out-dated
calculation. Previous examples identified by NRC inspectors indicated a declining
trend in the performance of operability determinations on shift.
E2.5 System Walkdowns (37550) '
a. Insoection Scope
,'
The team performed a walkdown of the three subject systems and other selected
plant areas to determine the overall material condition of equipment and
maintenance of housekeeping. In addition, the team walked down several portions
of the spent fuel pool cooling system, component cooling water system, and
instrument air system.
b. Observations and Findinas
The team found the housekeeping was generally very good. The team noted that ;
the system engineers and design engineers were both knowledgeable of their
systems. The engineers demonstrated their knowledge during the walkdown by
explaining component deficiencies in detail and relating to the team specific
operational problems with system operation. The material condition of system
components was noted to be very good with little evidence of boric acid leakage ;
and few deficiencies noted during the walkdown. The team noted that several '
minor system leaks had been previously identified by licensee personnel which
'
indicated that a very gcad threshold for deficiency identification had been
established.
,
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The team reviewed the system engineers' notebooks for the three systems selected.
The team noted that these notebooks were maintained in a well organized manner,
and the separate sections were tabbed for easy reference. The safety injection
system engineer kept the trend data and system walkdown sheets current, and had
a sufficient breadth of material to support the stated description of system engineer
responsibilities.
The team asked the safety injection system engineer what the maintenance rule
performance goals and actual system performance was for the safety injection
system. Both the present and former system engineers knew that the safety
injection system performance was exceeding the goal by a wide margin. However,
neither engineer could readily identify the actual system performance statistics
without speaking with the maintenance rule coordinator. While the team did not
view this as a significant weakness, it did indicate that in this case the system
engineers did not have ready access to current maintenance rule performance
statistics for their system.
Safety Iniection Svstem Enaineer System Walkdown
The team noted that the safety injection system engineer had been assigned to this I
system 8 weeks prior to the inspection. During this period, the system engineer
had conducted only one joint walkdown with the previous system engineer. The
system engineer conducted system walkdowns approximately weekly, but
management only required these walkdowns biweekly. The system engineer's
supervisor had participated in one of these walkdowns.
During the walkdown with the team, the system engineer did not tour the
1988 foot elevation of the auxiliary building and was, therefore, unaware of a
flange leak on the suction line between the refueling water storage tank and the
common suction header supplying the eight emergency core cooling pumps. When
asked by the team, the prior system engineer stated that walkdowns had included i
portions of the 1988 elevation of the auxiliary building, but had never included the l
- igh radiation area encompassing the pipe chase area. The system engineer
.
.. dicated that these walkdowns took frorn 1.5 to 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> each, but that during
some weeks the system engineer would take credit for system engineer presence in
the field supporting maintenance as the system walkdown for the week. With the
exception of the 1988 elevation of the auxiliary building, the system engineer's j
walkdown was adequate.
The team discussed with licensee management their expectations for system .
engineering walkdowns. Management stated that they expected the system
'
engineers to perform walkdowns in all areas containing system components,
although less frequently for high radiation areas due to exposure concerns. l
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Residual Heat Removal Temocrary Shieldina
!
- he team noted that temporary shielding had been erected on te ot-leg suction
piping for both trains of residual heat removal cooling and askeo >ut this situation l
and potential impact on system operability. The system engineer stated that this l
shielding was installed in 1991 per a temporary shielding request. The team j
reviewed the shielding request and scaffolding permits which controlled the erection !
of scaffolding used to support the shielding off of the system piping. The team )
noted that the scaffolding permits did not address potential static loads which might l
be applied if the plastic straps which held the shielding to the scaffolding should ;
fail. Licensee personnel acknowledged this deficiency in the scaffolding evaluation l
and inspected the erected scaffolding and temporary shielding. Licensee personnel
found that portions of the shielding were not secured by tie wraps as specified in
the evaluation and decided to remove the scaffolding pending completion of a new
evaluation.
The licensee completed a subsequent evaluation which determined that the secured
and unsecured shielding would not have adversely affected safety related piping
underneath the scaffolding. The team determined that the erected scaffolding and !
shielding had not been reviewed by engineering personnel and the system engineer j
was not knowledgeable of the condition of this temporary shielding even though it
had been installed for several years. The team considered the temporary shielding .
controls to be weak for not requiring an engineering review of erected temporary i
shielding and periodic inspections of installed temporary shielding. The licensee j
subsequently revised Procedure AP 25A-700,"Use of Temporary Lead Shielding,"
to require periodic inspections, verify shielding installation conformed with the
engineering disposition, and evaluation of the need for permanent shielding if
temporary shielding is installed for 6 months.
c. Conclusions
The team found the housekeeping was generally very good. The team noted that,
in general, system engineers and design engineers were very knowledgeable of their
system. The material condition of system components was noted to be very good
with little evidence of boric acid leakage and few deficiencies. A very good
threshold for deficiency identification had been established. System walkdowns by
the safety injection system engineers did not include all plant areas were system
components were located.
The team considered temporary shielding controls to be weak for not requiring an
engineering review of erected temporary shielding and periodic inspections of
installed temporary shielding. The residual heat removal system engineer was not
knowledgeable of the condition of temporary shielding even though it had been
installed for several years.
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E2.6 Ennineerina Work Backloa (37550)
a. Insoection Scope
The team discussed the status of the engineering backlog with the Assistant to the
Vice President of Engineering. The discussions included actions taken by the
engineering organization to reduce the backlog.
b. Observations and Findinas
The licensee's engineering backlog program was managed by the Assistant to the
Vice President of Engineering. The team interviewed the program manager and
found him to be knowledgeable of his responsibilities, but noted that no one had
been assigned backup responsibilities for this effort. This observation was
compounded by the fact that this program was not proceduralized, and that the
data was manually collected and tracked. Therefore, the team considered the
program to be very susceptible to personnel changes in the organization. In
addition, it was noted that the open item information collected had not been trended
to determine the overall impact the open items had on the engineering department
workload.
The licensee's eng;neering backlog listed only 65 open items. The team found this
number to be artificially low because the licensee's threshold for backlog item.s was
j high (i.e., several categories listed backlog criteria as high as 1 to 3 years old). The
licensee explained that when the program was initially started in 1992, the backlog
, criteria was set high intentionally so as to identify those items which were the
, oldest, while keeping the number of backlog items at manageable levels (i.e., with
these backlog criteria, the licensee engineering backlog, at the time, was greater
than 700 open items). However, the team noted that by 1994 the licensee had
significantly reduced their engineering backlog, but had failed to adjust the backlog
criteria. The failure by the licensee to reduce the threshold of the backlog criteria
was considered a weakness.
To better understand the work load on engineering personnel, the team questioned
the number of open items presently assigned to the department. At the time of this
inspection, there were approximately 1508 total open items. To determine the
impact of the opca items and to assess the safety significance of items still open,
the team reviewed a number of the open items listed (plant improvement requests,
corrective work requests, licensee event reports, etc.). The team determined that
the open items had been appropriately categorized and given the appropriate
prioritization for correction and closecut.
A number of closed items were also reviewed. Licensee actions in closing these
items were considered to be appropriate.
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Finally, the team interviewed members of the engineering staff with regard to work
'
backlog. Open items were tracked by engineering supervisors at the group level.
Engineers appropriately scheduled and worked on open items according to their
prioritization and procedural requirements.
The licensee indicated that according to their records, the overall number of open
items that are tracked has been generally on the decline. However, performance
improvement requests were the only open item group that had showed a steady
increase. The licensee attributed this to a lower threshold for issuance of these
reports and a heightened awareness by plant personnel due to increased training in
this area.
c. Conclusions
The licensee managed the engineering open item workload appropriately, but the
licensee backlog program was found to be behind the industry standard due to the
lack of a formalized program, high threshold for backlog criteria, and the failure to
trend the impact of the backlog on engineering personnel workload.
E2.7 Surveillance Testina
a. Insoection Scope
The inspector reviewed Technical Specification surveillance requirements for the
three systems selected and the most recently completd surveillance tests for each
of these surveillance requirements, j
l
b. Observations and Findinas )
!
The surveillance for the systems selected accomplished the Technical Specification
surveillance requirements and were performed at the correct periodicity. Exceptions
are noted below:
Imorocer Verification of Emeraency Core Coolina System Throttle Valve Mechanical
Stoo Position
Technical Specification Surveillance Requirement 4.5.2.g required the licensee to
verify the correct position of each mechanical position stop for the listed emergency
core cooling system valves every 18 months. This verification ensures that
sufficient cooling flow is available for post-accident conditions. The licensee i
accomplished this surveillance requirement by performing Procedures STS EM-001,
" Emergency Core Cooling System Throttle Valve Verification," Revision 11, and
STS BG-004, " Chemical and Volume Control System Seal Injection and Return Flow
Balance," Revision 4. These procedures required workers to measure the valve
stem height for the valves specified in the Technical Specification.
24
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The team asked how the surveillance procedures verified the position of the
mechanical position stops. The 12 EM (Safety injection) system valves listed in
Technical Specification 4.5.2.g, and Valve DGV-202, did not have mechanical
position stops, but were locked in place using a locked chain as specified in
Procedure AP 21G-001," Control of Locked Component Status," Revision 7. Seal
injection valves BGV-198, BGV-199, BGV-200, and BGV-201 had valve stem
locknuts to secure the valve in position, but they were not required to be tightened
or verified during performance of the surveillance test. In addition, the team noted
that the procedure contained a drawing of the valve which did not indicate the
presence of a locking nut.
The team considered the surveillance procedure to be deficient for not including the
specific design attributes of the mechanical stops and specific action necessary to
verify the correct position of the stops. In response to this concern, the licensee
checked the locknuts, and found them tight. The team interviewed two non-
licensed operators who had recently performed this surveillance procedure, and
found that the operators could not recall whether they tightened the locknuts during
this surveillance, or not. The system engineer also interviewed another non-
licensed operator who had recently performed this surveillance and also found that
the operator could not recall tightening the locknuts. The failure of
Procedure STS BG-004 to require the test performer to tighten the locknuts for
these valves is a violation of Technical Specification 6.8.1.a (50-482/96021-06).
Imoroner Essential Service Water Underaround Pioina Pressure Test
The team reviewed Performance Improvement Request 95-2326, which was
initiated on September 20,1995, to request a change in the test method for
essential service water system underground piping pressure tests. The description
of the problem stated that past performances of Test Procedure STS PE-049C,
" Essential Service Water System Underground Piping Leakage Test," Revision 1,
had proven to be very cumbersome and manpower intensive. This test was written
to satisfy the requirements of ASME Section XI as implemented by the licensee's
inservice inspection program for this Code Class 3 system. The test method being
used included the installation of blank flanges, isolating the system, and
determination of the rate of pressure loss. Because this portion of pipe is buried
underground, the initiator requested that the optional testing requirements in
Article WA-5244 of the ASME Code be considered for alternative testing of buried
components. Article IWA-5244 contains three options that are based on system
redundancy and piping isolation abilities:
(a) In non-redundant systems where the buried components are isolable
by means of valves, the visual examination VT-2 shall consist of a
leakage test that determines the rate of pressure loss. Alternatively,
the test may determine the change in flow between the ends of the
buried components. The acceptable rate of pressure loss or flow shall
be established by the owner.
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4
(b) In redundant systems where the buried components are nonisolable,
the visual examination VT-2 shall consist of a test that determines the
change in flow between the ends of buried components. In cases
where an annulus surrounds the buried components, the areas at each
end of the buried components shall be visually examined for evidence
of leAage in lieu of a flow test.
(c) In non-redundant systems where the buried components are
nonisolable, such as return lines to the heat sink, the visual
examination VT-2 shall consist only of a verification that the flow
during operation is non impaired.
In the evaluation for this request, the engineer concluded that each of the two trains
of essential service water could be considered a non-redundant system. This
interpretation determined that each train provided cooling water only to the loads )
associated with that train (i.e., Train A of essential service water supplies cooling )
water to Train A heat loads, and Train B of essential service water supplies cooling j
water to Train B heat loads, with no other cooling water supply to the separate l
trains). This interpretation was not based on an ASME Code definition or an official j
ASME Interpretation. !
i
As a result of the evaluation, the engineer further concluded that paragraph (c)
of Article IWA-5244, could be applied to the buried portions of the essential
, service water system. This conclusion resulted in revisions to Test
Procedure STS PE-049C, "A Train Underground Essential Service Water System
Piping Flow Test," and development of new Test Procedurs STS PE-049D,
"B Train B Underground Essential Service Water System Piping Flow Test," which
eliminated the previous method of performing the visual examination VT-2 (i.e.,
determination of the rate of pressure loss) and implemented visual examination VT-2
that consisted only of a verification that the flow during operation is not impaired.
Section 9.2.1.2, " Essential Service Water System," of the Updated Safety Analysis l
Report stats that the essential service water system consists of two redundant )
cooling water trains. The team considered the licensee's interpretation of system
non-redundancy to contradict this statement in the licensing basis.
The 10 CFR 50.59 screening for the test procedure change indicated that
Chapter 9.2 of the Updated Safety Analysis Report was reviewed. The screening j
did not discuss the discrepancy regarding redundant versus nonredundant j
definitions for the essential service water system trains. The licensee did not j
submit a request for NRC review and approval of the alternative test method, i
Neither did the licensee revise Chapter 9.2 of the Updated Safety Anansis Report to
indicate that the essential service water system trains could be considered
nonredundant systems. Therefore, the team considered that the screening for the
proposed change to the underground piping test procedures was deficient for not
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identifying that a change to the Updated Safety Analysis Report or inservice
inspection program (Technical Specification 4.0.5) was involved. This deficiency is
contrary to the requirements of 10 CFR 50.59 and is considered to be the fourth
example of the apparent violation (50-482/96026-02).
The revised Procedure STS PE-049C was used for the system pressure test
performed for the third 40-month period in the first 120-monthinterval. The test
was completed on January 17,1996. Likewise, Procedure STS PE-049D was
,
performed during January 1996. Performance of the revised tests resulted in the l
l
f ailure to comply with the requirements of Section XI of the ASME Code for buried l
piping in redundant systems and non-compliance with Technical Specification 4.0.5.
l
During the exit meeting, the licensee disagreed with the tearn's conclusion that this
matter was a violation. The licensoe stated that since neither the AMSE Code nor
the Technical Specifications defined the term " redundant"; therefore, it was
appropriate for them to do so. The licensee's inservice inspection engineer had
l attended industry working group committee meetings, which discussed pressure
testing and the definition of redundant and non-redundant systems. The licensee )
l referred to the 1995 Addenda to the 1995 Edition of Section XI of the ASME Code,
! Article IWA-5244, which had been changed to differentiate test methods based
only on whether the piping is isolable or non-isolable, and removed references to
l
- redundant or nonredundant. The inservice inspection engineer utilized this
l knowledge when interpreting these requirements for underground piping pressure
testing. In addition, the onsite Authorized Nuclear Inservice inspector had reviewed '
the change to the test procedure and had no comment. However, the inspection
l team noted that the NRC has not yet endorsed the 1995 Addenda and the
Authorized Nuclear Inservice inspector has no responsibility under 10 CFR 50.59.
l c. Conclusions
in general, the team found that the surveillance for the systems selected
accomplished the Technical Specification surveillance requirements and were
performed at the correct periodicity. However, the team identified one violation
associated with an inadequate procedure to verify emergency core cooling throttle
l
valve mechanical position stops, and an example of an apparent violation regarding
! pressure testing of essential service water system underground piping.
1
l E2.8 Industrv Event Assessment and Lessons Learned
a. Insoection Scope
The team reviewed two industry events to determine the licensee's action to
prevent similar problems. Industry documented failures of 4.16 kV General Electric
Magne-Blast circuit breakers to properly close, and of improper refurbishment of
l 4.16 kV breakers by overhaul vendors, were selected for review due to generic
applicability to the plant.
27 l
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b. Observations and Findinas
The team tound that the licensee had received reports of these events and had
taken corrective actions to prevent occurrence of these problems at Wolf Creek.
Preventive maintenance procedures and procurement documentation had been
reviewed by licensee personnel and appropriate revisions made to identify and
correct similar problems.
E3 Engineering Procedures and Documentation
E3.1 Review of Desian Basis Documents
a. Lniggection
r Scoce
The team reviewed the design basis documents for the essential service water
system, the residual heat removal system, and the safety injection system to verify
the validity of the design basis and determine the ease of retrieving the information.
b. Observations and Findinas
The team reviewed the design basis notebook for the essential service water
system and determined that the notebook had been approved in May 1993 and had
not been updated since then. The team noted a statement in the notebook that
when the notebook was to be used for design input, the user should take into
account the changes issued after the approval date of the notebook. At the time of
the notebook approval, the notebook had been a controlled document.
The team reviewed interoffice Correspondence ED 96-0047, dated September 17,
1996, concerning design basis notebooks. The letter stated that due to downsizing
of engineering and the need to reorganize the work effort, design engineering had .
identified that the notebooks were an opportunity to reduce the demand on )
engineering services. Some of the licensee's actions were to keep the notebook for j
information only and not maintain it as a controlled document. In addition, the !
licensee decided that the system description docurnents would be used to keep !
design basis information in the future. The licensee further stated that the extent of
information added to the system description would vary depending on the ;
judgement of the responsible engineer. The design engineering manager stated that i
there was no need for the notebooks since all of the engineers were very
experienced and knew where to find the design basis information.
Since the design basis notebooks provided information to support the design and
licensing basis and provided the location of other design bases documents, the team I
considered that uncontrolled and out-dated notebooks hindered the control of design
basis information. This conclusion was supported by the fact that no other
controlled document provided this information. The team also noted during the
inspection, there were times when the licensee had difficulty retrieving design basis
information. The team considered the licensee's control of design basis information
to be weak for not providing a centrallocation for the design basis information.
l
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c. Conclusions
Uncontrolled and out-of-date design basis notebooks hindered the control of design
basis information. The licensee's control of design basis information was found to
be weak,in that, it did not provide a centrallocation for the design basis
information. Licensee personnel had difficulty retrieving some design basis
information.
E5 Engineering Staff Training and Qualification
E5.1 System Enaineerina Staff Trainina and Qualification (37550)
a. inspection Scope
A review was performed of the system engineering training program. The team
reviewed Administrative Procedures AP-23-006," System Engineering Program, "
Revision 3, and AP 30F-OO1,"Engir:eering Support Personnel Training and
Qualification Program," Revision 2. The team discussed the training requirements
with a number of system engineers, and members of their direct management,
during individual interviews. In addition, the team reviewed the training records for
all of the system engineers.
b. Observations and Findinqs
The team found the guidance for system engineering training and management
expectations provided in the licensee's administrative procedures to be general in
nature. Training requirements for engineers newly assigned to the system
engineering department, were developed by the engineer's immediate supervisor,
and were found to consist of " Qualifying Activities," which included " Evaluation of
Nonconforming Conditions of Installed Plant Equipment," (i.e. operability
determinations) " Engineering Calculations," "Unreviewed Safety Question
Determination," etc. Specific training on assigned systems was not required and
was left to each engineer's discretion to take system-specific courses that
periodically were offered for operations personnel. With regard to those situations
in which system engineers were assigned to a specific system, but were later given
responsibility for another system, the team noted that little guidance on training was
available other than for " Qualifying Activities." Finally, none of the procedures
were found to specify a time period for completion of training requirements nor
were there any minimum criteria for system engineer acceptance in response to
this concern, the system engineering management issued a performance
improvement request.
In spite of the overall general guidance, the team found that the system engineer's
knowledge of each of their assigned systems was excellent. This was due, in part,
to a significant nurnber of engineers having been involved in operator systems
training prior to entering the system engineering program. In addition, the system
engineers and their immediate supervisors displayed excellent initiative to improve
their knowledge.
29
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F(,r example, the system engineers interviewed were knowledgeable of industry
problems and maintained periodic contact with other utilities and equipment
vendors. The system engineers also periodically walked down their systems in
accordance with a system walkdown schedule that had been reviewed and
approved by their immediate supervisors. The system engineering supervisors
encouraged their personnel to attend technical presentations, classes, and meetings I
held by vendors or other utilities. One specific example of the initiative taken by the 1
system engineering supervision involved the reactor coolant system engineer, who
had been recently assigned to take responsibility for this system. His supervisor
arranged a visit to the Callaway plant, which had an identical reactor coolant
system and was in an outage. This afforded the system engineer an opportunity to ,
walk down the reactor coolant system and become familiar with his system which I
he might not have been able to do at Wolf Creek until their next assigned refueling ;
outage. l
Finally, almost all system engineers were found to have completed the appropriate
" Qualified Activities" training as indicated by their departments training records.
Those cases where engineers had not completed their assigned training was due
specifically to the fact that they had recently been assigned to their present
position.
C. Conclusions
System engineering knowledge was found to be excellent and was based on the ,
initiative taken by system engineers and their immediate supervisors, and not by "
any specific guidance provided in administrative procedures available. Training
guidance was found to be too general. Specifically,it did not provide a minimum
standard for system engineer training or knowledge.
E6 Engineering Organization and Administration
E6.1 System Enaineerina (37550)
a. Inspection Scope
The team interviewed the system engineering manager, three group supervisors,
and seven system engineers. The team focussed on licensee management
expectations of the system engineers and the system engineering program. This
included the method in which these expectations were communicated to the system
engineers, the mechanics of how plant problems were identified and corrected, and
the adequacy of communication between the system engineering department and
other plant organizations such as operations and maintenance. Additionally, the
system engineers were questioned on technicalinformation and outstanding
deficiencies for their assigned systems, including actions they were taking to
resolve those deficiencies.
30
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b. Observations and Findinas
The licensee management expectations of the system engineers and the
system engineering program were delineated in licensee Administrative
Procedure AP 23-006, " System Engineering Program," Revision 3, and
Administrative Instruction Al 23-002," System Engineering Plant Walkdowns,"
Revision O. Licensee management also communicated their expectations verbally
either directly or through the group supervisors.
As stated previously in this report (Section E5.1), the team found that the guidance
provided in the administrative procedures and instructions were general in nature.
More specific guidance was verbally provided to the system engineers, at the group
level, by their appropriate supervisors.
The system engineers stated that although engineering management expectations
were general in nature, they believed that the guidance being provided presently
was an improvement over the lack of guidance that existed in 1995. This
improvement was in part the result of Self Assessment Reports SEL 95-039,
" System Engineering," dated January 19,1996, and SEL 96-025, " System
Engineering Self Assessment Effectiveness Follow-up," dated September 16,1996.
The system engineers indicated that with a clearer definition of their job scope, they
have a better understanding as to what they are required to do and which type of
activities they can defer to another organization. The team found that system
engineers understood their management's expectation in which they would be the
" experts" of their assigned systems and take " ownership" of their assigned
responsibilities.
In accordance with the procedural guidance, system engineers also had developed
primary trending parameters, and walkdown guidelines for their assigned systems,
which were reviewed and approved by their group supervisors. However, the team
noted that the consistency of how these two aspects of the system engineers
workload were being performed was not closely monitored by engineering
management. In addition, the system engineers used system notebooks in an
inconsistent manner. Nonetheless, the system engineers knowledge of their
individual systems was excellent Operations and maintenance planning personnel
considered the system engineers as the " experts" of their assigned systems and as
the focal point for any questions on these systems. Operations personnel indicated
that they had confidence in system engineering personnel to provide them the
appropriate information to make operability determinations.
System engineers displayed " ownership" of their system by following maintenance
activities being performed on their assigned systems. Plus, system engineers
periodically reviewed corrective work requests to identify if any applied to their
assigned system. As mentioned in Section E2.6, system engineers demonstrated
this ownership during the system walkdowns with team members.
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The team noted that the system engineering program did not specify the need for
backup system engineers for the safety-related equipment. The licensee had an
unofficial system engineer backup program, but it did not have any basic training
criteria or knowledge expectations, in addition, some of the system engineers were
unaware that they had been assigned as backup system engineers. and others were
not aware that any backup system engineers had been assigned to their system.
Finally, other plant personnel were unaware as to whom were the backup system l
engineers, and what systems they were responsible for. This is considered to be a
weakness in the system engineering program, and behind industry standards.
c. Conclusions
Overall, the system engineers were found to be knowledgeable of management
expectations and their responsibilities. Licensee management communication of
system engineering expectations has improved. The lack of assigned backup
system engineers was considered a program weakness.
E6.2 Desian Enaineerina (37550)
a. Inspection Scoce
The team conducted interviews with personnel from the maintenance planning, and
operations departments to evaluate tht, extent and effectiveness of design
engineering communications. The team also reviewed a number of change
packages and performanca improvement requests that required engineering
involvement, in an effort to determine how technical issues were resolved.
b. Observations and Findinas
The team identified that cooperation and communication among the design
engineering department and operatbns, and maintenance planning departments
were good. Engineers indicated that management encouraged identification of plant
problems. This has contributed to the increase in the number of performance
improvement requests.
The team found that the performance improvement requests and change packages
reviewed had technical resolutions with proper engineering justifications and that
the proposed corrective actions were adequate.
The team noted that engineers were appropriately utilizing available design basis
documents to determine if a proposed change was within the original design basis.
All personnel interviewed were aware that the design basis notebooks were not
controlled documents, and they only used them as reference documents.
c. Conclusions
The team concluded that the licensee was effectively implementing their program to
respond to requests for engineering resolution of plant problems.
32
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=
l
E7 Quality Assurance in Engineering Activities
a. Insnection Scope (37550)
I
The team reviewed four recent quality assurance self assessment reports related to
engineering activities. Self Assessment Report SEL 96-033," Licensee Event Report
Program," dated October 2,1996, SEL 96-025, " System Engineering Self
Assessment Effectiveness Follow-Up," dated September 9,1996, and SEL 95-056,
" Auxiliary Feedwater System," dated January 9,1996, were reviewed to evaluate
the effectiveness of the licensee's controls in identification and resolution of plant
problems. Although not complete, the inspection team reviewed the assessment
plan and preliminary findings for an auxiliary feedwater functional assessment.
b. Observations and Findinns
The team found that the self assessments were broad in scope and provided I
meaningful findings and recommendations for potential program enhancements. As l
an example, the auxiliary feedwater system self assessment resulted in a number of
improvement recommendations. These recommendations encompassed more than
enhancements to system performance and reliability but system engineering
program enhancements also. One such improvement recommendation included
placing the site wide trending program in a centralized location (e.g., trending data
is located in several groups and information exchanged is not formalized). Other
recommendations included a review of spare parts availability. Although
improvements since the previous self assessment (SEL 95-039) had occurred, the
system engineering self assessment identified weaknesses in management and
,
supervisory oversight of the system engineers. The self assessments resulted in the
- issuance of a number of performance improvement requests to address the
weaknesses identified.
The team found that the auxiliary feedwater system functional assessment plan
included similar items that the team was reviewing in addition, some of the initial
findings from this self assessment effort were similar to those identified in this
report.
c. Conclusions
The team concluded that the licensee's self-assessment reports were effective.
E8 Miscellaneous Engineering issues
E8.1 (Closed) Inspection Followuo item 50-482/9504-03: Use of gear operator stop nut
- for actuator braking. ;
The licensee contacted the valve operator manuf acturer who reviewed the
licensee's procedures for setting the stop nuts and lirnit switch settings and
concurred with the licensee's actions. The load applied to the stop nuts was within
rated design load.
.
33
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.
E8.2 (Closed) Licensee Event Report 50-482/96001: Loss of circulating water due to
icing on traveling screens.
This event was discussed in NRC Inspection Report 50-482/90-03 and was the
subject of a violation as listed in NRC letter EA96-124, dated February 29,1996,
item 06014. No new issues were revealed by the licensee everit report and
followup on the licensee's corrective actions will be performed during the review of
the violation.
E8.3 (Closed) Licensee Event Report 50-482/96002: Loss of essential service water
train due to icing on trash racks.
This event was discussed in NRC Inspection Report 50-482/96-03and was the
subject of two violations as listed in NRC letter EA96-124, dated February 29,
1996, items 02013 and 04013. No new issues were revealed by the licensee
event report and followup on the licensee's corrective actions will be performed
during the review of the violations.
V. Manaaement Meetinas ;
l
X1 Exit Meeting Summary
The team presented the inspection results to members of licensee management at the
conclusion of the inspection on October 25,1996. An exit meeting was held via '
teleconference on November 8,1996. The licensee acknowledged the findings presented.
The overall scope and results of the inspection were discussed with Mr. Terry Damashek,
on December 31,1996.
The licensee did not identify that any propriety information was reviewed by the team.
l
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~ - . - - ._. -.
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ATTACHMENT
,
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SUPPLEMENTAL INFORMATION
i
PARTIAL LIST OF PERSONS CONTACTED
Licensee
- G, Boyer, Director, Site Support
T. Damashek, Supervisor, Regulatory Compliance
R. Flannigan, Manager, Nuclear Engineering
a T. Garrett, Manager, Design Engineering i
l B. Grieves, Supervisor, Systems Engineering I
'
T. Hood, Supervisor, Design Engineering
i N. Hoodley, Manager, Support Engineering
I
R. Hubbard, Superintendent, Operations
- O. Maynard, Chief Administrative Officer j
- B. McKinney, Plant Manager i
- T. Morrill, Manager, Regulatory Services !
7
'
R. Muench, Vice President Engineering !
G. Neises, Supervisor, Reactor Engineering ;
'
D. Neufeld, Acting Manager, Integrated Planning and Scheduling !
-
W. Norton, Manager, Performance improvement and Assessment
i' K. Scherrch, Supervisor, Systems Engineering
R. Sims, Manager, Systems Engineering
l- J. Stamm, Supervisor, Safety Analysis
l' C. Warren, Chief Operating Officer
C. Younie, Manager, Operations
. i
NRC )
1
S. Freeman, Residunt inspector
i
j INSPECTION PROCEDURES USED
IP 37550 Engineering
j IP 37001 10 CFR 50.59 Safety Evaluation Program
IP 92903 Followup - Engineering
.
e
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_ . _ _ _ . . _ . _ . _ . _ . - . . - _ . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ .- _ -_ _
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ITEMS OPENED AND CLOSED
Ooened
~
. 50-482/96021-01 VIO . Inadequate Cortrol of Design Bases (Section E1.2)
,
60-482/96021-02 APV Four Examples of the Failure to Properly Perform Safety
Evaluations (Sections E2.2, E2.3, E2.3, and E2.7)
, 50-482/96021-03 APV Failure to disable centrifugal charging pump while in cold
- shutdown (Section E2.3)
l 50-482/96021 04 APV inadequate Corrective Action for Screening Technical
i Specification Clarifications (Section E2.3)
1
- 50-482/96021-05 VIO Unsupported Operability Determination for Containment
Cooler Flow (Section E2.4)
1
50-482/96021-06 VIO Inadequate Procedure for Verification of Emergency Core
Cooling Throttle Valves Mechanical Position Stops
j (Section E2.7)
Closed
50-482/95004-03 IFl Use of Gear Operator Stop Nut for Actuator Braking
l (Section E8.2)
]
50-482/96001 LER .oss of Circulating Water due to Ice (Section E8.3) i
l 50-482/96002 LER Loss of Essential Service Water train due to Ice
(Section E8.4)
i
j LIST OF DOCUMENTS REVIEWED
j Unreviewed Safety Question Determinations
i Number Title
'
59 93-0211 Main Steam Isolation Actuator Upgrade Modification, Revision O
i
j 59 94-0174 Deletion of Reporting Requirements from Updated Safety Analysis i
Report for Seismic Monitors, Revision 0
,
!
)
l
) 59 95-0003 Reactor Coolant Pump Flywheel Inspection Clarification, Revision O l
,
l 59 95-0016 Spent Fuel Pool Surveillance Level Indicator, Revision O
59 95-0034 Fire Area Combustible Load Evaluation, Revision 0
!
2
1
.- - .- ._
l
i
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59 95-0046 Optional Opening Between Room 1203 and Roc,m 1204, Revision O
l
59 95-0057 Minimum Acceptance Criteria for Centrifugal Charging Pump B, l
Revision 0 !
59 95-0061 Transient Cable Separation Criteria, Revision O
59 95 0063 Biennial Relevancy Procedure Review Requirements, Revision 0
59 95-0109 Auxiliary Feedwater Pump Turbine Exhaust Line Upgrade, Revision 0
59 95-0129 Emergency Diesel Generator Design E.xplanation, Revision 0
59 95-0160 Auxiliary Feedwater Flowrate Revision, Revision 0
59 95-0151 Emergency Core Cooling System Flowrate Revision, Revision 0
59 95-0156 Boron injection Tank Recirculation Pump Removal and Removal of
Thermal Relief Valve, Revision 0
59 96-0032 Operation with Polypropylene Filter Membrane Material in Spent Fuel
Pool, Revision 0
59 96-0034 Delete Reporting Requirements for Meteorological Tower j
Instrumentation, Revision 0 '
59 96 0038 Use of Safety injection Pump for Boration in Mode 6, Revision 0
59 96-0086 Downgrade of Reactor Coolant Pump #1 Seal Leak Off Pressure
indicator, Revision 0
59 96-0109 Highpressure Feedwater Heater Bypass Test, Revision O
59 96-0115 Delete Program Descriptions from Updated Safety Analysis Report,
Revision 0 I
59 96-0143 Revise Updated Safety Analysis Report to Refle t use of Auxiliary
Feedwater in Residual Heat Removal Process, Revision 0
59 96-0148 Revise Scaffolding Procedure, Revision 0
59 96-0155 Clarification of Regulatory Guide 1.144, Revision 0
3
.
.
Updated Safety Analysis Report Change Requests Associated With
Technical Specification Amendments
Number Title
Amendment 89 Updated Safoty Analysis Report Change Request 95-137, dated
12/1/95, Borated Water Sources
Amendment 91 Updated Safety Analysis Report Change Request 95-138, dated
12/1/95, Refueling Water Storage Tank Boron Concentration
Amendment 93 Updated Safety Analysis Report Change Request 96-004, dated ,
1/11/96, Relocate Time Response Tables to Updated Safety Analysis !
Report
Amendment 94 Updated Safety Analysis Report Change Request 96-104, dated
9/17/96, Operation of Emergency Fuel Oil Transfer System I
Updated Safety Analysis Report Change Requests
Number Title
87-022 Corrections to Typographical Errors in Chapter 6, dated 7/15/87 l
1
96-031 Surveillance Frequencies fo'r Main Dam, Saddle Dams, and Baffle
Dikes, dated 2/16/96
96-094 incorporate Technical Specificsiion interpretation, dated 8/29/96
96-095 Incorporate Technical Specification Interpretation, dated 8/29/96
96-096 incorporate Technical Specification Interpretation, dated 8/30/96
96-104 Revise Emergency Diesei Generator Transfer Pump Logic, dated
9/17/96
i
96-118 Revise Spent Fuel Pool Rack Information, dated 9/26/96 )
l
91-047 Correction to Updated Safety Analysis Report Change Request ;90-114, dated 7/10/91 j
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,
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Regulatory Screenings
-
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Number Title
t
j
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- 05622 Revision 0, Motor Operated Valve !
I !
j 05720 Revision 0 and Revision 1, Pressure Locking Modification
- l
- 05782 Revision 2, Turbine Driven Auxiliary Feedwater Pump Resistor Modifications *
'
!
05846 Revision 0, NK Battery Replacement '
i I
05900 Revision 0, Pressure Locking / Thermal Binding Evaluation !
.
05906 Revision 0, Centrifugal Charging Pump High Temperature Alarm
i
t
05927 Revision 0, Low Flow Cavitation Limit Exceeded
I
i
06023 Revision 0, Pacific Valve Configuration Change ,
] 06025 Revision 0, Drain Holes in Code Relief Valves
1
06107 Revision 0, Relief From American National Standards Institute Code
i Hydrostatic Test Requirements
}
d
i 06183 Revision 0, Delete Thermal Relief Valves from Component Cooling Water
j System
06189 Revision 0, Battery Charger Alarm Setpoint
06252 Revision 0, Turbine Driven Auxiliary Feedwater Pump Valve Stem
Replacement
06285 Revision 0, Revised Thermal Design Flow
06304 Revision 0, Load Drop for New Fuel Storage Facility
06394 Revision 0, Safety injection Pump Rework
06445 Revision 0, New Safety injection Pump A and Rotating Element B Approval
2121 Revision 5, Flow Element EM FE0928 ALARA Concern i
3749 Revision 1, SMB-00 Torque Switch improvement
4055 Revision 4, Valve EM HV8807A & B Speed Reduction
4139 Revision 6, Motor-operated Valve - Concerns with EM HV8814A/B
5
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.
4
,
4145 Revision 4, Adjust Torque Switch Settings on EM HV8924 ;
i 1
4148 Revision 7, Motor-operated Valve - Disposition for EM HV8801 A/B & EM
j HV9903A/B
4150 Revision 5, Valve EM HV8835 Motor-operated Valve Disposition
,
4385 Revision 2, Main Control Board Switch Engraving Discrepancies i
,
4394 Revision 13, Target Rock Valves Replacement
i 4537 Revision 4, Boron injection Tank Recirculation Removal
i
'
6424 Revision 1, Fabrication of Thrust Collar Spacer for PEM01 A
6457 Revision 0, Safety injection Pump Motors PEMC1B Bolts Modification
- Industry Technicalinformation Program Reports
i
! Number Title
l 02102 Liberty Technologies, 10-2-92: 10 Code of Federal Regulations Part 21
Notification, Stem Material Constants And Torque Calibrator Effects impact
! Votes Testing Accuracy, Potential For Overthrust
j O2340 NRC Information Notice 93-37: Eyebolts With Indeterminate Properties
- installed in Limitorque Valve Operator Housing Covers
J
- O2371 Limitorque Maintenance Update 92-02
- Motor Pinion Keys, Motor
!
'
Performance, Declutch Tips, Torque Switch Repeatability, Actuator
Nameplate, Actuator Wiring
!
. 02372 Limitorque Maintenance Update 92-02: Motor Pinion Keys, Motor
Performance, Declutch Tips, Torque Switch Repeatability, Actuator
Nameplate, Actuator Wiring
02373 Limitorque Maintenance Update 92-02: Motor Pinion Keys, Motor
Performance, Declutch Tips, Torque Switch Repeatability, Actuator
Nameplate, Actuator Wiring
6
.
.
Calculations
!
Number Title
C-1989-130 Seismic Reanalysis of Refueling Water Storage "anks, Revision 2
EF-M-014 Ultimate Heat Sink Thermal Analysis Review for Power Uprate,
Revision 1
EF-M-029 Minimum Essential Service Water Temperature Rise, Revision 1
EF-M-030 Determine Required Essential Service Water Warming Line Flow,
Revision O
EF-M-031 Determine Orifice Sizes for Ultimate Heat Sink Outlet, Warming Line
Outlets, and FE-3&4 Necessary to Ensure 5000 GPM Essential
Service Water Warming Line Flow and the Corresponding Maximum
Pressure Downstream of FE-3&4,
Revision O
l
EF-M-032 Determine Hydraulic Grade Line Elevation Required at the Essential l
Service Water Warming Line Branch, Revision 0
l
EF-M-033 Evaluate if 1" Thick Plate is Acceptable for EF-FE-03 & EF-FE-04,
Revision O
EF-M-034 investigate Design for Ultimate Heat Sink Discharge Orifice Plate on
Essential Service Water System, Revision 0
EF-M-035 investigate Design for Warming Line Discharge Orifice Plate on
Essential Service Water System, Revision 0
1
EF-M-036 Determiration of Maximum Lake Temperature for Operation with {
Warming Flow, Revision O
'
EF-M-037 Summary of Document Control Procedure 06349 M-11EF01 Flow
Diagram Changes, Revision 0
1
ECCS-5 Centrifugal Charging Pump "A" Net Positive Suction Head l
Determination During Cold Leg Recirculation, Revision 0
ECCS-6 Centrifuga' Charging Pump "B" Net Positive Suction Head
Determination During Cold Leg Recirculation, Revision O
ECCS-7 Centrifugal Charging Pump "A" Net Positive Suction Head
Determination During Hot Leg Recirculation from Residual Heat
Removal Sump "A," Revision O
7
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'A.4
_ 4.m.-- 4-n-.--m+. 2-+e- km esM, -
- -"#--.*4mnE-d hwq s hM -.4 64.) h..a 4 -A= ~ Akm.+m4.~4,
, 1
i
.
ECCS-8 Centrifugal Charging Pump "B" Net Positive Suction Heaa l
Determination During Hot Leg Recirculation from Residual Heat
Removal Sump "A," Revision O
i
ECCS-9 Refueling Water Storage Tank to Safety injection Pump A - Criteria ,
Calc. (1 - 10), Revision 0 - '
ECCS-10 Residual Heat Removal Sump A to Safety injection Pump A Suction - l
Mode F, Revision 0
1
ECCS-11 Residual Heat Removal Sump A to Safety injection Pump B Suction - 1
Mode F, Revision O j
l
ECCS-17 Maximum Head Loss from Refueling Water Storage Tank to Either
Centrifugal Charging Pump During injection Phase of SIS, Revision O l
I
ECCS-32 Containment Sump "B" to Safety injection Pump "B" Inlet, Mode E,
Revision O
ECCS-36 Refueling Water Storage Tank to Safety injection Pump "B" Suction
Mode A, Revision O
ECCS-47 Safety injection Pumps Net Positive Suction Head from Refueling
Water Storage Tank, Revision O
EF-35 ESW Pump Head Requirement, Revision 2
EJ-29 Residual Heat Removal- Flow Orifice Sizing, Revision O
EJ-30 Residual Heat Removal Pumps A&B Net Positive Suction Head,
Revision 1
EJ-35 Residual Heat Removal Pump Minimum Flow Recirculation Line Orifice
Sizing, Revision 0
EJ-37 Residual Heat Removal Cold and Hot Leg Recirculation Orifices,
Revision O
!
EJ-38 Containment Recirculation Sump Screen, Revision O
EJ-40 Containment Recirculation Sump Screen Fluid Velocity, Revision O j
EJ-M-001 Verification of Relief Valve Capacity for Valves EJ8708A&B,
Revision O
EJ-M-017 Potential Susceptibility for Pressure Locking of Motor-operated Valves
EJHV8813A&B, Revision 2
8
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.
.
EJ M-019 Sizing of Expansion Pipe for Valves EJHV8811 A&B for Pressure
Locking Concerns, Revision 1
EJ-MH-OO1 Heat Transfer for the Evaluation of Thermal Binding and/or Pressure
Locking of Valves EJ-HV8716A&B, Revision O
EJ S-OO3 Min. Wall Thickness Evaluation, Revision 1
1 -HBC-W Essential Service Water Discharge Piping Design Pressure and
Minimum Wall Thickness Determination, Revision 1
IMS-01 Missiles, Revision O
PB-01 Total Pipe Break Summary, Revision 1
BN-20 Refueling Water Storage Tank Level Set-Points, Revision 1
Modifications
Number Title
03377 Seismic Reanalyr.is of Refue! Water Tank, Revision O
03838 EF/EA Cross Tie Piping Modification, Revision O
Temporary Modification Order i
Number Title
96-018-EJ Installation of Pressure Gauge Downstream of Valve HV8840
96-024-BB Eliminate Nuisance Alarm of annunciator D074, Revision 2
i
'
96-038-FP Replace Plant Diesel Fire Pump with Temporary Pump While Fire Pump is
Repaired, Revision 1
96-040-SE Eliminate inadvertent alarm of Control Room Annunciators 828 and 83C,
Revision O
96-020-AB Install Temperature Monitoring Equipment on the Main Steam Isolation Valve
Accumulators, Revision O
96-021-BB Protect Vessel Hew # Seismic Support Plate from Excessive Leakage from the
Vessel Head Vent h..es, Revision O 'l
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Self Assessment Reports
Number Title
95-056 Auxiliary Feedwater System
95-039 System Engineering Self Assessment
96-025 System Engineering Self Assessment Effectiveaess Follow-Up
96-033 Licensee Event Report Program
Drawings
Number Title
M-12BB01 P&lD Reactor Coolant System, Revision 15
M 12BG03 P&lD Chemical & Volume Control System, Revision 16
M 12BN01 P&lD Borated Refueling Water Storage System, Revision 08
i
M-12EJ01 P&lD Residual Heat Removal System, Revision 15
M-12EM01 P&lD High Pressure Coolant injection System, Revision 16
M-12EM02 P&lD High Pressure Coolant injection System, Revision 09
M-12EMO3 P&lD High Pressure Coolant injection System Test Line, Revision 00
Reportability Evaluation Request Form I
Number Title j
i
96-035 Mechanical Position Stops on BG Valves, dated October 23,1996 !
!
Procedures
.
Number Title
28D-001 Self Assessment Process, Revision 2
05-004 Specifications, Revision 1
05-003 Design Document Change Notice, Revision 1
10
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0
05C-002 Engineering Evaluation Requests, Revision 0
05 002 Dispositions and Change Packages, Revision 2
05-001 Change Package Planning and Implementation, Revision 2
211-001 Temporary Modifications, Revision 1
AP23L-001 Lake Water Systems Corrosion and Fouling Mitigation Programs,
Revision 0
SYS EF-205 ESW/ Circ Water Cold Weather Operations, Revision 1
STS EF-100A ES'N System inservice Pump A and ESW A/ Service Water Cross
Connect Valve Test, Revision 17
STS EF 100B ESW System inservice Pump B and ESW B/ Service Water Cross
Connect Valve Test, Revision 18
STS EF-001 Essential Service Water Valve Check, Revision 7
STS IC-917 Analog Channel Operation Test Essential Service Water To Air
Compressor Isolation, Revision 5
STS IC-602A Slave Relay Test K602 Train A Safety injection, Revision 8
STS IC-603A Slave Relay Test K603 Train A Safety injection, Revision 14
STS IC-608A Slave Relay Test K608 Train A Safety injection, Revision 11
STS IC-609A Slave Relay Test K609 Train A Safety injection, Revision 10
STS IC-927 ESW to Air Compressor High DP isolation, Revision 3
STS IC-918 Channel Calibration Essential Service Water to Air Compressor
Isolation, Revision 4
STS AL-005 Auxiliary feedwater Auto Pump Start and Valve Actuation, Revision
11
STS KJ-001 B Integrated D/G and Safeguards Actuation Test Train B, Revision 14
AP 14A-003 Scaffold Construction and Use, Revision 3
AP 21G-001 Control of Locked Component Status, Revision 7
STS BG-004 Chemical and Volume Control System SealInjection and Return Flow
Balance, Revision 5
11
. _ _ _ . _
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o
STS EM-001 ECCS Throttle Valve Verification, Revision 11
MGE LT-012 SMB 000 Removal / Replacement, Revision 1
EMG ES-12 Transfer to Cold Leg Recirculation, Revision 7 j
AP 02 009 Chemistry Surveil!iance Program, Revision 2
l
STS EM-0038 ECCS (Safety injected Pump) Flow Balance, Revision 0 ,
l
STS EM-003A ECCS (Centrifugal' Char;;.ng Pump) Flow Balance, Revision 0 l
l
STS CR-001 Shift Logs for Modes 1,2, & 3, Revision 33
STS BG 002 ECCS Valve Check and System Vent, Revision 8
STS EM-003 ECCS Flow Balance, Revision li
STS IC-902A Actuation Logic Test Train A Residual Heat Removal Suction Isolation
Valves, Revision O
STS IC-902B Actuation Logic Test Train B Residual Heat Removal, Revision 0
STS KJ-001 A Integrated D/G And Safeguards Actuation Test - Train A, Revision 14 l
STS KJ-001 B Integrated D/G And Safeguards Actuation Test - Train B, Revision 14
STS IC-740A Residual Heat Removal Switchover to Recirculation Sump Test - Train
A, Revision 9
STS IC-740B Residual Heat Removal Switchover to Recirculation Sump Test -
Train B, Revision 9
Work Requests and Work Packages
Number Title
110110 Motor-operated valve motor insulation found designated incorrectly
104812 Residual Heat Removal Pump Mechanical Seal Leakage
104898 Replacement of Relief Valve EJ8856A
106028 Residual Heat Removal Heat Exchange A Shell to Waterbox i3olting Torque
Verification
10/013 Valve EJV0053 Needs Lubrication of Stem
12
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9
107292 Screens require refurbishment due to corrosion
108111 Essential service water pump motor oil level low
109892 Running of Residual Heat Removal Pumps Below 1700 gpm for Extended
Periods of Time
109954 Inspect Pump Internals Due to Material Found in Valve ME8956C
110193 Wall thickness due to corrosion
110524 Installation of Temporary Gauge @ EJV0063 Downstream of HV8840
110622 Valve EJHCV0606 Leaks By (open)
110955 Essential service water pump operation below flow limits
110959 Residual Heat Removal Pump A Run at Flow Rates Below 1700 gpm
113208 Essential service water pump casing line leaking
I
113614 Leaking valve
113731 Valve EJ HCV-8890BWill Not Open
114876 Verify Shell to Waterbox Bolting Torque for EEJ01 A (open)
115491 Check Valve EJ8730B Not Fully Seating (open)
108477 Essential service water pump prelube tank level indicator failed
109280 Cross tie valve failed leakage test
111729 Replacement of handle on essential service water tank screen
110136 Valve actuator shaft sheared off
Performance improvement Requests
Number Title
96-1488 Drawing change not properly removed from document control file
96-0634 Limit switch rotors not set correctly
96-0500 Drawing not added to ver. dor manual
13
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96 1617 Questions related to essential service water icing event
96-1542 Non safety-related sealant used
96-1288 Confusion in throttle valve position i
96-0659 Multiple failures of actuator shear pins !
96-0365 Level indicator problems
96 1684 Inservice Testing stroke time failure i
96-1214 Valve exceeded maximum alert stroke time
96 1741 incorrect stroke time in procedure !
I
981836 Corroded bolt holes on essential service water tank basket
l
96-1395 Difficulties encountered with controlotron operation
96-0737 Severe corrosion on essential service water piping and valves
96-0579 Severe corrosion on essential service water strainer backwash piping
96-2502 Valve failed stroke time test
96-1953 Fuse blocks found swapped
l
961902 Procedure conflict with updated safety analysis report
96-267b USAR Statement on ECCS Water Hammer
i
96-2729 Missing Internal Missiles Design Basis Calculs'.:on Reference j
96-2733 Questionable Use of a Pipe Whip Assumption in a Design Basis
Calculation
95 0428 Industry event evaluation regarding St Pump Runout Potential
96-2710 Mechanical Position Stops on BG Valves
94-0427 Low Flow Cavitation Limit Exceeded
i
94-0092 Limitorque Maintenance Update 92-02
94-0090 Limitorque Maintenance Update 92-02
94-0089 Limitorque Maintenance Update 92-02
14-
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95-0910 CCW Return Thermal Relief Valve Not Reseating
i
95-2901 Plant Modification Prepared Without Referring to Interim Drawing !
Changes
96-1014 Excessive Valve Local Leak Rates
l
94-0825 Potential for inadvertent Safety injection Actuation During
i
Surveillance Testing l
1
96-0308 Generic Letter 96-01
95-0625 Mitigation and Evaluation of Pressurizer Thermal Transients Caused
by insurges and Outsurges
95-0336 Lifting of Residual Heat Removal Relief Valves EJ8856A, B & EJ8842
96-0384 Thermal Binding Issue w/ Regard to Motor-operated valve EJ
HCV8840
!
15
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