ML18066A213
ML18066A213 | |
Person / Time | |
---|---|
Site: | Palisades |
Issue date: | 05/18/1998 |
From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
To: | |
Shared Package | |
ML18066A209 | List: |
References | |
50-255-98-03, 50-255-98-3, NUDOCS 9806290074 | |
Download: ML18066A213 (20) | |
See also: IR 05000255/1998003
Text
U. S. NUCLEAR REGULATORY COMMISSION
Docket No:
License No:
Report No:
Licensee:
Facility:
Location:
Dates:
Inspector:
Approved by:
9806290074 980518
POR
AOOCK 05000255
G
REGION 111
50-255
50-255/98003(DRS)
Consumers Energy Company
Palisades Nuclear Generating P!ant
27780 Blue Star Memorial Highway
Covert, Ml 49043-9530
March 25 - April 10, 1998
R. Westberg, Reactor Engineer
John Jacobson, Chief, Lead Engineers Branch
Division of Reactor Safety
EXECUTIVE SUMMARY
'
Palisades Nuclear Power Station
NRC Inspection Report 50-255/98003
This inspection reviewed the unresolved items and inspection follow-up items identified by the
Design Inspection conducted from September 16 through November 14, 1997.
Engineering
Good progress had been made in addressing the individual issues. from the Design
Inspection; however, the collective significance of the issues was still being reviewed.
A violation was identified for a recent failure to scope and include in the inservice
testing program, eight valves with specific functions in shutting down the reactor to a
cold shutdown condition, in maintaining the cold shutdown condition, or in mitigating
the consequences of an accident.
Failure to follow procedures resulted in multiple violations:
Five examples were identified where recent safety related calculations were
not revised when analytical inputs changed or were found to be in error as
required by procedures.
Engineers failed to document justification of the acceptability of scaffolding
installed adjacent to the safety related safety injection and refueling water tank
and in the east engineering safeguards (ESG) room adjacent to safety related
piping as required by the procedure.
An unsecured operations storage cabinet was found within nine feet of safety
related valves CV-737 and CV-0747A in the west engineering safeguards room
which was less than the procedure required 11.5 feet (cabinet height +5 feet).
Test results could not be located to verify that testing had been completed
during the 1995 refueling outage for overcurrent relays for supply
breakers 152-105 and 152-106 to Bus 1 C as required by the procedure.
A violation was identified for problems with the.original plant design:
Two vent pipes, which connected the containment sump to the 590 ft elevation
of the containment, did not have screens installed which were specified by the
original design drawings. This piping configuration resulted in a pathway for
debris to enter the recirculation system without being filtered by the
. containment sump screens with .a potential to clog the c_9ntairm1ent§Rr~y _____ . __ ..
nozzles.
Instrument tubing to the HPSI and LPS flow transmitters did not have the one
inch per foot slope specified by the original design drawings.
2
A deviation from a commitment to Regulatory Guide (RG) 1.97 was identified when
CCW flow could not be measured from 0-110 percent of flow using the listed
temperature instruments because their indication range was 0-180 °F and recent
sensitivity studies indicated that the outlet temperature of CCW from the shutdown
cooling heat exchanger would be 184 °F.
A deviation from a commitment to RG 1.6 was identified when a design change
moved the backup power source to a redundant power source, which resulted in
Bus Y-01 being able to automatically transfer between two safety related busses .
3
Report Details
This inspection reviewed the items identified in the Palisades Nuclear Power Station, Design
Inspection (NRC Inspection Report No. 50-255/97201) conducted from September 16 through
November 14, 1997. The 18 unresolved items and 13 inspection follow-up items identified in
the report are discussed below.
Ill.
Engineering
E1
Conduct of Engineering
E 1.1
Licensee Review of Collective Significance of the Issues
a.
Inspection Scope (37550)
The licensee assessed the issues identified in the Design Inspection Report and
issued internal commitments to address the programmatic significance in the areas of
design control, calculation control, and setpoint control. In addition, the impact of the
specific and programmatic inspection findings were also evaluated against the NRC's
October 9, 1996 request for information pursuant to 10 CFR 50.54(f) regarding
adequacy and availability of design basis information. The inspector reviewed the
response to the Design Inspection report, the AE Inspection Actions Matrix dated
March 20 and April 3, 1998, and documentation of corrective actions taken to date.
b.
Observations and Findings
c .
The inspector determined that the assessment of the collective significance of the*
issues identified in the Design Inspection Report was ongoing. While certain actions
had been planned, such as the improvements to the Calculation Control Program,
FSAR Verification and Validation Project, Setpoint Methodology and Control Program,
and the Fuse Control Program, the scheduled completion date of these improvements
was December 15, 1998.
Based on a review of the inspection findings and their comparison to the response to
the 1 O CFR 50.54(f) letter, the licensee concluded that their original response
remained complete and accurate. However, to enhance knowledge of the plant's
design basis, 1 O additional Design Basis Documents and three safety system design
confirmations similar' to the NRC's safety system functional inspections were planned.
A final review of the adequacy of the 50.54(f) response was scheduled for completion
by December 15, 1998.
Pending NRC review of the results of the collective significance and planned
programmatic improvements, this was considered an Inspection Followup Item
-(50-255/98003-01 (DRS)).
_ _ _ _
_ ______________ _
Conclusions
4
Good progress had been made in addressing the individual issues from the Design
Inspection; however, the collective significance of the issues was still being reviewed.
Miscellaneous Engineering Issues (92903)
E8.1
(Open) Unresolved Item 50-255/97201-01 The licensee received a revised
minimum flow requirement of 1600 gpm from the pump manufacturer. The
team's review of the licensee's completed flow model calculation will be an
Inspection Followup Item.
This item will remain open pending licensee completion of evaluation of the effects of
higher predicted temperature on the CCW system and subsequent NRC review.
E8.2
(Open) Unresolved Item 50-255/97201-02 It appeared that the requirements of
10 CFR 50, Appendix B, Criterion Ill, "Design Control," were not met in this case in
that the design basis for the CCW system, as defined in 10 CFR 50.2, did not
encompass the entire range of bounding temperatures.
This item will remain open pending licensee completion of evaluation of the effects of
higher predicted temperature on the CCW system and subsequent NRC review.
E8.3
(Closed) Unresolved Item 50-255/97201-03 Failure to perform IST in accordance with
TSs for RV-0939.
RV-0939 is one of three relief valves inside containment for the CCW system. CCW
piping inside containment is not required during an accident and is classified as non-
Q, non-safety related. The system is Class JB rated at 125 psig with flanged joints
and rubber gaskets. As a result of its function, the ISl/IST programs have classified
the CCW system and related components, which includes RV-0939, as non-class and
excluded it from the inspection and test requirements of the ASME Code.
Although RV-0939 is not required to be in the IST program, it is inspected,
maintained, and its set point verified by preventive maintenance activity PPAC
CCS043 on a ten year interval, which is essentially the same as the requirements of
the Code, ASME OM-1987, Part1. This item is closed.
E8.4
(Closed) Unresolved Item 50-255/97201-04 Requirements of AP 9.11 were not fully
met in that EA-GW0-7793-01, Revision 0, did not contain full substantiation of the
conclusion.
Administrative Procedure 9.11, "Engineering Analysis," Revision 9, Section 2(d),
stated that the analysis section of the engineering analysis (EA) will qualitatively and
quantitatively (if applicable) present an argument which substantiates the conclusion
()f tile EA and responds to the analysis objective.* EA-GW0-7793-01, "CCW Piping------
lnside Containment HELSA," Revision 0, did not contain the necessary analysis to
support the conclusion that the CCW piping inside containment was not affected by
high-energy line break accidents. During the inspection, ES-GW0-7793-01 was
5
revised to included a discussion of the walkdown analysis used to support the EA's
conclusions.
The inspector determined that CCW piping inside containment was not required during
an accident and was classified as non-Q, non-safety related. In addition, calculation
ES-GW0-7793-01 was classified as non-safety related. No Violation of NRC
requirements was identified. This item is closed.
E8.5
(Closed) Unresolved Item 50-255/97201-05 Failure to met a commitment to RG 1.97
in that the installed CCW temperature indicators were not capable of monitoring the
full temperature range of the CCW system.
The inspector reviewed RG 1.97, UFSAR Appendix 7c, "Regulatory Guide 1.97
Instruments," and Condition Report (CR) C-PAL-97-1363E, Draft.
RG 1.97 described a method acceptable to the NRC staff for complying with the
Commission's Requirements to provide instrumentation to monitor plant variables and
systems during and following an accident in a light-water-cooled nuclear power plant
and stated a range for CCW flow instrumentation of 0-110 percent of flow.
NRC letter to Consumers Power Company dated July 19, 1988, entitled "Palisades
Plant - Response to Generic Letter 82-33 Conformances to Regulatory Guide 1.97,
"Instrumentation for Light -Water-Cooled Nuclear Power Plants To Assess Plant And
Environs Conditions During And Following An Accident," allowed use of temperature
instruments to monitor CCW flow.
UFSAR Appendix 7C, Regulatory Guide 1.97, Rev 3, Parameter Summary Table,
Type D Variables, Item D31, stated that the range of these temperature instruments
used to measure CCW flow (TE-0912 and TE-0913) was 0-180 °F; however, recent
sensitivity studies indicated that the outlet temperature of CCW from the shutdown
cooling heat exchange would be 184 °F.
Failure to measure CCW flow from 0-110 percent of flow using temperature
instruments with sufficient indication range is a deviation from a previous licensing
commitment (50-255/98003-02(DRS)).
E8.6
(Closed) Unresolved Item 50-255/97201-06 Failure to perform IST in accordance with
TSs which requires testing of valves which perform a safety function.
The inspector reviewed CRs C-PAL-97-1592, C-PAL-98-0427, C-PAL-98-0431,
C-PAL-99-0433, and C-PAL-98-0477.
The inspector also reviewed License Event
Report 97-013, "Failure to Closure Test Two Check Valves Results in a Violation of
Technical Specification 6.5.7."
Technical Specification 6.5.7 contained requirements for implementing an inservice
testing (IST) program for ASME Code Class 1, 2, and 3 components, as required by
the ASME Code. American Society of Mechanical Engineers (ASME) Boiler and
Pressure Vessel Code Section XI, IWV-1100, "Valve Testing," states that valve testing
6.
shall be performed in accordance with the re*quirements stated in OM-10.
Section
1.1, "Scope," of OM-10 states the active and passive valves covered are those which
are required to perform a specific function in shutting down a reactor to cold shutdown
condition, in maintaining the cold shutdown condition, or mitigating the consequences
of an accident.
As of November 10, 1997, check valves CK-ES3339 and CK-ES3340 in the minimum
flow recirculation piping from the discharge of each high pressure safety injection
pump had a safety function to close to prevent the potential overpressurization of the
pump suction piping. As of March 17, 1998, check valve CK-DMW400 in the flow
path from the primary system make-up storage tank T-81 to the condensate storage
tank T-2 had a safety function to open to supply make-up to the condensate storage
tank. As of March 17, 1998, control valves CV-1813 and 1814 in the containment
purge and ventilation system had an active safety function to close to provide a
containment isolation function. As of March 26, 1998, control valves CV-1501, 1502,
and 1503 in the plant heating system had an active safety function to close to provide
a containment isolation function.
Failure to properly scope and include valves CK-ES3339, CK-ES3340, CK-DMW400,
CV-1813, CV-1814, CV-1501, CV-1502, and CV-1503 in the IST program was a
Violation of Technical Specification 6.5. 7 (50-255/98003-03).
EB. 7
(Closed) Unresolved Item 50-255/97201-07 Requirements of Procedure 9.11
regarding revising engineering analyses were not followed.
The inspector reviewed CR C-PAL-97-1558, "Nonconservative Input to ESR Heatup
Cale EA-D-PAL-93-27F-01" and Administrative Procedure No. 9.11, "Engineering
Analysis."
Administrative Procedure 9.11, "Engineering Analysis," Revision 9, Section 6.1.5.c,
stated that an analysis shall be revised if analytical inputs changed.
In May of 1991, deficiency report F-CG-91-072 identified that Calculation EA-FC-573-
2, "Calculated Required Air Flow for Inverter/Charger Cabinet Cooling Fan," assumed*
an ambient air temperature of 94 °F instead of the design basis temperature of 104
°F. F-CG-91-072 was closed in October 1994 without calculation EA-FC-573-2 being
revised.
Failure to revise Calculation EA-FC-573-2 when the analytical input for design basis.
temperature was found to be in error was a Violation of 10 CFR 50, Appendix B,
Criterion V (50-255/98003-04a(DRS)).
E8.8
(Closed) Unresolved Item 50-255/97201-08 Analysis were not revised when analytical
inputs changed as required by administrative procedure 9.11
7
The inspector reviewed CRs C-PAL-97-1636, C-PAL-97-1603, C-PAL-97-1670, and
Administrative Procedure No. 9.11, "Engineering Analysis."
Administrative Procedure 9.11, "Engineering Analysis," Revision 9, Section 6.1.5.c,
stated that an analysis shall be revised if analytical inputs changed.
An assumption regarding pipe break size was not updated in EA-A-NL-92-185-01,
"Worst Case Operating Conditions for the LPCl/SDC System MOVs," to determine the
effect of the motor operated valves to close against the break when a more
conservative pipe break assumption was used in a later analysis EA-C-PAL-95-1526-
01, "Internal Flooding Evaluation for Plant Areas Outside; Containment," Revision 0.
Failure to update EA-A-NL-92-185-01 when a change regarding pipe break size was
assumed is an example of a violation of 10 CFR 50, Appendix B, Criterion V
(50-255/98003~04b(DRS)).
Assumptions 5.9 and 5.10 of EA-A-NL-92-185-01, which stated that the HPSI and
LPSI flows to the loops were approximately equal under post-accident conditions were
not revised when flow values calculated in EA-SDW-95-001, "Generation of Minimum
and Maximum HPSl/LPSI System Performance Curves Using Pipe-Flo, " Revision 2,
found that Assumptions 5.9 and 5.10 were incorrect. Failure to update
EA-A-NL-92-185-01 when assumptions regarding flow rates were found to be incorrect
is an example of a violation of 10 CFR 50, Appendix B, Criterion V (50-255/98003-
04c(DRS)).
The required LPSI injection check valve flows identified in EA-E-PAL-93-004E-01, "IST
Check Valve Minimum Flow rate Requirements to Support Chapter 14 Events, "
Revision 0, were not revised after a new flow value was calculated in EA-SDW-95-
001.
Failure to revise EA-E-PAL-93-004E-01 when the analytical inputs changed or
were found to be incorrect was a further example of violation of 10 CFR 50,
Appendix B, Criterion V (50-255/98003-04d(DRS)).
E8.9
(Closed) Unresolved Item 50-255/97201-09 Procedures MSM-M-43 and 1.01 and the
"Palisades Ladder Control Policy for Operating Spaces" were not followed.
The inspector reviewed Maintenance Procedure MSM-M-43, "Scaffolding," Revision 2,
Palisades Administrative Procedure 1.01, "Material Conditions Standards and
Housekeeping Responsibilities," Revision 11, and CRs C-PAL-97-1417, C-PAL-97-
1585, C-PAL-97-1586, C-PAL-97-1587, and C-PAL-97~1601 ..
Section 5.3.1 of MSM-M-43, "General," required that in addition to other requirements
of this procedure, scaffolding constructed in the vicinity of safety related equipment
shall not be used in any plant location which contains safety related equipment
without prior engineering and approval and justification documented in Attachment 1,
"data Sheet," Step 2.6. It also required that the responsible engineer provide
justification and approval for any scaffold which deviates from the seismic
requirements of this procedure, and document justification and approval in Attachment
. 1, "data Sheet," Step 2.6.
8
As of October 6, 1997, engineers had not reviewed the acceptability of scaffolding
installed adjacent to the safety related safety injection and refueling water tank. In
addition, on October 30, 1997, engineers had not reviewed the acceptability of
scaffolding installed in the East engineering safeguards (ESG) room adjacent to safety
related piping.
Failure to review and document the acceptability of scaffolding installed in the vicinity
of safety related equipment was an example of Violation of 10 CFR 50, Appendix B,
Criterion V (50-255/98003-05(DRS)).
Appendix 2 of Procedure 1.01 required that unrestrained and potentially damaging
items which can topple should be separated from operable safety related equipment
by a minimum horizontal distance equal to the height of the item plus five feet. During
a plant tour on October 30, 1997, the inspectors observed an unsecured operations
storage cabinet within 9 feet of safety related valves CV-0737 and CV-0747A in the
West engineering safeguards room which was less than the required 11.5 feet (6.5
feet + 5 feet).
Failure to adequately maintain the required separation distance between an
unsecured operations storage cabinet and safety related piping and valves in the
West ESG room was an example of violation of 10 CFR 50, Appendix B, Criterion V
(50-255/98003-06(DRS)).
EB.10 (Closed) Unresolved Item 50-255/97201-10 A portion of the containment sump,
designed to exclude debris from the ECCS pump suction piping, was not constructed
in accordance with the design drawings.
The inspector reviewed drawing M-74, "Underground Piping Reactor Building," Sheet
1, Revision 10, drawing C-155, "Reactor Building Refueling Cavity and Sump Liner,"
Sheet 2, Revision 12, and UFSAR Section 6.4.2.3 which stated that the design of the
spray nozzles was reviewed to confirm that the spray nozzles are not subject to
clogging from debris entering the recirculation system through the containment sump
screens. In addition, the inspector reviewed CRs C-PAL-97-1571 and C-PAL-97-
1354.
During the Design Inspection, two vent pipes were identified which connected the
containment sump to the 590-ft elevation of the containment, bypassing the
containment sump screens. The design drawings specified screens on these two vent*
pipes; however, none were installed. Since the maximum predicted containment flood
level was-597-ft which was two-ft above the top of these vent pipes, this piping
configuration resulted in a pathway for debris to enter the recirculation system without
being filtered by the containment sump screens. The licensee performed an
operability assessment as part of C-PAL-1571 and concluded that the system was
operable. Temporary modification TM-97-046, was installed on October 29, 1997 ._to
add screens to the top of these vent pipes. Failure to correctly implement the design
for the containment sump as specified in drawings M-7 4 and C-155 and UFSAR
Section 6.4.2.3 was a Violation of 10 CFR 50, Appendix B, Criterion Ill (50-255/98003-
07a(DRS)).
9
E8.11 (Closed) Inspection Follow up Item 50-255/97201-11 Review of licensee "extent of
condition" review relative to rubber piping expansion joints used as penetration seals.
The inspector reviewed CR C-PAL-97-1627, "Inadequate Fire Barrier Evaluation." A
review for similar conditions disclosed the existence of two similar fire barriers with
rubber expansion joints in the floor of the CCW room above the West safeguards
room. However, these expansion joints had been evaluated and the results
documented in their respective engineering analyses. This item is closed.
E8.12 (Closed) Inspection Follow up Item 50-255/97201-12 Verify revision of setpoint
methodology guide EGAD-PROJ-08 and training of engineers.
During the Design Inspection, EGAD-PROJ-08,. "Design & Maintenance Guide on
Instrument Setpoint Methodology," Revision 1, was approved and issued to provide
guidance for instrument setpoint methodology. All Safety & Design Review Group
Engineers were briefed as to the need to utilize this guidance. This item is closed.
E8.13 (Closed) Unresolved Item 50-255/97201-13 A portion of the instrument tubing to the
HPSI and LPSI flow transmitters was not installed in accordance with the design
drawings.
During a walkdown of the SI system, the inspectors observed that transmitters for
containment spray flow, FT-0301 and FT-0302, and shutdown cooling heat exchanger
flow, FT-0306, were properly mounted below their flow elements, but the process
tubing was observed to be inadequately sloped back to the transmitters. Additionally,
a walkdown performed by the licensee at the team's request during an in-containment
inspection revealed that the process lines to the HPSI cold-leg. flow transmitters
FT-0308, FT-0310, FT-0312, and FT-0313 and the LPSI flow transmitters FT-0307,
FT-0309, FT-0311, and FT-0314 were also installed with inadequate slope. The
inspectors were concerned that inadequate slope in instrument tubing could contribute
to significant instrument uncertainty by entraining unequal amounts of air in either leg
of the transmitter, causing erroneous readings.
The inspector reviewed Drawing J-F-020, "Instrument Installation Notes - Flow,"
Revision 0, and Drawings J-F-152, "Flow Instrument Above Line WNents - Liquids,"
Revision 1 and J-F-153, "Flow Instrument Above Line WNents - Liquids," Revision 0.
J-F-020 specified a 5-ft minimum drop leg before tubing is sloped to the meter to
accommodate instruments mounted above flow elements and J-F-152 and 153
specified the installation of flow transmitters with a tubing slope of one inch per foot of
instrument tubing run. The inspector also reviewed CRs C-PAL-97-1561 and C-Pal-
97-1664.
Subsequent to the Design Inspection, the results of additional walkdowns determined
that the HPSI and LPSI flow transmitters were properly installed in accordance with
J-F-020; however, failure to properly implement the design basis for HPSI flow
transmitters FT-0308, FT-0310, FT-0312, and FT-0313 and LPSI flow transmitters
FT-0307, FT-0309, FT-0311, and FT-0314 and install instrument tubing with a one-
inch
10
per foot slope as specified in Drawings J-F-152 and 153 was a further example of
Violation of 10 CFR 50, Appendix B, Criterion Ill (50-255/98003-07b(DRS)).
E8.14 (Closed) Unresolved Item 50-255/97201-14 Calculations EA-ELEC-LDTAB-005 and
EA-ELEC-VOL T-13 were not updated to document changes to plant parameters.
The inspector reviewed CR C-PAL-97-1619, "Electrical Engineering Cales Not
Updated to Reflect Changes in Plant Loads" and Administrative Procedure No. 9.11,
"Engineering Analysis, Revision 9.
Administrative Procedure 9.11, "Engineering Analysis," Revision 9, Section* 6.1.5.c,
stated that an analysis shall be revised if analytical inputs changed. The team noted
that EA-ELEC-VOL T-13, "Palisades Loss of Coolant Accident with Offsite Power
Available," Revision 0, had not been revised since 1993 and that load magnitudes
identified in EA-ELEC-LDTAB-005, Revision 4, and EA-SDW-95-001, Revision 2 had
not been included.* Failure to revise Calculation EA-ELEC-VOL T-13 when load
magnitudes used as input to this calculation changed was a further example of
Violation of 1 O CFR 50, Appendix B, Criterion V (50-255/98003-04e(DRS)).
E8.15 (Open) Inspection Follow up Item 50-255/97201-15 The licensee stated that
evaluation of the effects of hot piping would be included under A-PAL-97-062.
The licensee will complete their Cable Ampacity Sizing Program by September 5;
1998, which will identify any cable degradation due to the close proximity of hot piping
and any degradation of fire stops due to local heat sources. This item remains open.
E8.16 (Closed) Unresolved Item 50-255/97201-16 Failure to meet a commitment to RG 1.6
in that an automatic transfer of loads between redundant power sources was created.
Licensee letter to the NRC dated January 24, 1978, stated that the recommendations*
of Regulatory Guide 1.6 would be implemented, in that, no provision would exist for
automatically transferring loads between redundant power sources. NRC Safety
Evaluation Report dated April 7, 1978, confirmed this commitment. Facility Change
(FC)-364, "Feeder Change for Instrument Bus Y-01, Revision 0, implemented .this
commitment and powered bus Y-01 from Motor Control .Center (MCC) 1 and non-
safety related MCC 3; however, a subsequent modification, FC-854, moved the
backup power source back to MCC-02. This modification also installed fuses between
each of the MCCs and transfer switch Y-50; therefore, there was not a single failure -- -
vulnerability.
FC-854, moved the backup power source from MCC 3 to MCC 2, a redundant power
source, which resulted in Bus Y-01 being able to automatically transfer between two *
safety related busses via transfer switch Y-50, which was a deviation from a previous
licensing commitment (50-255/98003-0B(DRS)).
E8.17 (Open) Unresolved Item 50-255/97201-17 No system analysis existed to show that all
the Class 1 E 120Vac loads had adequate voltages.
11
The licensee will perform a bounding analysis by August 15, 1998, to confirm that
Class 1 E 120Vac loads have adequate voltage during accident conditions. This item
remains open.
E8.18 (Closed) Unresolved Item 50-255/97201-18 Overcurrent relays for supply breakers
152-105 and 152-106 to Bus 1 C had not been calibrated and tested as required by
the surveillance test program.
The inspector reviewed CR C-PAL-97-1568 and the related operability assessment.
Periodic and Predetermined Activity APS025, "Bus 1C Relay Testing," required testing
of the overcurrent relays. During the 1995 refueling outage work order 24416160 was
issued dated June 28, 1995 to test the overcurrent relays for supply breakers 152-105
and 152-106 to Bus 1 C. During the Design Inspection, the licensee discovered that
no test results could be located for these relays. Plant records indicated that these
relays had not been tested since 1992; however, the operability assessment in
C-PAL-97-1568 found them operable based on low or lack of drift between
documented calibrations and a lack of TS requirements for testing periodicity. Failure
to calibrate th.e overcurrent relays for supply breakers 152-105 and 152-106 to Bus 1 C
was a further example of Violation of 10 CFR 50, Appendix B, Criterion V
(50-255/98003-09(DRS)).
E8.19 (Open) Unresolved Item 50-255/97201-19 The design-basis lifetime for Agastat relays
as stated by the manufacturer had not been correctly implemented in the facility.
During the A/E inspection, the licensee made an operability determination based on
the E7000 series relay's similarity to the 7000 series relay. The operability
determination concluded that the relays were operable. The licensee will complete
their analysis of 7000 series and E7000 series in safety related applications by July
15, 1998. This item remains open.
E8.20 (Closed) Unresolved Item 50-255/97201-20 Failure to enter an LCO during battery
charger switching evolution.
The inspector reviewed CR C-PAL-97-1537, Operating Procedure SOP-30, "Station
Power," Revision 20, and Technical Specification 3.7.1 h.
Battery charger 1 was supplied from MCC 1 and battery charger 3--was supplied fr6-m
MCC 2. Administrative controls limited the operation so that only one charger per
battery was in service. This prevented a common-mode failure from affecting both
emergency busses. The supply to 125Vdc bus 2 was similar, with battery charger 2
fed from MCC 2 and battery charger 4 fed from MCC 1. Operating Procedure SOP-
30, "Station Power," Revision 20, required the battery chargers to be operated in pairs
(1 and 2 or 3 and 4). During the Design Inspection, the inspectors noted that TS
3:7.1 h
required two station batteries and the DC systems (including at least
one battery charger on each bus) to be operable when the primary
coolant system was above 300 °F.
12
The Station Blackout (SBO) calculations verified that the Class 1 E batteries had the
capacity to meet SBO loads for a period of four hours. In addition, in the event of a
loss of coolant accident coincident with loss of offsite power with emergency
generators available, one charger for each battery will be energized automatically to
supply DC loads. Therefore, the station batteries will carry full load for approximately
10 seconds during this design basis accident and then they would be supported by
the battery chargers. The time period when neither battery charger is connected to
the 125Vdc bus during charger realignment would be expected to be shorter than the
time period in the design basis when the batteries are expected to carry full load.
Because of the short duration where the batteries carry full load, the batteries remain
On December 27, 1995, a TS change request was submitted which revised the
definition of 125Vdc bus operability based on specific bus voltages. In anticipation of
the related TS amendment, operating procedure SOP-30 was revised to require an
LCO entry whenever realigning battery chargers, an action more conservative than
required by the existing TSs. The amendment was never issued. On January 26,
1998, the TS change request was resubmitted as part of the Improved Technical
Specifications Program.
No violations of* NRC requirements were identified, this item is closed.
EB.21 (Open) Inspection Follow up Item 50-255/97201-21 Battery loading concern during
LOOP/LOCA with single failure loss of AC power
The licensee will complete a formal analysis of battery loading considering the battery
chargers are in their alternate alignment, a combined event of a LOCA/LOOP, and
single failure of AC power by January 15, 1999. This item remains open.
EB.22 (Open) Inspection Follow up Item 50-255/97201-22 Potential non-conservative TS Section 4.7.2c.
During the Design Inspection, an operability determination was made concluding that
the 4-hr Station Blackout station battery load profile enveloped the 2-hr Design Basis
Accident load profile. The licensee will complete a formal analysis of battery loading
considering the battery chargers are in their alternate alignment, a combined event of
a LOCA/LOOP, and single failure of AC power by January 15, 1999. This item
remains open.
EB.23 (Open) Inspection Follow up Item 50-255/97201-23 The team identified discrepancies
concerning EA-ELECT-FL T-005 as part of an inspection follow up item.
The licensee plans to revise EA-ELECT-FL T-005, to correct the deficiencies by
January 15, 1999. This item remains open~
_ .. __
EB.24 (Open) Inspection Follow up Item 50-255/97201-24 Lack of analysis to ensure that
adequate voltages would exist at the load terminals of the batteries.
13
The licensee will perform a bounding analysis to identify the worst-case minimum
voltage levels at the load terminals to assure that minimum load voltage requirements
are met by November 15, t998. This item remains open.
E8.25 (Closed) Unresolved Item 50-255/97201-25 It appeared that the requirements of
10 CFR Part 50, Appendix B, Criterion 111, "Design Control," were not followed in that
the design basis for the solenoid valve coils was not implemented in the plant.
The team questioned the capability of solenoid valves to operate at voltages of 87
Vdc as stated in DBD 1.01, "Component Cooling Water System," Revision 4. The
licensee determined that the DBD was incorrectly worded and that the correct
solenoid capability was90-140 Vdc. Upon further review, the licensee identified that
improperly rated coils, rated 102-126 Vdc, were installed in solenoid valves SV-0918
and SV-09778. Engineering Assistance Request (EAR) 97-0652 was initiated to
replace the coils.
Subsequent fo the inspection, the licensee determined that there was no impact on
the mitigation of an accident if solenoid valves SV-0918 and SV-09778 failed to open
due to low voltage, since the closed position was both the failed position and the
required safety position. In addition, ASCO catalog No. NP-1 stated that all ASCO
valves are tested to operate at 15% under the nominal voltage
No violations of NRC requirements were identified, this item is closed.
E8.26 (Closed)* Inspection Follow up Item 50-255/97201-26
Battery calculation
discrepancies.
The discrepancies identified were minor in nature and did not affect the conclusions of
the analyses. Supplied voltages remained within the equipment rating and the station
batteries were not affected. This item is closed.
E8.27 (Closed) Inspection Follow up Item 50-255/97201-27 Section 3.0 of the Acceptance
Criteria and Operability Sheet for Procedure R0-128-2 referred to TS Sections 3. 7 .1
and 4. 7 .1.11, and that these references would only be correct when the proposed
improved TSs, which have been submitted to NRC for approval, became effective.
On January 26, 1998, a request for improved technical specifications was submitted
which specified testing the diesel generators to the load intervals programmed by the
sequencer and eliminated specific references to the sequence time intervals. This
item is closed.
E8.28 (Open) Inspection Follow up Item 50-255/97201-28 Discrepancies in station battery
test procedures RE-83A and B.
The licensee will revise surveillance tests RE-83A and B as appropriate to support the
1998 refueling outage. The licensee will also review DC system requirements by
December 15, 1998. This item remains open.
14
E8.29 (Closed) Inspection Follow up Item 50-255/97201-29 The 1 O CFR 50.59 safety
evaluations were adequate, except for two examples:
Safety Reviews 95-1431 and 95-1432, dated July 7, 1995, for FES-95-206 stated that
the battery duty cycle service test duration for station batteries ED-01 and ED-02 was
changed from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The licensee noted that TS Section 4.7.2.c was
affected by this design change. However, the USQ evaluation, Question 2 of Section
II, was not checked "Yes" for a TS change. TS 4.7.2.c required that a 2-hour battery
test be performed; while design analysis ELEC-LDTAB-009 and FSAR Section 8.4.2
required a 4-hour battery duty cycle. The licensee has submitted a proposed TS
change to reflect the proper battery test duration and issued CR C-PAL-97-1551 to
address this discrepancy.
The preparer of the safety review did not consider that a TS change was necessary
for FES-95-206 to eliminate reference to a specific duty cycle time since that TS
change was planned to be submitted under the Improved Technical Specifications
Program. The required TS change was subsequently submitted on January 26, 1998,
as part of that program.
The safety review documentation for TM-96-027 stated that the FSAR was not
reviewed. Administrative Procedure 3.07, "Safety Evaluations," page 12, required that
the FSAR be reviewed and that those sections reviewed be noted on the safety *
review sheet. The licensee initiated C-PAL-97-1439 to evaluate this discrepancy.
The safety review, PS&L Log No. 96-05508, for temporary modification, TM-96-027,
"Install 152-Spare #5 Breaker in 152-113 Cubicle," was approved via telecon. It
inappropriately indicated that the FSAR had not been reviewed when in actuality, the
- FSAR was reviewed and found not to discuss the level of detail contained in the TM,
that is, auxiliary contact configuration. The safety review was correctly revised and
refiled with the original TM ..
This item is closed.
E8.30 (Open) Unresolved Item 50-255/97201-30 Discrepancies had not been corrected and
the FSAR had not been updated to ensure that the material in the FSAR contained
the latest material.
Section 6.7 classified the CCW penetrations as Class C-2, which was defined
as penetrations with lines not missile protected. However, EA-GW0-7793-01
stated that. the entire CCW system (both inside and outside containment) was
missile protected. The licensee issued FSAR Change Request 6-143-R20-
1427 to state that the CCW penetrations were not vulnerable to internally
generated missiles.
The CCW system was not designed to be missile protected. The statement in
EA-GW0-7793-01 refers to the fact that due to system configuration the
system is effectively protected from missiles, i.e., not vulnerable. The licensee
15
issued a FSAR change clarify this point. This portion of the unresolved item is
closed.
Section 8.4.2.2 stated that the station batteries would be tested to Institute of
Electrical and Electronics Engineers (IEEE) 450-1975. However, battery
Testing Procedures RE-83A, Revision 9, and RE-838, Revision 9, referred to
IEEE 450-1995. FSAR Change Request 8-126-R20-1249 had been initiated,
but the licensee did not intend to act on this change until approval was
received from NRC of a related proposed TS change.
The TS change request , which cites IEEE 450-1995 for battery testing, was
submitted to the NRC on January 26, 1998. This portion of the unresolved
item is closed.
Table 5. 7-8 listed the seismic design value for the station batteries and racks
as "later" instead of including the actual values of the batteries installed by
FES95-206. The licensee issued EAR 97-0636 to evaluate this discrepancy
and revise the FSAR.
Table 5.7-8 was designated as containing the original seismic design values.
The use of the term "later" was used in the original FSAR' because at that time
there was a planned upgrade to install a second redundant electrical train and
the seismic criteria were not available. The licensee will remove the word
"later" as a clarification and maintain the table as the original seismic design
criteria. This portion of the unresolved item is closed .
The remaining portions of this unresolved item remain open. For the UFSAR
deficiencies identified relative to the DC system, the licensee will review DC system
requirements by December 15, 1998.
E8.31 (Ooen) Unresolved Item 50-255/97201-31 Documentation discrepancies were
identified in the design basis documents (DBDs).
Design Basis Document Change Requests were generated and will be incorporated
into the DBDs by December 15, 1998. This item remains open.
V.
Management Meetings
X1
Exit Meeting Summary
The inspector presented the inspection results to members of licensee management at the
conclusion of the inspection on April 10, 1998. The licensee acknowledged the findings
presented.
The inspectors asked the licensee whether any material examined during the inspection
should be considered proprietary. No proprietary information was identified.
16
PARTIAL LIST OF PERSONS CONTACTED
Licensee
D. Rogers
General Manager - Plant Operations
G. Szczotka Manager NPAD
D. Malone
Configuration Control Manager
N. Haskell
Licensing Director
K. Haas
Engineering Director
S. Wawro
Director Maintenance and Planning
K. Toner
Licensing Supervisor
R. Westerhof
Configuration Control
R. Brzezinski
Design
Nuclear Regulatory Commission
J. Lennartz
Senior Resident Inspector
INSPECTION PROCEDURES USED
Engineering
Follow up on previously identified items.
Closed
50-255/97201-03
50-255/97201-04
50-255/97201-05
50-255/97201-06
50-255/97201-07
50-255/97201-08
50-255/97201-09
50-255/97201-10
50-255/97201-11
ITEMS OPENED, CLOSED, AND DISCUSSED
IFI
Failure to perform IST in accordance with TSs for RV- 0939.
Requirements of AP 9.11 were not fully met in that EA-Gwo*-
7793-01, Revision 0, did not contain full substantiation of the
conclusion.
Failure to met a commitment to RG 1.97 in that the installed
CCW temperature indicators were not capable of monitoring the
full temperature range expected for the CCW system.
Failure to perform IST in accordance with TSs which requires
testing of valves which perform a safety function.
Requirements of Procedure 9.11 regarding revising engineering
analyses were not implemented.
Analysis were not revised when analytical inputs changed as
required by administrative procedure 9.11
Procedures MSM-M-43 and 1.01 and the "Palisades Ladder
Control Policy for Operating Spaces" were not followed.
A portion of the containment sump, designed to exclude debris
from the ECCS pump suction piping, was not constructed in
accordance with the design drawings.
Review of licensee "extent of condition" review relative to
17
50;.255/97201-12
- IFI
50-255/97201-13
50-255/97201-14
50-255/97201-16
50-255/97201-18
50-255/97201-20
50-255/97201-25
50-255/97201-26
IFI
50-255/97201-27
50-255/97201-29
IFI
Opened
50-255/98003-01
IFI
50-255/98003-02
DEV
50-255/98003-03
50-255/98003-04
50-255/98003-05
50-255/98003-06
rubber piping expansion joints used as penetration seals.:.
Verify revision of setpoint methodology guide EGAD-PROJ-08
and training of engineers.
A portion of the instrument tubing installation to the HPSI and
LPSI flow transmitters was not installed in accordance with the
design drawings.
Calculations EA-ELEC-LDTAB-005 and EA-ELEC-VOL T-13 were
not updated to document changes to plant parameters.
The safety evaluation performed for FC 854 did not identify that
prior NRC approval was required.
Overcurrent relays for supply breakers 152-105 and 152-106 to
Bus 1 C had not been calibrated tested as required by the
surveillance test program.
Failure to enter an LCO during battery charger switching
evolution.
The design basis for the solenoid valve coils was not
implemented in the plant.
Battery calculation discrepancies.
Section 3.0 of the Acceptance Criteria and Operability Sheet for
Procedure R0-128-2 referred to TS Sections 3.7.1and4.7.1.11,
and that these references would only be correct when the
proposed improved TS.
The 10 CFR 50.59 safety evaluations were adequate, except for
two examples.
Pending NRC review of the results oft.he programmatic
improvements and the 10 CFR 50.54(f) comparison
Deviation from a RG 1.97 commitment.
Failure to properly scope valves CK-ES3339, CK-ES3340, CK-
DMW400, CV-1813, CV-1814, CV-1501, CV-1502, and CV-1503
and include them in the IST program
Failure to follow procedures and update calculations when
analytic inputs changed.
Failure to follow procedures and review and document the
acceptability of scaffolding installed in the vicinity of safety
related equipment
- -
Failure to follow procedures and adequately maintain the
required separation distance between an unsecured operations
storage cabinet and safety related piping and valves in the West
ESG room
50-255/98003-07a
Failure to correctly construct a portion of the containment sump.
in accordance with the design drawings.
50-255/98003-07b
Failure to correctly install instrument tubing for the HPSI and
LPSI flow transmitters with the correct slope.
50-255/98003-08
DEV
Deviation from a RG 1.6 commitment.
50-255/98003-09
Failure to test overcurrent relays as required
18
Discussed
50-255/97201-01
50-255/97201-02
50-255/97201-15
50-255/97201-17
50-255/97201-19
50-255/97201-21
50-255/97201-22
50-255/91201-23
50-255/97201-24
50-255/97201-28
50-255/97201-30
50-255/97201-31
IFI
IFI
IFI
IFI
IFI
IFI
IFI
iFI
Review of the licensee's completed flow mode.I calculation.
The design basis for the CCW system, as defined in 1 O CFR
50.2, did not encompass the entire range of bounding
temperatures.
The licensee stated that evaluation of the effects of hot piping
would be included under A-PAL-97-062.
No system analysis existed to show that all the Class 1 E 120-V
ac loads had adequate voltages.
The design-basis lifetime for Agastat relays as stated by the
manufacturer had not been correctly implemented in the facility.
Battery loading concern during LOOP/LOCA with single failure
loss of AC power
Potential non-conservative TS Section 4. 7 .2c.
The team identified discrepancies concerning EA-ELECT-FLT-
005 as part of an inspection follow up item.
Lack of analysis to ensure that adequate voltages would exist at
the load terminals of the batteries.
Discrepancies in station battery test procedures RE-83A and B.
Discrepancies had not been corrected and the FSAR had not
been updated to ensure that the material in the FSAR contained
the latest material.
Documentation discrepancies were identified in the design basis
documents.
19
AE
ccw
DEV
EAR
EOG
FC
IFI
MC Cs
TS
LIST OF ACRONYMS USED
Architect/Engineers
American Society of Mechanical Engineers
Component Cooling Water
Design Basis Document
Deviation
Engineering Analysis
Engineering Assistance Request
Emergency Service Water
Facility Change
High Pressure Safety Injection
Inspection Follow-up Item
In-service Testing
Loss of Cooling Accident
Loss of Offsite Power
Low Pressure Safety Injection
Motor Control Centers
Quality Assurance
Station Blackout
Technical Specification
Unresolved Item
Violation
20