ML18066A213

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Insp Rept 50-255/98-03 on 980325-0410.Violations Noted.Major Areas Inspected:Engineering
ML18066A213
Person / Time
Site: Palisades Entergy icon.png
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

PDR

REGION 111

50-255

DPR-20

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.

ES

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

operable.

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

IP 37550

IP 92903

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

URI

URI

URI

URI

URI

URI

URI

URI

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

URI

50-255/97201-14

URI

50-255/97201-16

URI

50-255/97201-18

URI

50-255/97201-20

URI

50-255/97201-25

URI

50-255/97201-26

IFI

50-255/97201-27

URI

50-255/97201-29

IFI

Opened

50-255/98003-01

IFI

50-255/98003-02

DEV

50-255/98003-03

VIO

50-255/98003-04

VIO

50-255/98003-05

VIO

50-255/98003-06

VIO

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

VIO

Failure to correctly construct a portion of the containment sump.

in accordance with the design drawings.

50-255/98003-07b

VIO

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

VIO

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

URI

IFI

IFI

URI

IFI

IFI

IFI

IFI

iFI

URI

URI

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

ASME

ccw

DBD

DEV

EA

EAR

EOG

ESW

FC

HPSI

IFI

IST

LOCA

LOOP

LPSI

MC Cs

QA

SBO

TS

URI

VIO

LIST OF ACRONYMS USED

Architect/Engineers

American Society of Mechanical Engineers

Component Cooling Water

Design Basis Document

Deviation

Engineering Analysis

Engineering Assistance Request

Emergency Diesel Generator

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