ML20138J622

Insp Rept 50-440/96-16 on 961028-970124.No Violations Noted. Major Areas Inspected:Engineering & Plant Support

ML20138J622
Person / Time
Site:	Perry
Issue date:	02/04/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20138J615	List: ML20138J612 ML20138J622
References
50-440-96-16, NUDOCS 9702070390
Download: ML20138J622 (9)

See also: IR 05000440/1996016

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U. S. NUCLEAR REGULATORY COMMISSION

REGION lll

Docket No.: 50-440

License No.: NPF-58

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Report No.: 50-440/96016(DRS)

Licensee: Cleveland Electric lliuminating Company

Facility: Perry Nuclear Power Plant

Location: P. O. Box 97, A200

Perry, OH 44081

Dates: October 28,1996 through January 24,1997

Inspectors: David Butler, senior Reactor Engineer

Doris Chyu, Reactor Engineer

Approved by: Ronald Gardner, Chief

Engineering Specialist Branch 2

Division of Reactor Safety

9702070390 970204

PDR ADOCK 05000440

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EXECUTIVE SUMMARY

, Perry Nuclear Station Unit 1

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NRC Inspection Report 50-440/96016(DRS)

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This regional inspection reviewed the licensee's efforts to address Appendix R hot short

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vulnerabilities and corrective actions to violation 50-440/94006-06B. The following

strengths and weaknesses were identified: r

Enaineerino

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The licensee's initial evaluation of Information Notice 92-18 adequately utilized the

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information available at that time. However, during the August 1994 refueling outage,

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the opportunity to identify MOV susceptibility to hot shorts was missed because the

design effort did not consider Appendix R requirements. Since re-identification of this

concern in early 1996, the licensee has takers adequate steps to resolve this issue

(Section E2.1).

Plant Suooort

The licensee's identification of discrepant fire seal configurations was good. An

unresolved item was identified pending the licensee's evaluation of the qualification of the

as built fire seals (Section F2.1).

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Report Details

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jll. Enaineerina

E2 Engineering Support of Facilities and Equipment

E2.1 Motor-Ooerated Valve (MOV) Performance Followino a Control Room Fire

a. Insoection Scoce

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On July 18,1996, the licensee identified that a postulated fire in the control room

could render several safe shutdown related MOVs inoperable due to hot shorts.

The licensee issued LER 50-440/96006, Revision 0, dated August 19,1996,and

Revision 1, dated December 2,1996, documenting this finding. The inspectors

, reviewed the following documents:

* Licensee responses to information Notice (IN) 92-18, " Potential for Loss of

Remote Shutdown Capability During a Control Room Fire,"

* Safe Shutdown Capability Report,

o Electrical drawing for proposed MOV modifications,

• ONI-C-61, " Evacuation of the Control Room," and

• Integrated Operating Instruction (101)-11, " Shutdown From Outside Control

Room."

b. Observation and Findinas

On February 28,1992, the NRC issued IN 92-18. This IN described an unanalyzed

condition regarding fire protection and a plant's safe shutdown capability when

reactor operators were forced to evacuate the control room. This fire could cau e

hot shorts, such as short circuits between motor-operated valve control circuit

conductors and their control power source, to initiate spurious operation of certain

MOVr before the operators shifted control of the valves to the remote / alternate

shutdown panel. Motor thermal overload (TOL) protection may be bypassed, set

high cr set with a longer tripping time to allow for additional valve duty cycles

and/or reversing of the MOV during stroking. The IN identified that MOV torque

and limit switches would not electrically disconnect a stroking valve due to the hot

short bypassing the limit and torque switches. This had the potential to cause

mechanical damage to the valve and/or damage the motor.

in April 1992, the licensee's response to IN 92-18 concluded that the electrical

design would protect the MOV against hot short and/or physical damage to the

valves. The design used dual-element time delay fuses, which were normally in

circuit, for MOV electrical protection. These fuses were sized to provide motor

overload protection and to allow sufficient margin for MOV operation. In addition,

based on design stem factors and motor efficiencies available from the MOV

manufacturer during this time period, the licensee concluded that the stall thrust

values were smaller than the weak link 'it values. Therefore, a valve could

withstand a locked rotor condition withe valve mechanical damage until the fuse

opened during the hot short condition. F .swever, the licensee did identify that

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several fuse current ratings could be lowered to optimize the opening time (15-20

seconds) during a locked rotor condition, if a fuse opened, operators could replace

the fuse at the MCC and regain control of the MOV.

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During the August 1994 refueling outage, the licensee modified several MOVs to

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increase their operational capability in response to NRC Generic Letter (GL) 89-10.

In addition, several of the fuses identified in 1992 were replaced. At that time, the

licensee had obtained site-specific data for stem factors from MOV testing. Using

site-specific stem factors and the available motor efficiencies, the licensee

determined that most of the stall thrust values for the modified MOVs were still

smaller than the weak link limit values. For the modified MOVs which had thrust ,

values greater than the weak link limit values, the licensee determined that the

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pressure boundary was not violated if a stall condition occurred. However, the

licensee did not look at the functionality (if the valves could be manually operated)

of the valves after experiencing a stall condition. The MOVs were modified

according to approved design change processes; however, the licensee failed to

consider Appendix R program requirements when addressing GL 89-10.

In early 1996, the licensee started to reevaluate the applicability of IN 92-18 due to

hot shorts concerns that were identified at other plants. In addition, other utilities

had identified through their testing that motor efficiencies were higher than

predicted in the past. Perry recalculated the stall thrust values. Of 54 MOVs

required to shut down the plant during a control room fire,27 valves were

identified to be susceptible to hot short conditions. Of the 27 valves,21 valves

had stall thrust values greater than the new weak link limit values.

The licensee used an ASME code screening methodology and determined that 12 of

the 21 valves would have stall thrust values greater than the new weak link limit

values. The licensee reperformed the weak link analysis for the 12 valves using

realistic factors such as stem nut engagements, torque-to-thrust conversion factors,

temperature factors, etc. At the completion of this effort, eight valves still had stall

thrust values greater than the new weak link limit values. The affected valves

were:

• 1E12-F0023, Residual Heat Removal (RHR) Isolation to Reactor Vessel Head

Spray Valve,

e 1E12-F0024 A, RHR Return to Suppression Pool Isolation Valve,

• 1821-F0019, Main Steam isolation Valve Before Seat Drain Valve,

• 1E51-F0063, RCIC Steam Supply Containment isolation Valve,

• 1E51-F0064, RCIC Steam Supply Containment isolation Valve,

• 1G33-F0004, Reactor Water Cleanup Suction Containment isolation Valve,

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e 1P57-F0015A, Safety-Related instrument Air (IA) Containment Isolation

Valve, and

• 1P57-F0020A, Safety-Related lA Containment Isolation Valve.

The licensee planned to modify these valves during the next refueling outage

starting September 1997 or the next outage of sufficient duration. In addition, the

licensee provided guidance and training to the control room operators of actions to

be taken in case of a hot short condition. A review of the licensee's compensatory

actions was performed by the NRC on September 10,1996, as documented in NRC

Inspection Report 50-440/96006. That review concluded that the compensatory

measures were acceptable.

The potential spurious operation with mechanical damage to certain Appendix R

designated valves could result in the loss of safe shutdown capability during a

control room fire.10 CFR 50, Appendix R, Section Ill.G 2, required, in part, that

alternative shutdown capability be provided where the protection of systems whose

function is required for hot shutdown does not satisfy the requirements of

Section lit.G.2.10 CFR 50, Appendix R, Section Ill.G.3, required, in part, that for

cables or equipment that could prevent operation or cause maloperations due to hot

shorts of systems necessary to achieve and maintain hot shutdown conditions

which are located within the same fire area outside of primary containment be free

of fire damage. From August 1994 through July 1996, the licensee failed to

provide adequate protection for equipment necessary to achieve and maintain safe

shutdown conditions during a control room fire. The failure to meet Appendix R

requirements for alternative shutdown capability is considered an apparent violation

of 10 CFR 50, Appendix R, Sections Ill.G.2 and Ill.G.3 (eel 50-440/96016-01).

c. Conclusions

The inspectors concluded that the licensee's initial evaluation of IN 92-18 was

based upon the use of fuses and at-that-time available design values for valve r. tem

friction coefficients. The inspectors determined that during the August 1994

refueling outage, the opportunity to identify MOV susceptibility to hot shorts was

missed. Although several MOVs were modified due to GL 89-10 issues, the design

effort did not consider Appendix R requirements. Since re-identification of this

concern in early 1996, the licensee has taken adequate steps to resolve this issue.

IV. Plant Suppgr1

F2 Status of Fire Protection Facilities and Equipment

F2.1 Qualification of Fire Seal Material

a. Insoection Scone

On October 22,1996, licensee quality assurance personnel identified that

construction gap fire seals and compartmentalized fire seals were not installed as

depicted on design drawings and in the qualification test report. The inspectors

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toured several of the affected areas and reviewed Potential Issue Form (PlF)

No. 96-3243; Fire Test Configuration for BISKO Three Hour Fire Seal Report

No. 3001-03-B; Design Guide UOO1 and Specification SP-2000, " Penetration Seals,

Raceway Fire Barriers and Radiant Heat Energy Shields;" and associated design

drawings.

b. Observations and Findinas

The quality assurance personnel identified through a routine audit that the

construction gap fire seals and compartmentalized fire seals were not installed

according to BISCO design drawing Nos. 4549-07A-005-2 and 4649-07A-015-3.

According to a general note on Drawing No. 4549-07A-005-2, construction gaps

wider than 4 inches shall be divided longitudinally by using marinite boards. The

licensee identified that the marinite boards for two construction gaps were installed

latitudinally rather than longitudinally.

In addition, according to Drawing No. 4549-07-015-03, for an opening of greater

than 6.25 square feet, the compartment framing materials were to be anchored to

the surrounding concrete walls. The framing materials mainly consisted of unistruts

and unistrut brackets. The licensee identified that the fire wall separating the

intermediate Building and the Fuel Building on elevation 574 was not

compartmentalized according to the drawing. The entire opening was divided into

smaller compartments and the framing materials were not anchored to the wall. (

The test configuration documented in BISCO Three Hour Fire Seal Report

No. 3001-03-B used a 48 by 48 by 12 inch concrete tesi slab with a 30 by 30 inch

opening. A 4 by 24 inch cable tray with 40 percent cable fill was installed in the

opening. The opening and cable tray were filled with BISCO SF 20 silicone foam to

a depth of 9 inches. The entire assembly was then subjected to a three hour fire

test and a hose stream test. The test results were satisfactory; however, this

tested configuration was different than the as-built configuration.

The licensee initiated a PlF and engineering analysis to evaluate the as-built

configuration. This item is considered an unresolved item (URI 50-440/96016-

02(DRS)) pending NRC review of the licensee's evaluation.

c. Conclusions

The inspectors concluded that the licensee's effort in the identification of this

discrepancy was good.

F8 Miscellaneous Fire Protection issues

(Closed) 50-440/94006-06B(DRS).: Slow corrective actions for inoperable

Appendix R required emergency lights. The inspectors reviewed the open PlF list

and work requests for Appendix R required emergency lights and identified no

outstanding items. In addition, since December 1994, the licensee had assigned a

system engineer to oversee the Appendix R emergency light program. The system

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engineer rewrote several surveillance and maintenance procedures to provide better

guidance for determining operability of Appendix R lights.

The inspectors reviewed Periodic Test Instruction (PTI)-R71-P0003,4, 5, and 6,

"Self-Contained Emergency Lighting Unit Discharge Test," Revision 1. However,

the procedure did not contain a quantitative acceptance criterion for determining

battery operability. The vendor's recommendation for final battery terminal voltage

for a 12-volt emergency light battery was 87.5 percent (10.5 volts) of the initial

voltage following a 8-hour discharge test. If the result was less than 10.5 volts, a

vendor representative was to be contacted. The only acceptance criterion in the

procedures required that the lamps be verified "on" at the conclusion of the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

discharge test.

The licensee indicated that PTI-R71 series procedures were developed according to I

EPRI Nuclear Maintenance Application Center (NMAC) guidance. In Section 5.3.2  ;

of the NMAC document,it discussed the acceptance criterion as a verification that I

all lamps were still energized at the end of the discharge test. The guidance also l

took into effect that the low-voltage cut out protection was in the circuit to protect

the battery. The NMAC document indicated that if the low voltage cut out

actuated (the lamps would be off) before the specified test duration, a battery

capacity problem existed and should be corrected before the test was repeated, in

Section 6.1.2, the document discussed a low voltage cut out setpoint of

75 percent 15 percent (9.6 to 8.4 volts). However, discharging to this voltage

could result in cell reversal and battery damage. In both Secticas 5.3.2 and 6.0,

the NMAC document stated that the manufacturer's literature should be reviewed U

for specific guidance because the information in the NMAC document may not fully

address the requirements of a particular type or model of emergency lights. The i

inspactors concluded the guidance set forth in the NMAC document was reasonable  !

for maintaining emergency lights.

The licensee determined that the ernergency lights were equipped with non-

adjustable low-voltage cut out protection. The setpoint for such protection was

based upon battery and relay coilimpedance, and circuit component tolerances. A

nominal setpoint for a lead acid battery was 8 to 9 volts. The vendor indicated that

a lead acid battery would become susceptible to cell reversal and potential damage

if the battery approached to 2.5 volts. Based upon the low-voltage cut out

protection, the inspectors concluded that the existing acceptance criteria for the

8-hour discharge test was acceptable.

The inspectors consider the corrective actions for this item to be adequate. This

item is closed.

V. Manaaement Meetinas

X1 Exit Meeting Summary

On October 31,1996 and January 24,1997, the inspectors presented the

inspection results to licensee management. The licensee acknowledged the findings

presented.

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The inspectors asked the licensee whether any materials examined during the .;

- inspection should be considered proprietary. No proprietary information was

identified.

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PARTIAL LIST OF PERSONS CONTACTED

Licensee

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D. Haviland, Civil / Structural Design Lead I

K. Jury, Compliance Supervisor

L. McGuire, Electrical Unit Supervisor

J. Perry, Lead Auditor, Quality Assurance

INSPECTION PROCEDURES USED

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IP 37551: Onsite Engineering

ITEMS OPENED, CLOSED, AND DISCUSSED

l Ooened

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50-440/96016-01 eel Appendix R required MOVs were susceptible to a hot short

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50-440/96016-02 URI Differences between the as-built fire sea! configurations and

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tested configuration

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50-440/94006-06B VIO Inadequate Breaker Lubrication and Slow Corrective Actions

for Inoperable Appendix R Required Emergency Lights

LIST OF ACRONYMS USED

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ASME American Society of Mechanical Engineering

eel Escalated Enforcement item

EPRI Electrical Power Research Institute

GL Generic Letter

IN information Notice

101 Integrated Operating Instruction

LER Licensee Event Report  !

MOV Motor-Operated Valve

NMAC Nuclear Maintenance Application Center

PlF Potential Issue Form

PTl Periodic Test instruction

RCIC Reactor Core Isolation Cooling

RHR Reaidual Heat Removal i

URI Unresolved item

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Attachment A

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D. C. Cook Calculation ENSM 961213AF, Revision O

Allowable Centrifugal Charging Pump Degradation

l 1. Please provide the basis for the assumption that the CCP miniflow paths are

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isolated when the suction is aligned to the RWST.

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2. Please provide the basis for the assumption that control valves ORV-200 and ORV- I

251 are fully open.

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3. Although this calculation accounts for pressurizer pressure instrument uncertainty,

1 it does not appear to account for the uncertainty in the instruments used to record

j the data. Please provide additional information regarding this issue.

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4. Please provide the Unit 2 pre-1990 operability review results.

5. Please provide additional information regarding piping configuration input into the

Proto-Flo code.

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J- 6. Please provide additional information regarding fluid viscosity inputs into the Proto-

Flo code.  ;

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i 7. Please provide additional information regarding initial RWST level assumptions. l

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8. Please discuss the sensitivity of flowrate to developed head and how this was

factored into the calculation.

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Attachment B

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D. C. Cook Calculation RD-96-02, Revision 0

Offsite and Control Room Thyroid Doses From

Containment Bypass Associated With a Charging Pump in ECCS Mode

1. Please provide the basis for the assumption that the leak persists for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. Please provide additional documentation which supports the chosen operating point

of filtered and unfiltered control room inleakage used in the calculation.

3. Please provide additional discussion of the purpose and effect of doubling the

"LEAKRATE" term in the code.

4. Please discuss whether the contribution from ESF leakage was included in the

control room thyroid dose calculation.

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l 5. Please provide RD-94-01, "Offsite Doses Due to FCCS Leakage."

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6. Please provide RD-88-01, Revision 2, " Control Room Dose to Operators Following a

LOCA."

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