ML20140C538

Insp Rept 50-346/97-03 on 970118-0307.Violations Noted.Major Areas Inspected:Licensee Operations,Maint,Engineering & Plant Support

ML20140C538
Person / Time
Site:	Davis Besse
Issue date:	04/07/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20140C531	List: ML20140C528 ML20140C533 ML20140C538
References
50-346-97-03, 50-346-97-3, NUDOCS 9704170091
Download: ML20140C538 (22)

See also: IR 05000346/1997003

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U. S. NUCLEAR REGULATORY COMMISSION - ,

REGION 111

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Docket Nos.: 50 346 '

License No.: NPF-3

EA No.: 97-112

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Report No: 50-346/97003 .

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Licensee: Toledo Edison Company

Facility: Davis-Besse Nuclear Power Station

Location: 5503 N. State Route 2

Oak Harbor, OH 43449

Dates: January 18, through March 3,1997 and

March 7,1997 I

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inspectors: S. Stasek, Senior Resident inspector

K. Zellers, Resident inspector

Approved by: John M. Jacobson, Chief,

Reactor Projects Branch 4

9704170091 970407 "'

PDR ADOCK 05000346

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EXECUTIVE SUMMARY '

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Davis-Besse Nuclear Power Station

NRC inspection Report 50-346/97003

This inspection included aspects of licensee operations, maintenancs, engineering, and

plant support. The report covers a six-week period of resident inspection.

Operations l

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A February 12, reactor downpower was conducted in a controlled, well managed ]

manner (Sections 01, M1.5). )

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Engineered safety feature systems walked down this inspection period were i

determined to be functional and appropriately lined up in conformance with the

updated safety analysis report (Section 02.1).

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The NRC identified that in one case, operations personnel failed to enter all required

Technical Specificatiun action statements upon determination that an amplifier

associated with nuclear instrument Ni-6 was inoperabis. This was considered one

example of a violation 10 CFR 50, Appendix B, Criterion V (Section 04.1).

Maintenance

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Overall, maintenance and surveillance activities reviewed this inspection period

adequately tested and maintained the associated equipment. However, some

weaknesses were noted related to the administrative control of some maintenance

and testing activities (Section M1.1).

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Two cases were noted where the control of equipment restoration following

maintenance was weak. The fact that associated sections of piping were slightly

drained during both work activities was not well communicated to operations for

consideration during the return to service process. In addition, the NRC identified

that work was performed outside the scope of the maintenance work order in one

of the two cases and is considered a second example of a violation of 10 CFR 50,

Appendix B, Criterion V (Sections M1.2, M1.3).

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The NRC identified that operators had performed procedural steps out of sequence

while surveillance testing the #3 service water pump. This is considered a third

example of a violation of 10 CFR 50, Appendix B, Criterion V (Section M1.4).

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When the licensee identified that a technical specification regtdied leakage rate test

of the decay heat removal watertight enclosure had not beer, performed following

the last opening of the enclosure on May 24,1997, appropriate followup actions

were taken including initiation of a plant shut down. The hRC subsequently issued

a notice of enforcement discretion which allowed the licensee to terminate the

shutdown and retum the unit to power operations. Howev6r, the failure to perform

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j , the leak rate test was considered a violation of the plant technical specifications

(Section M1.5).

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The licensee identified that a requirement to perform an independent verification

that the fitting used to secure an inspection port on the decay heat removal

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watertight enclosure was properly installed following its use was not included in all

necessary procedures / documents. This matter met the criteria for enforcement

discretion specified in Section Vil of the NRC's enforcement policy, and therefore

this matter was considered a non-cited violation (Section M1.5).

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During a routine plant walkdown, the NRC identified that a service water strainer

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blowdown valve was not closed as expected. Licensee personnel had previously

! noted similar cases as well. At the end of the inspection period, troubleshooting of

! the affected control circuitry was in process (Section M2.1).

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Administrative controls associated with the plant spray shield program appeared

incomplete. Informal guidance only was used to ensure that plant personnel would

j properly secure a spray shield following completion of activities that breached the

spray shield (Section M3.1).

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Plant personnel response to a component cooling water temperature controller

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failure was good (Section M4.1).

Enaineerina

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The licensee identified severalinstances where the design of the reactor coolant

pumps was not in conformance with 10 CFR 50, Appendix R requirements.

Specifically, small sections of oil lines were determined to be located such that

postulated oil leakage from those lines may not have been captured by the installed

oil collection system. It was subsequently determined that no significant safety

consequences would result from the identified configurations. This matter is

currently under NRC review (Section E1.1).

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The NRC identified that a service water pump performance curve included in the

plant " pump curve" book did not accurately reflect the currently installed pump

(Section E3.1).

Plant Suonort

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Appropriate ALARA and contamination controls were noted during the performance

of maintenance activities (Section R1).

- Radiological surveys accurately reflected current area conditions (Section R1)

. The Technical Support Center and Emergency Control Center appeared properly

maintained in a standby condition (Section P2).

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Socurity monitoring equipment was observed to be functional and in a good state of

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repair (Section S2).

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Fire response equipment was appropriately staged and available. Fire brigade

members staged their protective clothing and other equipment prior to assuming

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their shift duties (Section F2).

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, , Report Details

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. Summary of Plant Status

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The plant operated at nominally full power during the inspection period with the exception

of one significant downpower. On February 12, a shutdown por Techrucal Specification .

(TS) 3.0.3 was commenced when the licensee identified that a TS required surveillance j

had not been performed. During the shutdown evolution, the NRC issued a notice of

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onforcement discretion (NOED), allowing termination of the shutdown. Lowest reactor ,

power reached was 10 percent with the main turbine having been tripped. The unit was

! subsequently returned to full power.

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) 1. Operations  !

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l 01 Conduct of Operations i

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Plant operations appeared to be effectively controlled overall. Operations shift t

personnel were appropriately aware of and controlled surveillance and maintenance ,

activities that could affect safety significant equipment. Shift turnovers and shift '

briefings effectively disseminated important operational information to operating

crews. Information tags were hung on control room controllers / indications as

needed to communicate needed operational information. In general, operators '

adequately maintained the unit log so that significant shift activities and the status

of important equipment were appropriately documented. One reption was noted -

and is further discussed in Section 04.f below. The February h., reactor ]

downpower was conducted in a controlled, well managed mannei. 1

02 Operational Status of Facilities and Equipment

02.1 Engineered Safety Features System Watkdowns (71707)

The inspectors walked down the accessible portions of the following engineered

safety features (ESF) and important-to-safety systems during the inspection period:

- emergency diesel generators #1 and #2 '

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station blackout diesel generator

- auxiliary feedwater - trains 1 and 2

. high pressure injection system - trains 1 and 2

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decay heat removal system - trains 1 and 2 l

+ containment spray system - trains 1 and 2  !

. motor driven feed pump

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instrument air dryers  ;

- service water system - trains 1 and 2, and installed spare i

. containment hydrogen dilution system

Equipment material condition was found to be excellent in all cases. Pump / motor l

fluid levels were within their specified bands. Local and remote controllers were I

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[, correctly positioned and required instrumentation appeared functional. Auxiliary

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' equipment necessary for system operability was also functioning properly. In

general, housekeeping was found to be acceptable. Few oil and fluid leaks were

i noted during the walkdowns. System lineups and major flowpaths were verified to

i be consistent with the updated safety analysis report (USAR). The inspectors

j identified no substantive concerns as a result of the walkdowns.

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04 Operator K.-df-g and Performance

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04.1 Technical Snacification Action Statement Not Anoropriatalv Entsted  ;

s. Insoection Scone (71707)

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l The inspector reviewed operator followup actions to a failure of nuclear

j instrumentation NI-6 lower core detector amplifier on January 31,1997.

b. Dhastyations and Findings

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Operations personnel declared Ni-6 inoperable when a step change in its output

, was observed. Subsequently, the assistant shift supervisor, a senior reactor

4 operator (SRO), reviewed TS 3.3.1.1, . Reactor Protection System instrumentation,"

and identified that the subject failure adversely impacted one channel of both high

j flux and fluxNe/ta flux #/ow reactor trips specified in associated TS Table 3.3-1.

l Both table entries referenced an action' statement 2 which required the channel be

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tripped or bypassed within one hour and the quadrant power tilt be monitored at

i least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Those actions were then initiated as required.  !

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l The inspector noted however that an additional TS Table 3.3-1 entry addressing the  ;

high flux / number-of-reactor-coolant-pump-on reactor trip feature was also i

!' applicable. Accordingly, companion action statement 3 should have been invoked )

{ in addition to action statement 2. Action statement 3 also specified that the i

4 affected channel be tripped or bypassed within one hour- .  !

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Although operations personnel had met the requirements of action statement 3 by

implementing action 2, they had failed to recognize and log all TS limiting

conditions for operation that were affected by the Ni 6 failure. l

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j Plant administrative procedure DB-OP-OOOO5 (Revision 05), " Operator Logs and

! Rounds," specified in Section 6.2.2.d that, "The following are entries which shall

be recorded in the Unit Log: . . . Entering / Exiting a Technical Specification Action

{ Statement." Since operations personnel did not recognize that TS 3.3.1 action  ;

j statement 3 was applicable, and therefore failed to make the necessary unit log l

} entry, this constituted an example of a violation of 10 CFR Part 50, Appendix B,

j Criterion V, " Instructions, Procedures, and Drawings" (50-346/97003-01s(DRP)). j

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Operations personnel displayed a lack of attention to detail when determining

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appropriate TS actions to take. Some applicable action statements were not

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ll. Maintenance i

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l Conduct of Maintenance

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l M1.1 Maintenance and Surveillance Testina Activities

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The following maintenance and testing activities were observed / reviewed during the  ;

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MWO 3 97-4602-01 Containment isolation Valve MU-6419 VOTES l

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MWO 1-96-0708-00 Replace DH-100, Leak Test Valve for Pressurizer

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Auxiliary Spray Line

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MWO-3-97-2047-01 Containment Sump Pump Outside Containment 1

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Discharge Isolation Valve VOTES Testing I

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DB-PF-03030 Service Water Pump 3 Quarterly Test

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DB-MI-03201 Channel Functional Test and Calibration of

SFRCS Channel 1 Pressure inputs

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DB-OP-03006 Miscellaneous Instrument Shift Check

j Overall, maintenance and surveillance testing activities reviewed this inspection

period were in accordance with applicable FSAR requirements and adequately

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tested and maintained the associated equipment. However, as discussed below,

some weaknesses were noted related to the administrative control of several

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M1.2 Weak Control of Makeuo Line Restoration Activities

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j a. Insoection Scone (62707) 1

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l The inspector observed the performance of portions of maintenance work order  !

(MWO) 3-97-4602-01, " Containment isolation Velve MU 6419 VOTES Testing," on

February 4,1997,

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F- b. OWations and Findinas

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i The inspector noted that depressurization of associated piping to support MU-6419

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testing was achieved by opening a drain valve. Fluid leakage from the drain valve

was collected with a contaminated water collection rig that was attached to the

i drain valve with transparent tubing. The inspector observed that air bubbles were

gurgling up the transparent tubing, indicating that air was potentially displacing

, water in the high points of the piping. About a half gallon of water was

{' subsequently determined to have been drained from the piping.

After maintenance activities were completed, the inspector determined that

I operations personnel had not planned to fill ard vont the piping. The assistant shift

supervisor indicated that a fill and vont of the piping was not necessary because

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the line had only been depressurized and not actually drained, however, he

proceeded with plans to vent the line in response to the inspector's concerns.

When the line was filled and vented, no air was noted at the high point vent

connection, indicating the line had not been substantially drained. However,

followup discussion with the assistant shift supervisor revealed that he was

unaware that the line had been depressurized using the drain valve instead of the

vent valve and that a half gallon of water was lost from the line.

M1.3 Maintenance Activities Performed Outside Scone of Work Order

a. Insoection Scone (62707)

The inspector observed the performance of portions of MWO 1-96-0708-00,

" Replace DH-100, Leak Test Valve for Pressurizer Auxiliary Spray Line," on

February 19,1997.

b. Observations and Findinas

The function of the auxiliary spray line was to provide for pressurizer pressure and

temperature control using the decay heat removal (DHR) system during times when

the DHR system was in service. DH-100 was a leak test valve replaced as part of

the MWO; its replacement required that part of the auxiliary spray line be drained.

The inspector noted that a minimum of about six feet of a 1.5-inch line was drained

of water to accommodate the replacement of DH-100. However, the drained line

was not filled and vented following completion of DH-100 replacement activities.

When the inspector questioned the assistant shift supervisor during the restoration

process, the assistant shift supervisor indicated that he felt there was just a small

amount of air in the line following the maintenance; he also added that it would

have proven difficult to completely fill and vent the line due to a lack of a high point

vent in the affected piping. However, he subsequently requested engineering to

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review the matter to assure that the as-left condition was acceptable. In addition,

1* actions were undertaken to attempt to fill the line.

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Engineering thereafter responded that with that portion of line completely drained,

no operational or structural concerns would result. No damage to the system from

a waterhammer event would be expected because of the low process fluid

i pressures and temperatures, coupled with a robust piping system mechanical  !

design. The inspector agreed with the engineering conclusions. ,

) A review of the MWO package revealed that work had been performed outside the ,

j scope of the MWO. Specifically, two MWO steps were not completed as written,

i nor was the package updated to reflect the actual work accomplished. Step 3 of  !

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i the MWO stated " Refill, pressurize, inspect and verify no leaks." However, as

i stated above, the piping was not refilled following completion of maintenance.

y Step 1 of the MWO stated "Tagout and drain piping using isolation valves DH-178

, to isolate the LPl/ decay heat removal system and DH-2736 along with DH-2735 to

l isolate the RCS." However, the isolation boundary for the work had been changed

j earlier to eliminate tbs need to tagout and isolate DH-2735 without the MWO being

i appropriately revised to reflect that change. Additionally, an action plan that was

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originally intended to be used as part of the original scope of the maintenance

! activity, had provisions for signature approval authority, but those had not been

! signed-off.

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! The performance of work outside the scope of MWO 1-96-0708-00 is considered a

t second example of a violation of 10 CFR Part 50, Appendix B, Criterion V,

j. " Instructions, Procedures, and Drawings" (50 346/97003-01b(DRP)).

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! c. Maintenance Related Conclusions l

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j Equipment restoration control and communications weaknesses were identified in

i two cases. In one instance, personnel performed work outside the scope of the

i MWO, including a tailure to perform a fill of an auxiliary spray line following

i completion of maintenance, in the second instance, operations shift management

was not aware that a section of makeup system piping had been drained of a half

j gallon of water through a drain line.

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M1.4 Imoroner Seouencina of Test Procedure Steos

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l a. Insoection Scone (62707)

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i The inspector observed operations personnel perform portions of DB-PF-03030

l (Revision 00), " Service Water Pump 3 Quarterly Test," on February 24,1997.

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! b. Observations and Findinas

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.The inspector determined that part of the test had not been accomplished according

to administrative procedure DB-DP-00013 (Revision 04), " Surveillance and Periodic

Test Program." DB DP-00013 Section 6.3.7.h stated in part that, " Test

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, prerequisites and procedure steps shall be performed in numerical sequence . . .

unless optional sequencing is allowed by the procedure . . ."

Contrary to that requirement, DB-PF-03030 steps 4.7.2 through 4.7.5 were

completed before step 4.7.1, with no optional sequencing provided in the

procedure. This is considered a third example of a violation of 10 CFR 50,

Appendix B, Criterion V, " Instructions, Procedures, and Drawings" (50-346/97003-

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l The initial performance of the test failed the pump acceptance criteria due to a

service water strainer blowdown valve being inappropriately open. The blowdown

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valve was then shut, the test was reperformed, and acceptable test results were

obtained. (Reference report Section M2.1 for additional discussion related to

l troubleshooting of the service water strainer blowdown valve control logic)

} Additionally, the inspector identified that the service water pump performance

, curve used in the test procedure did not match the pump curve provided in the

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pump curve book. (Reference report Section E3.1 for details)

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l. c. Conclusions

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! Operations personnel failed to follow the stat.on test control procedure that required

! steps be completed in sequence. However, although certain steps were performed

} out of order, results of the test appeared to be valid,

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j M1.5 Decav Heat Pit Water Tiaht Enclosure Leakaan Test Not Performed As Reauired

a. Insoection Scone (71707)

l On February 12,1997, the licensee initiated a unit shutdown per Technical

Specification (TS) 3.0.3 following identification that a TS required surveillance ,

associated with the decay heat removal system had not been performed as j

required. While conducting the shutdown, the licensee concurrently requested a  :

notice of enforcement discretion (NOED) from the NRC to allow termination of the j

shutdown. At approximately 8:00 p.m. EST, NRR issued the NOED (No. 97-6- 1

003), and the unit was returned to full power operation. The inspectors reviewed

licensee actions following discovery of the missed surveillance, monitored portions

of the unit shutdown, and subsequently assessed the causes and contributing

factors leading up to the need for the NOED.

b. Observations and Findinas

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-The shutdewn had been initiated in response to plant identification that TS

surveillance 4.5.2.f to perform a vacuum leakage rate test on the decay heat

removal watertight enclosure had not been conducted when last required. The

enclosure, located in containment, houses valves DH-11 and DH-12 in the decay

heat removal / low pressure injection system that are credited for functioning up to

seven days following a loss of coolant accident (LOCA). The vacuum leakage rate

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~, test verified that water intrusion into the enclosure due to post-LOCA flooding

would be limited to an acceptable amount. The TS specified that the test be

performed periodically as well as after each opening of the enclosure. On May 24,

1996, during startup from the last refuel outage, an inspection port on the top of l

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the enc!osure was opened to allow a visual inspection of the enclosure's interior per

ASME Section XI and NRC requirements, and a leakage rate test was not

subsequently performed.

The licensee then determined that the leakage rate test could not be conducted

with the unit at power and a shutdown was then commenced per TS 3.0.3 to place

the unit in a condition where the test could be done. The inspectors reviewed the i

licensee's preparatory activities for the shutdown as well as observed portions of

the actual shutdown evolution itself with no substantive concerns noted. The

shutdown appeared to be conducted in a controlled, well managed manner.

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The licensee requested the NOED based upon, in part, the design of the Kamlock

type fitting utilized to secure the inspection port and its inherent leaktightness.

Regional as well as NRR management were involved with the discussions with the

licensee.

At the time the NOED was issued, the unit was at 10% power and the main turbine

had been tripped. The turbine was thereafter relatched and the unit retumed to

power. 1

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The inspection port ut!Uzing the Kamlock fitting was installed in 1986 as part of i

facility change request (FCR)85-303. The inspectors reviewed the safety i

evaluation (SE) associated with installation of the inspection port and noted that the )

SE indicated that since the Kamlock fitting was specifically designed for the J

purpose of sealing, the TS surveillance would not have to be performed after

opening / closing of the port. However, no followup action was taken to request a  ;

TS change to allow this exclusion. l

Alternatively, although not an optimum situation, the licensee could have continued

to ' implement the TS surveillance requirement without requesting a TS change. The

periodic leakage rate test could have been performed during each refueling outage

as before. Heatup of the unit could have then commenced and once in mode 3, the i

required interior inspection of the watertight enclosure could also have been

performed as before via use of the inspection port. However, since opening of the

inspection port constituted opening of the watertight enclosure as specified by the

TS, a shutdown /cooldown of the unit would have had to be performed to support

re-performance of the leakage rate test. Once the test was reperformed, the unit

could have been restarted without having to reopen the inspection port again. I

At the time of discovery, TS 4.5.2.f required that the decay heat removal system 1

be demonstrated operable "by performing e ;scuum leakage test of the watertight

enclosure for valves DH-11 and DH-12 that asures the motor operators on valves

DH-11 and DH-12 will not be flooded fet at least seven days following a

LOCA: . . 2. After each opening of tta watertikht enclosure." Since the

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, inspection port was opened on May 24,1997, and a leakage rate test had not been

subsequently performed, the licensee was considered in violation of TS 4.5.2.f.

Although the required surveillance had not been conducted, the decay heat

remonitiow pressure injection system was not declared inoperable

(50-346/9700342(DRP)).

In addition, the SE also specified that " Confirmation of proper installation (of the

Kamlock fitting) shall be accomplished by independent verification." However, with

the exception of operating procedure DB-OP-06900, " Plant Heatup," this

requirement was not properly translated to all necessary documents. For instance,

no requirements for performing independent verification of the Kamlock fitting

installation following its opening with the plant at power were delineated.

Consequentially, on February 12, plant personnel made a containment entry to

verify the Kamlock fitting's proper installation, and during that evolution, the fitting

was opened /reciosed without formally performing an independent verification.

10 CFR 50, Appendix B, Criterion ill, " Design Control," specifies in part that

measures shall be established to assure that applicable regulatory requirements and

the design basis (including design changes) are correctly translated into

specifications, drawings, procedures, and instructions. Because the independent

verification requirements specified in SE 85-303 were not properly translated to all

appropriate plant documents, this is considered a violation of 10 CFR 50,.

Appendix B, Criterion Ill. However, because the licensee identified the violation

during the initial review of the TS surveillance issue, and as of the end of the

inspection period an evaluation was inprocess to determine the appropriate

procedures / documents within which the independent verification requirement

should be included, this matter is considered a non cited violation (50-346/97003-

03(DRP)), and a Notice of Violation will not be issued consistent with Section Vil of

the NRC Enforcement Policy.

c. Conclusions

Upon implementation of FCR 85-303, the plant failed to request a TS change to

eliminate having to do a leakage rate test on the watertight enclosure when the

inspection port was opened, nor was the current TS surveillance performed as

required. In addition, the SE requirement to perform independent verifications of

proper Kamlock fitting installation each time the inspection port was opened and

raciosed was not addressed in all appropriate instructions and procedures.

M2 Maintenance and Material Condition of Facilities and Equipment

M2.1 Service Water Strainer Blowdown Valve Failures

a. Insnection Scone (37551)

The inspector conducted a routine walkdown of the service water pump room on

February 10,1997.

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j J, b. Ohwations and Findinas

The inspector noted during the walkdown that the #2' service water strainer

blowdown valve, SW-1380, was open without the necessary conditions present for

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it to be open. SW-1380 was designed to open on a high service water pump

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discharge pressure or a high strainer differential pressure. Neither of these

conditions were present at the time.

Subsequently, the valve apparently reclosed prior to the plant engineer for the

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service water system being ablo to investigate. Because another similar situation

{ had recently occurred, the plant engineer initiated a work request to troubleshoot

j the valve's control circuit.

I- Additionally, on February 24,1997, the #3 service water strainer blowdown valve,

l SW-1381, was found partially open during performance of the #3 service water

pump quarterly surveillance test. This was the third time that a service water

i blowdown valve was unexpectedly found open. The licensee then initiated

potential condition adverse to quality report (PCAOR) 97-0235.

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! Subsequent troubleshooting efforts by pont personnel determined that an age- .

j related relay degradation in the control cirNitry was adversely affecting the logic's l

j response timing. Consequently, the station planned to replace the potentially i

degraded relay in the control circuits for all three service water blowdown valve

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! An engineering evaluation to address a worst case scenario irwolving a full open

blowdown valve, determined that, with an ultimate heat sink temperature less that

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50*F, up to 4000 gallons per minute (GPM) of service water flow diversion could

j be accommodated without adversely impacting system response to a LOCA. Since

1 maximum flow from a blowdown valve was estimated at less that 2000 GPM, and

! since the ultimate heat sink temperature was currently about 42'F, the inspector

j had no concems that the situation was an imminent safety concern.

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! However, inspector review of this matter was incomplete at the end of the

, inspection period. The relay degradation noted above could cause a blowdown

3 valve to remain in a full open position as noted in at least two previous cases, but

! did not appear to account for the one valve found in a mid-position. This matter is

I considered an unresolved item (50 346/97003-04(DRP)) pending completion of

I

inspector review,

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, M3 Maintenance Procedures and Documentation

! M3.1 Lack of Sorav Shield Proaram Administrative Controls i

l

a. Insoaction Scone (62707)

f The inspector observed portions of the performance of MWO-3-97-2047 01

'

" Containment Sump Pump Outside Containment Discharge isolation Valve VOTES

Testing," on February 3,1997.

.

b. Observations and Findinas

The inspector noted that disconnect cabinet CDF11C, located in #4 mechanical

penetration room, had its door open so that test equipment could be powered from

the cabinet's internal electrical outlet. However, the cabinet was marked with a

} " Spray Shield Required" placard that implied that the door was subject to water

i intrusion administrative controls. When station personnel became aware of the

l inspector's observation, potential condition adverse to quality report (PCAOR) 97-

0144 w9s initiated.

l 1

, 1

J

The maintenance craft had been using this outlet for several years. Apparently, 1

. they had received verbal guidance to have a person available to shut the door if l

l necessary, and to secure the door once the activity requiring use of the outlet was

completed. i

Subsequently it was determined that the cabinet was not required to be in the 1

i spray shield program and the spray shield placard was removed. Other plant l

equipment was found to have spray shield signs that did not require them as well.

} The inspector conducted a search of the plant procedures database to determine if

l any drawings or procedures related to the spray shield program. Drawing M-269

- documented the requirements for spray shield protection, however no guidance was

!

'

found that provided information as to how personnel should control the breaching

of spray shield enclosures.

I

c. Conclusions

i

Administrative controls governing the spray shield program appeared incomplete.

j informal verbal guidance was used to ensure that maintenance personnel would not

leave an open spray shield door unattended. Maintenance personnel failed to

question whether the breaching of a cabinet labelled as a spray shield was

appropriate.

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!* M4 Maintenance Staff Knowledge and Performance

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M4.1 Personnel Resoonse to a Comoonent Coolina Water Temnerature Controller Failure  !

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i s. Inanection Scone (71707)

!

l The inspector observed a component cooling water (CCW) heat excf. anger being

placed into service.

{

! b. Observations ad Findinas

1

l Operations shift personnel were investigating a CCW heat exchanger temperature

control valve problem that resulted in minor tempere'

s ucreases of components

i cooled by CCW. '

.  ;

i- Within minutes, operations dispatched an instrumentation and controls (l&C)

technician to investigate. The technician immediately determined that a linkage arm

had fallen off the temperature controller and then took actions to reattached it. The

j . temperature control circuit then started to function properly.

{

,

No temperature alarms were received when the CCW temperature excursion

occurred. Also, the CCW heat exchanger remained operable because the CCW

j temperature control valve would have opened as designed during design basis  :

accident conditions, which was consistent with its updated safety analysis report

(USAR) description.

l

I c. Conclusions:

!

I&C personnel conducted timely diagnostic and corrective action activities

{ for a failed CCW heat exchanger temperature controller.

, ,

MS Miscellaneous Maintenance issues (92700) (92902)  ;

M8.1 (Closed) LER 50-346/96-005-01, inadeouate Control of Heavy Loads in the

Containment Buildina: The event involved the inappropriate traversing of the

reactor vessel head lifting tripod over the open reactor vessel during the tenth  ;

refueling outage. The mattor was the sub}ect of previous NRC review which ,

'

concluded with the issuance of a violation (reference NRC Inspection Report

50-346/96005).  ;

M8.2 (Closed) LER 50-346/94-001-00, inoperable Emeroency Core Coolina System: This

LER documented a situation where containment radiation monitor RE2OO7 mode 6 ,

'

trip setpoint had been calibrated to a value greater than the allowable limit.

RE2OO7 provided one of four channels of containment radiation monitoring input to  !

the safety features actuation system (SFAS). The monitors were set to trip at  !

certain values during power operations and to trip at lower values when the plant [

was shut down (i.e., mode 6). ,

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l ~, The setpoints were established via facility change request (FCR) 82-0034.  !

l Technical Specifications required at least three of four containment radiation

'

monitors to be operable during core alterations. The licensee determined that core

!- alterations were conducted October 1 through October 7,1g84 with both RE2OO7 i

and RE2OO5 inoperable.

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l- The licensee determined having two containment radiation monitors inoperable did l

'

not adversely impact previous accident analyses. The slightly higher trip setpoint '

4

remained conservative to limits required to meet 10 CFR 100 offsite dose

j guidelines. The previously analyzed fuel handling accident took no credit for

3 automatic containment isolation, a function provided by SFAS. The setpoint was

subsequently adjusted to the appropriate value prior to personnel performing further

i core alterations.

f lil. Enaineerina

$ E1 Conduct of Engineering

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E1.1 Annandix R lasues Associated with Reactor Coolant Pumos

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. a. Insoection Scone (71707)

i ,

! The licensee recently identified several configuration problems associated with the

reactor coolant pumps (RCPs) involving a failure to meet Appendix R requirements.

The inspectors reviewed the licensee's followup actions.

b. Observations and Findinos

The licensee discovered that the RCP lubrication system piping was routed in such

a manner as to not have all potential leakage paths contained within the RCP's oil

collection system. Specifically the licensee determined that in one case oil pressure j

sensing lines associated with the lift oil system was located above the oil collection

trap. It was concluded that if a sensing line should leak while the lift oil system

was operating, a potential spray condition rould occur and result in the oil to not be

adequately collected.

l

The second case involved installation of a oil fill line whereby personnel could add  !

oil to the RCP reservoir from outside of a high radiation area. However,if the fill

line were to break while oil was being added, the leak could be directed outside of

the oil collection system.  ;

The third case involved a small length of drain line underneath of the lower oil .

reservoir that was located outside of the oil collection system. If the drain line I

were to leak or rupture, up to 25 gallons of oil could drain from the reservoir to

creas outside of the collection system.

These conditions were documented in licensee event report 50-346/97-004-00. It I

was subsequently determined that no significant safety consequences would result  ;

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[ from the identified configurations. Further inspector followup of these issues will i

j be conducted during review of the subject LER.

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E3 Engineering Procedures and Documentation

E3.1 Service Water Pumn Performance Curve Out-of-bate

i  ;

a. inanection Scone (37551)

'

The inspector observed portions of the performance of DB-PF-03030 (Revision 00),

" Service Water Pump 3 Quarterly Test," by operations personnel on February 24, '

1997.

l b. Observations and Findinas

.

l

The inspector discovered that the #3 service water pump curve in Attachment 5 of

l

DB-PF-03030 did not match the #3 service water pump curve that was in DB-PF-

l 06704 (revision 01), " Pump Curves," Drawing CC14.73.

I

The curve included in DB PF-03030 reflected data collected after the #3 service

i water pump shaft and casing were replaced in 1995 and appeared to represent the

l baseline parameters for the #3 service water pump currently installed. However,

]

the curve in the pump curve book was dated June 4,1991, and reflected a pump )

that was not currently installed.

.  :

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The inspector was concerned that the outdated curve in DB-PF-06704 could be

used by plant personnel to justify engineering and safety evaluations, and flowrate

4

determinations for radioactive affluent calculations. Pending further inspector

,

review of this matter to determine the extent to which the outdated curve was, or

could be used by plant personnel, this is considered an unresolved issue

(50-346/97003-05(DRP)).

E8 Miscellaneous Engineering lasues (92700) (92903)

E8.1 (Closed) Inspection Followuo item (50-346/95007-05(DRP)): Containment air

cooler (CAC) 1-3 electrical alignment not fully evident. By design, CAC 1-3 was

capable of being aligned to either train 1 or train 2 electrical supply. When aligned

to either train, the CAC AC electrical supply was directly monitorable via control

room indication. However, the associated DC electrical alignment was not evident.

1he AC electrical supply was swapped via manipulation of breaker switches. The

DC electrical supply was swapped at the same time the AC supply was swapped ,

via pickup / dropout of relay auxiliary contacts. l

Following discussions with engineering personnel and review of associated Institute I

of Electrical and Electronics Eogineers (IEEE) Standards 279,338 and 352, the i

inspectors determined that the electrical design for CAC 1-3 was in conformance

with standard industry design specifications and the subject relays and contacts

were appropriately tested periodically.

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E8.2 (Closed) Insoection Followun item (50-346/94004-03(DRP11: Nuclear

instrumentation divergence events. The licensee was unsuccessful in determining i

,

specific root cause for the subject divergence events. However, the number of

l

divergence events, their magnitude, and duration, continued to decline over the last

,

two operating cycles possibly due to current core design, loading, and operating

l . strategies. No significant adverse effects as a result of the current trend is .

i

foreseen. This matter therefore is considered closed.

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l E8.3 (Closed) Insoection Followun item (50-346/93017-02(DRP)): Testing problems

2

associated with component cooling water isolation valve CC1495. The testing

!. inadequacies associated with the subject valve were subsequently corrected.

During inspector and licensee review of this matter, a question was raised

l. concerning the setup of air operated valves in general. No additional failures

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associated with the setup of AOVs were identified. This specific matter is ,

considered closed, however, the inspectors will continue to assess the proper i

functioning of AOVs during future inspection activities.

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E8.4 (Closed) LER 50-346/96-002-01. Potential Loss of Remote Shutdown canability

Due to MOV Fire Induced Damaae: This event report documented that a licensee

.,

review determined that several valves were susceptible to failure similar to that

i

described in NRC Information Notice 92-18. NRC review of this matter was

i previously completed with issuance cf escalated enforcement (reference NRC '

j inspection Report 50-346/96008).

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$ E8.5 (Closed) LER 50-346/96-004-00. Inadeouste Compensatorv Actions for Thermo-Lao

! for Radiant Enerav Shields: This event report discussed a condition where Thermo- 1

3 Lag was used as radiant energy shields in containment and containment annulus.

l Although the use of Thermo-Lag in applications requiring non-combustible material

i

was subsequently determined to be unacceptable, appropriate compensatory

} measures were not initiated. NRC review of this matter was previously completed

.

with issuance of a Notice of Violation (reference NRC Inspection Report '

50-346/96008).

,

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IV. Plant Sunoort

R1 Radiological Protection and Chemistry (RP&C) Controls

The inspectors noted that good ALARA and contamination control practices were

utilized during maintenance activities observed during the period. Maintenance

personnel were cognizant of area dose rates and attempted to utiliza existing

structure for additional shielding when necessary. Radiation protection personnel

were observed to provide appropriate contamination control support.

The inspectors independently verified on a sampling basis that area radiation levels

were consistent with current surveys, radiation areas and high radiation areas were

properly posted and controlled, and contaminated areas were adequately roped off

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, and marked to prohibit station personnel from inadvertently crossing control

! boundaries. I

P2 Status of EP Facilities, Equipment, and Resources

I

l- The inspectors walked down the Technical Support Center and Emergency Control l

'

Center during the inspection period. Both emergency response facilities appeared i

j. well maintained in an appropriate standby condition with associated equipment

functional. Personnel access to the facilities was controlled per the licensee's  !

program.

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. 82 Status of Security Facilities and Equipment

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i Protected area ingress and egress monitoring equipment was observed to be i

j functional and in a good state of repair. Protected area lighting and intrusion  !

4

detection equipment were also noted to be properly functioning. Guard force  ;

members were observed to conduct normal rounds and security alarm response  ;

j activities in a controlled and timely manner.  ;

I (

F F2 Status of Fire Protection Facilities and Equipment  ;

!,

! Required fire protection equipment appeared to be appro,4iately staged and

! available in their designated plant locations. Fire brigsde members adequately

j staged their protective clothing and other necessary iPe fighting equipment before

i assuming their shift duties. Fire brigade team minimum manning requirements were

j routinely satisfied throughout the inspection period.  !

$

[ V. Manaaement Meetings  !

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X1 Exit Meeting Summary

The inspectors presented the preliminary inspection results to members of licensee

management at the conclusion of the inspection on March 3,1997. A followup meeting

was held on March 7,1997, where the final inspection results were discussed with

licensee management representatives. The licensee acknowledged the findings presented.

The inspectors asked the licenses 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

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licenses

J. K. Wood, Vice President, Nuclear

J. H. Lash, Plant Manager

R. E. Donnellon, Director, Engineering & Services

. T. J. Myers, Director, Nuclear Assurance

I L. M. Dohrmann, Manager, Quality Services

R. C. Zyduck, Manager, Design Basis Engineering

,

J. L. Michaelis, Manager, Meintenance

J. L. Freels, Manager, Regulatory Affairs

i M. C. Beier, Manager, Quality Assessment

i W. J. Molpus, Manager, Nuclear Training

L. D. Hughes, Manager, Supply

'

D. L. Esholman, Manager, Operations

i R. J. Scott, Manager, Radiation Protection

i G. A. Skeel, Manager, Security

H. W. Stevens, Manager, Nuclear Safety & Inspection

l J. W. Rogers, Manager, Plant Engineering

! D. M. Imley, Superinte.. dent, Operations

! D. H. Lockwood, Supervisor, Regulatory Affairs

j G. R. McIntyre, Supervisor, SYME

D. R. Wuokko, Supervisor, Licensing

M. J. Roder, Supervisor, OWCU

J. E. Reddington, Superintendent, Maintenance

! R. B. Coad, Superintendent, Radiation Protection

] T. J. Chambers, Shift Manager, Operations

4 R. B. Ewing, D. B. Supply

l D. Converse, D. B. Business Services Manager

l M. K. Leisure, Senior Engineer, Licensing

J. W. Marley, Senior Engineer

'

i C. Kraemer, Engineering, Regulatory Affairs

l G. M. Wolf, Engineer, Licensing

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INSPECTION PROCEDURES USED

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IP 37551: Onsite Engineering

IP 61726: Surveillance

$ IP 62707: Maintenance

IP 71707: Plant Operations

IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor

Facilities

IP 92901: Followup - Operations

1

IP 92902: Followup - Maintenance

, IP 92903: Followup - Engineering

l lP 93702: Prompt Onsite Response to Events at Operating Power Reactors

i

<

) ITEMS OPENED, CLOSED, AND DISCUSSED

Opened '

'

50-346/97003-01 VIO Three Examples of Failure to Follow Procedure

50-346/97003-02 VIO Leakage Rate Test Not Performed

} 50-346/97003-03 NCV Independent Verification Requirements Not Translated to All

j Appropriate Plant Documents

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50-346/97003-04 URI Service Water Blowdown Valves Found Open

50-346/97003-05 URI Outdated Curve Found in Pump Curve Book

Closed

[

i 50-346/95007-05 IFl Containment Air Cooler (CAC) 1-3 Electrical Alignment Not

.

Fully Evident <

i 50-346/94004-03 IFl Nuclear Instrumentation Divergence Events

50-346/93017-02 IFl Testing Problems Associated With Component Cooling Water

Isolation Valve CC1495

i 50-346/96-002-01 LER Potential Loss of Remote Shutdown Capability Due to MOV

l Fire Induced Damage

i 50-346/96-004-00 LER inadequate Compensatory Actions for Thermo-Lag for Radiant

Energy Shields

50-346/96-005-01 LER Inadequate Control of Heavy Loads in the Containment

1

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Building

50-346/94-001-00 LER inoperable Emergency Core Cooling System

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LIST OF ACRONYMS USED

w

AFW Auxiliary Feedwater

AOV Air Operated Valve

ASME American Society of Mechanical Engineers

CAC Containment Air Cooler j

CFR Code of Federal Regulations '

CNRB Company Nuclear Review Board

l

CS Containment Spray

dpm disintegrations per minute

ECCS Emergem:y Core Cooling System 4

EDG Emergency Diesel Generator

ESF Engineered Safety Feature

EVS Emergency Ventilation System

FCR Facility Change Request

GPM Gallons per Minute

HPl High Pressure injection

I&C Instrumentation and Controls

IFl inspection Followup Item

IR inspection Report

LAR Licensee Amendment Request

LER Licensee Event Report

LPl Low Pressure injection

MCC Motor Control Conter

MOV Motor Operated Valve

MWO Maintenance Work Order

NCV Non-Cited Violation l

NOED Notice of Enforcement Discretion

NRC Nuclear Regulatory Commission

NRR Nuclear Reactor Regulation  !

OSC Operations Support Center

PCAOR Potential Condition Adverse to Quality Report i

QA Quality Assurance i

OC Ouality Control

i

RCP Reactor Coolant Pump  !

RCS Reactor Coolant System

RO Reactor Operator l

RP Radiation Protection

SFRCS Steam / Feed Rupture Control System

SRB Station Review Board

TS Technical Specification

VIO Violation

22