ML20153G193

Insp Rept 50-320/88-05 on 880227-0401.Violations Noted.Major Areas inspected:defueling-related Operations,Abnormal Plant Events,Implementation of Radiological Controls & General Plant Housekeeping

ML20153G193
Person / Time
Site:	Three Mile Island
Issue date:	05/02/1988
From:	Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20153G169	List: ML20153G165 ML20153G180 ML20153G193
References
50-320-88-05, 50-320-88-5, NUDOCS 8805110166
Download: ML20153G193 (11)

See also: IR 05000320/1988005

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U. S. NUCLEAR REGULATORY COMMISSION .

REGION I

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' Report No. 50-320/88-05  !

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Docket No. 50-320  :

License No. DPR-73 Priority --

Category C l

Licensee: GPU Nuclear Corporation  ;

P. O. Box 480 ,

Midoletown, Pennsylvania 1*057  :

Facility'Name: Three Mile Island Nuclear Station, Unit 2  :

Inspection At: Middletown, Pennsylvania

Inspection Conducted: February 27 - April 1, 1988

Inspectors: R. Conte, Senior Resident Inspector i

• T. Moslak, Resident Inspector  ;

A. Sidpara, Resident Inspector '

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• Reporting r.spector  !

Approved by: 1 E S/2,/5'y f

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C. Cowgill, Chief, Reactor Projects Section 1A Date '

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Inspection Summary:

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Areas Inspected: Routine safety inspection by site inspectors of routine plant i

operations, defueling related operations, abnormal plant events, implementation  !

i of radiological controls, and general plant housekeeping.

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Results: One violation was identified for failure to perform the procedural pre-

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requisites for controlling hotwork activities in the Decontamination Facility in

the reactor building. Failure to do these prerequisites resulted in fires in the

. facility. Overall, housekeeping in the reactor building and in the balarge of the >

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plant areas has significantly deteriorated. Inspectors will monitor licensee's  ;

progress in improving this aspect of the fire protection and industrial safety

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DETAILS i

1. Plant Operations  !

1.1 Defueling' Operations

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During this reporting' period, defueling crews continued to use the drill- i

ing machinery to separate the in-core instrument guide tubes (IIGT) and

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support posts'from the Lower Core Support Assembly (LCSA). Problems that

initially slowed this process have been resolved and significant progress

has been made. ,

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Thirty-four of forty-eight support posts have been drilled-out of the

uppermost plate, the Lower Grid Rib Section (LGRS). Present plans call

for continuing to use the drilling machine to drill out selected support

posts then using this machinery, to make severance cuts of the grid in-

tersections of the LGRS. Performing these cuts will result in the LGRS

being cut into one large square section and twelve smaller sections.

Upon removal from the reactor vessel, the larger sections will be stored i

in a Core Flood Tank that has been reconfigured to accept sections of i

the LCSA. In series with the drilling operations, visual examinations,  !

using underwater cameras, and radiation surveys of internal components

are being conducted to assess the strategy for their eventual removal .

from the reactor vessel, j

On March 29, 1988, a small specifically outfitted robotic submarine began  :

removing debris from the pressurizer. The debris is picked up by slave j

manipulators on the device and deposited into a waste receptacle. Some t

i material was placed in the receptacle, but a loss in visibility, caused i

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by the device's propellers stirring up fine material, has slowed progress. i

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1.2 Decontamination / Dose Reduction Activities i

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l Removing thin layers of concrete (scarification) from reactor building i

basement walls has been completed. Robots performed this task. Trans-  !

fers of the concrete dust / water slurry will be made from the basement

to a sludge receiving tank in the auxiliary building for eventual pro-

cessing and shipment of the solidified waste to a commercial burial site,

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! Preparations are being made to fill the concrete block wall with water, }

located in the reactor building basement, in an effort to leach out the +

imbedded contamination. Subsequent to performing this task, the con- '

taminated water will be processed through the Submerged Demineralizer

System to remove the radioactive contaminants.

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Scabbling, steam vacuuming, and hands-on decontamination continue in the

auxiliary and fuel handling buildings. To date, 116 of 143 cubicles have  ;

been decontaminated to end point criteria. On March 29, 1988, approxi- i

j mately seven cubic yards of concrete were poured into the floor of the  !

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seal injection valve room to reduce the dose rate in that cubicle. This

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additional shielding, ranging in depth from seven inches to eleven inches,

appears to have reduced the general area dose rate in the cubicle from

about 60 R/ hour to about 1 R/ hour.

2. Fires in Reactor Building Decontamination Facility

2.1 Background

On February 22 and 27, 1988, fires occurred in the Decontamination (Decon)

Facility a temporary structure located on the 347-foot elevation of the

reactor building. The fires resulted when sparks from a plasma arc cut-

ting job ignited various materials installed and staged in the facility.

In both occurrences, the portable fire extinguishers issued to the fire-

watch failed to operate. However, both fires were quickly extinguished

using other methods. In the first incident, February 22, when a dry

chemical extinguisher was found to be spent, the torch operator extin-

guished the flames by "patting-out" the flames with a leather glove he

was wearing; and, in the second incident, February 27, when the CO2 ex-

tinguisher failed to discharge, personnel obtained a dry chemical port-

able extinguisher from another work area and used it to extinguish the

fire. Subsequent to extinguishing the first fire, the torch operator

and his firewatch left the reactor building because they felt nauseated

from smoke inhalation and reported to the first aid department for ex-

amination. No injuries or illnesses resulted from the second fire.

Critiques were conducted and Incident Reports were prepared for both

fires; however, following the second fire, site management stopped all

cutting work in the facility until an evaluation of the facility was

performed by the fire protection engineer and corrective actions result-

ing from that review were completed.

On February 29, 1988, the licensee's fire protection engineer inspected

the facility and identified deficiencies in the construction, conduct

of operations, and housekeeping conditions in and near the facility; and,

provided recommendations for upgrading the facility and re-training per-

sonnel in controlling hotwork operations. The appropriate departments

took actions to modify the facility, improve the housekeeping, and pro-

vide job briefings. The fire protection engineer presented the job

briefings to about 144 personnel on the requirements specified in the

control of hotwork procedure (4000-ADM-1100.05) for performing cutting

operations. The job briefings provided specific information to plasma

arc firewatch personnel, torch operators, task supervisors, and person'el

monitoring the work from the coordination center on pre-work requirements,

work area maintenance requirements, post work inspection requirements,

and outlined the actions required to be taken if a fire is discovered.

To verify implementation of the recommendations, inspections of the

facility were performed by a fire protection engineer and a safety

specialist. The facility was determined to be safe for conducting hot-

work on March 7,1988.

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During the critique of the first fire, the licensee learned that person-

nel were-not complying with the notification and reporting requirements

of the fire protection plan. Personnel standing fire watches in the

facility had not notified the control room of this fire in a timely man- i

ner, apparently because the firewatch assumed that there was no reason  !

to notify the control room of such occurrences. As a result, the Inci-  !

dent Event Report (No.88-014) for this fire that occurred on February  ;

22, 1988, was not prepared until February 24, 1988, with the critique

being held on February 25, 1988. This finding was assigned to management ,

to take actions to assure that task supervisors and crafts personnel

advise the control room of all fires and provide input to generate In- '

cident/ Event Reports in a timely manner. When the second fire occurred, '

the notifications and reports were expeditiously made within the licen-

see's organization and the NRC Resident Office was contacted. .I

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2.2 Inspector Findings

The inspector evaluated the conditions that resulted ir the fires in the  !

Decontamination Facility and licensee actions taken to prevent a recur- T

rence through discussions with licensee representatives, reviews of [

relevant documentation, and examination of conditions in the reactor i

building. From this evaluation, the inspector determined the following. '

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Fire protection engineers had not been in the reactor building to  !

assess fire hazard conditions to the Decon Facility from October

1987, until after the second fire, February 29, 1988. This lack *

of involvement is considered to be a contributing factor to the '

deteriorated housekeeping conditions in and around the facility. ,

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The initial engineering review on the Unit Work Instruction (UWI I

No. 4730-3100-86-C1415) did not adequately assess the combustibility .

of the materials used in the construction of the facility or the '

materials used in support systems; i.e., the fabric tool window and  ;

the ductwork of the ventilation system, respectively. '

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The quality of the corrective actions taken after the first fire

were not adequate to prevent another fire. I

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A sense of complacency regarding poor housekeeping conditions in

and around the Decon Facility was developed by task supervisors,

torch operators, and firewatch standers that resulted in cutting l

and welding permits being issued without the prerequisite fire

protection measures being implemented.  ;

With respect to the last item, the inspector established that licensee

personnel failed to implement the specific requirements of procedure  !

4000-ADM-1100.05, "Control of Hotwork," which states, in part, in Section i

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"The supervisor in charge of e welding, cutting, or grinding job shall

ensure the Job Site has been physically inspected prior to starting the

job. The person assigned by the Supervisor to inspect the area initials

and signs the rear of the permit. The Supervisor's signature on the

front of the permit documents the inspection verified the items on the

rear of the job in addition to the following prior to starting the job.

a. The job area shall be clear of transient combustibles such as trash

and rags within a 40 foot distance to protect against ignition by

slag or sparks.

b. All fixed combustibles, machinery, equipment, and cable in trays

subject to possible ignition or damage by sparks or slag shall be

protected by appropriate noncombustible guards or covers.

c. Fire protection and suppression equipment installed in the job area

shall be verified to be in service.

d. Appropriate portable fire extinguishers shall be provided at the

immediate job site. Size, type, and number of extinguishers should

be obtained to suit the hazard in any individual job area, but shall

include a minimum of one ABC dry chemical unit, and shall be in

addition to the normal compicment of extinguishers installed in the

area."

Failure to implement this procedure is contrary to the requirements of

Technical Specification (TS) 6.8.1, which references Appendix A of Regu-

latory Guide 1.33, Revision 2, February 1978." Administrative procedures

that implement the plant's fire protection program are referenced in the

subject Regulatory Guide (320/88-05-01).

The inspector noted that a more in-depth review in the front end pro-

cesses, i.e., design review and procurement of materials and support

equipment used in the construction and operation, could have reduced the

probability of such fires occurring. The inspector determined that the

licensee failed to address the quality of the initial engineering review

for the construction of the Decon Facility as part of their post-fire

evaluations. The failure to make this determination to improve the

quality of future fire protection engineering reviews is considered a

weakness in the licensee's post-fire critique process.

On April 1,1988, the inspector entered the reactor building to assess

radiological, fire protection, and industry safety conditions. The re-

suits of this assessment are addressed in paragraph 3.0 of this report.

However, with respect to housekeeping and fire protection conditions,

the inspector concluded that housekeeping is very poor.

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Specifically, the inspector observed that transient combustible materials;

i.e., paper towels, card board tags, plastic bags, and plastic sheeting

were carelessly scattered around work areas, and low usage spaces on the

305-foot elevation, 347-foot elevation, and on top of the "A" D-ring.

Through discussions with the lead fire protection engineer concerning

the controlling of transient combustibles into the reactor building, the

inspector determined that the licensee document the transient combustible

materials taken into the building to verify that the maximum permissible

combustible loading of the reactor building is not exceeded but does not

have lower administrative action levels to assure that unused combustible

material is removed from containment in a timely manner.

The fire protectior. engineer stated that this item will be given consi-

deration in the upgrading of the implementation of the fire protection

program at TMI Unit 2.

3.0 Reactor Building Conditions

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(Closed) Unresolved (320/87-14-02)- Reactor Building Housekeeping

On Friday, April 1,1988, inspectors toured the reactor building to

evaluate fire prevention, industrial safety, and radiologic.a1 conditions.

The inspectors determined that housekeeping on all elevations had de-

teriorated in that paper towels, cardboard tags, plastic bags, and other

transient combustible materials were scattered in work areas and low

usage areas. The inspectors observed that the high rad waste storage

area was congested and that the waste drum used for disposing of plastic /

tape materials from protective clothing was filled to overflowing. There

were kick boards missing from the open batch on the 347-foot elevation

and a loose locking device on a temporary stairwell. Appropriate licen-

see managers were briefed on these findings. Subsequent to their dis-

cussions, licensee personnel took quick action to correct the deficien-

cies identified by the inspectors. The corrective actions to improve

the conditions in the reactor building will be evaluated during the in-

spector's next entry into the reactor building.

The inspector verified that the licensee took action to address specific

inspection findings identified in Inspection Report No. 50-320/87-14 and,

accordingly, this item is closed. However, the inspector determined that

the licensee should upgrade the overall implementation of the housekeep-

ing program in the reactor building and in the balance of plant areas

(see Section 4). To improve and to correct the deficiencies identified

during the current inspection period, these actions will be evaluated

on a regular basis as part of the routine inspection effort (320/88-05-

02).

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4.0 General Balance of Plant Conditions

The inspector, on March 9, 11, and 25, 1988, toured various areas in the

auxiliary / fuel handling building and turbine building. The licensee was

advised of the findings identified in these tours. As a result of the tours

and followup discussions with the licensee, the following findings / conclusions

were made.

4.1 Radiological Areas

Behind door AX-105 (Motor Control Center Area), several pieces of con-

taminated equipment were found in the corner without any radiation zone

barrier. Licensee advised that due to low level contamination and the

fact that the equipment was bagged and had proper radiation survey data,

no additional barrier was necessary. The inspector verified that ade-

quate measures were in place.

In the change room, near Locker No. 38, a barrel designed for clean trash

contained some anti-C's. Also behind the same locker area, more anti-C's

were noticed. The inspector was advised that, as a normal practice, the

barrels, even those marked with "clean trash," are sorted as a precau-

tionary measure and the anti-C's, if found, are removed. The inspector

had no further questions.

Several bags containing contaminated articles were found all over the

area. They were properly surveyed and had proper radiation tags. The

situation, however, created a housekeeping problem. The licensee has

initiated stricter administrative controls. A few days later, the in-

spector toured the same area and noted significant improvement.

Several radiation tags were found loose on the floor. Loose tags could

mean either the contaminated equipment has been disposed of but not the

tag or the contaminated equipment is still in the area without the tags

and, therefore, could cause further contamination probleas. Licensee

has counseled the supervisory staff to be aware of such situations and

ta take appropriate actions.

The inspector also witnessed the tool decontamination operation inside

a Plexiglas glove box fabricated by the licensee in the past. The glove

box system includes a solvent for decontaminating the tools, High Ef-

ficiency Particulate Absolute (HEPA) filter on top of the glove box, as

well as a vacuum unit to remove any airborne contamination. The glove

box has four gloves, two on opposite sides, as well as an access door

on one side. The following observations were made.

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The decontamination operation was in progress and the vacuum pump

was not running. The inspector noted that licensee stated that the

pump should be on anytime the glove box is in use,

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The operator was not aware of "y operating procedure, inspection,

or testing requirements for the glove box. He had received on-the-

job verbal. instructions only.

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The operator, when questioned on the pump operation requirement

turned the unit on, which created a very unpleasant odor. The pump

was turned off immediately.

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There was some solvent in the tray inside.the glove box. The

operator was not aware of the type of solvent being used.

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The-inspector also found two cracks on two sides of the glove box.

These cracks represented a potential leak path.

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The inspector asked the group radiological control supervisor about

the decontamination procedure,-solvent, inspection or testing pro-

cedure, etc. None was available.

Further review by'the inspector identified that there is no operating

procedure for the glove box. Also the glove box has never been inspected

or tested. There is no requirement regarding solvents. There is no

formal training on operating or maintaining the glove box. The inspactor

was also informed by the licensee that one of the cracks was through the

Plexiglas wall and the glove box has been put out of service until it

is repaired and required operating and testing procedures are developed.

The inspector also learned that the solvent being used is known as "simple

green," which is non-toxic, non-flammable but very effective for decon-

taminating the tools.

Since there are other means of protection in place, such as continuous

air monitors, area radiation monitors, routine radiological surveys, as

well as controls on work activities, no immediate radiation hazard ex-

isted. The licensee, however, took the appropriate actions to correct

the problem.

4.2 Temporary Shielding

The inspector reviewed the licensee's procedure on the control of tem-

porary shielding (TS). At the present time, the TS's, once installed,

are tracked on the quarterly audit program. The audit addresses the TS's

on a random basis and basically verifies the location, date installed, I

and date audited. The audit does not include an engineering assessment

. of the TS, it's adequacy, effectiveness, or any planned actions to

eliminate them. Since the quarterly audit does not cover all the TS's

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and since there is not any criteria to cover all of them in a certain

r time frame, it is possible that some of the TS's may be left unaudited.

The other radiological controls that are in place, however, are adequate

l to detect any significant change in the radiological conditions (refer-

j ence paragraph 4.1) and, therefore, does not cause any significant

i radiological hazard. The licensee, however, has planned to reassess

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their TS program and then implement the necessary corrective actions to ,

strengthen the current program. The inspector had no further questions

on this matter. ,

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4.3 Defueling Test Assembly  !

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, The inspector toured the defueling test assembly (OTA). No work was in

progress at that time and the access door was open and the DTA was un-

manned. According to the licensee, the access to the DTA is controlled

and the door should have been closed when it is unmanned. The licensee

has reinforced this administrative control. "

The floor opening on the top of the vessel at the DTA was found to be

inadequately guarded. One of the protective sides was removed and the

safety barrier (rope) was at the top. Therefore, the potential for

anyone falling into the opening existed. The licensee corrected the  ;

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problem immediately by installing another barrier (chain) at the bottom.

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At the DTA, the floor on the reactor vessel loop is made out of removable  :

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wooden pieces. One of the pieces was found to be unsafe. The licensee  !

corrected this problem right away and also inspected and strengthened

the rest of the floor area as appropriate. The inspector had no further

questions on this matter. ,

4.4 Liquid Radwaste Instrumentation

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At the waste control panel area on the annunciator panel, a high

level WDS-T-1A alarm was blinking. The actual level indication was  !

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9.3 feet, which is lower than the set point, as shown in procedure '

{ 4215-RPR-3233.01. Following investigation by the licensee, it was

found to be only a spurious alarm. The observed condition could I

not be repeated,

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The auxiliary building sump level is indicated by a temporary gauge  ;

installed under mechanical modification No. 138. The calibration >

requirement was not clear and the actual sump level is determined l

by subtracting ten from the indicated readout and then sividing by [

eight. This gauge makes it harder to readily determine the auxili-

, ary building sump level. The temporary installation was made ap-

! proximately four years ago. The inspector was informed that the

] gauge is periodically inspected and calibrated. An engineering  :

! evaluation is also performed periodically to assure its reliability. ,

! The licensee, however, is looking for a different kind of level (

) indicating gauge as the permanent installation.  ;

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Miscellaneous waste hold-up tank WDL-T-2 has a temporary level gauge. '

The original gauge has been out of service for about four years,

j The calibration sticker on the temporary gauge was found to be blank.

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The inspector was informed that the gaugo will be recalibrated. l

The ink on the calibration sticker was faded even though it was-

calibrated per established frequency.

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The inspector had no further questions regarding this matter. I

4.5 Posting of Radiation Surveys t

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The inspector reviewed radiation surveys in the auxiliary building con-

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trol point book and found a total of nine out-of-date surveys. The  ;

maximum elapsed time was about two weeks. The inspector also noted the '

daily check sheet for this time were signed off indicating all the radi- '

ation surveys at the control point were verified to be current. j

At the end of the tour, the inspector was advised by the GRCS that the  ;

current surveys were conducted, but they were locked in a file cabinet "

and not distributed at the control point.

The inspector was informed by the licensee that there was no new Radi-  :

ation Work Permit (RWP) was to be issued, since there was not any sig-  !

nificant change in the radiological conditions as indicated on the cur- i

rent radiation surveys. l

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Licensee's procedure 9200-ADM-4110 .04, "Radiation Work Permit," para-  !

graphs 4.3.5.1 and 4.3.5.2 require distribution of radiation surveys once  ;

i they are reviewed and approved by the GRC5. Also, per procedure 9200- l

., ADM-4200.01, "Radiation Surveys," paragraph 4.4.6, the GRCS is to ensure '

that surveys are performed as scheduled and that, at the end of each .

shif t, the schtdule (for radiation surveys) should be checked for com- I

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pleteness and status system updated upon completion of the radiation *

surveys. The licensee has corrected this problem by providing additional  ;

instructions to GRCS regarding items to be checked when signing off the l

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check sheets.

, 4.6 Miscellaneous Findings  !

. In addition to the findings identified in the above paragraphs, the '

inspector made the following observations during his plant spaces review.  ;

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In the fuel handling building, one blower with No. U4713-100, which  ;

is used to maintain negative pressure in the annular space between i

i the auxiliary building and the fuel handling building was operating *

with excessive vibration and potentially could cause a safety hazard '

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for the personnel working in the area. The inspector was informed  !

! by the shift supervisor that the blower has been operating in that i

condition for a long time; however, the maintenance department will  :

be requesced to evaluate the blower. ,

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Radiation tag, dated October 13, 1987, was found on the floor near

pump DC-P-1A.

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One radiation caution tag marked "Hand Tool," dated March 15, 1988,

was found on the floor near SF-C-18, but no hand tool was in that

area.

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The lube oil site glass on nuclear service pump NS-P-1B does not

have the protectivn cage. The glass, if broken, potentially could

cause loss of lube oil.

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Near door No. AX-112, an oil leak was found through an oil lubri-

cating device mounted approximately five feet above the ground level.

A followup tour indicated that this problem was corrected.

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A pair of anti-C's was found on the floor near the demineralizer

filter system unit on the 305-foot elevation.

Following the above-mentioned plant walkdowns, the inspector briefed

several licensee staff members. The licensee was responsive and took

immediate corrective actions or planned future actions emphasizing safety

significance.

The inspector had no further questions regarding these items.

5.0 Exit Interview

The inspectors met periodically with licensee representatives to discuss

inspection. On April 8, 1988, the site inspectors summarized the inspection

findings in a meeting with the following personnel:

P. Barth, Fire Protection Engineer

J. Byrne, Manager, TMI-2 Licensing

E. Gee, Industrial Safety and Health Manager

S. Levin, Defueling Director

T. O'Connor, Lead Fire Protection Engineer

W. Potts, Site Operations Director

R. Rogan, Director, Licensing and Nuclear Safety

F. Standerfer, Director, TMI-2

D. Turner, Director, Radiological Controls, TMI-2

D. Tuttle, Manager, Radiological Controls Field Operations, TMI-2

At no time during the inspection was written material provided to the licensee

by the inspectors.

Unresolved Items are matters about which information is required in order to

ascertain whether they are acceptable, violations, or deviations. Unresolved

items discussed during the exit meeting are addressed in Section 3.