IR 05000344/1990019

Insp Rept 50-344/90-19 on 900611-15.Violations Noted.Major Areas Inspected:Followup of Items of Noncompliance, Occupational Exposure During Outages & Keeping Occupational Exposure ALARA

ML20055H571
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/11/1990
From:	Coblentz L, Tenbrook W, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20055H568	List: IR 05000344/1990019 ML20055H567 ML20055H569
References
50-344-90-19, IEIN-90-008, IEIN-90-031, IEIN-90-31, IEIN-90-8, NUDOCS 9007270010
Download: ML20055H571 (19)

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i APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION'

REGION V

Report No.. 50-344/90-19 License No. NPF-1 Licensee:

Portland General Electric Company 121 SW Salmon Street Portland, Oregon 97204 Facility Name:

TrojanPlant.

Ic9ection at:

Rainier, Oregon Inspection Conducted:

Ju J1-15,1990

' Inspected by-

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W. KF'ienB ook,, adia on Specialist

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p. Q6blen

,RafiationSpecialist D&te 51gned

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I[78 Approved by:

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i G. Pi Rihas Chief

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Dat'e Signed I

ReactorRadIologicalPro[ectionBranch Summary:

Areas Inspected:

Routine, unannounced inspection cf followup items, fol'10wup-of items of noncompliance, occupational. exposure during outages, and keeping

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occupational exposure As Low As Reasonably Achievable (ALARA).' Inspection procedures 90713, 92701, 92702, 83728, and.83729.were used.

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Results: The' licensee's programs for radiationLprotection during outages were maintaining previous levels of performance.

Several improvements in tie areas of work control and planning were identified during review of the-ALARA.

i program (Section 3).

One violation-was identified.that involved exceeding.

I several Limiting Conditions of Operation that require the containment atmos)here particulate radioactivity monitoring instrument (PRM-1A) to be

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opera)1e.. A second violation identified the failure to properly report the discovery of filter paper depletion in PRM-1A.

A third violation identified

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the lack of procedures required to ensure operability of PRM-1A.

For-

-discussion of the violations, see Sectioh 3 of this report.

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9007'70010 900711

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DETAILS.

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Persons Contacted Licensee t

C.-Allen, Quality Assurance:

J. Cross,, Branch Manager, Nuclear Regulation S. Bauer

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Vice-President Nuclear L

N. Dyer, Supervisor,~HealthPhysics-

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l G. Huey, Supervisor, Radiation Protection l

M.- Murdock, Supervisor, Rt :ioactive Material l

B. Naik, Quality Assurance

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R. Nelson, Manager,-Nuclear Safety Regulations Division G. Rich, Branch Manager,: Radiation Protection

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.i W. Robinson,-Plant Manager l

C.' Seaman, General Manager, Nuclear Quality Assurance J

T. Walt, Vice President, Nuclear, Acting

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J. Whelan, Manager, Maintenance W.- Williams,. Regulatory Compliance-

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R. Barr,_ Senior Resident Inspector B.Olson,ProjectInspector The individuals listed above attended the exit meeting on June 15, 1990.

u The inspectors met-and held discussions with additional members of the-licensee's staff during the inspection.

2.

Review of Periodic Reports (90713)

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t The inspectors conducted an.in-office review of the Operational l

Environmental Radiological Surveillance Program 1989' Annual Report,

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submitted in accordance with-the requirements of Technical Specification:

(TS) 6.9.1.4 and 6.9.1.5.

The report provided data, interpretations analyses of radiological environmental samples and measurements in

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l accordance with the program described in TS 3/4'.12.

Results of the 1989 t

land use. census, results of licensee participation in the Environmental

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Protection Agency Laboratory Intercomparison Program, and a summary of 1989 Quality Control Analyses were presented.. Comparison with previous environmental surveillance reports' supported the conclusion that-plant-related airborne radioactivity, direct radiation,in the-and aquatic activity among other dose pathways, were not detected environme,nt.

The report summarized data in accordance'with the format of

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TS. Table 6.9-1 and summarized the program in accordance with the format of Regulatory Guide 4.8(1975).

The presence.of. Cs-137 and Cs-134 was noted ia milk samples from one of-the four sampled dairies, Location 63.

Levels up to 51 picocuries per.

liter were detected.. The licensee attributed this activity to Chernobyl

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fallout and the tendency of goats to concentrate activity.in milk (less milk for equal-consumption, as compared to cows)..The inspector noted that although activity was detected in 12 of 23 samples at Location 63,

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.no activity was detected at Location 17a,:also a goat dairy.: The l-

. licensee stated that Location 63 used a higher percentage of locally radionuclidesinthe-goats!oatdairy. milk indicated that the licensee's facility.

grown feed than the other The absence of s1 ort-lived-

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lE was.an unlikely source for the detected activity.

The presence offlow'

i level' Cs-137 and Cs-134 in the milk.had been consistent since 'a dramatic

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spike was observed in April 1986,.following_the Chernobyl incident..

g Sample analysis systems appeared to achieve-lower limits of. detection i

(LLDs) at or below the levels required by the TS.-

No samples indicated-activity-in excess of-the reporting limits specified in TS Table: 3.12-2.

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The licensee.had maintained their prior level of performance in l

environmental monitoring.

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3.

Followup (92701)

t-Oaen Item 50-344/89-07-01 (Closed):. This item concerned the adequacy ofl tie licensee's. quality assurance and quality control of chemical

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measurements.

Multipoint calibrations were used for most instruments,

~ Atomic absorbtion/ emission spectrophotometers were calibrated with a:

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single point standard and blank;.however,-the calibrated range was very

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near.the LLD, and any measurement in excess.of the calibrated range was

. diluted and reanalyzed within the calibrated range.--The inspector verified that the in-line ion chromatography system was validated and verified for acceptable precision, accuracy and-sensitivity.

Control standards were prepare'd using reagents. independent of those used

for instrument calibrations.

Control check logs and com chartswerekeptpursuanttoChemistryProcedursCP-54',guterizedcontrol-Performance Trending Program Data Collection." ' Anal tical. verification using spikes.

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withthellcensee'edbyall; technicians,y and blinds analyz was implemented in accordance s Chemistry Manual.and CP-54.

The licensee had implemented a measurement control ~ program 1 incorporating

l the items noted lacking in Inspection Report 50-344/89-07.

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0)en Item 50-344/89-30-02 (Closed):

This item concerned the-adequacy of r

tle single point sampling configuration of PRM-1 when monitoring.the 54" diameter purge duct.

In=the FSAR, paragraph 11.5.2.2,'the. licensee commits to meeting the recommendations'of ANSI 13.1-1969, which

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recommends,.in Appendix A, " Guides for Sam minimum of six. sample points for ducts 50"pling from Ducts and Stacks," a and larger in diameter.

However, based on an engineering analysis that' included experimenta1L samplin medium)g across the-duct at' standard flowrates (using helium as the

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, the' licensee determined that the-use-of a single sample probe.

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was adequate.

This analysis cited the provision from= Appendix:A of the'

previously mentioned standard, that " fewer withdrawal points may be used if careful studies show that uniformity'of. composition exists throughout!

the cross section of the duct." The inspector concluded that, based on the licensee's analysis, use of a. single point sampling configuration to monitor the purge duct was--justified.

Open Item 50-344/89-30-03 (0 pen):

This item concerned the need for evaluation of unmonitored. leakage to the environment through the 54"

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Th?ge valves when PRM-1 is-aligned to monitor containment atmusphere.

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pr s item had been entered into the licensee's Commitment Tracking List-fo: completion by March 20, 1990; however, at the time of this

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inspection no evaluation of changes in leakrate at-conditions of' lower -

i positivedIfferentialpressure'hadbeenperformed,andnoquantification

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of the magnitude of possible unmonitored releases through this pathway had been conducted. - This matter will be addressed during a subsequent

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Open Item 50-344/89-30-07 (0 pen):

This item concerned the status of the;

-Ticensee's Radwaste Action Plan.

The licensee's revised Process Control

Program was near completion.

The new program would incorporate at description of the radwaste solidification processes with a reference-index for specific procedures and parameters applicable'to each process.

Cargo vans for interim storage of dry active waste had been obtained However, storage-of waste liners within shields exposed to weathering had

'l continued, since insufficient van storage was available, and many-shields l

were'too massive'to be safely placed in a van.

The-inspector noted the-.

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current inventory of radwaste as the outage ended. This item radwaste inventory, will-be reviewed during a subsequent-inspe,.and the ction.

Information Notices 50-344/IN-90-08 (Closed), 50-344/IN-90-31 (Closed),

j 50-344/IN-90-31 (Closed): - These Notices involved Kr-85 hazards from^

spent fuel, radwaste form and containers, and sources of radiation-1l exposure from spent fuel' pools. - The referenced Notices were distributed

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to cognizant personnel for evaluation.

Open Item 50-344/90-01-30 (Closed): - This item concerned the discovery, y

on January 30, 1990, that the. filter paper was depleted in the

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containment atmosphere' particulate radicactivity monitoring. system -

(PRM-1A).

The event was initially re report under 10 CFR 50.72.b.2.iii.C. ported to NRC in a 4-hour telephone Subse-the licensee determined the event not to be reportable, quently,inded the report.

and resc No-written followup was submitted.

The inspectors reviewed licensee procedures technical specifications (TSs) andwrittenrecords,includingPGEInternal-

Event Reports90-025,and 90-031.

In addition,.the inspectors interviewed personnel involved in recording ~and assessing-

circumstances surrounding the event.

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Background and Sequence of Initial Events j

PRM-1 has four channels which monitor containment in non-accident i

conditions: - PRM-1A' monitors airborne particulate; PRM-18 monitors

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iodine; PRM-1C monitors-low-level noble-gas and PRM-10 monitors high-level noble gas.

DuringnormaloperatIon,PRM-1is.alignedto monitor containment atmosphere as an input to Reactor Coolant System Leakage Detection.

Duringcon'cainmentpressurereductions through the lydrogen vent system (known in the industry as " burps"),

PRM-1 is aligned to monitor the hydrogen vent system.

During containment aurges, conducted only while shut down, PRM-1 is aligned to monitor t1e purge duct.

In either of the two latter. alignments, a high radioactivity alarm from any one of the four channels will

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' initiate a Containment Ventilation' Isolation signal (CVI),

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i automatically ensuring that-all containment ventilation valves are closed; in this capacity PRM-1 provides an input _to the Engineered Safety-Feature Actuation System (ESFAS).

l 1855 on January 30,'1990

At approximately(P0T) 26-2 on the Auxiliar, during perforo nce of Periodic Operating Test

'l monitoring system (PRM-2), a " paper tear"y Building vent exhaust

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alarm was received.

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auxiliary operator (AO) performing the test contacted the shift chemist, who replaced the filter paper; the PRM-2A " paper tear" alarm, however, would not-reset. The A0 and the shift chemist then inspected the per

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supply for PRM-1A to verify proper installation of-the paper supp in t

PRM-2A.

At this time'it was observed that the paper supply for P-1A-

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was depleted.

Although this condition should have occasioned both a

" paper tear" alarm and a " low paper" alarm on the PRM-1 Peripheral Entry Controller:(PEC),noalarms-hadoccurred. The Control Room 0)erator was notified, and the shif t chemist replaced the filter paper in

)RM-1A.

Corrective' Action In response to the inoperable alarms, the hydrogen vent system was tagged

.out and Chemistry began taking grab samples every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,. in accordance

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with TSs 3.3.3.11,'3.3.2, and 3.4.6.1.

Troubleshooting of the alarms

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revealed-a-broken wire on the " paper tear" alarm bypass switch, and a

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broken microswitch for the " low paper"' alarm.' The licensee noted that,

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s since the microswitch serves as the mechanical stop for a' spring-loaded -

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arm when the supply roll is removed, the microswitch had probably been

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brokenduringthegreviouspaperchangeout.

A cause for the broken wire

on the " paper tear alarm bypass switch was=not determined.

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'he " paper tear". alarm was verified to The broken' wire was re

.be operating properly". paired, &

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low pe:gr"y Modification (TM)90-003 was written on Tempo m l

PRM-1A, removing the alarm from service, since-a replacement-microswitch was not readily available.

On February 2,1990, with TM c

90-003 in effect, PRM-1A was returned to operable: status.

TSs 3.3.3.11,s

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3.3.2,-and 3.4.6.1 were exited.

l In addition to the immediate corrective actions described above, the L

licensee proposed revising. Maintenance Procedure 2-32, " Post Accident ~

' Airborne Monitoring Radiation Monitor Calibration," to add semiannual

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testing of the " low paper" and " paper tear"! alarms for PRM-1A and-PRM-2A.

This revision had not been approved by the date of the inspection; The

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inspectors also noted that a policy had beenLinforma11yiestablished for chemistry technicians to verify the filter _ paper supply on a weekly basis, at the time of' collecting the weekly composite sample.

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L Licensee Evaluation Since no record had been maintained of filter paper changeouts, the.,

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licensee attempted to determine how long PRM-1A had been without-filter

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paper by asking chemistry technicians w1en the last changeout had beenL

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Internal Event Report 90-025 stated that;although all chemistry

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technicians were questioned, n'one could recall when paper was last changed.

Using the linear footage of a replacement roll of paper and l

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assuming a rate of movement of 1" per hour, the licensee calculated that j

paper s1ould have been changed every 30 to 40 days.

No further attempt i

was made by the licensee to deteimine how long the depleted paper

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condition had existed.

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In addition, the licensee cited a passage from NUREG-1022, " Licensee

Event Report System," Appendix II, " Questions and Answers from the LER i

Workshops," which states in part:

"In general for the purpose of evaluating the reportability of situationr tcund during surveillance tests, it should be assumed

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that the situation occurred at the time of discovery, unless there l

is firm evidence to believe otherwise."

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The licensee concluded that the fi'.ter paper depletion in PRM-1A should

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be assumed to have occurred when found,'and that the event, therefore,

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was not reportable.

The 4-hour report was subsequently rescinded, and no 30-day LER was submitted.

Finally, the licensee concluded that the event had no safety

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significance, based on several assumptions:

1) that all releases are evaluated independently by a weekly composite sample 2) that a-particulatereleasewouldbeaccompaniedbyagasrelease;and3)that the iodine and noble gas channels were operable and would have isolated

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the release path should an abnormal condition have occurred.

In addition, the licensee noted that all release paths have HEPA filtration L

that removes mest of the particulate activity.

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f Subseouent Events

On Fet.'uary 16, 1990, as described in PGE Internal Event Report 90-031,

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PRM-1A was found to have a filter paper speed of 2.5" per hour, rather than 1" per hour, as had been assumed at the time of discovery of depleted filter paper. As in the earlier case, the hydrogen vent system was tagged out and Chemistry began taking grab samples every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to comply with TSs 3.3.3.11, 3.3.2, and 3.4.6.1.

After experimentally determining that the PEC could be reset to maintain paper speed at 1" ger l

I hour, a deviation to Operating Instruction 2-5 " Radiation Monitoring, was processed, to require the proper paper spee,d setting on-future startups of the unit. The TS action statements were then exited.

The licensee concluded that faster paper speed was not safety-significant, since PRM-1A was used for

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. analyses, and since upward or' downward. qualitative and not quantitative trends could still be observed.

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The event'was determined not to be reportable.

A concurrent licensee examination of PRM-2A revealed that its filter i

paper was not moving at all, due to problems with the filter paper i

advance mechanism.

A maintenance request was issued to repair PRM-2A.

NRC Evaluation of Safety Sionificanc_e Since the " low paper" alarm was used in the )ast as the basis for filter

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paper replacement, the licensee had not esta)11shed an independent.

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I periodicity for how often paper should be changed. The estimate of 30 to i

40 days was baead on a paper speed of 1" per hour discovery that paper

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saeed had actually been running at 2.5" per hour Indicated that paper cianpouts should have been occurring about every 2 weeks.

The fact that

no caemistry technician could recall the last performance of this routine i

evolution, combined with the inoperable status of both paper alarms,The

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suggested the paper may have been depleted for an extended period.

i inspector asked whether this possibility had been addressed.

A licensee representative stated that one chemistry technician had recalled

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discovery, paper in PRM-1A about 4 months prior to the Januarybut had consid replacing 30, 1990,

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changeout.

The inspectors' efforts to determine the time of filter paper depiction

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l in PRM-1A were hampered by the inoperability of the strip chart recorder (RR-4043)associatedwiththeunit.

The drive cable for the recorder had i

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l been broken since October 19, 1989. -This condition prevented the print

head from moving, rendering the strip chart portion of RR-4043 t

l ineffective.

RR-4043 was repaired on February 27, 1990.

The inspectors reviewed records for containment pressure reductions recordedonOperationInstruction(01)10-3,"ContainmentHVAC" Attachment A, discovering filter paper depletion in PRM-1A DurIn

" Containment Pressure Reduction Checklist."

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days prior to eight l

hydrogen" burps"hadbeenperformed,foratotalofover1$0hoursduring i

which some combination of hydrogen vent valves had been open.

A significant increase in average PRM-1A background counts was recorded on

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Attachment A for hydrogen " burps" performed after February 2 1990 when PRM-1A was declared operable and known to have filter paper Instal, led.'

No corresponding increase was observed in the low-level noble gas channel (PRM-1C).

A graph of these background counts is presented in Attachment I to this report.

The inspectors noted several reasons that the liceaee's use of Appendix II of NUREG-1022 might be considered inappropriate as a basis for not

reporting the event.

First, a surveillance was not in progress on PRM-1A

when paper was found to be depleted.

Second, no scheduled maintenance or r

surveillance to verify the condition of the. filter paper was routinely

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Instead, the licensee relied upon the paper alarms.

Finally,

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the licensee did not a*Lempt to establish the time of paper depletion in PRM-1A by other methods, such as examininn procurement records or

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- inventoriesoffiltergaper,evaluatingfluctuationsinbackgroundcounts recorded for hydro burps " or reviewing control room or A0 logs for thelastrecordedgenlow paper alarm on PRM-1A.

The licensee's evaluation stated that the iodine and noble gas channels

were operable, This statement did not address the bases for TS 3.3.2,and sho l

had occurred.

" Protective and Engineered Safety Features Instrumentation," which states l

in part:

"The OPERABILITY of these systems is required to provide the

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overall reliability, redundancy and diversity assumed available in the i

facility design for the protection and mitigation of accident and

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transient conditions." The statement also did not note that TS 3.3.2 requires all four channels to be operable.

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Furthermore, the licensee's evaluations of safety significance did not address the requirement of TS 3.4.6.1, " Reactor Coolant System Leakage Detection Systems " that the particulate channel of PRM-1 te operable.

The licensee's Final Safety Analysis Report (FSAR), Section 5.2.5.4,

" Sensitivity and Response Time," states in part:

"Tne Containment-particulate monitor'is the me t sensitive instrument of those available for detection of reactor coolant leakage into the Containment.

This instrumer,t is capable of detecting particulate radioactivity in concentrations as low as IE-11microcuriespercubiccentimeter(uCi/ce)ofContainment air....

"The Containment radioactive gas monitor is inherently less sensitive (threshold at IE-7 uCi/cc) than the Containment air particulate monitor..."

The inspectors also noted that the Victoreen technical manual for PRM-1 stated that sensitivity of the p. articulate channel was calculated using the sample flow and filter speed parameters and that detectable range varied with filter speed.

AppendixC,"CalIbrationData."FigureB

" Sensitivity Data," showed that the efficiency, detectable range, an.d sensitivity of PRM-1A were each based on a filter speed of 1" per hour.

The impaired sensitivity of PRM-1A for reactor coolant system leakage detection due to the absence of paper or fast paper speed was not addressed as safety significant in PGE Internal Event Reports90-025 and 90-031.

As a result of the inability to confirm when PRM-1A was last operable, the misuse of NVREG-1022 guidance, and lack of assessment of the affected.

TS and their bases, the licensee's evaluation of the event did not assess whether limiting conditions for operation had been exceeded.

Regarding the licensee's corrective' action, the inspectors noted that i

PRM-1A was returned to service the alarms on a periodic basis, prior to implementing procedures to testand with th due to unavailable parts.

Further, the policy for chemistry technicians to check the filter paper supply when taking the weekly composite sample l

had not been formalized.

Finally, the inspectors were unable to find any

evidence that paper changeouts since the event had been documented, i

Related Licensee Event Reparts (LERs)

LER 89-27 documented that PRM-1D had caused a CVI during a hydrogen

" burp." Although PRM-10 hcd no history of electronic spiking determined that the signal had seen caused by a " period, the l and th recorded activity level built u) over a 6-minute icensee.

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spurious electronic spike," after performance of POT-26-2 verified the operability of PRM-1A, IB, 10 10, and IE (accident channel). -The inspectors noted that POT 26-2didnotverifyfilterpapersupply.

Eight other LERs for CVIs from PRM-1 had been submitted in the 2 year period prior to January 30, 1990.

Causes of these CVIs included electronic spikes, testing errors, inadequate procedures, and actus1

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increases in containment airborne radioactivity levels.

Each of the four PPJi-1 channels used for normal operations had caused at least one CVI.

NRC Evaluation of Technical Specification Requirements TS 3.3.2, " Engineered Safety Feature Actuation System Instrumentation,"

includes the following Limiting Condition for Operation (LCO):

The Engineered Safety Feature Actuation System (ESFAG)

instrumentation channel, and interlocks shown in Table 3.3-3 shall be OPERABLE...

ACTION:

b.

With an ESFAS instrumentation channel inoperable, take the I

action shown in Table 3.3-3.

Table 2.3-3, Section 3.b, " Containment Ventilation Isolation," requires a minimum of one operable thannel in PRM-1A during Modes 1, 2, 3, and 4,

-and refers to Action 17, rhich states:

With less than the Minim e Channels OPERABLE, operation may continue provided the containment ves.9 1ation valves are mainta',ned closed.

TS 3.4.6.1 " Reactor Coolant System Leakage Detection Systems," includes thefollowlngLCO:

The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a.

The containment atmosphere particulate radioactivity mcnitoring systr<m...

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ACTION:

With one of the ab6ve required radioactivity monitoring a.

leakage detection systems inoperable, operations may continue up to 30 days provided:

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The other two above. required leakage detection systems are OPERABLE, and

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Appropriate grab samples are obtained and analy ed at'

least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

otherwise, be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />."

The inspectors 'noted that 10 CFR 50.36.c.2, " Limiting Conditions for Operation," states that LCOs are "... the lowest functional capability orperformancelevelsofequipmentrequiredforsafeoperationofthe facility." The inspectors concluded that the licensee s assumption that filter paper depletion occurred at the time of discovery, with no apparent attempt to assess the significance of operating outside the above-L ed LCOs, constituted inadequate attention to' standards of safety i

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delineated by TSs. The inadequacy of this assumption was accentuated by the fact that no chemistry technician could recall replacing the filter paper in PRM-1A for a period of approximately 4 months prior to January 30~, 1990, even though paper replacement should have occurred about every 2 weeks.

The. inspectors concluded, further, that operating for an extended period prior to January 30 1990 without maintainin andwithPRM-10. inoperable,gappropriategrahcontainmentven valves closed withouttakin samples every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TSs3.3.3.11,3.3.2,and3.4.ppearedtobeinviolationoftheLCOsof a

6.1(50-344/90-19-01).

NRC Evaluation of Licensee Procedures TS 6.8.1 states in part:

Written procederes shall be established, implemented and maintained covering the activites referenced below:

a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, November,1972.

Regulatory Guide 1.33, Appendix A, Part C, states in part:

Instructions for... startup, shutdown and changing modes of operation should be prepared, as approp,riate, for tie following systems:

22.

Process Radiation Monitoring System Part H of the same appendix states in part:

1.

Procedures of a type appropriate to the circumstances should be provided to assure that tools, gauges, instruments controls,

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andothermeasuringandtestingdevicesareproperly controlled calibrated, and ad;usted at specified periods to maintain ac, curacy.

Specific examples of such equipment to be calibrated and tested are:

readout instruments... alarm devices...

The inspectors noted that the licensee did not have procedures to periodically check oper.ibility of the " low paper" and " paper tear" alarm devices, or to ensure timely filter paper replacement, or to properly adjustpaperspeedduringunitstartups.

The inspectors concluded that the absence of procedures necessary to ensure operability of PRM-1A was directly related to the failure to promptly identify the condition of the monitor and appeared to be a violation of TS 6.8.1, as quoted above (50-344/90-19-02).

NRC Evaluation of Event Reportability i

10 CFR 50.73, " Licensee event report system," states in part:

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The holder of an opersting license for a nuclear power p(lant(licensee)shallsubmitaLicenseeEventReport LER) for any event of the type described in this paragraph within 30 days after the discovery of the event

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2.

The licensee shall report:

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B.

Anyoperationorconditionprohibitedbythe plant s Technical Specifications...

The inspectors noted that cperation with PRM-1A inoperable was prohibited by the TSs, as described above.

The inspectors concluded that the licensee's failure to submit an LER within 30 days after discovering the depletion of filter paper in PRM-1A without any substantial evidence that the paper had been recently changed appeared to be a violation of 10 CFR 50.73.a. as quoted above (50-344/90-19-03).

Open Item 50-344/90-12-01 (Closed):

This item concerned workers Incurring unnecessary radiation exposure due to poor planning of work in controlled areas, and due to the absence of.a mechanism to withdraw secondary maintenance requests (MRs) for which the primary MR had been-cancelled. The inspectors noted that a change had been made to Administrative Order (AO) 3-9, list the identification, numbers of all" Maintenance Requ cover sheet of primary MRs to related secondary MRs. This practice was also being incorporated into the MR computer tracking system, by a software change that " tagged" primary MRs with associated secondary MRs.

In addition organization, the inspectors noted efforts by the licensee's' planning to avoid a recurrence of the 1990 outage planning problems..

The first pre-outage meeting for 1991 was held on June 15, 1990.

A preliminary 1991 outage scoae memorandum had already been issued at'the time of the inspection with a detailed job breakdown comparable to that issuedonlyamonthprIortothe1990 outage.

The outline of pre-outage preparation milestones established for 1991 was considerably advanced from prior years.

The inspectors deteroined that these changes, if proparly maintained, would significantly inorove the capability of tne Radiation Protection Planning Group to maintain occupational exposures ALARA.

4.

Followup on Corrective Actions for Violations and Deviations (92702)

Open Item 50-344/89-30-01 (Closed): This violation involved failure to follow A0 3-9 " Maintenance Requests," during installation of a temporary l

modification (,TM) on the hydrogen vent system.

The item also identified several corresponding weaknesses in the work ::ontrol program.

The inspectors noted that the licensee's immediate corrective action had included retraining of workers, temporary pipe markings, and a system walkdown to ensure that the proper train of the hydrogen vent system was being monitored.

In December 1989 permanent labels were installed on hydrogen ventilation trains A and B at their point of connection to the:

containment purge exhaust ductwork. 'A revision was issued to A0 5-8, i

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" Temporary Modifications," to improve control of TMs issued for both trains of a system. At the time of the inspection, the licensee was

evaluating use of identical field and maintenance copies of MRs is an additional measure to improve work control.

All corrective action

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commitments stated in the licensee's Response to Notice of Violation, dated February 20, 1990, were completed.

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0)en Item 50-344/89-30-04 (Closed):

This item involved a deviation from tie n w for Inaaeq cy of m sy-072 as a means of sampling the hydrogen

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vent system during drogen " burps." TM 89-072 did not meet the criteria of ANSI N13.1-1969 or isokinetic sampling.

During the 1990 outage, the

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id was removed, and a permanent design change (ROC 88-023) was installed.

Final inspection and approval were in process at the time of the

inspection.

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The inspectors reviewed the drawings, safety analysis, and engineering design input record for ROC 88-023, and determined that the ANSI

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criteria for isokinetic sampling, d sample probeincluding piping i

N13.1-1969 configuration flow velocity distribution an e

confiouration,,f RDC 88-023 gives PRM-1 the capability to directly m were adequately addressed.

The inspectors noted that

installation o either hydrogen vent line.

In addition, RDC 88-023 addressed such

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considerations as design temperature and pressure conditions, rcquired

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materials, TSs test requirements, and maintenance recommendations.

All correctiveactIoncommitmentsstatedinthelicensee'sResponsetoNotice of Deviation, dated February 20, 1990, were met.

5.

Occupational Exposure During Outages (83729)

Audits and Appraisals The inspectors reviewed the final report for PGE QA Audit CKS-202-90, dated Mey 4, 1990, which examined programs for radiation. protection and control of nuclear material. (Auditor qualifications and preliminary-results of the audit were examined in April 1990, identified apparent and discussed in Inspection Report 50-344/90-12.).Theauditteam-minor discrepancies in radiation protection training,ive sources, and respiratory protection, posting of hot spots, control of radioact procedure adequacy.

Audit results were being appropriately addressed.

yet complete,of this inspection, some of the corrective actions were not At the time i

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The inspectors also examined preliminary results of an audit of liquid

and gaseous radioactive waste systems, for which the final report had not yet been issued.

This audit generated six findings and seven observations, documenting several noncompliances with regulatory guides, TSs, and the FSAR.

The licensee maintained its previous level of performance in this area,

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and the audit program was adequate in meeting requirements of ANSI /ANS-3.2/N18.7, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants."

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External Exposure Control The inspectors examined external doses received by.:.lected licensee and

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contract workers under the following Radiation Work Perm.ii;s (RWPs):

90-0505:

Preparation and termination of s%am generator eddy j

current testing.

I 90-0507i Eddy current testing of a B C and D steam generators.

90-0513:

Valve maintenance includin,g s,upport work.

90-0522:

Steam generator tube plugging, plug removal and sleeving.

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.'0-0802:

Radiation protection surveys, coverage and calibrations (in containment)

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90-0815:

Minor valve maintenance inside containment.

90-0848:

Valve maintenance on MO-8808A, B, C and D.

90-0719:

Refueling-Reactor assembly.

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l The inspectors examined dosimetry evaluations of prior dose history, i

administrative approvals for dose limit extensions, and actual

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accumulated doses for the workers selected.

Most of the workers had l

received doses in excess of 1250 millirem during the second calendar quarter, 1990.

The inspectors verified that each worker exceeding 1250 i

milliter in the calendar quarter had had their exposure history

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documented on NRC form 4 or ecuivalent, and that all doses received fell l

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within the 5(N-18) accumulatec dose limit, N representing the persons'

l age in years.

All selected workers had accumulated less than 2000 millirem during the second calendar quarter, 1990.

The inspectors conducted independent tests of the licensee's' digital

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alarming dosimeter (DAD) system established to inform workers of their doseinthefieldandtocontrolentriesintoradiologicallycontrolled

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areas.

The DAD system contained each employee's administrative dose limit keyed to the employee's thermoluminescent dosimeter (TLD)/ security

badge number, and provided a dose alarm setpoint and a display of

millirem dose during each entry.-

A simulated accumulated dose of 450 millirem to an employee with a 500 l

millirem administrative limit was entered into the computer database.

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The inspectors and radiation protection personnel tried.to gain access under these conditions and alarmed the dosimetry station.

This and other tests verified that the DAD system would deny access to the' controlled

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area for any person with an accumulated DAD dose within 50 millirem of their administrative dose limit.

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A simulated RWP was established in the database with a simulated accumulated dose 150% that of the simulated ALARA collective dose i

estimate.

The inspectors and radiation protection personnel attempted to

gain access undet the simulated RWP, but were denied.. This test verified l

that access limitations would be imposed on each RWP at 150% of its

estimated collective dose.

The alarm feature of the dosimeter was not tested, as it required actual radiation exposure.

Based on documentation and system parameters, the

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DAD was set to alarm during work when 85% of the remaining dose to the administrative limit was exceeded, j

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During review of doses incurred by workers, a DAD accumulated dose of 7025 millirem was observed in an RWP Access Summary List.

The inspectors i

immediately questioned the licensee about this value.

The full dosimetry record for the individual was obtained, revealing zero record dose for i

the TLD used.

The DAD transaction had also been corrected.

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i Radiation protection personnel observed that the 7025 value corresponded t

to tk individual's badge number.

The licensee explained that the DAD

terminal allowed manual entry of.an accumulated dose to permit dosimetry

clerks to enter the largest whole body dose received by an individual using multiple dosimetry.

As workers periodically logged themselves out

of the DAD system, there had been occasions where workers had

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inadvertently entered their badne numbers as accumulated dose.

Although

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this isolated instance had been identified and corrected in the DAD

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system, the 7025 millirem value had been transferred to another database, and had persisted there until the inspectors' observation.

The inspectors and radiation protection personnel discussed the potential for workers to inadvertently or intentionally enter improper accumulated i

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dose into the DAD terminals, resulting improperly controlled access which

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could lead to an exposure above administrative or regulatory limits.

The licensee stated they would restrict this function to dosimetry and

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radiation protection personnel.

The licensee's program for external dosimetry and access control had maintained the level of performance observed during prior inspections,

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andwascapableofmeetingsafetyobjectivesforprotectionofpersonnel

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from radiation exposure.

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Control of Radioactive Materials and Contamination, Surveys and Monitging The inspectors reviewed selected records of radiation, contamination, and

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air particulate surveys conducted during May and June 1990.

The surveys

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were consistent with licensee procedures, and in accordance with requirements of 10 CFR 20.201.b.

Weekly summary sheets indicated that all survey's required by RP-114, Mar.agement reviews were timely." Radiolo Schedule, had been conducted.

The inspectors conducted several tours of the' Containment Building, Auxiliary Building, and radioactive material storage areas, Dose rate-due for calibration July 19, 1990.

The inspectors observed the follow 015843, surveys were conducted using ion chamber survey instrument NRC items:.

Housekeeping practices were excellent. Work. areas and step-off pads in general, ware well-kept, with radwaste, debris, and laundry neatlysequesteredintoconvenientbagsandreceptacles.

startup, with Containment was particularly clean in anticipation o r

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control boundaries established to preclude generai access in areas that had received final cicanliness inspections.

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in general, Personneldosimetrydevicesandprotectiveclothinglskingequipment, were properly used by. workers.

Portal monitors, fr

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and radiation monitoring instruments were consistently used.

Radiological postings were consistent with survey map inforniation, and with dose rate surveys performed by the inspectors. Monitoring instrumentation was in current calibration and periodically performance checked.

During a backshift tour,itallic gaskets on the residual heat removalan inspect p( omress to replace flexThe task was performed on several scaffolding Hl) system.

p atforms in the RHR room on the 5' elevation of the Auxiliary Building,insideacontamInationarea.-Theinspectorobservedthat all personnel involved in the task were wearing protective clothing-required by the RWP, including plastic clothing worn by workers on the primary platform.

In addition personnel seemed well-prepared and knowledgeable regarding their p,ortions of the task, and had obtained necessary tools prior to entering the contaminated area.

The inspector observed, however, the following problems with contamination control:-

a Before detensioning the flange to open the RHR system a large yellow plastic bag containing absorbents was taped in, place under the flange.

The bag rested on the primary scaffolding platform, approximately 12 feet above the floor of the room.

The contractor radiatica protection technician (RPT) covering.

the task stated that the bag was in place to catch 4 or 5 gallons of water expected to leak from the opened RHR flange?

The inspector asked if the bag was expected to catch and contain all the leakage.

The RPT stated that. it didn't really matter, because with the contamination levels already present in the RHR room only make the ro,om ' cleaner" (less contaminated).anyRHRleakagethatesc

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Water leaking from the flange dripped both inside and outside the bag, and the maintenance technicians on the platform attemptedtoadjustthebagseveraltimestobettercontainthe leakage.

After the bag contaird severa' gallons, it began to leak.

Leakage frai bcth inside and outsile the ba primary platform, and dripped on the floor below. g wetted the The RPT placed additional absorbents on'the platform and on the floor.

When most of the dripping from the flange had sto) ped, Holding the RPT climbed up to the primary platform to remove the sag.

the bag (which now contained 4 or 5 gallons of RHR leakage) in one hand, the RPT attempted to descend the vertical ladder from the secondary scaffolding platform. When his free hand failed

to grasp the second rung of the ladder, the RPT fell backward i

off of the platform,RHR room, and dropp ng the yellow bag. landing heav

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the entrance to the The inspector rendered aid, but the RPT dec ared that he was all'

right, and continued to cover the task..

i Approximately ten minutes later, the-inspector observed the RPT cleaninkupthewateronthefloorundertheprimaryplatform.

The RPT s protective clothing.had been slightly disarrayed by

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f.he fall, and the inspector observed a portion of exposed skin i

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i between the hood and coveralls.

Drippage was still observable-from the flange and from the primary platform to the floor at this time.

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Although no_ contamination of the RPT or other personnel resulted t

from this incident, the inspectors determined that the practices i

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observed did not constitute proper contamination control,.and the i

backshift lead RPT was informed.

The RPT covering the task was

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promptly counseled.

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The inspectors reviewed Radiological Event Reports 90-13 through 90-38, to determi.1e whether the observed deficiencies were common practice.

Although 1,1 stances were noted involving personnel contamination, the incident described above appeared to be an isolated occurrence. The

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number of personnel contaminations for the 1990 outage was considerably.

reduced from numbers in previous years.

t maintaining its previcus level of performance in this area,peared to be With the exception of the occurrence noted, the licensee ap

and the associated radiation protection programs appeared to be accomplishing the licensee'ssafetyobjectives.

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6.

Maintaining Occupational Exposures ALARA'(83728)

A review of certain portions of the licensee's ALARA program was 4'

conducted in April 1990,ALARA deficiencies observed in tie area ofand the results re 50-344/90-12.

Specific

planning were ct.rrected by the licensee, and are discussed above under

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Section 3, Item 50-344/90-12-01. ~The inspectors examined additional

aspects of the ALARA program by observation, discussions with responsible

personnel, and review of applicable records and procedures.

ALARA Goals and Objectives

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The inspector compared the ALARA Status Report for June 13, 1990, to a

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similar report from April 18, 1990, to determine what changes had been made to dose estimates since the last inspection..This comparison

indicated that more than 50 new RWPs had been added in dose estimates of over 150 person-rem.

Inaddition,ratotalincrease fo

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the dose estimates for 17 RWPs had been substantially increased, for a total increase in dose estimates-of approximately 17 person-rem.

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l The licensee was maintaining its annual site goal of 350 person-rem, l

although the total dose'of 250 person-rem accumulated at the time of the

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I inspection indicated that the end-of-outage total would fall well below

theprojectionof315 person-rem.

A review of the current ALARA status

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report 'ndicated that a significant portion of this difference was due to l

fob cancellations and reduction in outage scope.[ob estimates is discussed Accurac i

j Review of the ALARA Planning 1991 Bud'et Request revealed several s>ecificobjectivesoftheRadiation rotectionPlanningGroup(RPPG)for t1e upcoming year.

Theseobjectivesincludedsignificantcomputer i

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upgradestoimprovetrackingofmaintenancerequestslmentaluseofwater

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radiation work l

permits, survey data}elding installation time up rades to. communication and personnel MPC-hours; exper l.

shields to reduce sh

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equipment used to support work inside the Radlolo ically Controlled Area; i

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and in)rovements to video equipment used in plann ng, training, and l

pre-jo) briefings.

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consideration to increase ALARA staffing.

One Health Physicist had been hired as a permanent addition to the RPPG, and was scheduled to begin i

work-in July 1990. In addition, requests had been submitted to improve'

the knowledge and industry awareness of RPPG members by participation in

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specific industry seminars and short courses, i

i ALARA Results

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Discussions with the RPPG Supervisor indicated that improved steam

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generator mock-up training had significantly improved the' ALARA awareness -

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of workers performing associated tasks.

Other ALARA initiatives that i

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resulted in substantial dose savings included use of SM-10 " probe l

l pushers" for remote operations inside the steam generator manway, continued use of glove-bags for the manway swing.and other tasks, an/

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installation of shield doors on the steam generator robotics.

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1990, to determine The inspectors reviewed RWPs completed as of June 13, jobs.

Eighty-two how accurately doses had been estimated for specific RWPswerelistedascomplete(notincludingcancelledRWPs);ofthis number, 7 RWPs had exceeded the original dose estimate by greater than 35 percent, and 38 RWPs had been completed for less than 65 percent of thc original dose estimate.

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t The inspectors determined that the substantial revisions t'o dose-estimates that occurred well into the outage, as well as the high degree of inaccuracy in estimating dose could be attributed largely to the poor

planning and ineffective schedullng identified in the previous inspection.

Again, the inspectors concluded that impt w d timeliaess of

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planning the 1991 outage, if consistently maintained;.Luld assist RPPG

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in effectively estimating dose.

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The licensee's ALARA program was adequate in meeting its safety objectives.

Overall awareness of the-importance of planning in maintaining exposures ALARA had improved since the previous inspection.

l 7.

Exit Meeting (30703)

The inspectors met with licensee management on June 15, 1990, to discuss the scope and findings of the inspection.

The inspectors emphasized the t

weaknesses in evaluating the duration and safety significance of filter paper depletion in PRM-1, identifying a potential deviation from FSAR commitments for containment particulate monitoring and a potential violation of 10 CFR 50.73 reportability requirements.

The inspectors-statedthattheirfindingswouldbesubjecttofurtherreviewbyNRC management.

Licensee representativ>es acknowledged the inspectors concerns, and committed to submitting an LER regarding the filter paper

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depletion.

In addition, the licensee stated that specific measures would be taken to preclude recurrence of the event, i

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(taken from Operating Instruction 10-3, " Containment HTRC," Attachment AJ

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