IR 05000373/1987020

Insp Repts 50-373/87-20 & 50-374/87-20 on 870701-15. Violation Noted.Major Areas Inspected:Solid Radwaste Mgt & Transportation Programs,Gaseous & Liquid Radwaste Programs & Operational Radiation Protection Program

ML20236N456
Person / Time
Site:	LaSalle
Issue date:	08/05/1987
From:	Greger L, Michael Kunowski, Paul R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20236N405	List: IR 05000373/1987020 ML20236N401 ML20236N423
References
50-373-87-20, 50-374-87-20, NUDOCS 8708110507
Download: ML20236N456 (14)

Text

.

%

U. S. NUCLEAR REGULATORY COMMISSION

REGION III

l Reports No. 50-373/87020(DRSS); 50-374/87020(DRSS)

Docket Nos. 50-373; 50-374 Licenses No. NPF-11; No NPF-18 Licensee: Commonwealth Edison Company Post Office Box 767  !

Chicago, IL 60690 l Facility Name: LaSalle County Nuclear Station, Units 1 and 2 l Inspection At: LaSalle County Station, Marseilles, Illinois j

,

Inspection Conducted: July 1-15, 1987 Inspector: R. A. paul r f'?

Date AA l Accompanying Inspector: M. A. Kunowski I T7

-

Date (Alh Approved By: L. R. Greger, Chief 7 Facilities Radiation Dat'e

'

Protection Section Inspection Summary Inspection on July 1-15, 1987 (Reports No. 50-373/87020(DRSS);

No. 50-373/87020(DRSS))

Areas Inspected: Routine, unannounced inspection of solid radwaste management and transportation programs, gaseous and liquid radwaste programs, and the operational radiation protection program. Also reviewed were past inspection findings, open items, licensee event reports, and the licensee's radiological controls during fuel transfer Results: One violation was identified in one area (failure to adequately perform calibration in accordance with procedures - Section 8).

8708110507 870006 3 PDR ADOCK 0500

.

-

i DETAILS

!

1. Persons Contacted

• L. Aldrich, Rad / Chem Supervisor A. Bailey, Operations Staff, Radwaste

• R. Bishop, Services Superintendent L. Blunk, Instructor, St.pport Services Training for Rad / Chem Technicians

• D. Brown, Superintendent, Quality Assurance (QA)

G. Diederich, Station Manage r J. Fulton, Instructor, Operations Training for Station Operators

.

'

• D. Hieggelke, Lead Health Physicist E. Huerta-Pavia, ALARA Coordinator J. Jalovec, Health Physicist l P. Leheney, Instructor, Operationi Training for Station Operators J. Lewis, Rad / Chem Supervisor Staff W. Luett, Rad / Chem Supervisor Staff

• P. Manning, Assistant Superintendent, Technical Services

• M. Myrick, Corporate, Nuclear Servicas Health Physics, Health {

j Physics Supervisor j

• P. Nottingham, Lead Chemist J

• J. Renwick, Production Superintendent  !

• M. Richter, Technical Staff, ATSS Compliance Group Leader j J. Schuster, Chemist L. Shearer, Operations Staff, Radwaste Coordinator C. Sprunger, Technical Staff, Environmental Quality Group j D. Trager, Group Leader, Support Services Training j

• Jordan, NRC Senior Resident Inspector R. Kopriva, NRC Resident Inspector The inspectors contacted other licensee and contractor personne l

• Denotes those present at the exit meetin i

2. General

]

This inspection, which began at 9:00 a.m. on July 1,1987, was conducted to examine the solid, gaseous, and liquid radwaste management programs, the transportation program, and aspects of the licensee's operational health physics program. Also reviewed were corrective actions for past inspection findings and open items. Several tours of controlled areas were made to review postings, access controls, effluent monitors and the radwaste shipping facilities. The inspectors performed direct radiation and surface contamination surveys during the tours. One violation concerning failure to follow a procedure was identifie _- _ _ - _ - _ _

_------

-

!

.

3. Licensee Action on Previous Inspection Find'ings (0 pen) Open Item (373/84006-03): Installation of-additional. permanent-shielding around several Unit I containment penetrations. The )

installation of the shielding.is awaiting completion of the modification of reactor level-sensing' lines. Temporary shadow shielding-and' access 1 control are being maintaine I (Clcsed) Open Item (373/85014-02; 374/85014-02): Alteration of.the'

Standby Gas Treatment System (SGTS) lines to increase elemental iodine transmission. .The. radiation sampling lines of the SGTS have:been

'i modified and new 2.0 SCFM pumps have been installed. The modified lines and new pumps will reduce plateout of iodine in the sampling. line Testing of the equipment has been-completed. Station Operations 'j Engineering and QA staff persons have approved routine use of the .. )

equipment. The Final Safety Analysis Report (FSAR)-will' be update (Closed) Open Item (373/85030-04; 374/85031-05): Development of an inspection program to verify stored container integrity while containers ,

are stored in' the Interim Radwaste Storage Facility (IRSF). As suggested i in NRC Generic Letter, 81-38, a quarterly visual inspection 'of container J integrity will be performed. The inspection will be conducted with.the )

IRSF crane and crane camera following Procedure LOS-WX-Q (Closed) Open Item (373/85030-05; 374/85031-06): Review the. potential for hydrogen generation from resin stored in the IRSF and develop a monitoring system if necessary. Using a computer program distributed b Electric Power Research Institute (EPRI), the licensee has determined that hydrogen gas generation in a 55 gallon drum of radwaste woul probably not result in a concentration greater than 3.3% after ten years of storage; therefore, a monitoring system is not necessary. The i determination was based on the isotopic. activity of two " worse-case" i

drums currently in storage in the Radwaste Building. Drums will not_be stored in the IRSF longer than five years. Hydrogen concentration in

~

drums and liners is limited to 5%. The EPRI report, EPRI NR-4938, Project 2558, was issued March 1987, and is titled " Methodology for '

Calculating Combustible Gas Concentration in Radwaste Containers."

(Closed) Open Item (373/85030-06; 374/85031-07): Further NRC review of 10 CFR 50.59 evaluation of the IRSF. The NRC inspectors reviewed the 10 CFR 50.59 safety evaluation of the IRSF during the September 1985, inspection (Inspection Reports No. 50-373/85030; No. 50-374/85031).

Further NRC review during this inspection (Inspection Reports No.'50-373/87020; No. 50-374/87020) show that, as required by 10 CFR 50.59 and 50.71, the facility description and other appropriate sections in the FSAR have been updated to include the IRS (Closed).0 pen Item (373/86023-01; 374/86022-01): Corrective actions to assure compliance with State of Washington transportation regulations on condition of DAW drums. The training procedures of stationmen who perform DAW compaction have been changed to emphasize inspection of the

_ _ _ - _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ - _ - - _ _ _ _ - _ _ - - _ _ _ _ _ _ _ _ -

- - - _ _ - ______ _ _ - _

h

.

.

DAW drums before and after compaction. In addition, an Operations ,

radwaste manager and a QC inspector inspect the DAW drums for damage on

'

the loading dock and on the truck after loading per Procedure i LTP-800-1 '

(Closed) Violation (50-373/86038-02; 50-374/86038-02): Failure to make a survey to ensure compliance with 10 CFR 20.103 limits. The licensee has strengthened the radiological air sampling program, issued a policy j requiring surveys of plastic face-shields before reuse, and instructed health physics foremen to ensure adequate pre-job and routine surveys are performe . . Organization, Management Controls, and Staffing The inspectors reviewed the licensee's organization and management controls for the radiation protection and chemistry programs, including '

changes in the organizational structure and staffing, effectiveness of procedures'and other management techniques used to implement these ;

programs, and experience concerning self-identification and correction of i program implementation weaknesse Since the previous inspection (Inspection Reports No. 50-373/87005; No. 50-374/87005) the only significant change in the Rad / Chem Department i involves replacement of the recently terminated ALARA coordinator by the current contamination control coordinator, who has assumed the duties of both coordinator position .

l The Chemistry Department currently is staffed with a lead chemist who )

has six years experience and meets the ANSI N18.1-1971, " Radiochemistry" requirements, four chemists with experience ranging from two to six years, {

j a laboratory foreman, and an engineering assistant with four years i experience. The chemistry staff appears stable and thi experience level is adequat No violations or deviations were identifie . Gaseous Radioactive Waste The inspectors reviewed the licensee's gaseous radwaste management program. The effluent records from July 1986 through June 1987 were selectively reviewed. The 1986 semiannual effluent reports were selectively reviewe The gaseous effluent release paths, control mechanisms and procedures, effluent quantification methods, dose projections, and dose calculations are unchanged from the description provided in Inspection Reports No. 50-373/85018; No. 50-374/85019. In these reports it was noted that containment vents and purges are normally exhausted via the regular vent stack; however, if filtration is needed the exhaust may be diverted through SGTS. During this inspection it was noted that although the exhaust can be diverted through the SGTS, under normal conditions the filtration is provided by the vent and purge filtration system, not the SGT The SGTS would be used only under emergency condition _ _ - _ _ - _ _ _ .

.

.

l Radioactive effluent release rates and offsite dose calculations were reviewed. In 1985, about 180 curies of noble gas and about 8.4E-03 curies of iodine-131 were released in gaseous effluents (both units). In 1986, about 3,000 curies of noble gas and about 6.5E-02 curies of iodine-131 i were released in gaseous effluents. The increase in 1986 was caused by a j minor fuel cladding problem in Unit 2 as noted in Inspection Reports

'

No. 50-373/86026; No. 50-374/86027. Through June 1987, 32 curies of noble gas have been released. Using the Offsite Dose Calculations Manual (ODCM) methodology the calculated maximum whole body and thyroid doses to l an individual beyond the site boundary from these releases in 1986 were ,

4.8E-03 mrem and 2.7E-02 mrem, respectivel l No violations or deviations were identifie . Liquids and Liquid Radioactive Wastes

The inspectors reviewed the licensee's reactor liquids and liquid i radwaste management programs, including: determination whether reactor liquids meet chemical and radiochemical requirements; determination whether liquid radioactive waste effluents were in accordance with regulatory requirements, adequacy of required records, reports, and notifications; and experience concerning identification and correction of programmatic weaknesse l In 1986, about 1.8E-02 curies of gross beta gamma activity (excluding ,

tritium) and about 1.4E-01 curies of tritium were released in liquid '

effluents from both units combined. Using ODCM methodology, the calculated maximum whole body and GI-LLI doses to an individual beyond

,

!

the site boundary from these releases were 2.7E-06 mrem and 2.0E-05 mrem i respectivel l l

The only liquid batch releases made in 1986 were in the third quarter; j five liquid batches have been made through June 1987. Selected radwaste discharge records were reviewed; no problems or deviations from the Technical Specification 3/4 " Radioactive Effluent" requirements were identifie No violations or deviations were identifie . Coolant Chemistry and Radiochemistry l Selected reactor sampling and analysis records were reviewed for compliance with chemical and radiochemical criteria contained in '

Technical Specifications 3/4.4.1 and 3/4.4.5, respectively. Test results reviewed for the period July 1986 to June 1987 include pH, conductivity, chloride, and dose equivalent iodine-131, no problems were note No violations or deviations were identifie .

. Calibration and Functional Tests of Gaseous and Liquid Process and Effluent Monitors The inspectors reviewed records .for monitors on the radioactive liquid systems (including the common radwaste discharge, the service water .J discharge for each reactor unit, and two RHR service water monitors'for each unit) and monitors on the radioactive gaseous systems (the SGTS,

.the vent stack, and the main steam lines).

The inspectors reviewed calibration records for these monitors, selectively reviewed records of monthly source checks, and made a-visual inspection of the-licensee's liquid,. gaseous, particulate, and iodine sampling stations and monitor locations. A visit was made to the control room to observe the monitor readout systems and to review set points. No significant problems were note The initial calibration of liquid monitors . involved determination' of an efficiency using a single concentration liquid cesium-137 source and

'

response to several' solid cesium-137 sources to determine linearity of l response and for use in subsequent calibration checks. For subsequent calibrations a solid cesium-137 source (decay corrected) is used to verify that response to a solid source has remained within:an established error band. This verification is performed using Procedure LCP-820-1 Sections 26.c and'd of this procedure require.that the cesium-137 source response be within plus or minus 20 percent of the decay corrected initial calibration response; responses outside this acceptance range require additional actions'specified in the procedur During a review of the "B" RHR Liquid Process Monitor calibration, it was found that the decay correction for the cesium-137 source was incorrectly computed (the "A" RHR monitor response readings were used to make the decay correction instead of the "B" RHR monitor readings).

Because of this error, the "B" RHR monitor response was. incorrectly determined to vary 13 percent, well within the accepted limits of the error band. However, had the source decay been properly computed and the correct decay value used, the "B" RHR monitor response would have l been outside the acceptance range (24 percent) and additional procedure I

steps would have been required.. Failure to properly verify that the cesium-137 solid source response was within the acceptance range of the decay corrected initial calibration response is a violation of Procedure LCP-820-13 requirements (50-373/87020-01; 50-373/87020-01).

A review of this matter indicated that the calibration was performed by an individual who previously had used the procedure to perform calibrations but who previously had not been involved in' calibration of these monitors or received On-Job-Training (0JT) for calibrations of the monitors. This matter was discussed at the exit meetin !

One violation was identifie _ _ _ _ . _- _- --__ ___

.

.

I

9.- ' Procedures for Controlling Releases -

'd The inspectors selectively reviewed revisions to the licensee's. liquid and-gaseous radwaste procedures. The procedures are listed below. No significant problems were identifie .. 4 LCP 140-7, Revision 12, Analysis.of Radwaste Discharge Tanks 1(2)WF05T .i and Determination. of Discharge Flowrate Using Cooling Pond Blowdow .I I

LCP-730-1, Revision- 1, Determination' of Station Vent Stack and Standby j Gas Treatment Stack Wide Range Gas Monitor -!

Effluent Release Rate Alarm Setpoint LCP-820-13, Revision 6, Efficiency Calibration of NaI Liquid Process Radiation Monitor LCP-820-23, Revision 1, Determination of Efficiency Curve for Radwast Liquid Process Monito I No violations were identifie . Solid Radioactive Waste

The inspectors reviewed the licensee's solid radioactive waste management '

program, including: determination whether changes to equipment and j procedures were in accordance with .10 CFR 50.59; adequacy of implementin '

procedures to properly classify and characterize waste, prepare manifests, and mark packages; overall per.formance of the process control and quality assurance programs; adequacy of required records, reports, and notifications; and experience concerning identificat ion and correction of programmatic weaknesses. Tours of. packaging and storage areas.and the radwaste solidification system control room were made. Audits'are discussed in Section 1 Solid radioactive waste consists of spent resins, filter sludges, evaporator bottoms, and dry active wastes (OAW). If compactible, DAW is placed into 55 gallon steel drums and compacted; if not compactible, DAW l

'

is packaged in 96-cubic foot steel bin The DAW compactor has self-contained filtered ventilatio Evaporator bottoms (wet waste) are solidified using a Stock Equipment cement solidification system. This system is described in Inspection Reports No. 50-373/86023; No. 50-374/86022. All bead resins.and some powdered resins are being dewatered by a vendor and shipped in~ High Integrity Container (HICs) and metal liners to off-site burial site The vendor dewaters the contents of 1A/1B/2A/2B Phase Separator Tanks, the URC Sludge Tank, and the Spent Resin and Sludge Tanks. The stock system has been modified to allow interfacing with the vendor's equipmen The contractor uses their own equipment, personnel and procedures to perform the dewatering. The dewatering is performed in accordance with

- _ _ - _ _ _ _ _ - - _ - - - _ _ _ . - - . _

- - _

.

.

a NRC-approved Topical Report and Process Control Program. All vendor:

equipmentLis located in the Radwaste Truck Bay. The dewatering takes place under both operating and health physics supervision; air samples are take During a previous inspection (Inspection Reports No. 50-373/86023; No. 50-374/86024) concerns with the incomplete solidification of drums processed by the stock solidification system were noted involving free-standing water in the drums. The licensee determined that the

primary cause for this problem was the result of operational problems with

'

the decant tank' level indicators. To resolve the problem, the licensee no. longer processes powdered resins from the decant tanks by means of the Stock system; the resins are currently vendor processe No violations or deviations were identifie . Transportation of Radioactive Materials The inspectors reviewed the licensee's radioactive materials transportation program, including: determination whether written implementing procedures are adequate, maintained c Jerent, properly approved, and acceptably-implemented; determination whether shipments are in compliance with NRC and DOT regulations.and the licensee's quality assurance program; determination if there were any transportation _ i incidents involving licensee shipments; adequacy of required records, 1 reports, shipment documentation, and notifications; and experience concerning identification and correction of programmatic weaknesse Audits and surveillance are discussed in Section 1 Shipments of low specific activity (LSA) waste to licensed burial sites are the major transportation activity. Contaminated solid trash (paper, plastic, wood, metal, clothing, etc.) is either packaged in 55 gallon steel drums and compacted or packaged in large metal boxes if the materials are not compactible. New DUT Specification 17-H drums are used which meet the DOT 7-A performance specification. Liquid' wastes consisting of resins, filter sludge, and evaporator bottoms are solidified'using vendor and Stock solidification systems. Selective records of radioactive shipments made during 1986 and 1987 to date were reviewed for compliance with 49 CFR 170-189 and 10 CFR 71; no problems were note No violations or deviations were identifie . Training and Qualification Effectiveness Liquid, solid, and gaseous radwaste processing, releases, and shipments are handled by operations and rad / chem personne Laboratory work and analyses associated with effluent releases, monitor calibrations, and processing of waste for solidwaste shipments are performed by chemists, rad protection foremen, health physicists, and rad / chem techs (RCTs). A formal training program is in place for only the RCTs, who perform most of the day-to-day lab work and analyses; no formal training program exists

L-----.----- __ _ - -- - - - . - _ _ - .- - - - _ - - - - - - - - - _ - - _ - - - - - - - - - - - - - - - - - - - _ 9

- - - - __ __ - - - - - _ _ _ . _ _ _

,

.

.

for chemists or health physicists. The'RCTs receive approximately )

14 weeks of generic training at the Production Training Center, two or J

,

three weeks of plant specific training,.and approximately 10 weeks of l

practical factor qualification. Written and practical exams are given .

during the training. The RCT candidate must be interviewed and approve by the rad / chem. supervisor prior to certificatio Stationmen, who compact dry active waste (DAW) in preparation for shipment, receive training.in radwaste and other station operations during a 4-day classroom training program and during practical factor qualification. Continued training is provided as.neede Non-licensed operators concentrate and solidify radwaste in 55-gallon drums, assist RCTs and vendor representatives during solidification of resins in liners, and perform other. duties in the Radwaste control roo '

Training in radwaste operations is received during approximately 14 weeks of classroom instruction and practical factor qualification. Written and practical exams are given during the training. Certified-operators receive approximately 4 weeks each year of continued trainin QC, operations managers, and rad / chem managers periodically review the activities of the stationmen, RCTs, and operators involved with radwaste processing. Trcining and qualification effectiveness and management-oversight appear adequate in the radwaste progra No violations or deviations were identifie . Internal Exposure Control 1 The inspectors reviewed the licensee's internal exposure control.and assessment programs, including: changes to procedures affecting internal exposure control and personal exposure assessment; determination whether engineering controls, respiratory equipment, and assessment of individual I intakes meet regulatory requirements; planning and preparation,for j maintenance and refueling tasks including ALARA considerations; and  !

required records, reports, and notification !

The licensee uses a vendor furnished whole body counter (WBC) for i baseline counting of incoming contractor personnel and for counting I station personnel at prescribed frequencies. Contractors are also counted upon completion of their outage activities at the statio The inspectors selectively reviewed WBC results for the first two quarters of 1987. No exposures greater than the 40 MPC-hour control measure were indicate The yearly calibration of the WBC was performed by the vendor in March 1987 using vendor calibration procedures. The calibration is performed using standard sources and test phantoms to. simulate counting l geometries. A review of the calibration results indicated the new counting efficiency remained essentially the same; no problems were note ;

!

l l'

L - - ---_ --- - . - - - - - - - - _ - - - - - - - - - _ - - - - - _ - - - - _ -_- _ 1

. - _ - _ - . _ _ - - _

.

14. Audits

.

Formal Quality Assurance Audits of radwaste and transportation activities l were performed by offsite personnel on March 24 and April 28, 198 These audits reviewed the activities and documentation associated with radioactive waste, radwaste personnel training, dewatering and solidification processes, radwaste shipment activities, and verification

,

'

of radioactive contents of batch radioactive liquid waste prior to ,

releas No significant problems were found; the audits indicated that j the station was effectively implementing the established requirements associated with radwaste and transportation activitie l In addition to surveillance checks of each radwaste shipment, monthly and special surveillance are also made of radwaste activitie One such surveillance identified problems concerning housekeeping and poor radiation protection practices involving improper frisking and use of l

protective clothing, problems that had been previously identified by NRC inspectors. These identified problems were discussed with the station manager; corrective actions were taken. The effectiveness of the corrective actions are required to be reviewed during subsequent audits / surveillanc The extent of audits, surveillance, and adequacy of corrective actions appear good, i

l No violations or deviations were identifie ,

15. Unit 1 Reactor Recirculation Pump Repair (RPR) '

The inspectors reviewed the planning and implementation of ALARA and l radiological and contamination control procedures / practices used during l

'

the removal and decontamination of a defective reactor recirculation pump. On July 13, 1987, the pump impeller and associated parts were removed from the pump bowl in the Unit I drywell and placed in the Unit 2 dryer / separator pit on the refuel floo Decontamination and repair or replacement of pump parts will be performed on the refuel floor. In 1 addition to the RPR, other Unit I drywell maintenance work involves a IA 1 reactor recirculation loop isolation valve (the 67 valve) and the inboard isolation valve and motor on the RHR shutdown cooling line (the 9 valve).

Planning A health physicist, mechanical maintenance staff, and other licensee representatives were sent to WNP-2 in Richland, WA, on two occasions in 1987 to gather information concerning lessons-learned from a similar RPR performed at WNP-2 in 198 During one of the lessons-learned visits, the licensee observed the removal, relocation, and decontamination of two defective reactor recirculation pumps. The WNP-2 staff provided the licensee with a video tape of the operation, records of exposure and contamination

_ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ - _____ _ ___________ _ ___

J

,

.

.- 1 surveys, ALARA considerations, and suggestions on methods of decontamination. At.LaSalle, this information was discussed at

.

periodic meetings of health ~ physics, construction and maintenance, and vendor / contractor staffs in preparation for the RP The licensee developed and used radiological control procedures an y ALARA measures.for the RP TLD and betagamma dose rate' readings- )

'

were performed on the recirculation pump. -The maximum' gamma dose rate was about 3 R/hr, and the maximum beta dose rate was about 40 rads /hr. During the pump removal, transfer,"and decon, rubber protective clothing and full-face' respirators were used, air. samples were taken, and continuous-HP coverage was provided. ALARA measure included the use of lead blankets'and plexiglases sheets as temporary shielding, beta eye shields, and alarming dosimeters' set a fractions of the allowable daily dose limi i Access Control -)

-l Workers dressed out, received dosimetry, and signed the appropriate- !

Radiation Work' Permit (RWP) in Unit 2 before ente' ring Unit Frisking booths were located in Unit 2. During the relocation of the impeller from elevation 740' to the refuel floor. (elevation 843'),

personnel were posted on intervening floors around the perimeter of the equipment floor hatches, and on the refuel floor-to prevent personnel exposure during transfer of the impeller. The impeller was transferred in a specially designed shielded containe ~ Radiation Work Permit l

'

A Radiation Work Permit (RWP) was written for the pump pull and relocation. Separate RWPs were written 'for the removal of-interferences from the drywell prior to. the RPR, the impeller -,

decontamination, and repairs on the 67 and 9 valves. Persons l entering the Unit I drywell area and the decontamination and repair '

areas on the refuel floor are assigned a dose-limit and given t

dosimetry after reading and signing the appropriate RWPs.

I Dosimetry l

Whole body TLD badges, 0-200 mR (or higher) gamma dosimeters, and )

alarming electronic dosimeters are assigned to all persons entering the drywell are In addition, 2 TLD ring badges were assigned to each person involved in the bagging of the impeller prior to relocation, and in the decon and refurbishing of the impelle Approximately, one person rem was received through July 15, 1987; l 6 person rems had been projected. The low person-rem total is attributable to inclusion of WNP-2 lessons-learned in the RPR l pre planning and to lower than anticipated dose rates during the RPR'

l and deco ,

!

l'

!

- ___ _

_ _ _ _ _ _ - _ _ - _ - _ - - -

.

,

.

!

Tour of Unit 1 Drywell and Refuel Floor An inspection. tour was conducted. in. the. Unit L1 drywell, Unit 2 j staging area, and the refuel floor to, examine RPR and decon 1 activitie The inspectors reviewed worker use of protective ~'

clothing, worker adherence to radiological control procedures, good health physics practices, use'of temporary' shielding, radiation protection job coverage, and the use of survey and monitoring instrument Based on the inspectors review of the RPR, it appears that the licensee's pre planning, ALARA considerations, and radiological controls were goo No violations were identifie . Control of Radioactive Materials and Contamination, Surveys, and Monitoring The. inspectors reviewed the licenseefs program for control of radioactive mattrials and contamination, including: adequacy of supply, maintenance, and calibration of contamination survey and monitoring equipment; effectiveness of survey methods, practices, equipment, and procedures; adequacy of review and dissemination of survey data; and effectiveness of-I methods of control of radioactive and contaminated material .l The licensee is currently operating several beta-sensitive Whole Body Friskers (WBFs) to identify personal skin and clothing contamination events. The WBF's appear to be the most sensitive, routinely-used monitoring system in the station for the identification of low-level personal contamination and are located-in'those areas of the station through which the majority of workers pass before egress to the gate house. Hand-held beta-sensitive friskers (HP 210's) and gamma-sensitive protal monitors (IRT's) are also used to identify personal contamination events; however, these devices are not as effective in use for detection:

of low level contamination than the WBFs. The HP 210s are located throughout the station and are primarily used to identify personal contamination problems at step-off pads; several are used for a final frisk before egress to the gatehouse. Two IRTs are located in the gatehouse and are used as the final plant egress survey, others are located in the service and turbine building During this inspection it was observed that while there are many  :

locations final personal surveys can be performed before egress to '

the gatehouse, under normal conditions only.cne location has a radiological control station (the "15 line" in the turbine building)

with an RCT normally in attendance. The other egress locations generally are equipped with HP-210s or IRTs and do not have an RCT in attendance. These egress locations do not provide as good a. quality personal contamination control as the "15 line." It appears that a disproportionately large number of these other egress locations exist at the La Salle Plant compared to other Region nuclear plants. This 'i practice could lead to degradation of the licensee's contamination i control program if significant numbers of personnel are allowed to exit the controlled area via these other egress locations. This matter was discussed at the exit meeting and will be reviewed further during future inspections (50-331/87020-02; 50-374/87020-02).

. _,

.

.

17. Surveillance; Independent Surveys; Plant Tours Based on several tours of the plant, the inspectors noted: (1) No persons were observed violating procedural or regulatory requirements; this includes observations of workers performing activities under the requirements of several different RWPs. (2) Independent radiation .

surveys performed by the inspectors indicated radiation areas were posted l as required. The results of fifty-four floor and horizontal smears indicated that no areas in excess of the licensee's limits for controlling contaminated areas were found. (3) Two of the three l Constant Air Monitors (CAMS) located on the refueling floor were inoperable; this condition was reported to the licensee; work orders for l repair were issued and the CAMS were operable before the inspection was completed. (4) At almost all frisker stations, instructions are posted that require persons who have detected personnel contamination to go to i the nearest telephone and notify the Rad / Chem departmen However, l l

protective shoe covers and gloves to minimize the potential spread of I contamination are not provided at most of the frisker stations. This issue had been discussed with the licensee during a previous inspection (Inspection Reports No. 50-373/85014; No. 50-374/85014) and was i discussed at the exit meeting for this inspectio It will be reviewed l at a future inspection, (50-373/87020-03; 50-374/87020-03). (5) The I licensee has placed steel grating mats at the entry to each of the WBFs '

with the intent that workers scrap their feet on the mats to prevent the buildup of dict on the foot pedals of the WBF. At one WBF, a worker was )

observed to alarm the foot detectors, step off, scrap his shoes on the l mat, and step back onto the WBF for a resurvey; the foot alarm did not i sound on the resurvey and the worker proceeded to cross the step-off pad. The use of the mats to scrap shoe contamination before WBF surveys hinders identification of possible plant contamination problems and causes the number of contamination events to be underreported. This l matter was discussed at the exit meeting and will be reviewed further l during future inspections (50-373/87020-02; 50-374/87020-02).

'

l 18. Exit Meeting The inspectors met with licensee representatives (denoted in Section 1)

at the conclusion of the inspection on July 15, 1987. Further discussions were held by telephone on July 22, 1987. The inspectors summarized the scope and findings of the inspection. The inspectors also discussed the I

likely information content of the inspection reports with regard to document: cr processes reviewed by the inspectors during the inspectio The licensee did not identify any such documents / processes as proprietar In response to certcin items discussed by the inspectors,-

the licensee: Acknowledged the violation (Section 8) Stated that a review would be made to evaluate the feasibility of reducing the number of egress locations before entry into the gatehouse, and increasing the number of radiological control egress l 13 l

l

. - - _ _ _ _ _

,

a-survey location In addition, when more WBFs are purchased, consideration will be given to' placing them at egress locations that currently have HP 210s (Section 16).

c. Stated that plans are being developed to provide shoe covers at selected frisker stations throughout the plant (Section 17).

l l

!

l

1

!

,

k

1l i

-

I l

,

i L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _