ML20207L031

Insp Repts 50-413/86-20 & 50-414/86-23 on 860513-16 & 0602. Violation Noted:Failure to Maintain Radiation Exposure Records Per Form NRC-5 Instructions

ML20207L031
Person / Time
Site:	Catawba
Issue date:	06/24/1986
From:	Hosey C, Revsin B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20207L025	List: ML20207L022 ML20207L027 ML20207L031
References
50-413-86-20, 50-414-86-23, NUDOCS 8607300173
Download: ML20207L031 (9)

See also: IR 05000413/1986020

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UNITED STATES

/ja CE7pf"o NUCLEAR REGULATORY COMMISSION

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101 MARIETTA STREET.N.W.

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Report Nos.: 50-413/86-20 and 50-414/86-23

Licensee: Duk'e Power Company

422 South Church Street

Charlotte, NC 28242

Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-48

Facility Name: Catawba 1 and 2

Inspection Conducted: Ma 13-16 and June 2, 1986

Inspector: / bN k

Dats Signed

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B. K. Revsifi V

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Accompanying Personnel: C. H. Bassett

Approved by: )GW

C. M. Hosey7 Section Chief

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Date' Signed

Division of Radiatiqn Safety and Safeguards

SUMMARY

Scope: This routine, unannounced inspection involved onsite inspection in the

area of radiation protection including: startup shielding surveys; internal

exposure control and assessment; external exposure control and personal

dosimetry; radioactive materials control, posting and labeling; solid radioactive

waste classification and characterization; transportation of radioactive

materials; and program for maintaining exposure as low as reasonably achievable

(ALARA).

Results: One violation - Failure to maintain radiation exposure records in

accordance with instructions contained on Form NRC-5.

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8607300173 860715

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• J. W. Cox, Technical Services Superintendent

• W. P. Deal, Station Health Physicist

• C. L. Hartzell, Compliance Engineer

• P. G. LeRoy, Licensing Engineer

1. • R. E. Sorber, Health Physicist, General Office

L. D. Schlise, Health Physics Supervisor

H. F. McInvale, ALARA Supervisor

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G. A. VanderVelde, Health Physics Supervisor

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M. C. Couch, Administrative Health Physics Supervisor

G. T. Mode, Health Physics Coordinator

1. T. Keane, Health Physics, General Office

1. W. M. Carter, Radiological Health Supervisor, Applied Science Center

! #P. S. Wingo, System Environmentalist, Applied Science Center

NRC Resident Inspectors

• P. K. VanDoorn, Senior Resident Inspector

*M. Lesser, Resident Inspector

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P. Skinner, Resident Inspector

• Attended first exit interview.

# Attended second exit interview.

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2. Exit Interview

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The inspection scope and findings of the May 13-16, 1986 inspection were

summarized on May 16, 1986, in the first exit interview with those persons

indicated in Paragraph 1 above. An unresolved item * (Paragraph 11)

!' concerning the density thickness at which the licensee was assessing worker

whole body exposure when the lenses of the eyes were not shielded by at

least 700 milligrams per square centimeter (mg/cm2) density thickness was

discussed in detail. The licensee acknowledged the inspection findings and

took no exceptions. The licensee did not identify as proprietary any of the

materials provided to or reviewed by the inspector during this inspection.

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l At a second exit interview conducted on June 2, 1986, and attended by those

l persons indicated in Paragraph 1 above, licensee management was informed

that the unresolved item would Se considered closed. The licensee was also

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informed that failure to maintain radiation exposure records in accordance

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l *An Unresolved Item is a matter about which more information is required to

determine whether it is acceptable or may involve a violation or deviation.

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with instructions contained on Form NRC-5 would be considered an apparent

violation of 10 CFR 20.401 (Paragraph 11). Licensee management took

exception to this finding based on their opinion that measurement of whole

body radiation exposure through a tissue equivalent density thickness of

1000 mg/cm 2 , rather than the required 300 mg/cm2, would not result in a

significantly different whole body dose.

3. Followup on Inspector Identified Problems (92701)

(Closed) Inspector Followup Item (50-413/84-41-01) Snubber Inspection in

High Radiation areas. The licensee has snubbers located in High Radiation

Areas which require periodic inspection as required by Technical

Specification 3.4.7. The licensee applied for an exemption from this

inspection requirement for certain snubbers but were denied on the basis of

ALARA practices being insufficient justification for deleting inspections.

Consequently, the licensee reanalyzed 11 snubbers as rigid supports and

issued Nuclear Station Modification Request No.10736 to implement the

modification. Completion is expected by February 1,1987.

4. Control of Radioactive Materials and Contamination, Surveys and Monitoring

(83526)

The licensee was required by 10 CFR 20.201(b) and 20.401 to perform surveys

and to maintain records of such surveys necessary to show compliance with

regulatory limits. Survey methods and instrumentation were outlined in the

Final Safety Analysis Report (FSAR), Chapter 12, while Technical

Specification (TS) 6.11 provided the requirement for adherence to written

procedures. Radiological control procedures further delineated survey

methods and frequencies.

During plant tours, the inspector examined radiation level and contamination

survey results outside selected rooms and cubicles. The inspector performed

independent radiation level surveys of selected areas and compared them with

licensee survey results. The inspector noted that locked high radiation

areas outside containment were maintained as required by TS 6.12. The

inspector noted that there were numerous contamination recepticles in use

about the Auxiliary Building and was informed by the licensee that they had

implemented a very vigorous program to minimize contaminated areas in the

plant, and consequently, whenever possible leakage was identified, it was

captured or run into a drain.

During tours of the plant, the inspector observed the use of survey

instruments by plant staff and compared plant survey instrument results with

readings taken by the inspector using NRC equipment. The inspector examined

calibration stickers on radiation protection instruments in use throughout

the facility.

No violations or deviations were identified.

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5. Transportation (86721)

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10 CFR 71.5 requires that licensees who transport licensed material outside

the confines of their plants or other places of use, or who deliver licensed

material to a carrier for transport, comply with the applicable requirements

of the regulations appropriate to the mode of transport of the Department of

Transportation (DOT) in 49 CFR Parts 170 through 189.

The licensee stated that at the time of the inspection, only two shipments

of radioactive waste had been made. They were Shipment No. CNS 85-08 made

on October 17, 1985 and Shipment No. CNS 86-10, made on February 27, 1986.

Both shipments contained 14 B-25 boxes of compacted dry active waste (DAW).

The inspector verified that the requirements for 49 CFR Parts 170 through

189 had been met for these shipments.

No violations or deviations were ' identified.

6. Solid Radioactive Waste (84722)

10 CFR 20.311 requires that the licensee maintain a tracking system for

radioactive waste shipments to verify that shipments had been received

without undue delay by the intended recipient. The inspector reviewed the

tracking metnodology used by the licensee and examined the documented

receipt acknowledgements for Shipment Nos. CNS 85-08 and CNS 86-10.

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10 CFR 20,311 requires a licensee who transfers radioactive waste to a land

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disposal facility to prepare all waste so that the waste is classified

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according to 10 CFR 61.55.

l 10 CFR 61.55(a)(8) states that the concentration of a radionuclide may be

l determined by indirect methods such as the use of scaling factors which

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relate the inferred concentration of one radionuclide to another that is

i measured if there is reasonable assurance that the indirect methods can be

correlated with actual measurements.

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The licensee had been using scaling factors developed for waste streams at

i its sister station, McGuire, since Catawba had not been operational long

l enough to develop plant specific scaling factors for its waste streams.

i licensee representatives stated that plant specific waste streams had been

! sampled in February 1986, and that vendor analyses of the samples had been

l completed. It was anticipated that the plant specific scaling factors would

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be available for use in January 1987. The inspector questioned the licensee

! about the lag time between sample analyses and implementation of new scaling

factors. The licensee stated that scaling factors were computed by the

General Office (GO) and since they also generated the scaling factors for

the other two plant sites, the GO preferred to put them all on a similar

schedule. The inspector stated that at the time of the inspection this area

was not an issue for the licensee since all shipments thus far and those

anticipated for the near future were DAW. However, with the upcoming

refueling outage scheduled for late summer, 1986, this situation could

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readily change. The inspector reviewed Procedure HP/0/8/1006/05, Periodic

Sampling of Radioactive Waste for 10 CFR 61 Scaling Factor Determination.

The inspector discussed with licensee representatives the program for waste

characterization. The licensee did not anticipate that waste solidification

would be performed frequently, and at the time of the inspection, some resins

were anticipated to be possible candidates for solidification. All

waste processing equipment would be vendor operated. In general, when

waste stability was required, the licensee planned to use approved

containers.

No violations or deviations were identified. ~

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7. As low As is Reasonably Achievable Program (ALARA) (83728)

10 CFR 20.1(c) specifies that licensees should implement programs to keep

worker's doses ALARA. FSAR, Chapter 12, also contains licensee commitments

regarding ALARA actions. >.

The licensee had dedicated six . full time individuals for their program for

maintaining doses ALARA. All work with estimated exposure greater than one

man-rem required ALARA preplanning. Through the preplanning review,

shielding recommendations were made, job surveys were delineated and overall

evaluation of the work to be performed was made from a radiological point of

view. Post-job ALARA reviews were required for jobs whose dose estimates

exceeded 50 man-rem and for those jobs where dose estimates were exceeded by

25 percent.

The inspector reviewed the minutes of -the ALARA committee meetings for

October 16, 1985, and January 2, 1986. The licensee stated that the

committee met on a quarterly. basis and was composed of representatives from

each plant work group. These members, while not from the ranks of upper

management, had the authority to make commitments for their group.

Plant management commitment to ALARA appeared to be strong. ALARA reports

are sent to plant management monthly and include such data as man-rem by

work group, station dose distribution by work group and a graphic

representation of estimated versus actual exposure received for the year.

Additionally, station section managers have had ALARA goals incorporated

into their performance appraisal goals.

The ALARA group had begun preparation for the refueling outage planned for

August 1986. The inspector reviewed work packages, WR No. 6335 SWR, Reactor

Coolant Pump Seal Work, Unit 1, and WR No. 5311 SWR, Unit 1, Conoseal Work,

that were under preparation.

Scheduling and planning for the outage were ongoing and as more information

of work to be performed became available, the ALARA group would continue to

develop work packages.

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The ALARA Section had also developed a library of slides and video tapes of

i various rooms showing valve locations and various job evolutions to orient ,

and familiarize workers with the jobs to be performed. While not required

e - by procedure, these visual aids were often presented at job preplanning

meetings.

No violations or deviations were identified.

8. Radiation Protection-Startup (83521)

The inspector reviewed Procedure TP/2/B/2200/01, Biological Shield Survey.

This procedure required facility surveys for gamma and neutr'on radiation at

the following power levels: 0%, 30%, 50%, 75% and 100%. Background surveys

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prior to fuel loading were also required. The licensee had completed the

i required surveys for background and 0% power levels. All values were found

i to be below the Lower Limit of Detection for the instrument used for

measurement except for the area of the Auxiliary Building on the 543 foot

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level in Mechanical Penetration Room No. 227. Gamma radiation levels were

1 found to be 0.1 mR/ hour due to piping from the floor drain tank from Unit 1.

The licensee stated that these surveys would be completed as power ascension

continued.

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l No violations or deviations were identified.

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l 9. Internal Exposure Control and Assessment (83725)

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The licensee is required by 10 CFR 20.103, 20.201(b), 20.401, 20.403 and

20.405 to control uptakes of radioactive material, as ess such uptakes, and

l keep records of and make reports of such uptakes. FSAR, Chapter 12, also

includes commitments regarding internal exposure control and assessment.

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Due to the limited operating history of the facility and the active

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containment program in operation, the licensee has had few occasions for

use of respiratory protective equipment. The inspector reviewed the daily

exposure computer printout (REC) and found the highest Maximum Permissible

. Concentration (MPC) hours assigned to an individual was 1.64. The inspector

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reviewed the methodology the licensee had in place for MPC hour tracking and

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determined that until an individual accumulated 0.1 MPC hours, the computer

I program is set up so that the hours do not appear on the REC. However, a

printout of MPC hours by individual Health Physics (HP) number showed that

all these data were present in the individual's dose file. The inspector

l reviewed procedure HP/0/B/1000/04, Preparation of Health Physics Radiation

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Work Permits and Standing Radiation Work Permits, May 5, 1986 which provided

! the guidelines for issuance of respiratory protective equipment.

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No violations or deviations were identified.

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10. Staffing (83722)

HP staffing levels were discussed with the Station Health Physicist (SHP).

At the time of the inspection,104 staff positions were authorized,103 of

which were filled. Ir4 addition, 19 HP contractor technicians were onsite to

augment the experience level of the staff. At present, 27 HP technicians

were ANSI qualified. ' Shift coverage is provided by three to five technicians

along with one ANSI qualified technician. The primary responsibility of

the HP shift crew was liquid and gaseous releases.

Duke Power was in the process of reducing the size of their construction

group and as a consequence, 18 of the 104 staff persons at Catawba had come

from the construction crews. It was anticipated that 10 more would be added

to the HP staff in the future. At the time of the inspection, the training

for these individuals consisted of Independent Radiation Worker (IRW)

training. IRW training consisted of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classroom instruction

followed by three weeks of on-the-job training during which time trainees

were task qualified. It was anticipated that as slots opened at the

Training Center, these individuals would rotate through the regular HP

, technician training program. The SHP stated that for the upcoming outage

for Unit 1, these individuals would be assigned to decontamination work.

2 Also, 60 contract HP technicians will be hired for the outage.

No violations or deviations were identified.

11. External Occupation'al Exposure Control and Personal Dosimetry (83724)

a. 10 CFR 20.202 requires that appropriate personnel monitoring devices be

worn by personnel likely to receive exposures in excess of 25 percent

of the limits specified by 10 CFR 20.201 or who enter a high radiation

area, and to require the use of such devices. During tours of the

facility the inspector observed personnel wearing monitoring devices as

, required.

b. 10 CFR 20.401 requires that each licensee maintain records showing the

radiation exposures of all individuals for whom personnel monitoring is

required and that such records be kept on Form NRC-5 in accordance with

the instructions contained in that form.

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Form NRC-5, Item 5 requires that unless the lenses of the eyes are

protected with eye shields having a tissue equivalent density thickness

of at least 700 mg/cm 2, dose recorded as whole body dose should include

the dose delivered through a tissue equivalent absorber having a

density thickness of 300 mg/cm2 ,

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Licensee representatives stated that thermoluminescent dosimeters

(TLDs) were provided to the plant by the dosimetry laboratory of the

Applied Science Center of the Training and Technology Center located

near Huntsville, NC. All dosimetry processing was performed at the

center which maintained its own staff with additional technical

assistance provided by the G0 health physics group. The licensee

informed the inspector that Teledyne Isotopes TLDs and processing

equipment were in use by the Center and that these TLDs utilized

calcium sulfate as their phosphor.

The inspector visited the dosimetry laboratory and was informed that

the TLD card utilized in the badges had eight possible readout areas,

four main read out areas and four backup readout areas. Filtration of

the TLD card was provided by the badge case such that the effective

density thicknesses at which radiation levels were measured were:

Area 1 - 7 mg/cm2

Area 2 - approximately 1000 mg/cm2

Area 3 - approximately 1600 mg/cm2

Area 4 - approximately 1600 mg/cm 2

The licensee stated that the ratio of Area 2 to Area 4 was used to

provide information concerning the hardness of the gamma radiation

field and that this ratio was utilized through various algorithms to

treat the data from the four areas. Skin dose was assigned from Area 1

after manipulation by the algorithm and whole body or deep dose was

assigned - from Area 4 which, after treatment with the algorithm,

provided a whole body dose equivalent to that delivered through a

density thickness of 1000 mg/cm2 ,

The inspector questioned the licensee about density thickness of eye

protection worn by workers at the plant. The licensee stated that

safety glasses were required to be worn at all times while in the plant

which included work in containment. The licensee had determined that

both safety glasses and goggles had a density thickness of

approximately 200 mg/cm2 and that respirator face plates had a density

thickness of 260-270 mg/cm2 The inspector asked the licensee how

whole body dose through a tissue equivalent absorber of 300 mg/cm2 was

assessed since the eye protection provided the workers did not meet the

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700 mg/cm 2 as specified by Form NRC-5. The licensee responded that

whole body dose was assessed at 1000 mg/cm2 and that no assessments

were performed at 300 mg/cm 2 . The licensee also stated that each of

the Duke Power facilities had an ongoing beta radiation program

designed to characterize the beta spectrum at each unit of each

facility and that this data permitted them to conclude that whole body

dose measured at 1000 mg/cm2 was not significantly different from whole

body dose measured at 300 mg/cm2,

The inspector reviewed the results of the half value layer (HVL)

studies performed at the various facilities and found that at 300

mg/cm2 attenuation, all but approximately one to twelve percent of the

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beta radiation dose was attenuated. Consequently, for the worst case,

the maximum that whole body dose could have been underestimated was 12

percent. The licensee reviewed their dosimetry records since the

installation of the Teledyne system in 1976 and found that using this

conservative percentage of 12, no individual was likely to have

received an overexposure to the whole body. The inspector informed the

licensee that failure to assess whole body dose through a tissue

equivalent absorber of 300 mg/cm2 when the lenses of the eyes were not

shielded by at least 700 mg/cm2 would be considered an apparent

violation of 10 CFR 20.401 (50-413/86-20-01).

12. IE Information Notices (92717)

The following IE Information Notices were reviewed to ensure receipt and

review by appropriate licensee management.

85-46 Clarification of Several Aspects of Removable Radioactive

Surface Contamination Limits for Transport Packages

86-22 Underresponse of Radiation Survey Instrument to High

Radiation Fields

86-23 Excessive Skin Exposures Due to Contamination with Hot

Particles

13. Plant Statistics .

a. Collective Dose

For 1985, the collective dose for the facility as measured by TLD was

61 man-rem. A collective dose of 97 man-rem had been projected for the

year.

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For 1986, a collective dose of 540 man-rem was projected, and as of

February 28, 1986, 19 man-rem had been received. The increase in

expected man-rem for 1986 is due to the refueling outage scheduled

later in the year for Unit 1.

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