ML20205E939

Insp Repts 50-338/86-07 & 50-339/86-07 on 860317-21. Violation Noted:Failure to Perform Surveys & Dose Rates on External Surface of Package of Radioactive Matl

ML20205E939
Person / Time
Site:	North Anna
Issue date:	04/28/1986
From:	Hosey C, Revsin B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20205E923	List: ML20205E919 ML20205E925 ML20205E939
References
50-338-86-07, 50-338-86-7, 50-339-86-07, 50-339-86-7, NUDOCS 8608190041
Download: ML20205E939 (11)

See also: IR 05000338/1986007

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UNITED STATES

[gn Mo o ' NUCLEAR REGULATORY COMMISSION

y" , & REGION 11

y j 101 MARIETTA STREET, N.W.

• t ATLANTA, GEORGI A 30323

%*****/

Report Nos.: 50-338/86-07 and 50-339/86-07

Licensee: Virginia Electric and Power Company

Richmond,'VA 23261

-Docket Nos.: 50-338 and 50-339 License Nos.: HPF-4 and NPF-7

Facility Name: North Anna 1 and 2

Inspection Conducted: March 17-21, 1986

Inspector: - -

44RN) b

B. K'. Rev sN y Date Signed

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Approved by:

C. M. Hosey, Sect \on Chief Date Slgned ,

Division of Radiat) ion Safety and Safeguards

SUMMARY

' Scope: This routine, unannounced inspection involved 38 inspector-hours 'onsite

during normal hours in the areas of radiation protection including control of

radioactive materials, contamination surveys and monitoring; external exposure

control and personal dosimetry; internal exposure control; facilities and

equipment; solid radioactive waste; transportation of radioactive materials; and

program for maintaining exposures as low as reasonably achievable (ALARA).

Results: Two violations - failure to perform surveys, and dose rates on the

external surface of a package of radioactive material offered to a carrier for

transport in excess of Department of Transportation (DOT) limits.

8608190041

DR 86080s -

ADOCK 05000338

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• E. Wayne Harrell, Station Manager

• E. R. Smith, Jr. , Assistant Station Manager

• A. H. Stafford,. Superintendent, Health Physics

• 0. E. Hickman, Jr., Supervisor, Health Physics

• R. T. Johnson, Supervisor, Quality Control

• J. Leberstien, Licensing Coordinator

• F. L. Thomasson, Corporate Health Physics

R. R. Irwin, Supervisor, Health Physics

T. G. Johnson, Assistant Supervisor, Health Physics

R. Evans, Shift Supervisor, Health Physics

T. Peters, Shift Supervisor, Health Physics

H. F. Kahnhauser, Assistant Supervisor, Health Physics

J. Johnson, Health Physics Technician Trainee

Other licensee employees contacted included six technicians, and two

security force members.

NRC Resident Inspectors

M. Branch, Senior Resident Inspector

L. King, Resident Inspector

• Attended exit interview

2. Exit Interview

The inspection scope and findings were summarized on March 21, 1986, with

those persons indicated in Paragraph 1 above. Two violations, (1) failure

to comply with Department of Transportation requirements for radiation

levels on the external surfaces of packages (Paragraph 10), and (2) failure

to perform radiation surveys (Paragraph 6), and one unresolved item *,

assessment of occupational exposure delivered through a tissue equivalent

absorber of 300 milligrams per square centimeter (Paragraph 6), were

discussed in detail. Licensea management took no exceptions. The licensee

did not identify as proprietary any of the material provided to or reviewed

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by the inspector during this inspection.

3. Licensee Action on Previous Enforcement Matters

This subject was not addressed in the inspection.

• 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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4. Training and Qualifications (83723) '

a. Radiation Protection and Chemistry Technician Training Program

The licensee was required by Technical Specification (TS) 6.3.1 to

qualify radiation protection and chemistry technicians in accordance

with ANSI N18.1-1971. The inspector discussed with the Health Physics

Superintendent (HPS) the progress of the accreditation process by the

Institute of Nuclear Power Operations (INP0) for these programs.

The HPS stated that the programs had been submitted to INPO in 1985 and

that during the first week of February 1986, an INPO accreditation

visit had been made to the site. The licensee expected to have both

health physics (HP) and chemistry technician training programs

accredited by mid-1986.

b. Advanced Radiation Worker Training

The licensee was in the process of developing a training program for a

category of radiation worker called Advanced Radiation Worker (ARW).

This program was designed to take station employees other than HPs and

provide them with a more extensive background in HP. At the time of

the inspection, only quality maintenance teams had been eligible to

receive ARW training and three training sessions had been held. The

licensee stated that the intent of the program was to increase worker

involvement in radiation protection and to relieve the HP staff by

providing positive control over work by the ARW in the Radiation

Control Area (RCA) for jobs performed where general area whole body

exposure rates are less than three Roentgens per hour. These workers

would then provide job coverage but only after initial radiological

conditions had been determined by HP.

The HPS stated that the program was still in the developmental stages

and that work was continuing toward developing the guidelines and

controls that would be used to implement the program.

c. Staffing

TS 6.2.2 specified minimum plant staffing. Final Safety Analysis

Report ( FSAR) , Chapters 12 and 13, outlined further details on

staffing. The HP Superintendent stated that authorization for 12

additional HP staff had been approved which would bring the total staff

to 92 persons. At the time of the inspection, 78 positions were filled

and recruiting efforts were underway to man the newly acquired

positions.

For the ongoing Unit 2 refueling outage, 164 contract HP personnel had

been secured to assist in providing work coverage. One hundred of

these individuals were technicians and 64 were decontamination workers.

No violations or deviations were identified.

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5. 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 as necessary to show compliance with

regulatory limits. Survey methods and instrumentation were outlined in the

FSAR, Chapter 12, while Technical Specification (TS) 6.11 required adherence

to written procedures for all operations involving personnel radiation

exposures.

a. Frisking

During tours of the plant, the inspector observed the movement of

workers and material f om contamination control areas to clean areas of

the plant to determine if proper frisking was performed by workers and

if proper direct and removable contamination surveys were performed on

materials. The licensee had recently abandoned the practice of wearing

shoe covers in the RCA and as a result, had detected an increased

number of shoe contaminations at the frisking station at the exit to

the RCA. The licensee stated that these events had led to a change in

the practice of wet mopping RCA floors to using Masslin wipes which had

been effective in reducing the numbers of shoe contaminations being

detected at the exit,

b. Surveys

10 CFR 20.203(b) stated that each radiation area shall be conspicuously

posted with a sign or signs bearing the radiation caution symbol and

the words, Caution, Radiation Area.

10 CFR 20.202(b)(2) defined radiation area to mean any area, accessible

to personnel, in which there exists radiation, originating in whole or

in part within licensed material, at such levels that a major portion

of the body could receive in any one hour a dose in excess of five

millirem or in any five consecutive days, a dose in excess of 100

millirems.

During plant tours, the inspector examined radiation level and

contamination survey results outside selected cubicles and on survey

maps in the HP office. The inspector obtained independent radiation

level measurements of selected areas in the auxiliary building and in

tha restricted area outside the auxiliary building to verify that areas

were properly posted. The inspector noted that locked high radiation

areas inside and outside containment were maintained as required by

TS 6.12.

c. Instrumentation

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

instruments by plant staff and compared plant survey instrumer.t

readings with measurements made by the inspector using NRC equipment.

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The inspector examined calibration stickers on radiation protection

instruments in use by licensee staff and friskers located throughout

the plant.

No violations or deviations were identified.

6. External Occupational Dose Control and Personal Dosimetry (83524)

10 CFR 20.101(b)(3) required licensees to determine an individual's

accumulated occupational exposure to the whole body on an NRC-4 Form or-

equivalent prior to permitting the individuals to exceed the limits of

10 CFR 20.101(a). The inspector reviewed selected occupational exposure

histories of individuals who exceeded the values in 10 CFR 20.101(a) and

deterrined that exposure histories were being completed and maintained as

required.

10 CFR 20.202 required each licensee to supply appropriate personnel

monitoring equipment and to require the use of such equipment. During tours

of the plant, the inspector observed workers wearing thermoluminescent

dosimeters (TLDs) and pocket dosimeters as required.

10 CFR 20,201(b) specified that each licensee shall make or cause to be made

such surveys as may be necessary for the licensee to comply with the

regulations and are reasonable under the circumstances to evaluate the

extent of radiation hazards that may be present. 10 CFR 20.201(a) defined

survey to mean an evaluation of the radiation hazards incident to the

production, use, release, disposal, or presence of radioactive material or

other sources of radiation under a specific set of conditions.

The inspector reviewed the following Radiation Work Permits (RWPs)

associated with the Unit 2 refueling outage:

RWP 86-SP-200, Remove and Replace Snubbers in U-2 Reactor Containment

RWP 86-SP-228, Remove Primary Manways and Diaphragms, "C" Steam

Generator (S/G)

RWP 86-SP-241, Remove Primary Side Diaphragms, Install Ventilation

on "C" S/G

RWP 86-SP-249, Remove Diaphragms, Install Nozzle Covers, Set Up

Equipment "C" S/G

RWP 86-SP-253, Perform Eddy Current Testing Primary Side "C" S/G

RWP 86-SP-202, Remove Primary Manways and Diaphragms "B" S/G

RWP 86-SP-233, Inspect "B" S/G, Primary Side

RWP 86-SP-240, Perform Eddy Current in "B" S/G

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RWP 86-SP-288, Remove Insulation and Primary Manway from "A" S/G

RWP 86-SP-289, Remove Primary Side and Diaphragms, Install Ventilation

on "A" S/G

RWP 86-SP-309, Perform Eddy Current in "A" S/G

The inspector noted that for S/Gs "C" and "A," direct beta radiation surveys

had not been obtained after S/G manway removal nor after diaphragm removal

and that eddy t irrent testing had been performed without ascertaining the

beta radiation hazard. The licensee explained that during the outage, the

"B" S/G had been opened first and that beta surveys had been performed.

They reasoned that since all three S/Gs were exposed to the same reactor

coolant, surveys on "C" and "A" were not mandatory. The inspector reviewed

the survey map of the bowl of "B" S/G taken after manway and diaphragm

removal. The measured beta dose rates were as follows:

Hot Leg Cold Leg

1 foot outside manway 3 rad 4 rad

At manway (inside) 1? rad 8 rad

1 foot inside manway 0 rad 2 rad

3 feet inside manway 6 rad 0 rad

The inspector questioned licensee representatives as to the reason for the

apparently anomalous beta radiation results. The licensee had no

explanation as to why beta radiation levels were higher outside the steam

generator than at some locations inside the steam generator. The inspector

stated that even though all three S/Gs were exposed to the same source term,

there could be no guarantee that radiation levels would be similar, and in

view of the paradoxical results from the survey of S/G "B," measurement of

beta radiation levels in S/Gs "C" and "A" would be necessary to evaluate the

hazard that may have been present. The inspector stated that failure to

perform surveys for S/Gs "C" and "A" was an apparent violation of

10 CFR 20.201(b) (50-338,339/86-07-01).

10 CFR 20,401(a) required each licensee to maintain records showing the

radiation exposures for all individuals for whem personnel monitoring was

required by 10 CFR 20.202 and that such records shall be kept on NRC-5 Form

or equivalent in accordance with the instructions contained in that form.

Item 5, Instructions for Preparation of NRC Form 5, stated that unless the

lenses of the eyes are protected with eye shields having a tissue equivalent

thickness of at least 700 milligrams per square centimeter (mg/cm2), dose

recorded as whole body dose should include dose delivered through a tissue

equivalent absorber having a thickness of 300 mg/cm2 or less.

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The licensee stated that the TLD system in use at the facility had four main

areas, Areas 1, 2, and 4 of which were normally used to determine beta dose

and gamma dose and provide for the estimation of the energy of the gamma

radiation. Area 3 was identical to Area 4. The licensee stated that whole

body dose was assigned from Area 4 which has a density thickness of 2750

mg/cm2 due to aluminum and copper filters added to 1000 mg/cm2 CaSO. which

corrected for the over response of the phosphor at low energies.

In the S/G work that had been performed during the ongoing refueling outage,

the eyes of the radiation workers had been protected by air supplied hoods

which had a density thickness of 75 mg/cm2 The inspector queried licensee

representatives regarding whole body exposure assessment delivered through a

tissue equivalent absorber having a density thickness of 300 mg/cm2 or less

as specified by Form 5 instructions. The licensee stated that ratios of

Area 2 to Area 4 of the TLD ribbon indicated that the gamma radiation

generally seen at the facility was sufficiently energetic that any resulting

dose would be measured as deep dose on Area 4. The licensee also stated

that due to studies that had been performed by the Corporate HP group, the

conclusion had been reached that the observable beta radiation seen at the

facility was not energetic enough to penetrate tissue to a depth greater

than 300 mg/cm 2 and that whole body dose from this source would, therefore,

be negligible.

The inspector reviewed the licensee's study data, noting that the analysis

of plant samples utilized data from the licensee's Surry Power Station. The

licensee stated that the studies had originally been performed to

characterize the beta spectrum at Surry; however, samples had also been

obtained at North Anna, the results of which fell within the ranges of the

Surry results. Consequently, it had been assumed that the results from

Surry data were applicable to North Anna as well. The Surry study concluded

that the beta spectrum characteristic of the facility was most typically

repres;nted by T1-204 which had a maximum beta energy of 765 kev and an

average beta energy at 267 kev.

Review of the Surry study revealed several factors:

a. Methodology utilized to calculate maximum and average beta energies

from smears appeared nonconservative in that verification of the

methodology using pure sources of Co-60, Cs-137, and Sr/Y resulted in

underestimates of both the maximum and average beta energies,

b. Attenuation studies performed on the smears showed that all beta

activity was not attenuated by a density thickness of 300 mg/cm2 ,

c. Isotopic analyses performed on smears included only those nuclides that

could be easily measured by gamma spectroscopy. Consequently, the

contribution of isotopes such as Sr and Y were not included in the

energy evaluation.

d. Smears used in the study were taken from various areas in the plant but

it was not clear if the sample points represented systems most likely

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to contribute to the worker's beta exposure, e.g., steam generator

diaphragm, tube sheet, etc.

e. It was not clear whether an empirical verification of the theoretical

treatment of the data had been performed.

The inspector stated that even though T1-204 might be most representative of

the beta energy spectrum seen at Surry, North Anna would need to perform an

analysis of their own data to verify that similar results were applicable to

their f acility, particularly in view of the fact that the percent composi-

tion for the nuclides- in the Surry study varied so widely and that the

questions raised above would need to be addressed. Further, that since

radionuclides such as Cs-134, Cs-137, Sr-89, Sr/Y and Co-60 had maximum

beta energies capable of penetrating to a density tnickness greater than

300 mg/cm 2

, their contribution to whole body dose required assessment. In

addition, the necessary frequency for reverification of the beta

contribution to worker whole body exposure should be considered.

Since in addition to high energy beta radiation, some low energy gamma

radiation might be expected to contribute to the whole body dose measured

between 300 mg/cm2 and 1000 mg/cm 2 , the licensee should make the

determination as to whether the TLD system currently in use is adequate to

assess whole body dose delivered through a tissue equivalent absorber having

a density thickness of 300 mg/cm2 or less. Pending further review of the

data by the licensee, the inspector stated that this would be considered an

unresolved item (50-338,339/86-07-02).

7. Internal Exposure Control and Assessment (83525)

The licensee was required by 10 CFR 20.103, 20.201(b), 20.401, and 20.405 to

control uptakes of radioactive material, assess such uptakes and keep

records of and make reports of such uptakes. FSAR, Chapter 12, included

commitments regarding internal exposure control and assessment.

During plant tours, the inspector observed the use of temporary ventilation

systems and containment enclosures. The inspector discussed the use of this

equipment with HP staff. The inspector observed the use of respirators by

workers performing various work activities in Unit 2 containment and in the

auxiliary building, and observed the storage of respirators and verified

that storage conformed to manufacturer's recommendations. The inspector

reviewed the air sample data and Maximum Permissible Concentrations (MPCs)

calculations which had been used to control respiratory protection require-

ments for the kWPs enumerated in paragraph 6.

No violations or deviations were identified.

8. Solid Waste (84722)

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

radioactive waste shipments to verify that shipments have been received

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

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tracking methodology utilized by the licensee and examined the documented

receipt acknowledgements in the shipping files for selected shipments made

in 1986.

10 CFR 20.311 required a licensee who transfers radioactive waste to a land

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

according to 10 CFR 61.55.

10 CFR 61.55 specified the parameters applicable for waste classification.

The inspector reviewed procedure No. 3.2.8, Radwaste: Packaging and

Shipment of Radioactive Waste. The licensee's program for development of

scaling factors to relate the inferred concentration of one radionuclide to

another that is measured was examined to verify that the in ferential -

determination could be related to actual measurements. Scaling factors that

were developed from smears and used for dry active waste classification for

1986 were examined and were found to be within a factor of ten. The

licensee stated that they were in the process of changing vendors for the

computer code that performs the waste classification and generates the

radwaste manifest. This change was anticipated to take place during 1986.

No violations or deviations were identified.

9. Maintaining Occupational Doses ALARA (83728)

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

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

regarding ALARA actions.

The inspector reviewed recent changes to administrative procedures that

implemented the elements of ALARA and determined that in February 1986, the

requirements for a quorum for the ALARA Committee had been relaxed. A

review of attendance at ALARA Committee meetings from January 1985 through

February 1986 showed that the September 1985 meeting had been cancelled for

lack of a quorum and that in general, attendance by some Committee members

was sporadic.

The licensee stated that recently, ALARA Coordinators had been appointed for

each department and would -act as the focal point for ALARA issues for their

respective departments. In turn, these individuals would sit on the ALARA

Committee.

The inspector reviewed several ALARA Post Job Reviews (Nos. 85-AE-0076 and

85-AE-88) from the Unit I refueling outage in November 1985. The licensee

stated that post job review was required when the collective dose for a RWP

exceeded ten man-rem and the prejob man-rem estimate was in error by more

than 25 percent. The inspector stated that the post job reviews appeared

cursory in scope in that they were essentially check lists so that the

problems that had been encountered were never explained or documented. The

ALARA Coordinator stated that plans were underway to improve that aspect of

the ALARA program.

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The inspector discussed the goals and objectives for 1986 with licensee

representatives and reviewed the man-rem estimates for 1986. A collective

dose of 485 man rem had been projected for the facility for 1986 and on

March 19, 1986, the facility surpassed this goal.

No violations or deviations were identified.

10. Transportation (86721)

10 CFR 71.5 required 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, shall comply with the applicable

requirements of the regulations appropriate to the mode of transport of the

Department of Transportation in 49 CFR Parts 170 through 189.

The inspector reviewed selected records of radioactive waste shipments

performed in 1986 and verified that the requirements of 49 CFR Parts 170

through 189 had been met for those shipments.

49 CFR 173.441(a) stated that except as provided in 49 CFR 173.441(b), each

package of radioactive materials offered for transportation shall be

designated and prepared for shipment so that under conditions normally

incident to transportation, the radiation level does not exceed 200 millirem

per hour at any point on the external surface of the package.

On March 5,1985, radioactive materials shipment No. 86-RNS-14 containing

boxes of equipment, unused resin and two pressure vessels (demineralizers)

was shipped offsite to the owner of the equipment. Preshipment survey of

the external surfaces of the packages disclosed that each of the pressure

vessels possessed a " hot spot," an area approximately two inches in

diameter, with a radiation level of 200 millirem per hour as measured by

Teletector flo. 35646. The measurements were taken at one inch from the

surfaces of the containers. The shipment vehicle was an open, flat-bed

truck and prior to leaving the site, the packages were covered by a

tarpaulin.

On March 6, 1986, the equipment owner notified the licensee that one

pressure vessel had a radiation level in excess of 200 millirem per hour on

the external surface of the package upon arrival at their facility. On

March 11, 1986, three licensee representatives traveled to the vessel

owner's facility to resurvey the piece of equipment in question. Using the

same Teletector (No. 35646), dose rates were found to read 225 millirem per

hour at one inch from the surface of the vessel and 250 millirem per hour

with the detector in direct contact with the exterior surface of the

pressure vessel. The second pressure vessel was not resurveyed. The survey

performed before the shipment left the licensee facility and the resurvey

performed on March 11, 1986, were performed by the same individual. The

exception of 49 CFR 173.441(b) were not applicable to this shipment. The

inspector informed the licensee that dose rates on the external surface of

the package in excess of 200 millirem per hour was an apparent violation of

10 CFR 71.5 (50-338,339/86-07-03).

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11. IE Information Notices (92717)

IE Information Notice-85-92, Surveys of Wastes Before Disposal From Nuclear

Reactor Facilities, was reviewed to ensure its receipt and review by

appropriate licensee management.

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12. Enforcement Conference

An Enforcement Conference was held at NRC Region II on April 9,1985, to

discuss the transportation of radioactive materials. The following persons

were in attendance:

1. Virginia Electric and Power Company

W. L. Stewart, Vice President, Nuclear

E. R. Smith, Jr., Assistant Station Manager, Nuclear Safety and

Licensing

A. H. Stafford, Superintendent, Health Physics

-W. W. Cameron, Director Health Physics, Corporate

N. E. Clark, Licensing

0. E. Hickman, Jr., Supervisor, Health Physics

D. J. Van de Walle, Supervisor, North Anna Licensing

A. C. Rowe, Manager, Security

2. Nuclear Regulatory Commission

J. P. Stohr, Director, Division of Radiation Safety and Safeguards

J. C. Bryant, Senior Resident Inspector, Oconee

G. L. Troup, Senior Radiation Specialist

B. K. Revsin, Radiation Specialist

L. Trocine, Enforcement Specialist

D. R. McGuire, Chief, Physical Security

G. F. Gibson, Physical Security Inspector

W. J. Tobin, Senior Security Inspector

During the meeting, licensee personnel presented discussions of the

transportation event described in Paragraph 10 covering: 1) sequence of

events, 2) root cause of the event 3) short term corrective actions and

4) long term corrective actions.

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NRC representatives discussed the sensitivity of transportation problem and

NRC enforcement policy in this area.

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