Information Notice 1993-39, Radiation Beams from Power Reactor Biological Shields

Radiation Beams from Power Reactor Biological Shields

ML031080043
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	05/25/1993
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-93-039, NUDOCS 9305200025
Download: ML031080043 (10)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 May 25, 1993 NRC INFORMATION NOTICE 93-39: RADIATION BEAMS FROM POWER REACTOR

BIOLOGICAL SHIELDS

Addresses

All holders of operating licenses or construction permits for nuclear power

reactors, PAurpo le

This Information notice isto alert addressees to narrow, intense beams of

radiation that can stream Into accessible areas of a drywell through

penetrations in the biological shield of a boiling-water reactor (BWR),

potentially causing personnel exposures above regulatory limits and exposing

environmentally qualified (EQ) equipment located Ina drywell to high levels

of radiation. It isexpected that recipients will review the information for

applicability to their facilities and consider actions, as appropriate, to

avoid similar problems. However, suggestions contained in this information

notice are not NRC requirements; therefore, no specific action or written

response isrequired.

cdPI

eSrDi ofs CcUms~tames

During startup of the Philadelphia Electric Company (the licensee) Limerick

Generating Station, Unit 1,on July 7, 1992, operators could not properly

operate a main steamline sample-flow isolation valve (HV-041-IF084) from the

control room. Members of the licensee Operations Group determined that the

valve was an inboard primary containment Isolation valve and generated an

action request to troubleshoot the valve. The valve ison the 303-foot

elevation of the drywell (at an azimuth of 280 degrees). The electrical

junction box for the valve ison the 296-foot elevation at about the same

azimuth inthe drywell and isdirectly across from the reactor water-level

Instrument-line penetration in the biological shield.

The licensee revised an existing radiation work permit, originally prepared

for the inspection of systems inthe Unit I drywell, to include trouble- shooting of the valve. The licensee stated that itroutinely performs such

inspections during startup from a refueling outage.

From July 7 through 9, 1992, six separate work crews performed troubleshooting

and repair work inthe drywell with the reactor operating at a maximum of

about 10 percent of its rated power.

During the first and second entries into the drywell on July 7 and 8, 1992, personnel worked only on the 303-foot elevation of the drywell, where

9305200025 14 a- P t o+;%t)

IN 93-39 May 25, 1993 radiation levels were generally low and where there were no biological shield

penetrations in the immediate work area. However, during the third and fourth

entries on July 8 and 9, 1992, personnel went to the 296-foot elevation and

either passed in front of or worked in sight of tRe reactor water-level

instrument-line penetration (Figures 1, 2, and 3) . Unknown to those working

on this elevation, a narrow, intense, beam of radiation passed from reactor

water-level instrumentation penetration N16-D directly across the drywell, striking the inner drywell wall in the immediate vicinity of the work area.

The diameter of the beam ranged from about 0.15 meter [0.5 foot] at the

penetration to 0.3 to 0.6 meter (1 to 2 feet] at the drywell wall. The NRC

determined that licensee radiological controls personnel did not know that the

third work crew had gone to the 296-foot elevation. Although a radiological

controls technician (RCT) accompanied the fourth work crew, the beam was not

detected during this entry.

During the fifth entry, on July 9, 1992, the work crew, accompanied by a

radiological controls technician, entered the 296-foot elevation and worked in

sight of penetration N16-D. While working, one worker's dosimeter alarmed;

apparently, the beam struck the dosimeter. The RCT conducted a radiation

survey, detected the beam, and immediately evacuated the work area.

The licensee detailed survey found, at the extremity of the work area, radiation levels of about 30 mSv per hour [3 rem per hour] (gamma); and, at

the point where the beam emerged from the penetration, levels of about

150U mSv per hour [150 rem per hour) (gamma) and greater than 50 mSv per hour

(5 rem per hour] (neutron). Radiation dose rates in the work area, readily

accessible to personnel, and attributable to the beam, ranged from about 30 to

about 250 mSv per hour [about 3 to about 25 rem per hour] (gamma) while the

maximum general area radiation dose rates were 10 mSv per hour [1 rem per

hour] (gamma) and 5 mSv per hour [500 millirem per hour] (neutron). These

last maximum general area dose rates were used as the basis for radiation work

permit requirements for work in the area. Figures 1, 2, and 3 show the

location of the penetration, the approximate path of the beam, and the work

location on the 296-foot elevation of the Unit I drywell. After the licensee

detected the beam and reviewed the situation, a sixth work crew entered the

296-foot elevation to restore electrical connections to the isolation valve

while remaining out of the path of the beam.

Specific licensee actions in response to this event are described in

Attachment 4.

From the Philadelphia Electric Company's presentation to the NRC

October 14, 1992

IN 93-39 May 25, 1993 Upon reviewing this event, the NRC concluded that a significant potential

existed for personnel to receive exposures i,.excess of regulatory limits on

the 296-foot elevation of the drywell. This conclusion is based on the

following: (1)personnel entered an area (the 296-foot elevation where the

beam was later found) without the knowledge of radiological controls

personnel; (2) the licensee failed to detect the beam using its normal

radiation survey p ocedures and techniques; and (3) the licensee did not

anticipate such beams. The NRC also found that such beams may exist at other

facilities, pr.ticularly at BWRs.

The Limerick event Indicates that licensees may not be adequately considering

the effects of radiation beams with respect to environmental qualification of

equipment exposed to such beams. The NRC established environmental design

criteria to ensure that all safety-related equipment is capable of performing

its safety function or remaining in a safe mode under all conditions

potulated to occur during its Installed life. these criteria are

Incorporated Into requirements such as Section 10 CFR 50.49 of Title 10 of the

SQ.LELL~AJeit&R99eU141J~n1 (10 CFR Part 50).

Worker entry intn a BWR drywell or pressurized-water reactor containment at

power involves a challenging environment for radiological controls and

monitoring. These include (1)the possibility of high levels of airborne

radioactivity, (2)high gamma and neutron radiation dose rates, (3) the

potential for large radiation dose rate gradients, Including relatively small, Intense beams of radiation, which may change location as rod positions

and (4) the difficulty of detecting and characterizing small beams of change, radiation, using routine survey procedures and instruments. If appropriate

radiation surveys inside containment at power have not been performed during

previous reactor startup, the potential exists for significant undetected and a

untharacterized radiation dose rates from radiation beams. Each of these

fdctors presents significant personnel exposure control problems.

IN 93-39 May 25, 1993 This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: R. L. Nimitz, RI

(215) 337-5267 J. M. Bell, NRR

(301) 504-1083 William Ruland, RI

(215) 337-5227 Attachments:

I. Figure 1, 'Limerick Generating Station

Unit I Drywell*

2. figure 2, 'Limerick Generating Station, Unit 1, Drywell 296 Elevation Survey Data"

3. Figure 3, 'Limerick Generating Station, Unit 1, Section Along Azimuth 2806*

4. Licensee Actions

5. Ist of Recently Issued Information Notices

-

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_ -i_

- _

Figure 1: Limerick Generating Station Unit I Drywell

-

Source: Philadelphia Electric Company's Briefing of NRC, 10/14/92

.

C3

0 R/hr Gonrn

0 = Renm/hr Neutron

(7=

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'E D r

9-

180 '0

0w

i

Figure 2: Limerick Generating Station Unit I Drywell

296' Elevation Survey Data

Source: Philadelphia Electric Company's Briefing of IRC,

10/14/92

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EL. 2M--9 -

-lNo OD AL rob

SCAVE -;I w

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Figure 3: Limerick Generating Station Unit I Drywell

Section Along Azimuth 280-

Source: Philadelphia Electric Company's Briefing of NRC. 10/14/92

Attachment 4 IN 93-39 May 25, 1993 Page I of 2 LICENSEE ACTIONS

After detecting and characterizing the beam, the licensee took the following

corrective actions:

1. Immediately evacuated the area and had radiation protection supervisors

and station managers evaluate the situation.

2. Prevented personnel from entering the beam path.

3. Sent personnel dosimeters for evaluation and performed dose assessments.

4. Studied each individual's activities in the drywell to determine

individual exposures and which individuals may have been exposed to the

beam. The licensee believes that the maximum individual exposure from

the beam was about 300 inillirem and that no one's radiation exposure

exceeded regulatory limits.

5. Prepared a radiological occurrence report.

6. Had its Independent Safety Engineering Group (ISEG) perform a root- cause and barrier analysis of the event. The Group recommended that the

licensee review EQ concerns. The licensee concluded that reactor power

did not increase while personnel were in the drywell.

7. Had the ISEG evaluate the event during which it found that:

• the beam should have been anticipated,

• the beam probably resulted from the unique geometric arrangement of

the low-pressure coolant injection (LPCI) piping, the shield

penetration, and the core peak axial power location, and that

• moving the rods downward to increase reactor power caused the

location of peak axial power to move downward in the core, increasing the angle of the beam unward and causing the beam to pass

over the top of a LPCI line, located in front of the penetration, and Into the work area (see Figure 3). (Although operators changed

the reactor power level between entries, power level changes were

controlled so that no changes occurred while personnel were in the

drywell.)

8. Had the ISEG review the potential for beams at other biological shield

penetrations. The ISEG:

• concluded that, owing to the unusual circumstances, the subject

penetration was the worst case for occupational radiation

protection, and

Attachment 4 IN 93.39 May 25, 1993 * found that the recirculation inlet nozzle penetration had the

highest associated radiation dose rates for EQ considerations. (The

dose rates at this latter penetration were used for bounding EQ

calculations).

9. Confirmed previous EQ evaluations but found that more detailed reviews

were needed.

10. Compared measured dose rates as a function of distance from penetration

N16-!) with dose rates calculated earlier by the architect-engineer, establishing that the measured dose rates associated with the

penetration did not decrease as rapidly as indicated by the

calculations.

11. the licensee planned to complete the following long-term corrective

actions:

• Prepare a special procedure for drywell entries at power.

• Require a higher level of station management approval for work in

the drywell at power.

• Prepare special guidance on survey techniques and Instruments for

surveying penetrations in the biological shield.

• Emphasize work-area boundaries In preparing radidtion work permits, performing ALARA reviews, and performing pre-Job briefings.

• Emphasize the need for the control of reactor power levels during

work in the drywell.

• Increase health physics coverage during at-power entries to the

drywell.

• Improve shielding, access control, or both, for high-radiation areas

resulting from radiation beams.

• Include review of this event in the training of health physics, operations, and radiation workers.

Attachment 5 IN 93-39 May 25, 1993 Page 1 of I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Informatlon

Notice No. Subject Issuance Issued to

93- 30 Inadeq~ute Testing of OS/24/93 All holders of OLs or CPs

Engineered Safety for nuclear power reactors.

Features Actuation

System

93-3 7 Eyebolts with Indeter- 05/19/93 All holders of OLs or CPs

minate Properties In- for nuclear power reactors.

stalled in Limitorque

Valve Operator Hlousing

93 - 36 Notifications, Reports, 05/07/93 All U.S. Nuclear Regulatory

and Records of Misadmin- Commission medical

istrdtions licensees.

93-35 Insights from Common- 05/12/93 All holders of OLs or CPs

Cause Failure Events for nuclear power plants

(NPPs).

93 34, Potential for Loss of 05/06/93 All holders of Ols or CPs

Supp. I Emergency Cooling for nuclear power reactors.

Function Due to A

Combination of

Operational and Post- Loca Debris in Contain- ment

93 -34 Potential for Loss of 04/26/93 All holders of OLs or CPs

Emergency Cooling for nuclear power reactors.

Function Due to A

Combination of

Operational and Post- Loca Debris in Contain- ment

93-33 Potential Deficiency 04/28/g3 All holders of Ols or CPs

of Certain Class IE for nuclear power reactors.

Instrumentation and

Control Cables

Opera n gicse - -- --

CP - Construction Permit