Information Notice No. 93-39: Radiation Beams from Power Reactor Biological Shields
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.
Purpose
This information notice is to 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 in a drywell to high levels
of radiation. It is expected 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 is required.
Description of Circumstances
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-041-1F084) 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 is on the 303-foot
elevation of the drywell (at an azimuth of 280 degrees). The electrical
junction box for the valve is on the 296-foot elevation at about the same
azimuth in the drywell and is directly 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 in the Unit 1 drywell, to include trouble-
shooting of the valve. The licensee stated that it routinely performs such
inspections during startup from a refueling outage.
From July 7 through 9, 1992, six separate work crews performed troubleshooting
and repair work in the 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
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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 the reactor water-level
instrument-line penetration (Figures 1, 2, and 3)1. 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
1500 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 1 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.
____________________________
1 From the Philadelphia Electric Company's presentation to the NRC
October 14, 1992
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Discussion
Upon reviewing this event, the NRC concluded that a significant potential
existed for personnel to receive exposures in 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 procedures and techniques; and (3) the licensee did not
anticipate such beams. The NRC also found that such beams may exist at other
facilities, particularly 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
postulated to occur during its installed life. These criteria are
incorporated into requirements such as Section 10 CFR 50.49 of Title 10 of the
Code of Federal Regulations (10 CFR Part 50).
Worker entry into 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 change,
and (4) the difficulty of detecting and characterizing small beams of
radiation, using routine survey procedures and instruments. If appropriate
radiation surveys inside containment at power have not been performed during a
previous reactor startup, the potential exists for significant undetected and
uncharacterized radiation dose rates from radiation beams. Each of these
factors presents significant personnel exposure control problems.
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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.
ORIGINAL SIGNED BY
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:
1. Figure 1, "Limerick Generating Station
Unit 1 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 280�"
4. Licensee Actions
5. List of Recently Issued Information Notices.
Attachment 4
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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 millirem 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:
o the beam should have been anticipated,
o 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
o 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 upward 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:
o concluded that, owing to the unusual circumstances, the subject
penetration was the worst case for occupational radiation
protection, and
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Attachment 4
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o 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-D 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:
o Prepare a special procedure for drywell entries at power.
o Require a higher level of station management approval for work in
the drywell at power.
o Prepare special guidance on survey techniques and instruments for
surveying penetrations in the biological shield.
o Emphasize work-area boundaries in preparing radiation work permits,
performing ALARA reviews, and performing pre-job briefings.
o Emphasize the need for the control of reactor power levels during
work in the drywell.
o Increase health physics coverage during at-power entries to the
drywell.
o Improve shielding, access control, or both, for high-radiation areas
resulting from radiation beams.
o Include review of this event in the training of health physics,
operations, and radiation workers.
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