ML20003C858
| ML20003C858 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/13/1981 |
| From: | Book H, Wenslawski F, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20003C855 | List: |
| References | |
| 50-206-81-02, 50-206-81-2, IEC-80-03, IEC-80-14, IEC-80-18, IEC-80-3, NUDOCS 8103180591 | |
| Download: ML20003C858 (10) | |
See also: IR 05000206/1981002
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U. S. fiUCLEAR REGULATORY COM'4ISSION
OFFICE OF INSPECTION AtlD ENFORCEMEflT
REGION V
Report No.
50-206/81-02
Docket No.
50-206
License No.
Safeguards Group
Licensee:
Southern California Edison Company
2244 Walnut Grove Avenue
Rosemead, California 91770
Facility Name:
San Onofre Unit 1 (SONGS-1)
Inspection at:
Camp Pendleton, California
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Inspection conducted:
January 19-23, 1981
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~ Inspectors: b h dtdwo
g-l3- g i
G. P.
. Radiation Specialist
Date Signed
Date Signed
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Approved by:
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F. A. Wenslawski, Chief, Reactor Radiation Safety
Dhte S'igned
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Approved by:
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H. E. Book, Chief, Fuel Facility and Materials
Date Signed
Safety Branch
. Summary:
Inspection on January 19-23, 1981 - Report No. 50-206/81-02
Areas Inspected _:
Routine unannounced inspection by a regional based
< inspector of the radiation protection program during major outage conditions;
response to IE Circulars 80-03, 80-14, and 80-18; review of Licensee
Event Report 80-37; _ and followup on previous inspection findings. The
inspection involved 38 inspector-hours onsite.
' Results . Of the areas inspected . no items of 'oncompliance were identified.
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DETAll
1.
Persons Contacted
- J. G. Haynes, Manager, Nuclear Operations
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- H. L. Ottoson, Manager, Nuclear Engineering and Safety
- J. M. Curran, Plant Manager, San Onofre
- R. R. Brunet, Superintendent, Unit 1
H. B. Ray, Steam Generator Repair Project Manager
- R. V. Warnock, Radiation Protection Supervisor
M. Wharton, Supervising Engtheer, Unit 1
- W. G. Frick, Compliance Engineer
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- J. D. Dunn, Project Quality Assurance Supervisor
- D. D. Duran. Engineer
- f. J. Bennett, Radiation Protection Foreman
- Denotes those individuals attending the exit interview on January 23, 1981.
In addition to the individuals noted above, the inspector met with
other members of the licensee's and contractor's staffs.
2.
Licensee Response to IE Circulars _
IE Circular No. 80-03, " Protection from Toxic Gas Hazards", was
received and reviewed by the licensee. Primary responsibility for
perfomance of the NRC recommended action was assigned to the
corporate Nuclear Engineering Staff. The inspector reviewed a
draft report prepared by a contractor which addressed the toxic gas
issue. This report had also been reviewed by the onsite licensee
representatives. The report and the licensee's comments were
responsive to the circular guidance. A corporate Nuclear Engineering
staff representative inforned the inspector that their January 1,1981
date for issuance of an implementation schedule has been delayed to
April 1, 1981. Since appropriate corrective actions have nrt yet
been finalized, this matter will be reviewed in a subsequena inspection.
IE Circular No. 80-14. " Radioactive Contamination of Plant Demineralized
Water System and Resultant Internal Contamination of Personnel",
was received by the licensee.
Its review was assigned to an onsite
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individual.
From~ discussions with the assigned individual, the
inspector learned that no action had yet taken place. The licensee's
response to this Circular will be reviewed in a subsequent inspection.
IE Circular No. 80-18, "10CFR50.59 Safety Evaluation for Changes to
Radioactive Waste Treatment Systems", was received and reviewed by
the licensee. The licensee's recommended action documented in a
January 13, 1981 memorandum was to revise Station Order S01-A-110
" Organization and Responsibility of the Onsite Review Committee to
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include the guidance provided in this Circular. The inspector
discussed with licensee representatives the need to insure that
individueh who are in a position to make field changes to radioactive
waste treatment systems are aware of this guidance so that they may
initiate the required safety evaluation.
This item is considered
closed.
No items of noncompliance was identified in this area.
3.
Licensee Event Report
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On October 27, 1980 the licensee submitted a report pursuant to
Section 5.6.3b(3)(a) of Appendix B to Provisional Operating License
DPR-13 that thermal measurement data required in Section 3.1.1.a.(5)
of the Environmental Technical Specifications had not been collected
on seven occasions during 1979.
From discussions with the Supervisory
Research Scientist, the inspector confirmed that the reported
corrective action to replace the aged detectors and deploy duplicate
temperature sensors had been completed on July 3, 1980. The licensee
reported that the correctiva action has been successful.
The
inspector had no further questions regarding this matter.
4.
Licensee Action on Previous Inspection Findings
(Closed) (50-206/80-13-01) Noncompliance, failure to adhere to
Radiation Protection Procedure, S-VII-1.5 regarding smoking and
drinking water in controlled areas. The licensee includes a warning
regarding smc king in controlled areas in the training program.
There are no longer any functioning drinking fountains in the
controlled area.
During tours of the controlled area, the inspector
did not observe any indication of smoking.
(Closed)(50-206/80-17-01) Noncompliance, failure to label containers
of radioactive material. The inspector verified by record review
that the two laborers, three contract radiation protection technicians,
and the Radwaste Foreman had all received the specialized training
described in the licensee's September 8,1980 response letter.
No
additional instances of improperly labeled containers of radioactive
material were observed by the inspector during tours of the restricted
area.
(Closed)(50-206/80-33-01) Inspector identified item involving the
disposal of excavated material.
During December 1980, about 100
cubic yards of sand, black top, and concrete were removed from a
location near Unit I containment structure and dumped at a landfill
on federal property near Jap Mesa.
Based on review of licensee
survey data, the inspector determined that trace quantities of
radioactive material were probably present and that the aggregate
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sum of material present may aave exceeded the values expressed in
10CFR20.304, " Disposal by burial in soil".
The inspector brought
these observations to the licensee's attention at the exit interview
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held on December 18, 1980.
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The licensee responded to this observation by:
Prohibiting further disposal of excavated material originating
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at Unit 1 from being dumped at Jap Mesa.
Identification, posting, and contr'ol of the material already
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dumped at the mesa so as to prevent its dispersion.
Collection of 13 samples from the dumped material for relative
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counting.
Shipment of the highest activity sample to an
independent laboratory for analysis.
Performing a direct radiation survey of the dumped materials
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with a low level survey instrument.
On January 6,1981, the licensee notified Region V that based on
preliminary results of. activity as reported by their independent
laboratory, that all excavated materials which had been dumped
would 'e drummed and shipped to a licensed burial facility.
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During the course of this inspection, the inspector reviewed the
licensee's data, performed independent radiation surveys using an
Eberline PRM7 Micro "R" meter serial No. 453, calibrated December 15, 1980,
and collected samples for analysis by NRC laboratory facilities.
The highest concentration sample taken from the dumped excavated
material at Jap Mesa was reported by the independent laboratory to
have the following significant isotopic content.
Isotope
Activity pCi/ gram
40 K
16.6 + .8
54 Mn
2. 2 T .1
58 Co
3. 9 T . 2
60 Co
29 + T.0
134 Cs
8.8 + .4
137 Cs
26 + T.0
144 Ce
3.5 + .2
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These activities were used by the licensee to estimate the total
activity shipped to the burial facility.
The inspector reviewed the licensee's direct radiation survey
performed on January 7,1981, using a Ludlum Micro "R" meter prior
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to removal of the excavated materials. This survey indicated 8-12
pr/hr general background with twenty five readings taken on and
around the dumped material ranging from 8 to 25 pr/hr.
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Review of radioactive material shipping records indicate that 390
fifty-five gallon drums containing a total of 7.5 mci of licensed
material in 108 cubic yards of dirt were shipped from the dump site
to a licensed burial facility in the period January 9 to January 15, 1981.
The licensee's resurvey of the area after drumming indicated 8-12
pr/hr.
On January 20, 1981 the inspector performed an independent direct
radiation survey consisting of 30 locations in the general area
where the material had been dumped. This survey indicated radiation
levels from 5-10 pr/hr with no statistically significant increase
in the localized area from where the excavated materials had been
removed.
On January 21, 1981 the inspector collected one square meter surface
samples from the effected area and from an area considered to be
background. The licensee was provided a fraction of each sample-
for comparative analysis. NRC analysis of the samples performed at
Region V using the ND6600/ intrinsic germanium detector located in
the mobile van indicate that virtually all the excavated material
containing trace quantities of radionuclides had been effectively
removed. The residual activity is noted below:
Isotope
Activity pCi/ gram
54 Mn
.04 + .02
137 Cs
.53 T .27
60 Co
.57 I .29
109 Cd
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Jased on a weighted average technique, the licensee calculated that
the mix of isotopes present in the removed material represented 773
times the values specified in 10CFR20 Appendix C.
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Review of regulatory requirements expressed in 10CFR20.304, " Disposal
by burial in soil", indicates that if the licensee had bulldozed
the piles of excavated materials into the fill area as is the
common practice for clean fill from Units 2 and 3 a violation of
regulatory requirements would likely have occurred.
Since the material was not buried, did not exceed the regulatory
limits expressed in 10CFR20.105, " Permissible levels of radiation
in ' unrestricted treas", and was' completely removed in an expaditious
manner, no . item of noncompliance was identified.
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(0 pen)(50-206/80-26-09) Noncompliance, failure to make appropriate
measurements of radioactivity in the body and measurements of
radioactivity excreted from the body of those individuals involved
in handling the NFS-4 NAC-IE cask on September 5,1980.
On October 2, 1980
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NRC Region V issued an Immediate Action Letter confirming actions
the licensee agreed to take regarding the cask.
Item 3 of that
letter stated that the licensee would make such measurements as
necessary for them to evaluate the indiv'. duals' exposure in accordance
with 10CFR20.103.
In response to the above commitment, th'e licensee arranged through
the U.S. Department of I nergy for the individuals to receive a
comprehensive series of measurements at Oak Ridge Associated
Universities. The i::easurements were performed November 6 and
11, 1980. They included a physical evaluation, cytogenetic study,
whole body counting, and measurements of. radioactivity excreted
from the body.
Based on the results of these measurements, the
licensee concluded that neither individual received an intake of
radioactive material in excess of the regulatory limit. The
inspector reviewed the data and agreed that the licensee's conclusion
was appropriate.
(Health Physics Appraisal, Inspection Report No. 50-206/80-17)
In a letter dated September 30, 1980, the licensee responded to the
findings of that inspection. The inspector reviewed five commitments
presented in the response which were scheduled for completion in
January 1981.
The first two areas reviewed involve reorganization of the
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facility staff and separation of the Station Nuclear Chenistry
and Health Physics groups.
Reorganization of the facility
staff requires amendment of Section 6.2 of. the Technical
Specifications. As of January 26, 1981 a proposed amendment
-had been generated, reviewed and was expected to be forwarded
to NRR- for approval by January 30, 1981.
Station Order S0I-E-
211 which describes the duties and responsibilities for members
of the radiation protection organization has been draf ted.
' separation of the chemistry and radiation protection technicians
has taken place.
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The licensee has implemented Health Physics Procedure 501-VII 4.2,
" Bioassay Program".
Review of this procedure indicates that
it may not provide sufficient direction to insure compliance
with the requirements expressed in 10CFR20.103b(2). -This was
brought to the licensee's attention.
The licensee stated that an inventory, maintenance, and calibration
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program for health physics instruments will be established and
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implemented by January 1, 1981.
In a letter dated January 15, 1981,
the licensee informed Region V that this program will not be
in place until January 31, 1981.
Review of this commitment
will take place in a subsequent inspection (50-206/81-02-01).
A new high pressure baler was installed in the Auxiliary
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Building. The inspector reviewed Radiation Protection Procedure
S01-VII-1.56, " Compacting Low Level Radioactive Waste", which
diracts operation of that equipment.
No item of noncompliance was identified in this area.
5.
Radiation Protection Activir.ies
On January 19, 1981 after normal working hours, the inspector made
an unannounced radiation survey of the beach area adjoining the
Tsunami wall to determine compliance with 10CFR20.105. The survey
was performed using a PRM7 Mirco "R" portable survey instrument.
10CFR20.105(b)(2) permits a maximum radiation level of 100 mrem in
any sevel consecutive days (0.595 mr/hr) in an unrestricted area
unless otherwise authorized by the Commission. The maximum observed
radiation level was 0.1 mr/hr at a position approximately 50 feet
from the wall southwest of the Auxiliary Building. The 30 measurements
ranged from 0.007 mr/hr to the maximum 0.1 mr/hr.
The distribution
of-radiation levels appeared to indicate a localized source within
the restricted area near the Auxiliary Building.
The inspector discussed the survey results with licensee representatives
in terms of the ALARA criterion. Several licensee representatives
stated that they felt the probable source of radiation was the
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Waste Monitor Tanks.
On January 22, 1980 the inspector performed
an independent survey of the Waste Monitor Tanks and Auxiliary
Building area usinc a Keithley, Model 36100 Serial No. 9864 portable
radiation survey instrument, calibrated January 6,1981. The
average radiation level measured at contact with the exterior of
the Waste Monitor Tanks was less than 30 mr/hr. The concrete block
cubicle adjacent to the Auxiliary Building was found to have a
general area radiation level ranging from 50 to 350 mr/hr on its
roof.
The maximum obsey.'d contact measurement on the surface of
this roof was 550 mr/hr. 5!nce the entrance to this cubicle had a
radiation level 420 mr/hr, the inspector did not perform a survey
inside.
A licensee representatives stated that the cubicle contained
some fairly high level radioactive waste. Based on the survey and
discussions with licensee representatives, it appears that the high
level radwaste storage cubicle is responsible for most of the
radiation measured on.the beach. The inspector observed that the
licensee had placed 3/8 inch sheet lead on the roof of this structure
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since the August inspection; however, the storage of high level
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waste appears one area worthy of additional review from an ALARA
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point of view.
The inspector performed a tour.of the controlled area including-the
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containment on the evening of January 19, 1981.
During this tour,
the inspector made independent radiation measurements to verify
compliance' with posting and control of radfCive material requirements
expressed in 10CFR20.203, observed comfiiance with the licensee's
radiation protection Lprocedures, and re. viewed: licensee survey
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Radiation areas and radioactive materials were properly posted and
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controlled.
Considerable _ ALARA effort was observed in the containment
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.and steam generator work areas.
Radiation survey records used by the Chemistry Radiation Protection
Technicians to determine precautionary requirements for Radiation
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Exposure Permits were reviewed.
Surveys performed on December 17, 1980-
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January 10'and January 19,-1981 indicated beta-gamma smearable
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activity from 50,000~to 500,000 dpm/100cm on the. 38' and -10'-
elevations of containment.- These survey records did not include a
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measurement of alpha activity as required by Radiation Protection
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Procedure S-VII-1.13 Section- IV,10. This procedure requires an
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alpha count anytime the beta-gamma activity exceeds 22,000 dpm/100cm .
The surveys were performed by contractor Radiation Protection-
Technicians. The survey records.had'other minor errors and omissions
and ~did 'not include any evidence that they had been ~ reviewed by
station ~ personnel.
The inspector questioned the Senior Radiation Protection Technician
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1 associated with the cadre of contractor. technicians that had-performed
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these-surveys. This individual acknowledged that they had been
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trained in:the licensee's survey techniques including procedure
S-VII- 1.13 and-had errored with ' respect to' the surveys in question.
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The individual-stated that this same point had been b'rought to his
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The Radiation Protection Supervisor informed ^the inspector that his
- Acting Radiation Protection Foreman had identified this problem and
was in the process 'of preparing a memorandum to reemphasize proper
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survey techniques and documentation.
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0n the evening of January 21,_1981 while performing a survey:of the
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protected. area,' the inspector measured average' radiation levels of
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1.0 mr/hr inside the Security Escort Trailer.1.-This trailer is
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located in close proximity to the R.E.D.~ Building.
In discussions
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that the trailer is occuppied 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day normally oy two individuals
Recently, the dose received by these individuals had increased.
Based on review of survey records dated December 21 and 23,1980
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and discussion with licensee representatives, it appears that the
movement of steam generator grit tank, which read an average af
400 mr/hr on contact, from the containment on December- 21, 1980 to
the R.E.D. Building increased-the dose rate in the trailer. Surveys
performed on December 23, 1980 indicated a measurement of 4 mr/hr
was made outside the R.E.D. Building near the trailer.
Although no
record indicates survey results inside ,the trailer, a licensee
representative stated that he surveyed the trailer and posted it
with a sign requiring personnel dosimetry.
He stated that no
action was taken to reduce the radiation level inside the trailer
since he did not expect the grit tank to be in the R.E.D. By f.fing
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very long.
The inspector brought the licensee's attention to the apparent
unnecessary exposure being received by occupants of the Security
Escort Trailer. The licensee performed a survey of the situation
and had the trailer moved to a low background area within the
Protected Area. On January 23, 1981, the inspector resurveyed the
trailer and noted the average radiation level had decreased to 0.04
mr/hr.
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With regard to the steam generator repair project, survey data,
operational experience and personnel exposures were examined and
discussed with licensee and contractor representatives.
Several additional positive steps have been accomplished to reduce
exposures since the la'st ' inspection. These included revision of
security escort practices and installation of the tube sheet
shields. The revision of security escort practices is expected to
save 20 person-rem.
Installation f the tube sheet shield reduced
the A steam generator dose rate from 2.8 to 1.6 r/hr.
This will
result in considerable exposure savings.
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Additional area shielding has been installed and more channel head
surface shielding is planned.
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Review of ~ actual versus projected radiation exposures associated
with the steam generator repair project indicates that 820 person-
rem have been incurred through January 21, 1981. This is about 245
person-rem in excess of the November 16, 1980 dose projection. The
excess 245 person-rem were primarily.the result of program changes
and process system _ failures.
The licensee is experiencing considerable difficulty in the brazing
phase of the repair project.
If the previous dose accumulation
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versus dose projections results continue during the sleevirg phase
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then,-in spite-of the good ALARA engineering efforts, it appears
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the total dose projection of 1783 person-rem will be exceeded.
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No item.of noncompliance was identified in this area.
6.
Exit Interview
- The inspector met with the licensee representatives (denoted in
Paragraph 1) at the conclusion of the inspection on January 23, 1981.
The inspector summarized the scope and findings of the inspection.
In addition, the inspector brought to the' licensee's attention the
somewhat subjective observation that there appears to be a deterioration
in cooperation between licensee and contractor radiation protection-
personnel. The inspector noted that if this condition is real and
-is allowed to continue, the positive im.orovements in the radiation
-protection program could be effected.
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