IR 05000206/1980010
| ML13330A090 | |
| Person / Time | |
|---|---|
| Site: | 02700010, San Onofre |
| Issue date: | 07/02/1980 |
| From: | Book H, Curtis J, Weinslawski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML13330A087 | List: |
| References | |
| 15000027-80-30, 50-206-80-10, IEC-80-03, IEC-80-3, NUDOCS 8008270332 | |
| Download: ML13330A090 (7) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION V
50-206/80-10 Report N Docket N License N DPR-13 Safeguards Group Licensee:
Southern California Edison Company 2244 Walnut Grove Avenue Rosemead, California 91770 Facility Name:
San Onofre Unit 1 Inspection at:
Camp Pendleton, California Inspection conducted:
March 31, 1980 to April 8, 1980 Inspectors:
- /
30 }(,
,/J. R. Curtis, Radiation Specialist Date Signed Date Signed Approved by:
F. Wpnslawski Chief, Reactor Radiation Safet Date Signed Approved By:
"
L A
T -1 ec On H. E. Book, Chief, Fuel Facility and Materials ate Signed Safety Bran6 Summary:
Inspection on March 31 to April 3, 1980 (telecon 4/8/80) Report No. 50-206/80-10 Areas Inspected:
Routine, unannounced inspection of the licensees advanced planning and preparation efforts related to the refueling outage, investigation of licensee action in connection with a possible item of noncompliance related to a shipment of spent resin, followup on IE Circular 80-03 concerning protection -from toxic gas hazards and followup to licensee event report number 80-009. The inspection involved a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> onsite by one inspecto Results:
No items of noncompliance were identified in the areas covered during this inspectio RV Form 219 (2)
8008270 15131-
DETAILS 1. Persons Contacted
- R. Brunet, Superintendent, SONGS-i
- M. Sullivan, Supervisor, Chemical and Radiation Protection (CRP)
- G. Peckham, Chemical and Radiation Protection Engineer, Unit 1
- D. Duran, Chemical and Radiation Protection Engineer, Unit 1
- E. Bennett, Chemical and Radiation Protection Foreman, Unit 1 S. Meddling, Chemical and Radiation Protection Engineer R. Morgan, Chemical and Radiation Protection Technician, Unit 1 M. Wharton, Nuclear Engineer
- G. MacDonald, SCE, Quality Assurance W. Rutland, SCE, Quality Assurance, Unit 1 and other members of the SONGS-i staf *Indicates presence at the exit intervie. Refueling Outage - Advanced Planning and Preparation The inspector discussed general preparations and the status of specified assigned tasks related to the outage scheduled to start on or about April 11, 198 Topics covered were:
general staffing requirements in the Chemical and Radiation Protection area including assignments for shift coverage and special operations such as personnel dosimetry and access control; respiratory protective equipment use, training and maintenance; radiation safety training and unescorted access qualification training for station personnel, off-station SCE and contractor' workers; procurement and preparation of adequate protective clothing, supplies and equipment for the refueling outage operations; and special bioassay and whole body counting programs associated with the outag The inspector concluded that advanced planning and preparations for the radiation protection aspects of the outage were underway, additional staffing had been requested and was being arranged for, some specific tasks and shift coverage assignments had been made, refresher training in Radiation Protection and qualified escort training was being conducted at-an accelerated pace to provide training for personnel who will participate in the refueling outage, and equipment and supplies needed for the radiation protection aspects of the outage had been ordere Additional planning and status review meetings of the radiation protection staff were scheduled during the week prior to shutdow No items of noncompliance or deviations were identifie.
Status of IE Circular 80-0 The inspector discussed the status of IE Circular 80-03 regarding protection from toxic gas hazards with the SONGS-1 Nuclear Engineer who evaluates and schedules activity on such items. The circular was received.and routed to the appropriate SCE engineering group where the recommendations were reviewed. The licensee's action in this area is linked to projects prioritized under the TMI Lessons Learned Program (Establishment of On-site Operations and Technical Support Centers).
As a consequence, all planning, design and engineering effort related to changes in the ventilation systems for these centers and the control room are scheduled for after the present outage and prior to the January 1981 TMI Lessons Learned target dat No items of noncompliance or deviations were identifie. Review of Licensee Event Report 80-009 Re:
Inability to Perform Required Bethnic Station Survey The licensee reported on April 2, 1980 that due to severe storms and unusually heavy rains, diving surveys at only six of the required eleven bethnic stations were conducted. Technical specification requirement states that whenever primary data collection required by section 3 and 4 of the Environmented Technical Specifications is lost, invalid or cannot be obtained due to severe weather a 30-day written report shall be submitte The licensee reported this inability to obtain the required data for the first quarter in a timely manner and adequately identified the occurrence, the cause, and evaluated the impact of the loss of data. This.item is considered close No items of noncompliance or deviations were identifie. Investigation of Licensee Activity Related to a Spent Resin Shipment Involving Possible Noncompliance with Transportation Regulation A spent resin shipment was made from San Onofre Nuclear Generating Station (SONGS) to the Beatty, Nevada Burial Site on March 20, 198 Radiation surveys of the incoming shipment, made by an NRC inspector at the burial site, indicated radiation levels possibly in excess of regulatory limits. (See Attachment #1 for a copy of the NRC inspection report for the Beatty inspection). As a consequence, part of the inspection activity was devoted to gathering facts regarding the licensee's performance of radiation surveys of the spent resin shipment package and the transport vehicle prior to release, and other licensee action associated with the shipmen Interviews were conducted with licensee personnel who conducted the surveys, persons responsible for review of results, quality assurance personnel involved, and with engineering and supervisory personnel who participated in the spent resin shipment operation. Applicable procedures, records of surveys, quality assurance check sheets,
-3 telecon documentation records and other documents were reviewed and discussed. Findings related to specific areas of possible noncompliance are addressed in the following sections. The portion of the Beatty, Nevada inspection report (80-30, 15000027), refering to these areas of possible noncompliance are presented in an indented paragraph which precedes the findings of this inspectio a. Person using a certified Package shall be a registered user 10 CFR 71.12 (b)(1)(iii)
The cask was a model B-2 transport cask with certificate number 6144 and package identification number USA/6144/ The Certificate of Compliance dated January 23, 1980 approves the package for use under the general license provisions of 10 CFR 71.12 (b).
requires that a person using a package pursuant to 71.12(b) must register with the Director of Nuclear Material Safety and Safeguards prior to first use of the package. Southwest Nuclear Company had so registere Southern California Edison Company had not registere The shipment appears to be in noncompliance with 10 CFR 71.12 (b)(1)(iii). See attachment # (Note:
The conclusions of the NRC inspector at the waste disposal site were based on a list of certified users. That list did not reflect the information identified during this inspection and discussed below).
Authority to ship in a DOT specification container as specified in 10 CFR 71.12 was an item checked by the Southern California Edison Company (SCE) Quality Assurance personnel at the San Onofre Nuclear Generating Station (SONGS) site. The ID number of the B-2 container was not apparent on the container on arrival at SONGS and Southwest Nuclear Corp. was not listed on the NRC Certificate of Compliance as a registered user of the B-2 container I.D. 6144/ Prior to the shipment, Quality Assurance personnel from SCE called appropriate persons in NRC (a representative of the Transportation Certification Branch, NMSS), and DOT to get confirmation of the proper ID number and establish that Southwest Nuclear was an authorized user of the cask. Clarification of Southwest Nuclear's role as transporter/ broker and the status of SCE's requirements to be listed as a certified user were also requested. A written record of the telephone conversation and a "Rapifax" copy of an NRC letter identifying Southwest Nuclear was reviewed and discussed with the license The Region V office contacted the NRC, NMSS representative and confirmed the context of the telephone conversation between them and the SCE representatives. Based on the verbal discussions and approvals given SCE by NMSS, this matter is not considered to be noncomplianc b. Dose rate at any point on the external surface of the package should not exceed 200 mRem/hr. 49 CFR 173.393 A survey of the cask surface showed a maximum reading of 300 mR/hr using a G-M type survey instrument with NRC
- 000358 (Don Collins Minimonitor II) due for recalibration on or before May 28, 1980. The maximum reading was found on the top surface of the cask. The shipment appeared to be in noncompliance with 49 CFR 173.393 (i) which specifies that the dose rate at any point on the external surface of the package should not exceed 200 millirem per hou See attachment # The inspector interviewed the Supervisor of Chemistry and Radiation Protection, the Radiation Protection Engineer for the operation, the Radiation Protection Foreman and the technician who performed the surveys of the shipment at SONGS. The foreman and technician demonstrated the survey technique used for the March 20 shipment on a similarly loaded trailer. The inspector reviewed copies of reports of the surveys and a document describing the chronological sequence of events related to the March 20 shipment. The following paragraphs describe a summary of the licensee action related to the possible area of noncompliance referenced abov The licensee representatives, including the radiation protection technician who performed the survey and his immediate supervisor, indicated that the radiation survey of the resin shipment was done using standard techniques for this type of survey, including consideration of keeping personnel exposure ALARA. They indicated that the performance and results of the survey were subject to particular interest and attention by SCE Quality Assurance personnel because of the NRC/DOT/State of Nevada emphasis on the safe shipment of radioactive materials and by supervisory personnel associated with Southwest Nuclear who were sensitive to their contractual responsibility and the fact that this was the first of a series of SCE resin shipments scheduled for transport to the Beatty, Nevada disposal sit None of the ten measurements reported as "contact" readings on the release survey exceeded 100 mRem/h The area identified by the NRC inspector at the waste disposal site was rather small in size. It is apparent that the licensee's initial survey did not identify this area of the container. On March 22, 1980, SCE personnel went to the disposal site to re-survey the shipment. Using the same survey instrument initially used, the licensee obtained a maximum surface doserate of 180 mrem/hour. It is believed the difference between the NRC and licensee measurements can be attributed to the survey instruments used. The NRC instrument was a small hand held instrument with a small detector capable of measuring dose rates closer to the surface. It is interesting to note that a State of Nevada survey performed the same time as the NRC survey and using a third kind of survey instrument revealed 210 mrem/h Subsequent to the NRC inspection at the waste disposal site, a question was'raised regarding the applicability of the referenced DOT requirement (49 CFR 173 (i)).
Because the shipment was on an exclusive use vehicle, it is apparent that the limits specified in 49 CFR 173 (j)
are the applicable requirements. 49 CFR 173 (j) (2)
specifies a limit of 200 millirem per hour at any point on the external surface of the car or vehicle but applies to closed transport vehicles only. The shipment in question was an open transport vehicle and it was not clear if the 200 millirem limit applied to the surface of a package in the case of an open vehicle. The NRC requested an interpretation from the DOT on this point. The DOT interpretation provided to the NRC stated: "If a shipment is made on an open transport vehicle (i.e. any vehicle not meeting 173.389(q)), then no package in the shipment is allowed to exceed the radiation levels prescribed in 49 CFR 173.393(i). This applies whether or not the vehicle is restricted to exclusive use."
Thus the requirement originally referenced by the NRC inspector at Beatty was correc The substance of this potential item-of noncompliance was discussed within the Region V office and IE Headquarters. When all factors were considered, it was concluded that the matter was not noncomplianc This decision recognized that certain limited errors are frequently present in calibrated instruments and that additional disagreements can be introduced by different types of measuring instruments and different dimensions of the detectors involved. This situation had been further complicated by the fact that the area of higher radiation levels was a rather small, localized spo C. Radiation level at 2 meters (six feet) from the vertical planes projected from the outer edges of the vehicle shall not exceed ten millirem per hour. 49 CFR 173.393 A survey at six feet from the outer edges of the trailer showed a maximum reading of 22 mR/hr. The NRC #000358 instrument was used. The shipment appeared to be in noncompliance with 49 CFR 173.393 (j)(3) which requires that the radiation level at six feet from the vertical planes projected from the outer edges of the vehicle does not exceed 10 millirem per hou The inspector interviewed persons who participated in the radiation protection aspects of the spent resin shipment, reviewed survey records and other documents related to preparation of the shipment for release from the site. The licensee provided copies of.survey reports and a document describing the chronological sequence of events related to the March 20 shipment. The following paragraph describes a summary of the licensee's action related, to the possible area of noncompliance referenced abov The licensee representatives indicated that their preliminary survey of the loaded spent resin cask indicated that the radiation level at two locations six feet from the front portion of the transport container exceeded 10 mRem/hr. As a result of this finding, extra shielding in the form of approximately 1/2-inch thick steel plate was installed inside the transport containe SCE maintenance, radiation protection and quality assurance personnel were "held over" beyond their normal quitting time to complete the task of installing the shieldin The loaded trailer was moved away from the resin transfer area to an area of normal background and a second radiation survey was performed. Radiation levels in this release survey were measured and confirmed by SCE and Southwest Nuclear personnel using three separate survey instruments, two of these were identified by number and had been calibrated at SONGS on January 23, 1980. All radiation levels reported in this survey were below DOT regulatory limits and the shipment was release State of Nevada surveys performed at the disposal site showed maximum radiation levels six feet from the shipment to be 15 mR/h The disposal site representatives measured the maximum radiation levels and found levels of 12 mR/hr. There is no readily apparent reason for the differences in survey results other than the variations that can occur between instruments as discussed in paragraph (b)
above. It is possible that a shift in the resin within the container or supplemental shielding occurred, but this possibility was not established as fact. It is noted that the referenced DOT requirement specifies the 10 mrem/hr limit at "2 meters (six feet)."
The NRC inspector (and others stated above) performed the survey at the more conservative six feet distance. It is suspected that the radiation levels at two meters would have been less than those actually measured. Considering the variations of measured readings and the question of the distance measurement, it was concluded that this item would not be identified as noncomplianc. Exit Interview An exit interview was held at the close of the inspection, at which the scope and findings of the inspection were reviewed and discussed with licensee representatives. The licensee was advised that the findings of the inspection related to the spent resin shipment would be presented to the NRC Regional Office for consideration and review and that the issues of possible noncompliance with DOT and NRC regulations related to the packaging and shipment of the spent resin, shipped on March 20, 1980 would require consultation at levels of higher authorit hI I!,