ML20042G322
| ML20042G322 | |
| Person / Time | |
|---|---|
| Site: | 07001100 |
| Issue date: | 05/11/1990 |
| From: | Vaughan R ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Knapp M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 9005140088 | |
| Download: ML20042G322 (16) | |
Text
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ABB 1
ASC A BROWN BOVERI l
May 11, 1990 1
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Docket No. 70-1100 License No. SNM-1067 Dr. Malcolm R. Knapp, Director Division of Radiation Safety and Safeguards Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406
Subject:
Response to Notice of Violation (Inspection Report No. 70-1100/90-03)
Reference:
Letter, M. R. Knapp (NRC) to C.
R. Waterman (C-E), dated April 20, 1990 Dear Dr. Knappt Combustion Engineering has reviewed the Notice of Violation l
received with the Reference letter and our reply is provided herewith (Enclosure I).
Our review of the violations produced information, not made known at the time of the inspection, which we believe demonstrates compliance with requirements.
Therefore,ils some of the issued violations may be unwarranted.
The deta of this. additional information are set forth in our reply to each violation.
Further, Combustion Engineering will provide a summary of the results of our investigation relative to the unresolved item (discussed in Section 2.1.2 of the inspection report), once the investigation is completed.
ABB Combustion Engineering Nuclear Power CombuShon Engm4ng #c 1000 Prosts:t HA Road Terpwe (203) 0881911 900314cogg h hffoo Post co<e oo. sm re. (ra 2ensit
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1 Dr.'Malcolm R. Knapp Page 2 May 11, 1990 If I can be of further assistance on this matter, please do not hesitate to call me or Mr. J.
F. Conant at (203) 285-5002.
Very truly yours, 1
COMBU STION ENGINEERING, INC.
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Vaughan l
Plant Mana er Windsor Nu rF 1
Manufacturi REVtjdk Enclosure As stated xct G.
Bidinger (NRC)
J. Roth (NRC - Region I) 7
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ENCLOSURE 1 RESPONSE TO NOTICE Of VIOLATION (NRC INSPEC~ ION REPORT NO. 70 1100/90 03) t 1.
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6 Response to Notice of Violation (Inspection Report No. 70-1100/90-03)
Violation A 10CFR20,201(b) requires, in part, that each licensee shall make or cause to be made such surveys as (1) may be necessary for the licensee to comply with the regulations in this part and (2) are reasonable under the circumstances to evaluate the extent of radiation hazards that may be present.
10CFR20.101(a) requires, in part, that no licensee shall possess, use or transfer licensed material in such a manner as to cause any individual in a restricted area to receive in any period of one calendar quarter from radioactive material and other sources of radiation a total occupational dose in excess of 1.25 rems per calendar quarter to the whole body which includes the lens of the eyes and/or 7.5 rems per calendar quarter to the skin of the whole body.
10CFR20.101(b)(1) allows the licensee to permit an individual in a restricted area to receive a total occupational dose to the whole body not to exceed 3.0
+
rems per calendar quarter.
Contrary to the above, between September, 1989 and February, 1990, the licensee did not complete evaluations to:
1.
Show that adequate surveys were conducted in the Pellet Shop stack and load area to prove compliance with the dose limits of 10CFR20.101(a) and 1
(b);
2.
Determine the accuracy of beta dose measurements to the skin of the whole body, in this case, the face, and; 3.
Determine the accuracy of beta shielding of safety glasses used in the Pellet Shop to ensure compliance with whole body dose limits specified in 10CFR20.101(a) or (b).
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Response
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A review of this violation and the facility's operating history with regard to previous studies performed in this area has led Combustion Engineering to conclude that the violation is unwarranted.
During the period from February 1985 through June 1985, Combustion Engineering conducted an extremity monitoring-survey in order to obtain actual dose exposure information to the u
extremities and the unshielded faciel area next to the eyes of workers
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assigned to various work stations in the facility.
The survey was I
accomplished by assigning additional TLD badges to the participating workers which were worn at the extremity or on the facial area next to the eyes.
The I
work stations selected for the survey included the Batch Makeup and Powder Preparation Station, Pellet Pressing Station, Pellet Stacking and Rod Loading Area, Rod Inspection and Fuel Rod Prestacking Stations, and the Fuel Bundle Assembly Area.
Consideration for selection of workers who participated in the survey reflected normal work station assignments as well as consistently high attendance performance of the individuals.
Individual attendance records were l
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Selection of extremity areas i
for monitoring considered the specific work station.
Extremity areas included (depending on work station) the ankles, fingers and wrists.
The facial area immediately adjacent to the eyes of two workers a: signed to the Batch Makeup and Powder Preparation Stations, and the Pellet Pressing Stations was also i
monitored throughout the survey period in order to determine the dose to the facial area including the lens of the eyes in addition to the various extremity exposures.
Based on the results of this survey, it was concluded that individuals would not exceed 25 percent of the regulatory limits for extremity exposure.
Additionally, direct measurement of exposure to the facial area of monitored individuals was within regulatory limits.
As reported in NRC Inspection Report No. 70-1100/87 01, dated May 26, 1987, the survey data was reviewed as a part of an inspection of external exposure control practices at the facility.
This inspection and review also considered the placement of whole body dosimetry worn by individuals at the Pellet Stac(ing and Loading Area.
Based on review of the survey data, the inspector indicated it was appropriate for the licensee to assume that the external exposure control program maintained personnel extremity exposure within regulatory limits.
As reported in NRC Inspection Report No. 70 1100/89 80, dated October 2, 1989, two concerns regarding the manner in which exposure to workers' extremities and skin were being monitored were identified.
In res)onse to the first concern which pertained to extremity monitoring of wor (ers involved in the stacking and loading of fuel pellets, an additional ex)osure survey was performed which confirmed results of the 1985 survey t1at extremity monitoring was not required.
The second concern involved the practice of accepting reported shallow due i
exposure values without applying a correction factor for beta energy spectrum difference in the uranium isotopes and the dosimetry vendor's calibration standard and a correction factor to account for potentially higher exposure received by the skin of the face compared to that of the TLD worn under protective clothing.
In response to this concern, a study was performed by our dosimetry vendor to determine a correction factor to account for beta energy spectrum differences.
Based on this study, a correction factor is l
being applied to exposures of shallow dose reported since December 1989 to account for beta energy differences.
l As discussed during the February 1990 inspection, the need to evaluate previous exposure history records for possible application of the beta correction factor (s) is recognized and is being addressed separately in response to this unresolved item.
With regard to the concern for potentially higher dose received by the skin of the face.and lens of the eye, Combustion Engineering believes that exposure of personnel working in the fuel facility is within regulatory limits.
The l
results of the survey performed in 19S5 included dose measurements by TLDs (which were not shielded by protective clothing nor safety glasses) worn on the facial area did not indicate exposures which exceeded regulatory limits.
We further believe, that based on our own evaluation of the 1985 survey l
results and subsequent inspection and review by the NRC in 1987 of these same
results, that the survey was reasonable under the circumstances to evaluate the extent of radiation hazards that may be present at the facility.
Until the recent initiation of redeployment activities, the operations performed today are essentially the same as those performed during the time periods when the various studies were. conducted. As such, we believe that the study results obtained remain applicable today.
Nevertheless, an additional survey to measure dose received by TLDs placed above the Fuel Pellet Stacking Table was initiated on April 13, 1990. The survey results will be used to compare dose measurements for TLDs which are unshielded by protective clothing and TLDs which are covered by protective clothing. Dose measurement data is also being obtained by this test to compare attenuation effects of protective clothing and standard safety glasses. A survey period of six months is planned for the test program with interim evaluation at the end of each month.
Should these monthly evaluations indicate that the applied beta correction factor needs to be adjusted immediate corrective action will be taken.
Based on the above information, Combustion Engineering believes th&t it has performed reasonable surveys to assure that it is in compliance with 10 CFR 20.101(a) and 10 CFR 20.201(b).
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Violation B 10CFR20.408 requires, in part, that "when an individual terminates employment with a licensee, the licensee shall furnish to the Director, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, D. C.
20555, a report of the individual's exposures to radiation and radioactive material.
Such reports shall be furnished within 30 days after the exposure of the individual has been determined by the licensee or 90 days after the e
date of termination of employment or work assignment, whichever is earlier".
In addition, 10CFR20.409(b) requires, in part,.that "when a licensee is required pursuant to 20.405 or 20.408 to report to the Commission any exposure of an individual to radiation or radioactive material, the licensee shall also notify the individual.
Such notice shall be transmitted at a time not later than the transmittal to the Commission".
Contrary to the above, as of March 1, 1990, exposure records for seven individuals who had terminated employment between November 17 and 20, 1989 had not been furnished to the D'irector or the individuals as required.
ReSD0nse Termination reports for the subject seven individuals were forwarded to the Director, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission and to the termin ded individuals by March 13, 1990.
Combustion Engineering is now in compliance with regulatory requirements for termination exposure reports.
Revision of the procedure and instructions covering dosimetry records in general, and specifically the preparation and submittal of termination reports is in progress.
The revised procedure will clarify management review requirements and, where appropriate, responsibility for approval of dosimetry records. Additionally, a method for notifying the Manager, Radiation Protection and Industrial Safety of individual employee terminations will be i
I prescribed in writing.
This should enhance our ability to file timely reports in the future. We anticipate the revised procedure will be implemented during the second quarter of 1990 l
With respect to the concern expressed by the inspector at the exit meeting regarding Combustion Engineering's ability to satisfy the requirements of 10CFR 20.408 and 20.409 for 30 individuals who terminated in December 1989, these termination reports were appropriately submitted on March 19, 1990.
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Y1olation C Section 2.13 (Note: should read Section 2.1.3) of the NRC approved License No.
SNM-1067 license application (Part 1 - Criteria) states, in part, that the Program Manager, Radiological and Industrial Safety, is responsible for defining the radiological protection program.
This )rogram addresses the 1
safety criteria and procedures necessary to ensure t1e protection of employees.
Section 2.1.5 states, in part, that the Manager, Radiological Protection and Industrial Safety, is responsible for implementing the program defined by the Program Manager and reviews and approves safety-related operating procedures.
Eection 2.6 states that Nuclear Fuel Manufacturing Facility and Product Development Laboratory operations which involve licensed materials shall be conducted in accordance with written procedures.
Section 7.1.2 of Radiological Protection Instruction RPI-205, " Dosimetry Program",
q which involves licensed materials, is a safety-related procedure and was written to implement Program Document PR 6 that requires "special" dosimeters capable of monitoring beta, gamma and neutrons be issued to Radiation Protections technicians.
Contrary to the above, as of February 27,1990, "special" dosimeters capable of monitoring beta, gamma and neutrons had not been issued to the Radiation Protection Technicians working in the Nuclear Fuel Manufacturing Facility.
Response
The Notice of Violation (70-1100/90-03) states that "special" dosimeters capable of monitoring beta, gamma, and neutron radiation had not been issued to the Radiation Protection Technicians working in the Nuclear Fuel Manufacturing facility as of February 27, 1990.
In fact, Combustion Engineering began issuing "special" dosimeters capable of measuring beta, t
gamma and neutron radiation to Radiation Protection Technicians in January 1978. Neutron ("special") dosime.try was issued to all' Radiation Protection Technicians on a monthly basis until December, 1989, a period of almost eleven years.
l' Issuance and wearing of dosimetry for measuring neutron radiation was prescribed in Radiological Protection Instruction, RPI-205, Dosimetry Program and RPI-206, Radiological Restricted and Control Areas, which also required neutron dosimetry to be worn by personnel assigned to work at the Fluoroscope l
Fuel Rod Inspection Station and Low Level Weste Assay Unit.
In December, 1989, Procedure Change Requests were submitted to revise RPI-205 and RPI-206 to delete the requirements for wearing of neutron dosimetry.
These changes were initiated following a review of personnel exposure records, which revealed that no neutron exposure was ever assigned to any individual issued i
neutron dosimetry throughout the period of nearly eleven years. Additionally, neutron surveys confirmed insignificant dose rate measurements in the immediate vicinity of the Fluorosco)ic Fuel Rod Inspection Station and the Low Level Waste Assay Unit located in t1e Pellet Shop annex.
On December 8, 1989, the Program Manager, Radiological and Industrial Safety approved the Procedure Change Request to delete Paragraph 8.6.3 of RPI-206, thus eliminating all i
reference to personnel TLD neutron dosimetry in that procedure.
The Program l.
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Manager substantiated his approval action based on the documented " Review of Whole Body and Skin Exposure Records for Nuclear fuels Manufacturing -
Windsor." During the investigation of this NRC inspection item a copy of an approved Procedure Change Request for RPI-205 could not be found, although, a copy of the request as submitted by the originator at the same time as the request for change to RPI-206 is available.
The Procedure Change Request for RPI-205 specifically recommended deletion of Paragraph 7.1.2.
Had the. Program Manager disapproved of the recommended change, a copy of his review would have-been so indicated.- O'n December 12, 1989 the Program Manager, Radiological and Industrial Safety issued document PR-6, External Exposure Control Program which sets the requirements for the subject program.
Review of this document 1
confirms that use of "special" dosimeters capable of measuring neutron 2
radiation is not required.
Radiation Protection. Instruction RPI-205 has been corrected to delete the requirements for issuing neutron dosimetry to RP Technicians (and other workers) on a routine basis.
Combustion Engineering believes that the violation is unwarranted in that we have been operating in compliance with Program requirements.
The issuance of the violation resulted from confusion caused by an administrative oversight.
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Violation D Section 3.2.3 of the NRC-approved License No. SNM-1067 license application air) sampling techniques to obtain representative samples will(be verified (Part I - Criteria) states, in part, that the adequacy of the stack exhaust quarterly in the Manufacturing Facility.
Contrary to'the above, between June, 1988 and February 27, 1990, the adequacy of the stack exhaust air sampling techniques to obtain representative samples was not verified quarterly in the Manufacturing Facility.
Renonsk Combustion Engineering believes that we were, and continue to be, in compliance with the approved license application requirement that the adequacy of the (stack exhaust air) ly in the Manufacturing facility.
sampling techniques to obtain representative samples be verified quarter The verification is actually performed by the Radiation Protection Technicians on a daily basis when the facility is in operation.
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The verification is accom)11shed by reasuring the velocity of the air in each of the exhaust ducts of tie ventilation systems by readirg a calibrated manometer which is attached to a Pitot tube installed in the ventilation system exhaust duct near the sampling probe.
The air flow into the sampling probe is adjusted by controlling sample pump flowrate, if required, to assure isokinetic sampling by comparison with measured air velocity in the duct.
In accordance with ASNI N13.1-1969, isokinetic sampling combined with proper j
probe design is considered adequate to provide a representative sample.
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adequacy of the probe design was certified by an independent engineering firm t
l (Donovan,Hammich,andErlandsenAssociatesInc.)in1987,andtheprobeis l
fixed inside the duct.
The adequacy of the sampling technique is verified by creating an isokinetic sampling condition. As such, if isokinetic sampling is l
verified, the adequacy of the sampling is verified.
Records of the verification checks of the adequacy of the ventilation sampling L
technique are maintained with the sample result records used to calculate the quantity of uranium released through the ventilation systems.
The records provide the sampling rate in liters per minute and the ventilation duct i
velocity in feet per minute for each of the ventilation exhaust systems on-a 1
daily basis.
Proper sample probe design has been confirmed as the result of an independent engineering design review of the ventilation systems.
The engineering firm of Donovan, Hammich, and Erlandsen Associates Inc. performed a design review in June 1988 of each of the ventilation exhaust systems as installed in the fuel I
manufacturing facility.
This design review included measurement of actual air flow, confirmation of the accuracy of existing manometers, and inspection of the design and installation of sample probes and conformance of sample system installation to ANSI N13.1-1969. The report of this design review recommended redesign of the sample probe in the FA-2, Furnace Hydrogen Burnoff ventilation
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exhaust system and confirmed that the design and installation of sample probes in the other three ventilation exhaust systems were correct.
A new sampling probe for the FA 2 ventilation system was designed, fabricated and installed in June 1989.
Minor corrections to actual sample system tubing runs was also i
made to achieve.conformance to ANSI N13.1-1969. A copy of this design review report is available for review at the facility.
Prior to mid-1989, the requirement was met by obtaining ventilation exhaust system " grab samples" using a portable high volume (11 cfm) air sampler.
These samples were obtained from the roof of the facility, in the open air near the discharge of each of the ventilation exhaust ducts.
Air sample results were counted for activity and compared against the activity level of normal air sampling system results obtained on the same day as the portable
" grab samples".
" Grab sampling" was performed on a quarterly basis through March 1989.
This method had a number of disadvantages including personnel safety concerns as well as the questionable validity of samples due to possible dilution resulting from sampling outside the duct work.
As a result, the system design r3 view was conducted by the engineering firm of Donovan.
Hammich, and Erlandsen Associates Inc.
This engineering firm also confirmed that the method of verifying the adequacy of the sampling technique currently employed'is preferred over the " grab sample" approach. This firm also indicated that " grab sampling" could not assure the representativeness of the sampling technique.
In the course of reviewing this NRC inspection item and to assure that the system is maintained in a condition which enables it to perform properly, the Manager, Manufacturing Engineering also reviewed normal preventive maintenance requirements for inspection of mechanical and electrical components in each of the four ventilation systems.
These inspections are being performed on a semiannual basis which is considered adequate periodicity for the service rec uirements.
The Manager, Manufacturing Engineering also recommended an adcitional preventive maintenance check to compare installed airflow manometer I
instrumentation readings with readings obtained using a portable instrument, (Kurz Model 441S Velometer) in the exhaust duct on a quarterly basis. This l
recommendation has been fully implemented.
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Based on the above information, Combustion Engineering believes that it has performed, and continues to perform, the necessary verification of the adequacy of sampling techniques as required by Section 3.2.3 of the NRC approved License No. SNM 1067 license application and that the violation j
is unwarranted.
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Violation E Condition 14 of NRC License No. SNM 1067 requires, in part:
"If permanently mounted air sampling equipment is used to determine the breathing zone air concentration levels, the licensee shall evaluate its representativeness at least once every 12 months and whenever any licensed process equipment change is made".
Contrary to the above, aermanently mounted air sampling equipment, which was used to determine breatiing zone air concentration levels, had not been evaluated between January 1,1989 and March 2,1990.
In addition, many licensed process and equipment changes have been made since January 1, 1990 when redeployment of processing equipment from the Windsor Fuel Manufacturing facility started and none of the fixed location breathing zone air sampling stations were reevaluated.
Response
Prior to July 1988, it was Combustion Engineering's practice to calculate individual worker intake based upon results obtained from the fixed position air sampler system, in July 1988, it was decided to use individual personal lapel air samplers for the determination of intake for all personnel who work with unclad uranium.
It was further decided to continue to collect data from the system of fixed position air samplers, it was felt that the fixed position air samplers could prove useful in locating the source if elevated airborne contamination levels were experienced. An average of the results i
from all fixed position air sample stations was to be used to calculate intake for individuals, who entered the pellet shop without lapel air samplers, should the general air samplers. indicate the presence of airborne contamination.
In order to determine individual stay-time, all persons not wearing lapel air samplers were required to log in and out of the pellet shop.
While the general air samplers have not indicated airborne contamination levels which warranted assignment of a calculated intake for personnel not working with unciad fuel, we have been unnecessarily assigning calculated intake to these individuals.
The assignment of a calculated intake to persons not wearing lapel air samplers without evaluating the representativeness of the fixed position air samplers has been procedurally incorrect.
Before redeployment operations started in January 1990, we performed a modified enrichment clean-up of the Pellet Shop to minimize the potential for airborne contamination. As a precautionary measure we have continued our practice of using individual lasel air samplers for all workers performing disassembly of equipment or mac11nery who could potentially be exposed to airborne contamination during the redeployment project activity.
In addition, those workers handling unclad fuel pellets in the Fuel Rod Loading Area of the Pellet Shop not undergoing redeployment are also assigned intake based on l
lapel air sanplers, 1
L As a result of the continuing clean air environment in all areas of the Pellet L
Shop, we have stopped assigning a calculated intake to persons who enter the pellet shop without lapel air samplers in accordance with 10 CFR 20.103(a)(3).
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Persons who do not wear individual lapel air sam)1ers are not permitted to handle unciad fuel nor may they perform disassem)1y of equipment and machinery.
While Combustion Engineering does not believe that a violation is warranted in this situation, because the assigned intakes generating the concern were not required to begin with, we have nevertheless ceased such intake assignments to avoid confusion in the future. Assignment of personal intake is now being conducted in compliance with regulatory requirements.
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Violation F Section 2.1.1 of the NRC-approved License No. SNM-1067 license application (Part 1 - Criteria) states, in part, that the Plant Manager, Windsor Nuclear Fuel Manufacturing has the overall responsibility for the safe operation of Combustion Engineering's nuclear fuel manufacturing facility. His or her responsibilities include ensuring adequate training of the staff.
Section 2.1.10 states that the Technicians are responsible for the day to-day monitoring of operations at the Fuel Manufacturing facility and the Product Development Laboratories.
Monitoring is accomplished through the collection of data which allows the effectiveness of radiological, criticality, and industrial safety, environmental protection and emergency planning programs to be assessed.
Technicians also monitor the sroper implementation of Radiation Work Permits.
Section 2.2.10 states that t1e Technicians shall also complete a facility specific training program (s) in safety related areas within their area (s) of cognizance.
Contrary to the above, the Radiation Protection and Industrial Safety Technician training program that was developed to ensure proper training of these technicians was terminated approximately two months after the program began, the Technicians did not complete a facility specific training program, and the Plant Manager did not ensure that the technicians were adequately trained to fulfill their responsibilities to monitor operations at the fuel manufacturing facility and to monitor the proper implementation of Radiation Work Permits.
Response
Combustion Engineering believes that training provided to Radiological Protection and Industrial Safety Technicians was, and continues to be, adequate to ensure that the technicians are capable of carrying out their responsibilities for day-to day monitoring of operations at the Fuel Manufacturing facility and Product Development Laboratories and their responsibility for monitoring proper implementation of Radiation Work Permits.
I In this regard, Radiation Protection and Industrial Safety Technicians are required to attend and successfully pass the General Employee Training (GET) course per the Nuclear Fuel Manufacturing Training Program, PR ll.
This requirement was being met prior to formalization in the Training Program document issued in December 1989.
Training is initially given to all prospective Radiation Workers shortly after the individuals are employed.
Each of the RP Technicians is up to date with his or her Radiation Worker classification in the General Employee Training (GET) program.
The Technicians' current GET status is a result of training administered in 1989 for seven of the individuals.
The other two Technicians are current because their GET was administered in 1990.
The GET program includes site-specific I
information in addition to the other topics identified in Part I, Section l
2.5.1, of SNM-1067 and in PR-ll.
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Additionally, the Radiation Protection and Industrial Safety Technicians have-also been required to participate in a program that requires reading,
understanding, and sign off of the Radiation Protection Instructions (RPIs).
The RPI sign off is based on discussions between the Radiation Protection Supervisor regarding each RPI and the level of understanding by the individual Technician.
Since early 1989 other approaches have been utilized to acquaint the Technicians with the RPIs, get their comments concerning workability of these documents when initially issued and provide evidence that each Technician had read and understood the RPI contents. While these efforts did not involve formal classroom sessions, the individual reviews did serve the necessary purpose of causing the Technicians to understand and utilize the RPIs in their day-to day tasks.
Implementation of Radiation Work Permits is performed in accordance with RPI-204, Radiation Work Permits, which was last revised in May 1989.
At the time of the NRC inspection, seven of the nine Technicians had been signed off by the Radiation Protection Supervisor as having satisfactory knowledge of RPI-204 requirements. One of the RP Technicians is newly hired and is assigned duties under the instruction of an experienced RP Technician.
The other technician, whose individual sign-off record was not complete, states that he had read RPI 204, Radiation Work Permits when it was initially issued but at the time the sign-off records were being maintained for each RPI rather than by a record for each individual RP Technician. This has since been corrected and the RP Technicians have been signed off on an individual training record basis for RPIs.
The Science and Fundamental Engineering (SAFE) Radiation Protection Technician (RP) Training Program referred to in the Notice of Violation, is a program that was established to enhance the )rofessional development of the Radiation Protection and Industrial Safety Tecinicians. Completion of the SAFE Training Program was never and is not now required prior to normal work assignment.
The contents of this program range from basic mathematics to nuclear physics l
which form the basis for radiation protection fundamentals.
Training progress with the SAFE program stopped in the fall of 1989 primarily due to management changes in the Radiation Protection organization. No specific action was taken to terminate the training. When the program was initiated, it was being monitored by the then incumbent Manager, Radiation Protection.
The program was assigned by him as a self study effort for the Radiation Protection l
Technicians, essentially on their own time.
Lack of training progress with the SAFE program was identified by the Combustion Engineering Self Assessment Task Force prior to NRC inspection performed in February 1990.
The Self Assessment Task Force also made a recommendation to improve the Radiological Protection Technician training program.
This recommendation was based on the Task Force finding that documentation of RP Technician training was lacking.
Nuclear Fuel Manufacturing has been in the process of upgrading RP Technician training.
Since the SAFE program is a professional development training effort, it is not training required of RP Technicians to perform their basic tasks (i.e.,
day to day monitoring of operations at the Nuclear Fuel Manufacturing Facility and the Product Development Laboratories).
Job specific training for day-to-day activities is provided separately as discussed above.
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' Combustion Engineering recognizes the value of the SAFE program and the i
benefits from professional enhancement of the Technicians and, as a result, i
has reconfigured and restarted the program. The SAFE training has been l
divided into modules and is administered at the rate of approximately one module per month with formal classroom work, study assignments, and examinations. Since program restart, one module has been completed.
This j
activity was initiated based on the Self Assessment Task Force findings and the recommendation related to the RP Technician Training.
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Also in response to the Self-Assessment Task Force recommendation to improve the Radiological Protection Training Program, the Training Manager has proposed a number of other improvements.
The specific recommendations of the Training Manager are presently under review. This review is scheduled for completion by June 1,1990, and implemantation is scheduled to begin l
June 30, 1990. The Training Manager also took action to revise the overall Training Program document (PR-11) to require annual General Employee refresher training for RP Tec1nicians. This revision has been fully implemented.
As discussed above, all RP Technicians were, and are, up-to-date in the General Employee Training Program.
Based on the abovo information, Combustion Engineering believes that it has provided the necessary training to the RP Technicians to assure that they are capable of performing their assigned duties.
As such, Combustion Engineering does not believe that the violation is warranted, i
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