IR 05000267/1980013
| ML20002A853 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 10/06/1980 |
| From: | Baird J, Brown G, Everett R, Randi Neff, Selby J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20002A850 | List: |
| References | |
| 50-267-80-13, NUDOCS 8011210718 | |
| Download: ML20002A853 (44) | |
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S.. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT . REGION IV IE Inspection Report No. 50-267/80-13 Docket No. 50-267 License No. DPR-34 Licensee: Public Service. Company of Colorado P. O. Box 840 Denver, Colorado' 80201 Faci of Name: Fort St. Vrain-Nuclear Generating Station Appraisal at: Fort St. Vrain Site, Platteville, Colorado Appraisal Conducted: June 16-27, 1980 Y6G l 'L4bth l8 ffY Team Members: J.~B.Baird,RadiationSpepialist,NRC /Da(e (Team Leader) / l' C% /0 J
R.'J. Everett, RadiationSpepalist,NRC Aap 2 6 M.,#Lon n aAdd J. Pf. Selby, Manager, HP Technolog Battelle Aatt f M M 'L R' D'. Neff, PhD, Health Physicis Date Approved by: M DC/4% /d C G( D. Brswn, Chief, Fuel Facilities and Date / Mater.als Safety reanch .
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Inspection ~ Summary Appraisal onLJune 16-27, 1980 (Report No. 50-267/80-13) Areas Appraised: Announced appraisal of health physics program, including organization :nd management, personnel selection, qualification and training, internal'and external exposure controls, surveys and access controls, radio-active waste,.ALARA, facilities and equipment, and emergency response capabilities..The appraisal involved 272 appraiser-hours on-site by'two NRC Radiation Specialists and two NRC contract Health Physicists.
Results: Several significant. weaknesses in the health physics program were identified. These weakness are in the. areas of personnel selection, qualifica-tion and training (Section~ 2.1), radiation protection organization (Section 1.1), internal' radiation exposure centrol (Se'etion 3.2), airborne radioactivy and L personnei contamination surveys (Section 3.3.1), and radiation protection-instrumentation and area monitors (Section 3.3.2).
One apparent item of noncompliance was found (infraction qualification of Radiation Protection Manager not in accordance with Technical Specifications - Section 1.1).
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CONTENTS Summary.
1.0 Radiation Protection Organization , 1.1 Description 1.2 Scope of Responsibilities 1.3' Staffing 1.4 Review and Audit 1.5-Communications 1.6. Conclusions 2.0 Personnel Selection, Qualification and Training 2.1 Personnel Selection and Qualification 2.2 Personnel Health Physics Training 2.2.1 General Employee Training (GET) J 2.2.2 Job-Related Health Physics Training 2.3.3 Health Physics Technician Training
2.3 Conclusions 3.0 Exposure Control 3.1 External Exposure Control 3.2 Internal Exposure Control 3.2.1 Dosimetry Program 3.2.2 Exposure Limitations . 3.2.3 ' Respiratory Protection 3.2.4 Conclusions . t ,,, ,. - -. .
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.. -4 ' 3.3' Health Physics Surveys and Access Controls 3.3.1 Scope > 3.1 ~. l.1 Procedures and Basis 3.3.1.2-Responsibility 3.3.1.3 Types 3.3.1.4 Records 3.3.1.5 Conclusions
3.3.2 Instrument Suitability and Use 3.3.2.1 Inventory 3.3.2.2 Capability-3.3.2.3 Calibration, Response Checks, and Maintenance 3.3.2.4 Records 3.3.2.5 Tr.ining . 3.2.6 Contamination Control Instrumentation
4 ! 3.3.2.7 Area and Process Monitors ! 3.3.2.8 Process Monitors and CAMS ! 3.3.2.9 Radiochemical Laboratory Analyzers' t 3.3.2.10 Conclusions 3.4 Access Controls; 4.0" Radioactive Waste Management System 4.1' Program Responsibility 4.2 Waste. Processing Systems 4.2.1 Liquid Waste-Processing System . ! J u , - - e e,- -
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'4. 2.2 Gaseous Waste Processing System 4.2.3 Solid Waste Processing and Shipment 4.3 High Efficiency Air _ Filtration Systems 4.4 Accident Sampling and Monitoring 4.5 Conclusions 5.0 ALARA Program , 6.0 Health Physics Facilities and Equipment 6.1 Facilities 6.1.1 Radiation Protection and Radiochemistry 6.1.2 Conclusions 6.2 Protective Equipment 7.0 Emergency Response /Re-Entry Annex A Exit Interview Annex B Persons Contacted c Annex C Documents Reviewed f f
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- 6 SUMMARY The'Special Health Physics Appraisal was conducted during the period June 16-27,
' 1980, to evaluate the: adequacy and effectiveness of Fort St. Vrain Nuclear Generating Station's (FSV) overall health physics program.
The appraisal team consisted of two inspectors from the NRC Region IV office and two contractor personnel provided by Battelle-Pacific Northwest Laboratories. The appraisal included observation of work practices, review of selected procedures and representative records, together with interviews with Public Service Company of Colorado (PSC) personnel. The scope of the appraisal included: A.
Radiation Proctection Organization and Management B.
Personnel Selection, Qualification and Training C.
Internal and External Exposure Controls D.
Surveys and Access Controls E.
Radioactive Waste Management F.
ALARA Program G.
Facilities and Equipment.
H.
Emergency Response Capabilities Weaknesses in the FSV health physics program were identified in several areas.
Items identified which are considered to be significant weaknesses are as follows: 1.
Lack of personnel selection, qualification and training criteria which are adequate to ensure appointments to health physics staff-and Radiation Protectf.n Manager positions meet industry. standards and regulatory: requirements.
2.
Weakness in the organizational structure relative to establishment of a Radiation Protection Manager function that is independent of Station ~ divisions whose primary responsibility is operations.
3.
Omissions in the internal exposure control program relative to establishing and evaluating airborne radioactivity areas, and developing internal dosimetery procedures for evaluation of personnel intakes of radioactive materials.
Weakness in sampling and evaluating airborne radioactivity, and monitoring for personnel contamination at the reactor building and Station exit points.
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Deficiencies in the numbers, types and calibration of portable radiation survey instrumentation. Weakness in the testing and calibration of area and process monitoring equipment.
Additional weaknesses which are considered to be less significant but important-to the implementation of a quality health physics program are identified and-discussed in the respective report areas.
In' addition to the weaknesses described above, one apparent item of non-compliance with NRC requirements was identified as follows: Selection and appointment of a Radiation Protection Manager who did not meet the applied radiation protection work experience specified in Technical Specification AC 7.1.1 and NRC Regulatory Guide 1.8.
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1.0 Radiation' Protection Organization and Management
1.1 -Description.
The Fort St. Vrain Nuclear Generating Station (FSV) organization-in place p-at the initiation of the appraisal ~is depicted by the chart in Figure 1.
'This organization:is not the same as that specified in Figure 7.1-3 of the FSV Technical ~ Specifications; however, the appraisers noted that-licensee had' filed.for am2ndment of the Station operating organization .on January. 23, 1979. During the appraisal, a different functional organization for:the radiation protection program was proposed and des-cribed-to the. appraisers. This organization, shown in Figure 2, splits ' out the technical direction of radiation' protection from the day-to-day
applied health physics responsibilities at the Station. -Both responsibili- ' ties were previously' vested in the Health Physics Supervisor (Radiation Protection Manager) cosition which was vacated earlier in the year.
In the' proposed organization, the Radiation Protection Manager (RPM) reports to the Technical Services Supervisor, who reports to the. Manager , of Nuclear Production. -The Health Physics staff is directed by the [ Health Physics Supervisor reporting directly to the Operations Manager.
Under the HP. Supervisor are a Senior Health Physics Technician-and a Radiochemist, who coordinate the activities of six HP Technicians.and two , - Radiochemistry: Laborato ry Assistan'.1, respectively.
Chemistry is separated from Health Physics under the Operations Superintendent.
There are no other; L l organizational units.onsite or offsite that have radiation protection program responsibilities.
Public Service of Colorado (PSC) does not have I a radiation protection organization or individual at the corporate level.
The FSV organization. places the Health Physics Supervisor at the same reporting level as-the Maintenance and Operations Superintendents. The Appraisal Team considers this to be appropriate although there is some concern that the FSV organization may indicate an emphasis on operational , matters'with potential. conflicts with implementation ~of radiation pro-tection programs.
No-obvious conflicts were identified by the Team dur-ing the appraisal but interviews'with-the staff indicated that?some felt there may have been some such problems in the past.
This' concern is significantly aggravated by the' location of the RPM' function under the Technical Services Supervisor. This results in a reporting line apart from,the Operations Manager, but appears to subordinate the RPM
to a unit.that is traditionally operational support oriented and makes
it. unclear as to the relationship of the RPM to Health Physics and-Operations.. This could affect the RPM's effectiveness.in previding
[ technical' direction to the radiation protection' program and securing ' ! adequate Station management support to bear on radiation protection " program weaknesses.that might be identified.
During the appraisal the: Team discussed.this matter with FSV management andl pointed out that NRC Regulatory. Guide 8.8 recommends, in Section C.1.b.c, that the Radiation Protection Manager's-safety-related_ function and responsibilities can-best-be fulfilled if the position is independent of nuclear plant s.
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divisions, such as operations, maintenance or technical support, whose prime responsibility is cotainuity or improvement of pl. ant operability. FSV management stateo cuat it was felt that the RPM's . position was adequately _ independent of operations and any questions
about-the: RPM's role could be answered by issuing a written statement of the RPM's responsibilities and authority in tre radiation protection p rogram. The Team is of the opinion that the RPM's position in the organization needs to be carefully evaluated to determine independence of-the safety-related function from the Station's production related function.
1.2 Scope of Responsibilities The scope of responsibilities for the radiation protection program is defined in FSV Administrative Policy HP-01, " Health Physics Responsibilities and Authority" and Administrative Policy HP-02, " Plant Chemistry and Chemistry procedures." These Station policy procedures establish general responsibilities for carrying out health physics related activities in monitoring work in contaminated or radiation areas, Station effluents, disposals and transfers, documenting radiation protection activities, and developing procedures to accomplish special and routine health physics functions. Also the responsibility and authority for Health .' Physics to terminate work if hazardous conditions develop by notifica-tion to the Shift Supervisor, Superintendent of Operations or Operations
Manager is established.
Policy HP-02 describes t'e Health Physics Manual status in regard to radiation control precedures, HP training and ALARA requirements. The appraisers noted that the HP Manual, issued in 1975, is generally outdated in termiaology and technical content and should be revised to reflect currer.*_ pro' gram requirements.
~ Job descriptions for functional positions within the radiation protection program were reviewed to determine if the scope of responsibility was clearly definad.
Job descriptions for the Radiochemist and Health Physics Supervisor were available but no position' descriptions for the union HP technicians or the newly appointed RPM were available.
The two descriptions examined were somewhat outdated in statements of duties and reporting position titles.
Interviews with personnel indicated that job responsibilities were fairly well understood up to the recent organizational changes.
Based on these observations, the Team recommends h ii .t at pos t on descriptions be developed to clearly define the job position reporting lines and responsibilities, and procedures be developed to periodically-update the descriptions when organizational or functional changes are made.
Special attention should be given to defining the . RPM position reporting chain, responsibilities and interface with other Station. organizations on matters of radiation protection.
, i.3 Sta f fing The health physics staff consists of one Health Physicist / Health Physics Supervisor,fone Senior HP Technician and six HP Technicians.
The total ,
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, ' Station staff is approximately 175. -The radiochemistry group has a Radio-chemist.and two Laboratory Assistants.
Chemistry is split off from health-physics under operations. Neither of these groups cross over to perform health physics functions, however the HP Technicians do collect chemistry and radiochemistry samples.
No contract HP Technicians are used during + normal non-outage operations but may be used for outages. During the last refueling outage, six contract technicians from Allied Nuclear vere used to supplement the Station staff primarily for back shift and weekend coverage. Experience during the outage indicated that probably no additional outside help would be required-during similar outages in the future. The current health physics group staffing allows 24-hour per day health physics coverage'during normal operations.
There is no back shift health physi:s group supervision on duty but only experience technicians are allowed to be on shift alone.
The Appraisal Team noted-that there is no significant specialization within the health physics group in technical areas such as internal dosimetry, respiratory pro-tection, counting procedures, etc.
Also there is no permanently assigned clerk for maintaining health physics files and Station radiation pro-a tection records.
In the opinion of the Appraisal Team, health physics staffing is regarded as adequate but it appears that the staff efficiency and performance could be improved by use of specialist in technical areas, relieving the HP Technicians from sample collection duties and assigning properly HP trained clerical help to the health physics office.
1.4 Review and Audit FSV quality assurance (QA) audit and surveillance requirements related to radiation protection are established in the Station Technical Specifica-tion AC 7.1.3(g)2 and QAP-600, " Quality Assurance Plan-Chemistry, Health Physics and Environmental Monitoring." Audits and surveillances of health physics, chemistry, radiochemistry and radioactive waste activities are performed by an onsite QA organization. QA surveillances are performed on an annual schedule and audits are on a biennial schedule.
During the appraisal the following surveillance and audit reports were reviewed: QAA-601-79-1, August 14, 1979, Chemistry and Radiochemistry , QAA-6' _-79-1, August 14, 1979, Health Physics QAA-1501-79-2, November 5, 1979, Radioactive Waste QASP-601-80-1, June 2,.980, Radiochemistry QASP-601-79-1, June 5, 1979, Health Physics QASP-601-78-2, December 25, 1978, Health Physics QASP-1501-79-1, May 7, 1979, Radioactive Waste ,.l^
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From this review and discussions with the QA staff it appeared that the scope and depth of the QA surveillance and audit program has been generally adequate in determining compliance with program requirements and also useful in evaluating performance elements for which specific requirements may not have been established. Responses to requests for corrective action and deficiency reports resulting from the above audits and surveillances also appeared to be timely and followup for resolution of _ open items was generally adequate.
In the judgement of the appraisers, the quality assurance audit program for health physics activities is adequate.
1.5 Communciations Communication within the health physics staff and other Station organiza-tions was described as adequate by most of the people interviewed. The Health Physics Supervisor tours the plant, reviews logs and has verbal contact with shift HP Technicians on a regular daily basis.
The HP staff is also made aware of procedure changes and new NRC regulations as they occur. Health physics representation is present at the morning planning meetings and plant changes are posted in the health physics.
field office so that work involving health physics considerations is identified.
Also there is health physics representation at outage plan-ning and activity planning sheets are reviewed for Radiation Work Permit (RWP), survey and health physics coverage requirements.
In summary, no apparent problems in communications either within the health physics group or with other Station organizations was identified by the Appriasal Team.
1.6 Conclusions Based on the appraisal findings in this area, FSV must take the necessary action to assure that the RPM function is sufficiently independent of . operations and has adequate authority and status to carry out the safety-related responsibilities of this function to achieve a fully acceptable program. Other elements of this area appear to be adequate but improve-ment in the program could be obtained by providing position statements which clearly define job scope and responsibilities and maintaining these statements in current status.
2.0 Personnel Selection, Qualifications and Training 2.1 Personnel Selection and Qualification , The Fort St. Vrain technical specifications commit them to ANSI N18.1 and Regulatory Guide 1.8 for selection and qualification of technicians and the RPM.
The Health Physics Supervisor and RPM both, however, said they had been told'that FSV is not committed to ANSI N18.1 but that it was used as a reference when a new employee was Seing con-sidered.
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No formal, written selection' criteria are used for new employees at any level and'there is no listing of_qualific; tion wcept for pro-motion.
All HP Technicians start out with the title of Technician whether they are ANSI N18.1 qualified or not.
The Training Depart-ment.has a statement at the end of one of their outlines that.the ' person is ANSI N18.1 qualified after completing the material:in the outline. The outline, however, is not the same one being used by the Health Physics Department for technician training.
. Of the six HP Technicians in the department one apparently did not . meet ANSI N18.1 requirements but he was very new and was still being.
closely supervised by the Lead HP Technician and the other technicians.
The designated RPM apparently did not meet Regulatory Guide 1.8 requirements for experience but the HP Supervisor did.
2.2 Personnel Health Physics Training Health Physics training for Fort St. Vrain personnel is conducted at rr.mral stages in plant training' programs.
General Employee Training, Job Related HP Training, and HP Technician Training are all given based upon each individuals need for his job functions.
2.2.1 General Employee Training (GET) All employees and contractor personnel who need unescorted access to the. plant must go through the GET.
It consists of approximately.
four hours of lectures and tapes on general plant safety, security, and health physics. An exam is given at the end of training with 70% or greater to get the appropriate security badge for unescorted access.
The health physics part of the training. consists of three tapes done by NUS but at FSV and the tapes contain plant specific' material. The
instructor is from the FSV Training Department and he also gives more plant specific details between tapes.
The instructor also asks questions during training about material just covered on a tape.
The tapes are fairly well done but they do contain some questionable tae of terminology such as using mR and mrem interchangeably as well as exposure and dose. Emergency notification instructions were confusing and no phone numbers were given.
During discussion periods between tapes several questions were asked concerning health physics that the instructor could not answer.
Instructions were given on using and reading pocket ionization chambers > as well as FWP forms but no demonstration were conducted and no practice was given in either.
It appears that some classroom demon-stration and practice would be useful.
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l 2.2.2 Job-Related Health Physics Training .Each new permanent employee at FSV-gets eight hours of Job-Related Health Physics 1 Training, except-for clerical positions. This train-ing is directed toward the' areas of work for each individual and acts as an_ extension of the health physics training received in the GET.
This training is scheduled by-the Training Department but is con-ducted by a member of the Health Physics Department, usually the Lead HP. Technician. He teaches from an outline previously prepared by-Health Physics and kept on file in the Training Department.
In addition to the initial Job-Related HP Training each employee gets a two hour retraining course every two years.
Completion of the Job-Related HP Training makes an employee "H.
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Qualified" which means he can work.in radiation areas as a radiation worker, use protective clothing, use a portable survey meter, and act as an escort for visitors in the vital areas.
The Training Department has a " Training Programs Administrative Manual" which spells out the training required for all personnel, the material to be covered, and-the schedule for retraining. - Licensed operators i receive the Job-Related HP Training but are not considered to be qualified in-radiation monitoring.
2.2.3 Health Physics Technician Training
-Fort St. Vrain has a " Health Physics Technician Training Check-Off List" which is used to record training given new technicians and qualification in r,pecific job function.e.
The list is comprehensive and is reviewed and approved by the Health Physics Supervisor. A review of individual records however revealed that most of the items tended to be checked olf'all on the same day rather than a step by step progression through the list. There also is no coordination of the training qualification list with ANSI N18.1 requirements.
Also, all examinations are oral.
The Health Physics Supervisor stated that they are developing a retraining program for-HP Technicians that'will be in twelve modules given one each month. The_ modules and procedures for use are not yet available.
Training of HP Technicians is restricted to health physics instru- . mentation and procedures. There is no instruction in plant systems ~similar to' that' given licensed operators.
As'a result the HP Techni-cians are not thoroughly familiar with all plant systems.
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2.3 Conclusions Based on the appraisal findings in this area, personnel selection d qualification criteria must be established as necessary to ensure hat appointments to health physics positions meet industry standards and comply with NRC regulatory requirements to achieve a fully acceptable program.
Training is considered to be adequate but could be improved by more practical demonstrations in classes and training check lists should be coordinated with the qualification criteria referred to above.
3.0 Exposure Control 3.1 External Exposure Control External personnel radiation exposures at Fort St. Vrain are very low.
FSV uses a commercial dosimetry service on a monthly exchange frequency.
Beta gamma and neutron films are worn by all personnel who enter the reactor building while clerical people wear beta gamma only.
TLD finger rings are available and worn during appropriate activities such as portable survey meter calibration. Review of exposure records revealed that in the past two years there were two reported exposures between 100 and 250 Further investigation indicated that one of those was actually a mrem.
skin dose listed in the annual report as whole body.
The large majority of personnel exposures are reported as minimal.
FSV has an in-house computer sys';m for maintaining personnel exposure histories.
It is, however, not se'. up to yield the information normally obtained from computer exposure records such as quarterly, year to date, and lifetime dose.
It does serve to' generate the monthly reports on terminations and visitor doses.
The quarterly, year to date, and life-time doses are contained in the monthly Landauer dosimetry report and that file is kept in the Health Physics Supervisor's Office.
The HP Supervisor generates the annual personnel exposure report required by NRC requirements.
Individual personnel exposure files contain internal exposure records but not external. They also do not contain records of actions taken such as removal of reported dose for any reason.
An example of this is that for badge period 12-20-78 to 1-19-79 there were many reported exposures ranging from 20 mrem up to 190 mrem.
It was evidently established that the films had been x-rayed prior to use in the Security System X-ray.
A letter to this effect is in the file of Landauer reports but nothing appears in the in-house computer reports or in individual files.
Spiking of beta-gamma badges is done approximately every two months with the cesium-137 calibration source. There is no written procedure for this or any acceptance criteria.
There was some indication that + 10% was intended but no action taken when most fell outside of that range.
There has been no spiking of neutron film and no verification that film
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i;' -is the appropriate dosimeter for the n'eutrons encountered at FSV.
There is a testing program scheduled in the future to see if film is appropriate.
[ 'It was not possible to compare pocket chamber with film badge data since there were very few' readings on either.
. Pocket ionization chambers are used at FSV as backup for film badges of persons. entering the reactor building and for dosimetry of visitors. The i chambers are not well maintained and calibrated.
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There is no initial testing and acceptance for new purchases of chambers.
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The chambers are used with no protection cap on the electrode l.
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The chambers are calibrated only every six months and the acceptance criteria used are not in accordance with ANSI N322(1977).
i Calibration is done by exposing the chambers to 50% of full scale gamma dose and accepting all those which read 1 24%. If a chamber reads outside of this range it is retested.
If it falls inside.the 1 245 on the second test it is accepted.
ANSI N322 specifies 80%, 50%, and 20% of full scale' exposures should read . 91 thin + 10%. If FSV used these values it is estimated that l greater than 50% of their approximately 700 chambers on hand - 3, would fail.
! From a review of procedures and results it appears that the poor results may be from the calibration itself rather than from bad chambers.
Emergency personnel dosimetry at FSV is very questionable.
Film badges
would have to be mailed to Landauer for reading.
The only pocket chambers ' on hand are 0-200mR and a'few 0-600R.
There is an in-house TLD reader but it has not been set up for personnel dosimetry.
I In conclusions, the appraisal observations in this area indicate that
the licensee's program is acceptable in relationship to the exposure potential at the present time but could be improved as follows: .l 1.
Neutron energy specta in areas of potential personnel exposure should be evaluated for purposes of selecting personnel dosimetry devices.
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Procedures and criteria ~for personnel-dosimetry devices quality control should be developed and implemented.
3.
-The pocket chamber' program should be upgraded by improvement > ~in maintenance and calibration practices.
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Documentation of external exposures should be improved so that personnel files indicate any adjustments of exposure that may be made.
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Provisions for personnel dosimetry in the event of an accident should be examined and appropriate action taken.
3.2 Internal Exposure Control 3.2.1 Dosimetry Program health Physics Procedure HPP-2, " Bioassay Procedure" establishes the FSV bioassay program.
Onsite bioassay capabilities are limited to collection and analysis of urine samples from selected groups of employees.
Urinalysis is required when starting and terminating employment at the Station, and at regular intervals for most of the Station workers.
Currently ten groups of workers are in the program on a six months frequency.
Urinalysis for assessment of known or suspected acute intakes are performed as directed by Health Physics.
Radiochemistry performs the analyses with a Beckman LS-100 liquid scintillation spectrometer. Quenched standards are run to obtaia calibration curves for the raw urine samples and the calibration curve and external standard count is used in a computer program to determine tritium concentrations.
A screening level of 3 x 10 uCi/mi is used, below which no addition evaluation is done. Above this value the sample is recounted to confirm the results and additional evaluations may be performed. The relationship of this or other results to the intake limits of 10 CFR 20.103 is not documented in procedures and there have not been established any biological model or calculational procedures to evaluate the results in terms of dose or intakes.
In reviewing this area, an appraiser also noted that the samples are only counted for tritium and the presence of other beta emitting nuclides which might not show up in a whole body count is not determined. This appears to be easy to add to the counting procedure and the Team recommends this improvement be added to the program.
The licensee's procedures also establish the frequency of whole body counting for Station personnel.
Permanent employees receive an initial and a termination count, and thereafter an annual count.
The whole body counting is routinely performed in Denver at the laboratory of the Colorado State Department of Health.
Special counts may be performed at the Colorado State University's counter in Fort Collins.
Results of the whole body counting is filed and maintained in the individuals personnel dosimetry files.
As noted above, there are no internal dosimetry procedures which relate whole body counting results to intake limits or establish calcula-tional techniques as recommended in ANSI N343-1978, " Internal Dosimetry for Mixed Fission and Activation Products." This is considered to be
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a significant weakness in the licensee's program.
Another concern is the lack of a whole body counter at the FSV site, this appears to the Appraisal Team to be placing too much reliance on outside agencies for proper maintenance and availability of a whole body counter for emergency as well as routine operations.
Radiochemistry conducts an inhouse calibration and quality control program for the liquid scintillation counter, and participates in the EPA cross-check program.
During a visit to the whole body counter facilities, an appraiser determined that there is no FSV initiated quality control program for the whole body counter services supplied by the Colorado Department of Health and no auditing or other measures taken to assure the quality of results of whole body count-ing. This appears to be a weakness in FSV's internal exposure control program and appropriate steps should be taken to assure that whole body counting results quality is documented and adequate.
3.2.2 Exposure Limitation Administrative controls for airborne radioactivity are established in procedure HPP-9, " Establishing and Posting Controlled Areas," and pro-cedures for calculating maximum permissible concentration (MPC) hours personnel may be exposed to is contained in HPP-16, " Full Face Respira-tors, Self-Contained Breathing Apparatus and Half Mask Respirator," Thegg procedures define an unidentified radionuclide in air MPC of 3 X
uCi/mi based on 40 hours of exposure in any weekly period.
In reviewing these procedures, it was noted that there are no provisions for implementing the controls in radioactivity areas as contained in 10 CFR 20.103(b) and designating airborne radioactivity areas as required in 10 CFR 20.2C3(d)(1)(ii).
In addition, the Health Physics procedures do not relate tne exposure controls to the 40 MPC-hour control measure of 10 CFR 20.103(b)(2), and no provisions for evaluating exposure which might exceed this control measure and assuring against recur-rence have been established.
These findings show that the licensee has not established procedures to fully implement the requirements of 10 CFR 20.103.
3.2.3 Respiratory Protection The licensee's respiratory protection program is established by the following Station procedures: HPP-16, Full Face Respirator, SCBA and Half Mask Respirator HPP-48, Routine Maintenance and Inspection of Respiratory Equipment HPP-49, Respirator Facepiece Fitting HPP-30 Breathing Air System Sampling Procedure
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- , r ADM Policy HP-06 Administrative Policy for Use of Respiratory
Protective Devices i.
' A formally established program is in place and-the licensee pro-l . cedures.give credit for the protection factors contained in j~ Regulatory Guide 8.15.
Respiratory protective equipment for [ routine use is maintained under the control of health physics.
and is stored and issued from cabinets located on the turbine ' deck, Respirators.for emergency use are located in emergency.
, j kits and six other locations, in addition to airl!ne respirators
- in the control-room. Maintenance, _ inspection, cl eaning and disinfection of respiratory equipment is performed by a Maintenance
Technician assigned to the health physics area. -Inspection and maintensace records are maintained on cards for each device.
4-A review of these records by an appraiser indicated _that this ' element'of the program is well done'. FSV does not have an area set aside.for cleaning, disinfecting and drying respirators.
Cleaning and disinfection is' performed in a " fan room" which con- .
tains a sink and also serves as a storage and utility. room.
- Drying is accomplished by hanging the respirators up on lines in the HP instrument maintenance room. The Team considers this to be inadequate facilities to support a respirator program although it is recognized that usage is low during normal opera- , i tions. Decontamination of contaminated equipment is done in [ the decontamination facilities in the health physics area, Breathing air is supplied by two special oilless air' compressors i provided with RAF air purifying systems which remove moisture, contaminants and carbon monoxide from the air. The compressors ( - are located outside the reactor / turbine buildings in an area of uncontaminated atmosphere. Five 2400 PSI storage-bottles and SCBA tank filling equipment are located with the compressors.
There is routine surveillance of the air compressor and a quarterly surveillance ~of the air. quality to assure that breathing air meets at~least grade D standards.
No problems related to
breathing air were identified.
' Initial and annual respirator protection training is by use of film. video tapes and hands-on demonstrations.- Medical evaluation to determine capacity to wear respiratory protective equipment - is done as part of.the initial employee medical physical examina-tion and no annual' review of medical status is performed to determine if the employee is still medically fit to use respira-tors.
The appraisers pointed out-that this is contrary to Regulatory. Guide 8.15, which requires an annual review of the medical status of the user.
Initial fit testing:of respirators i-is by use of banana oil or irritant smoke challenge atmosphere . m m W We P G W r TV -? --vwT - t-- - F- -M - T- - '*DT
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19 is an qualitative test.
Fitting records for individuals are maintained on a " Respirator Fit Test" record.
No facilities for conducting quantitative testing as described in NUREG-0041, " Manual of Respiratory Protection Against Airborne Radioactive Materials," are available.
Field testing of respirator fit prior to use is by means of negative or positive pressure testing.
The appraisers recommend that FSV obtain and use equipment to conduct quantitative initial fitting of respiratory protective equipment.
An appraiser reviewed the supply and inventory of respiratory protection equipment and checked the equipment for NIOSH certifi-cation.
All of the non-expendable equipment is maintained on an inventory system.
Stocking of filter cart.idges for air purifying devices is on an informal basis.
Ali of the respirator protective equipment in use was noted to be NIOS. certified.
The supply of various types of equipment was noted to be rela-tively small but probably adequate for normal operations.
Ade-quacy of the supply for accident conditions is questionable and FSV should re-evaluate their needs in this area.
3.2.4 Conclusions Based on the findings in this area, internal dosimetry procedures for evaluating intakes of radioactive materials need to be develop-ed and procedures for establiehing and evaluating airborne radio-activity areas require improvement to achieve a fully acceptable program.
In addition, the following should be considered for improvement of the internal exposure control program: 1.
Providing quantitative fit equipment for evaluating and documenting the initial fit of personnel respiratory pro-tective equipment.
2.
Expanding the urinalysis counting procedure to screen bioassay specimens for beta emitters which may be present in addition to tritium.
3.
Providing an onsite whole body counting capability and establishing a quality assurance / quality control program for whole body counting.
. 4. '. Performing annual medical reviews of personnel medical fitness to wear respiratory p,tection equipment.
- I ' 5.
Providing facilities for clea Ting, disinfecting and drying of respiratory protection equfhment.
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6.
Re-svaluating the supplies of respirators in terms of support-ing anticipated and accident occurrences.
3.3 Health Physics Surveys and Access Controls 3.3.1 Scope 3.3.1.1 Procedures and Basis Fort St. Vrain's routine surveillance program is documented in the plant health physics procedures.
Survey activities are under the direction of the health physics supervisor.
Routine beta,- gamma, and neutron surveys are conducted in the plant daily, weekly and monthly as specified in Health Physics Pro-cedure HPP-1.
An appraisar reviewed all health physics proce-dures pertaining to the surveillance program and these procedures are listed in Annex C.
Procedures and records pertaining to
the radiation work permit program were also reviewed.
Independent surveys were made by the appraisers to verify reported levels.
The routine survey program is conducted primarily on the night shifts or the health physics staff using approved plant pro-cedures.
Special surveys and surveys taken prior tc issuance of a radiation permit are taken as directed by the health physics supervisor or 'the lead technician.
Procedures covering issuance'of a radiation work permit and health physics follow-up with changing work conditions were considered adequate.
In addition to the routine direct and indirect surveys for surface contamination on floors, equipment and tools, the licensee uses portal monitors and friskers to measure contamination on personnel leaving contaminated areas. Acceptable levels of contamination on personnel leaving these areas has not been documented in plant procedures.
Requirements that personnel use these devices are ' vague. Procedures require that friskers be made available at each level of the reactor building where contamination exists. This procedure is.not being carried out in all areas, instead the licensee supplies friskers and portal monitors at the level 5 and 7 exits to the reactor building. A third portal monitor is at the security house which is not in service and not required to be used.
The appraiser'also reviewed procedures and practices pertaining to personnel access to controlled areas.
The controlled areas , ' reviewed were: Radiation areas, high radiation areas. airborne radioactivity areas and radioactive material areas.
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21 3.3.1.2' Responsibility The routine survey program is conducted by the health physics staff and no outside organizations or consultants are utilized.
, The routine program is reviewed periodically by the health physics
supervisor and changes are made base upon recommendations of i thr. health physics ~ staff.
Routine and special survey data is reviewed by the lead technician and by the health physics ' supervisor.when levels approach administrative limits.
3.3.1.3 Types FSV measure airborne activity by' using high volume samplers and filter. media, continuous air monitors and tritium air monitors , (triton). Low volume se.mplets or lapel samplers for representa-tive sampling are.not utilized.
The high volume samplers are , used with a. type SG (H&V) filter, which consists of cotton, j glass and synthetic fibers. The particulate efficiency is not known, however, DOP. penetration through five filters is report- ,
ed by the manufacturer to be.1 .3%. All air samples are
- aken for ten minutes, giving a total volume of about 100 cubic
. , l Jeet. Filter papers are counted without delay.using a gas flow porportional counter in the health physics office.
Cgggter ggficienciesaredeterminedbycountingstandardsof Th and Tc prepared on a metal substrate. Health physics procedure HPP-12. specifies that all air samples be taken to radiochemistry for analysis.
Since the counting procedure, HPP-13, does not allow for the decay of natural. activity, the concentrations re-i porte? and evaluated are primarily natural activity.
After subtJaction of a background concentration (natural activity), the alpha and beta concentrations are added together to obtain a composite concentrat mixtureMPCof3X10{gnwhichiscomparedtotheunidentified - , uci/ml.
Gamma isotopi "*1 8i I filter papers revealed natural activities plus E8*b and i30
R Cs particulate activity.
Since the half-lives of these nuclides are 17.8 and 32.8 minutes, respectively, a ten minute sampling time is long compared to the half-lives and corrections need to be made.
' Noble gas activities are not measured directly in the reactor building. One continuous air monitor on the refueling floor and anotherinthebasementofthereactorbuildingmoggtorpargjgulate daughters of at.least two noble gas parents i.e.
' * * Thepredominantnoblegasintheprimarycoolantis{53" Xe, which decays to a stable daughter.
._ _ .. _.
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The collecti. iand measurement of iodine aerosols in various work areas is not rotanely performed.
The licensee can now use a high volume sampler with a charcoal cartridge to collect iodine aeroso)-
- would experience low efficiency due to high flow rates -arough the cartridge. A FSV representative stated that low volume samplers with charcoal cartridge adapters are on order.
A summary of the deficiencies noted in the air sampling program are as follows: 1) The minimum detectable concentration for alpha activity in air is not vifficient to detect concentrations at the applicable 10 CFR 20 limit.
2) The filter media used would result in high alpha absorptien when counted directly.
3) The counting procedure does not allow for decay of natural activity in the presence of fast decaying plant related activities 4) ine alpha and beta standards are not of the same matrix as the samples being counted, which would result in considerable error. 5) The evaluation of counting data is improper in that counts from the alpha and alpha plus beta plateau are added arrive at a mixture concentration.
6) No corrections are made for losses during sample collection.
7) Periodic measurements of iodine atmospheres are not made using proper equipment.
8) Noble gas concentrations are not measured directly at several locations within the reactor building.
It should be added that the presence of alpha contamination has not been i;.icated on swipe samples taken throughout the plant and the concentrations of airborne cor.taminants are believed to be quite low, except for an occasional small release in the reactor building. Nevertheless, an adequate air sampling program needs to be developed to demonstrate expected low levels and to evaluate occasional releases in the reactor building.
The appraiser reviewed neutron survey data and noted levels of a few millirem per hour measured at two penetrations.
Measurements were made by using a Eberline PNR-4 portable neutron survey meter with moderating sphere. Additional neutron shielding has been added at these penetrations. The neutron energy spectrum is not known therefore the adequacy of the NTA film personnel dosimeter cannot be determined.
The licensee plans to contract the measure-ment of the neutron spectra in the near future, although survey result indicate that neutron dosimetry is not required.
The licensee occasionally encounters relatively high beta dose rates and user the Eberline Pic-6A and R0-2 portable survey instrumente for surveys.
The Pic-6A is clearly not the instru-ment of choice since it was designed only to detect energetic betas. Sulfur-35 and other low energy emitters have been identi-fied in the plant.
The R0-2 does not appear to have sufficient range to cover the expected beta dose rates.
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As previously mentioned, friskers and portal monitors are placed ' at several locations for use by plant personnel exiting con-trolled areas. The suitability of the portal monitors for detecting beta / gamma levels of interest appears to be a general problem. The portal monitors are so constructed that beta detection appears unlikely. The alara point after a three ggcond count is ;.'t 14 sed _upon response to a ten microcurie Co source. The alarm point on the friskers (pancake GM detector) is set two-three times background.
This would cor-
respond to about 1000 dpm/100 cm at 10% deficiency if the
survey was properly performed.
It appears that sensitive hand and foot monitors will be necessary to detect beta / gamma con-tamination levels of interest.
- 3.3.1.4 Records The licensee uses a series of prepared survey maps to record survey data.
These maps become the permanent survey record.
A blank survey map is available for special areas or equipment surveys. Air sample data is recorded on a special form described in procedure HPP-12.
Survey maps are posted at the level 7 entrance to the reactor building and a copy filed for permanent record. The appraiser-noted no problems associated with survey documentation. The radiation work permit program and documentation is described in administrative procedures.
FSV has used the RWP sparingly since they do not have the radiation and con-tamination problems normally associated with a light water nuclear power plant.
3.3.1.5 Conclusions Based upon the above findings, improvements in the following areas are required ta have an acceptable program: 1.
The measurement and evaluation of airborne radioactivity needs considerable improvement in the form of adequate procedures, proper equipment and materials to collect and count air samples.
2.
Direct neble gas monitoring of several areas in the reactor building is indicated.
3.
Use of proper equipment to collect and measure iodine aerosols in several_ areas of the plant.
Other areas of the survey program appear acceptable, but the - following areas need to be improved: . , vr--
.
s 1.
Adequate portable instrumentation should be available to measure beta dose rates up to at least 50 rads per hour.
2.
The neutron energy spectrum should be determined and the < neutron dosimetry program evaluated as previously stated in Section 3.1.
3.
Improvements are needed in plant procedures and monitoring equipment in order to measure contamination levels of interest on personnel leaving controlled areas.
3.3.2 Instrument Suitability and Use Instrumentation (portable, semi-fixed and fixed), associated with the health physics program was discussed in depth with members of the Health Physics and Results staffs.
The appraisal involved a review of procedures, training, records and observations of calibra-tion techniques.
3.3.2.1 Inventory The inventory of health physics instrumentation was reviewed, together with the maintenance and availability experience. The - licensee has a reasonable selection of specified instruments, such as Ludlum Micro-R meters, Tritons and Victoreen R meters.
However, inventory of the main measurement tool for the Health Physics Technicians, the ionization chamber dose rate instrument, appears to be totally inadequate.
Currently, there are four F ,-6 and four Ro-2 dose rate instruments in the inventory.
7.o of the Pic-6's are assigned to emergency kits and three of the Ro-2's are out of service.
It is questionable whether this is an adequate number for routine operational purposes.
During an emergency, a sizeable number of instruments could be used.
Currently, the licensee owns four neutron instruments. However, due to quarterly calibration at an offsite vendor, the number available at any given time may be only two.
They currently own fcur E-400 and three E-500 GM radiation measurement instruments.
These are calibrated in units of mR/hr; however, the health physics procedures correctly indicace that these should not be used as dose rate instruments. The inventory of semi-fixed contamination control instruments includes four RM-15's and 13 RM-14's, together with 16 operational HP-210 probes for these instruments.
This nay be an inadequate number.
During the appraisal, one of the RM-14's was removed from the analytical instrumentation laboratory to be used in a survey program.
Part of the problem is that four of these instruments are currently dedicated to a special operational program in the
a l ' .
< basement of the reactor building where they are used to meascre buildup in lines as opposed.to contamination control.
The licensee currently owns three portal monitors and no hand and shoe counters. The air sample air monitors include two NMC air me-; tors (one out of service), one Eberline beta particulate ' monitor, an Eberline PING monitor, three Tritons (one out of service), five Staplexes (one out of service and two assigned to emergency kits).
It would appear that coupled with calibra-tion problems identified later and the existence of out of
service monitors and samplers that the station has inadequate ) capability to provide for special air sampling-air monitoring ! at work locations where airborne contamination could develop.
{ 3.3.2.2 Capability The capability of the health physics instruments is marginal to inadequate,particularlywhenlookingattheinstrumen5atinneeds after a serious accident.
The Pic-6 range is up to 10 R/hr.
They have no instruments that will achieve 10 R/hr as suggested in the lessons learned or ANSI N320, " Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation."
The Pic-6 does not provide beta capabilities, thus, the station is limited in beta capabilities to 5 rad /hr which is provided by the Ro-2.
Currently, they have no operational alpha survey instruments, although the station does own an instrument.
The performance capability of all the health physics instruments has been accepted without question. There is no acceptance testing on the part of the staff to determine whether the instru-ments meet applicable ANSI standards, including N13.10.
The instruments are not tested to determine whether they meet the vendor's specifications.
The GM and HP-210 probes are accepted without testing of the sensitivity of operating characteristics.
Thus, the measurements may vary from instrument to instrument.
3.3.2.3 Calibration Response Checks and Maintenance The calibration and maintenance progenm is well established by procedure. Apparently the Station does not experience any difficulty in the maintenance of 'tealth phsyics instruments with the exception of a detector problem that exists currently with the R0-2's (75% are_out of service).
Serious problems exist with the calibration of all health physics instrumentation.
The entire calibration proceaures and~ calibration equipment should be reviewed careful'.y with the criteria contained in [ ANSI N323, " Radiation Pre'.ection Instrumentation Test and Cali-J bration," to identify tra basic changes that are necessary in ' 'the program.
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Neutron calibration lis' performed by an offsite vendor.
The ' station does own a neutron source; however, it is not routinely used for calibration because of source size and lack of trace- - ability to_NBS.. According _ to information from the health physics . staff, the.offsite. vendor calibrates neutron instruments-only
at one point rather than two on each scale as suggested in ANSI N323. The vendor's sources were certified in 1962.
However, ' there is no information available that suggests that the Pu neutron source has been corrected for in-growth.
In an article from Mound Laboratory (the major supplier of Pu neutron sources), it was suggest thereis0.7%ggythat the in-growth may be as high as 2% if Pu present in the isotopic mix when the source is encapsulated.
A similar Pu source certified in 1962 had an overall in growth of approximately 20% through June of 1980; thus, it can.be seen that the' calibration of the neutron instru-ments would be significantly in error.
The photon calibration exhibits inadequacies in several areas.
The size of the large photon source results in a maximum calibra-tion rate of approximately 40 R/gr.
It should be noted that the Pic-6 has a range of 7 to 10 R/hr.
Thus, the Pic-6 should be limited to use only to those ranges for which it has been properly calibrated. Both the Pic-6 and the Ro-2 dose rate instruments are calibrated at only four points rather than two on each range as suggested in ANSI N323.
The calibration jigs for the instruments-and the position of the two photon sources permit as many as four variables to occur each time a calibra- , tion is performed on an instrument.
The calibration of Station instruments was compared to calibration of an instrument supplied by the appraisers. A considerable variation was noted between readings at a given point ~for these instruments.
The variation was not consistent either in magnitude or direction; thus, it would appear.that calibration jigs should be carefully developed that would permit _ accurate repetitive calibrations to be performed.
A better method of positioning the source containers should be developed to assure that.the source has the same configuration each time for the entire calibration range to permit accurate calculation of the dose rateu.
It should be noted th;t during the calibration performed for the appraisers the R chambec reading did not fall within the permitted range.
Currently no beta calibration is performed on operating instru-ments.
It was reported during the appraisal that one of the HP Technicians experienced high beta fields during an unusual - occurrence. -This would suggest that even thcugh there is no beta dose rate during normal operating conditions that personnel .-
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may be required to survey relatively high beta fields during an emergency.
Daily operational response checks cannot be performed as suggested in ANSI h323. One check source is available in the instrument storage room.
However, it does not satisfy the check source re-quirement which includes sufficient range to check each scale and a rejection criteria to indicate whether the instrument is operating properly. There are no check sources currently available within the reactor building.
Friskers are used at major work locations, such as the refueling floor; however, no check sources are pro-vided. At least one dose rate instrument is left on the refuel-ing floor. There is no check source for this instrument.
3.3.2.4 Records The radiation protection staff maintain a good set of records which include a summary cardex file which indicates the current calibration and operability status and a file of individual cali-bration sheets.
It would appear that the need for maintaining complete records should be reemphasized to the HP Technicians and Results staff.
For the March calibration of the HP-210 probes, 10 of 16 were not recorded in the cardex file.
It was noted that flow calibration date was not recorded on the two Staplexes included in the emergency kits.
In each case it was possible to verify that the calibration had been performed.
3.3.2.5 Training In the event of a severe emergency or even some of the unusual occurrences, non-uniform fields and beta fields may be experienc-ed by the HP Technicians. They currently have no experience dealing with high contamination radiation levels and training does not include realistic testing of their capabilities to handle these situations. Training does not include any information on beta interpretation nor on calculations necessary to correct nonuniform field readings.
It should be noted that errors as high as a factor of 100 can occur.
3.3.2.6 Contamination Control Instrumentation Contamination control instrumentation at the Station is limited principally to friskers and portal monitors.
It was concluded that the contamination control program using these instruments is weakened due to inadequate calibration.
The portal monitors attT37 entries into the reactor building were tested with a 0.5 uCi Cs source held directly on the surface of each probe.
None of the probes would respond.
The current procedure involves l l
. .
an electronic calibration coupled with a response check using a 10 mci cobalt source. This would suggest assuming a 5% geometry on.the opergtional test that these instruments only have ta alarm at 10 counts per minute.
There is no rejection criteria contained.in the procedures. A test of the portal monitor located in the security check point revealed 8 of 14 detectors did not work.
It was al-- determined that for that instrument the alarm circuitry can be 'eated in the case of contamination on an individual if he..
- out of the monitor as soon as it alarms but prior to completion of the counting period. Under these-circumstances, the. portal monitor will change from a red light to a green light and the alarm will stor, indicating a satisfactory count.
The. friskers were checked at several locations throughout the facility.
It appears that the response of these are fairly uniform. However, thcy are, as noted previously, not provided with sources to be used for a source check.
The HP-210 probes, coupled with the instrument are not given a proper set of cali-brations as suggested in ANSI N323 and the HP-210 probes are not provided a performance. test upon purchase, thus leading to potential non-uniformity in sensitivity and in operating enar-acteristics.
Current station procedures do not require the use of friskers for survey use in the reactor building even though they provide a great deal more sensitivity than the portal monitors. The station does not employ the use of hand and shoe counters in place of portal monitors which could provide similar or improved sensitivity over the friskers.
Area and equipment monitors are provided with a weekly check with an internal source to compare tracking.
Electronic calibra-tion is done every six months along with a source check that looks at one point on the low scale and one point on-the high scale , under a fixed geometry situation. However, it should be noted ! that this curren+ calibration procedure is in conflict with both the initial and annual calibration requirements of HPP-39.
Item 14 of-SR 5.4.9-A3 (Surveillance Testing of Area and Equirment Monitors) is in conflict with the requirements stated ir JPP-39.
The current method does not provide for a linearity test nor a sensitivity test since only the low and high ends are checked.
The GH tubes used in the area and equipment monitors are not pro-vided a performance acceptance test.
It was reported by Results ' personnel that if an area monitor currently is not calibrated that the detector could be replaced without going through even the overall calibration requirements of 3R 5.4.9-AE.
The complete calibration would be done at the regular interval.
This'is in violation of HPP-39.
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3.3.2.8 Process Monitors and Cams Generally, process monitors and CAMS receive an electronic , calibration coupled with an response check using a transfer radiation standard. -Thus, tests for sensitivity response , checks to various energies (different isotopes) and linearity of the detector are not performed.
One notable exception to this practice is the excellent calibration that has been developed for the process gas monitors.
Tbis c 7Coration involves the introduction of a known amount or standard radioactive gas _and ~ subsequent dilution to provide a check of the various ranges.
This procedure, which has only been recently adopted, resulted
in a determination that the sensitivity of the instruments had changed by 35%. As a result of one of the quarterly calibratious, an error in the electronic calibration was discovered.
There are two problems that exist with the process monitors and CAMS: 1) routine flow :.iibration is not performed, and 2) the instruments have not been checked for inleakage.
Inleakage in CAMS and other selected air monitors has been fo"nd to be as high as 100%. Very few of these instruments exhibit no inleakage when tested. Obviously, with inleakage the air concentration results are in question and where the sample is removed for further laboratory analysis the results continue to be in question.
The liquid waste process monitors are given a yearly electronic calibration and source check.
A calibration procedure is being i developed by the Radiochemical group to provide to provide a complete check of the' range and sensitivity of the liquid monitors.
. The Triton air monitors are well calibrated with known tritium gas aliquots.
The two new Eberline monitors (PING and particulate) are not cali-brated by the station. They are relying on the calibration provided by the vendor.
It would appear to be unwise to accept this calibraticn without a proper performance test by the health physics personnel.
The gas chromatography is provided with a weekly and monthly check with a single standard gas.
In 1974, a linearity check was performed with several gases.
However, this has not been performed since. This practice would appear to be inadequate.
3.3.2.9 Radiochemical Laboratory Analyzers The radiochemical laboratory analyzers have been placed on an ex-cellent background and calibration program. Control charts are
a
maintained for the analyzers along with computerized analysis of the performance in some instances.
3.3.2.10 Conclusions Based on the appraisal findings in this area, improvements in the numbers and/or calibration and operational check procedures for portable and semi-fixed radiation protection instrumentation is required to achieve an acceptable program.
Also with the exception of noble gas monitors, the calibration and testing of area and process monitors could be improved.
3.4 Access Controls FSV defines four types of controlled areas at the plant.
These are: radiation area, high radiation area, contaminated area and radioactive material area. Radiation areas are considered those areas where radia-tion levels exceed 2.5 mr per hour.
Procedores require these areas to be conspicuously posted and the area established with rope and stanchions.
Plant procedures require that high radiation area be posted and equipped with a control device or locked when not in use.
Ia addition, entry into high radiat.ies areas requires authorization by the shift supervisor and the health physics department and the use of a pcrtable radiation monitor-ing instrument.
Contaminatgd areas are defined as those with contamination in excess of 100 dpm/100 cm (B,G).
Controls consist of postings, the use of stanchions and rope to establish a perimeter and a piece of absorbent paper taped to the floor as a step-off pad.
Procedures require friskers at these contaminated areas.
In practice, friskers are placed at levels 5 and 7 access points to the reactor building.
The appraisers toured controlled areas on several occasions and noted no high radiation areas, one radiation area associated with a fuel cask and one airborne radioactivity area. The latter area was the gas compressor room which remains locked.
Contamination area controls were observed to be generally adequate. The exception being the availability of friskers at each control point.
Based upon the above findings, access controls appear to be adequat , however, the placement and use of friskers at each contaminated area control point should be required by plant procedures.
4.0 Radioactive Waste Management System 4.1 Program Responsibility The plant system desir,ned to store, process and dispose of gaseous, liquid and solid waste are described in Section XI of the FSV Safety Analysis Report (SAR).
Plant procedures have been developed and implemented to L-
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control the processing and disposal of radioactive waste.
The procedures reviewed in the aIpeaisal are listed in Annex C.
The operation of the rad waste system is the responsibility of the operations supervisor.
Routine operations are handled by equipment tenders and operators under the direction of the shift supervisor. An appraiser reviewed the major components, the operating experience, design objectives and the reported , releases from the plant. The plant surveillance of the reactor building vent filtration system and the plant emergency sampling capability was also reviewed and included in this section.
4.2 Waste Processing Systems 4.z.1 Liquid Waste Processing System The liquid waste system consists of a 1000 gallon waste sump, two waste filters, two 3000 gallon waste receiver tanks, two demineralizers and one 3000 gallon monitor tank. Releases are made, af:er lab analysis, from the monitor tank to the cooling tower blowdown lin2.
Releases are also made from the reactor building sump in the automatic mode at a rate less than 10 gpm.
Both release paths,gre monitored and restricted to a concentration of less than 2X10 ucif9. Dilution
by blowdown flow reduces the concentration below 1X10 uci/al which is the limiting concentration.
Routine plant operations do not generate significant liquid waste. The principal and controlling activity in liquid waste is tritium. The effluent monitor with its associated blocking valve, does not respond to tritium, therefore, the principal control is the laboratory analysis of each batch and maintenanceofb1cwdownflow.
Activity concentrations in liquid waste greater than 2X10 uci/ml can be recirculated through the demineralizers until concentrations are reduced.
The appraiser noted that demineralizer samples taken for analysis showed evidence of oil contamination.
The licensee stated that oils are not separated from liquid wastes and the demineralizers are changed when the WP is greater than 10 inches and the resin is no longer effective in removing activity.
Chemical and laundry waste input volumes are small and do not appear to have any significant impact upon the liquid waste system. The licensee stated that no design changes have been made to the system and the system has adequate capacity to handle anticipated liquid waste volumes.
The appraiser reviewed effluent reports for 1979-1980, and concluded that design objectives of the liquid waste system had been met.
Releases are well below.2 curies per year excluding tritium.
4.2.2 Gaseous Waste Processing System The gaseous waste system consists of two inlet headers, one for low activity gases that are filtered, monitored and released to the plant exhaust system and another for high activity gases that are piped to the waste gas vacuum tank. This gas is compressed by one of two compressors
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and stored in one of two waste gas surge tanks.
If low activity gas concentrations exceed present limits, two valves position to divert the gas to the waste gas vacuum tank. Higher activity gases stored in surge tanks are discharged when the tank pressure indicates full. A laboratory analysis of tank contents determines allowable release rates into the reactor building ventilation exhaust.
Stack monitors monitor the diluted release. During normal plant operation, only one surge tank is in use, leaving adequate capacity remaining for helium purification regeneration or other operations that produce a large gas volume.
The licensee stated that no design changes have been made to the Gas Water System and the system has adequate capacity to process and store anticipated gas volumes.
The appraiser reviewed the gas release permits and semi-annual effluent reports for 1979-1980, and concluded that design objectives for the gas waste system had been met.
The appraiser noted that the gaseous release permits had space for alarm set points for the vent monitors and these values did not change.
The monitors and constant alarn points are: RT-7324-1 - 77,000 cpm RT 7324-2 - 2,000 cpm RT 7325-1 - 3,200 cpm RT 7325-2 - 10 mr/ hour Investigation revealed that RT 7324-1,2 are noble gas monitors of different size and the alarm galves representated a release rate of 31.2 millicurie per second (8 Kr equivalent) in the vent.
This release rate would deliver 500 mr/ year at the Exclusion Area Boundry (EAB).
For particulate and iodine monitors RT uci/ml (37325-131 )larm set a
points represent a concentration of 7.4X10 I in the vent.
This concentration would yield a concentration at the EAB of MPC/700.
The low activity gas monitors RT-6314-1, 2 that are used to make decisions on treatment and storage of waste gas are also alarmed set very high.
131 RT-6314-1 - An alarg point of 10 cpm represents I concentration of 3.2X10 uci/ml at the monitor.
RT-6314-2 - A alarm point of 2.3X10 cpm represents a concentration (E8Kr) of 5.2X10 uci/ml at the monitor.
g It is apparent that alarm set points on vent monitors and the two process monitors above are based upon severe accident situations.
No analysis has been made of technical specification requirements on gaseous releases and corresponding alarm set points.
The appraiser saw no evidence that gaseous release rates limits specified in technical speci-fications had been exceeded due to limitations on the rate in which gases can be released from the surge tank, . . ., u . . , 33'
4.2.3 Solid Waste-Processing and Shipment t ' Fort St.'Vrain has never made a shipment of solid radioactive waste to a burial site and does not anticipate doing so in the near future.
They have about 116 fifty-five gallon drums of solid waste on hand stored in locked and marked semi-trailers.
The individual drums are , only numbered and the contents list.is maintained in the HP operations . office. No activity or isotope identification' tags are on the drums.
Compacting and resin bead transfer to drums are both done in the compacting area.
Compacting is done with no protective clothing and no air sampling. Resin bead draining and vacuum transfer to drums is done in protective clothing and no air sampling.
Smear surveys ' are done in the compacting area on a regular basis and after each resin handling job.
Total radioactivity content of resin beads-is done by Health Physics.
They take 2 - 7 grams of beads in a planchet and beta count with a gas proportional counter.
The efficiency factor used is that obtained from a plated 99-Tc standard with no correction for self absorption.
The value obtained could be off by an order of magnitude or more.
, . Activity content of drums containing unidentified fission products is determined by applying a gamma reading one foot from the drum surface to the 6 CE law. This was pointed out as being incorrect and the Lead HP Technician will develop a more accurate technique.
4.3 High Efficiency Air Filtration Systems The appraiser visited the reactor building exhaust filtration trains on June 25, 1980, and reviewed surveillance procedures and test data.
Each train consists of demisters, prefilters, high efficiency filters and charcoal beds. Testing and surveillance of these trains are technical . specification requirements and are tested in.accordance with ANSI 510-1975. -The appraiser noted oil on the The oil was identified as 60 weight mobil gear oil, floor of one train.
No. 632.
The oil is used to ' coat the prefilters. Tests performed on May 20, 1980, of the charcoal beds indicated acceptable efficiency.
The impact of excessive oil in the trains and possible adverse effects on the charcoal beds is not known. ~The item is considered.open pvading further discussions with the.NRC staff. The appraiser also reviewed surveillance activities , on-the waste gas' filters.
This system consists of a glass fiber pre-filter, a high efficiency filter and charcoal beds.
This system provides-filtration for low activity gases prior to discharge of the reactor building-vent.
The licensee stated that this system is checked weekly
by the operating: staff for adequate gas flow and acceptable pressure drops. No ~ formal testing and surveillance program exists for this system.
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4.4 Accident Sampling and Monitoring NUREG-0578 contains items that impact directly upon the health physics staff. These are post accident sampling, range of radiation monitors and in-plant iodine monitoring.
The appraiser reviewed the licensee's l letters of December 12, 27 and 28, 1979 in response to all NUREG-0!!78 sections. Emergency procedures were reviewed and future plans discussed with the plant staff. The decision.to move the radiochemistry lab from the reactor builing has been made and the new building nears l completion. Health physics procedure HPP-14 is concerned with sampl-ing of the primary coolant at the analytical instrumentation panel in the reactor building. The appraiser noted that the procedure did not l address the concentrations of radioactive materials expected after a design basis accident and the protective clothing and respiratory pro-tection that would be required to enter and collect a primary coolant sample.
Further, the procedure did not address supplies and protective equipment that may be necessary to limit personnel exposures.
FSV's analysis does not postulate any accident where their noble gas monitors would go off scale. An an interim measure, however they have installed a portable survey instrument which measures radiation levels from the vent stack via a collimated shielded enclosure external to the stack. By calculations the release rates in the stack are related to the portable instrument reading. A procedure is being finalized to cover this interim approach. Analysis of particulate and iodine activity in the vent stack during the design accident would require use of present monitors, RT-7325-1, 2 and multiple trips to remove aid replace filters and charcoal cartridges.
The filter and charcoal would be analyzed by the radiochemistry laboratory.
- The reactor building vent filters would need to be shielded in order i to reduce radiation levels for personnel pulling vent samples.
High i range containment monitors do not appear applicable to FSV'and the only change would be to area monitor RT-9325-1 which is near the main stack filters in the turbine building.
FSV's improved in plant' iodine monitoring capability has not been completed.
Portable or cart mounted iodine samplers are not commer-cial available at present.
Low volume samplers * hat can accomodate the standard charcoal cartridge are not available but presently on order. The licensee can use.the high volume campler now with the charcoal cartridge but flow rates will result in considerable error.
Gamma. spectra analysis, using Ge(Li) detectors, can be used to analyze charcoal samples.
One detector is now located near the security gate and the other detector will be in the new building.
4.5 -Conclusions Based upon the above findings, this portion of the licensee's program appears to Ne acceptable, but the following matters should be considered for improvenent.
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Develop an alarm set point procedure for all efflueni moi : tors that < . will incorporate technical specification requirements.
2.
Evaluate inpact of pref 11ter oil on efficiency of charcoal beds in reactor builotag' vent filtration trains.
3.
Complete in plant fodine monitoring capability by obtaining and using low volume samplers with charcoal cartridges.
4.
Revise HPP-4 to include protective equipment, supplies and respira-tory protection requirements for entries into the reactor building under accident conditions.
5.
Procure and install high range noble gas monitor in plant vent.
5.0 ALARA Program The Appraisal Team reviewed the licensee's administrative policies and implementation of measures to maintain occupational radiation doses as low as reasonably achieveable (ALARA) at FSV.
It was noted that there were no written administrative or HP procedures establishing the management policy and commitment to principles of ALARA except for brief statements in HP procedures.
In addition, it was found that there is no one individual or group at the Station or corporate offices that has been designated to develop ALARA programs, ensure their implementation and measure the degree of progress in achieving program objectives.
In discussions with Station personnel it appeared that ALARA principles are applied in every day operations in an informal way.
Good communiciatons, preplanning of activities .and supplemental shielding of potential exposure pathways appeared to con-tribute to potential dose savings.
The relatively low exposure potential at FSV mitigates somewhat the ck of a formal ALARA program but the Team recommends that a formal A..RA program be developed and implemented as in accordance with guidance contained in NRC Regulatory Guide 8.8, "Informa-tion Relevant to Ensuring That Occupation Radiation Exposures.t Nuclear Power Stations Will be as Low as Reasonable Achieveable."
6.0 Health Physics Facilities and Equipment
6.1 Facilities The appraisers visited and reviewed the facilities used by the health physics and radiochemistry staffs in carrying out their various functions.
Since little guidance is available to determine the adequacy of these facilities, appraisal findings are based upon appraiser judgement and comparisons with other nuclear power plants. The facilities reviewed were: The health physics field office and access control, health physics - w
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counting area, instrument calibration and storage, personnel decon, equip-ment decon, change rooms, office space, laundry, personnel dosimetry, respirator fitting, testing and storage, health physics training, radio-chemistry laboratory, sample storage and access to sampling areas.
6.1.1 ' Radiation Protection and Radicchemistry The health physics field office is located just outside the level 7 access to the reactor building.
The large windows would suggest an intent to observe personnel entering and existing the control point. The windows are now covered with various materials, there-fore little surveillance is taking place. The field office may be too close to potential sources of contamination.
Incidents in the past have caused the office to be evacuated. The outside hall way is the path for contaminated laundry and personnel.
Counting of swipes and air samples is performed in the same area and equipment could become contaminated from an incident in the reactor building.
An adjacent area contains office space, storage laundry, personnel decon and first aid.
Although the amount of laundry is small, the laundry operation presents a significant source of contamination and should be located in an isolated controlled area. The first aid area containing a cot and first aid supplies could be contaminated from an incident in the reactor building or from the laundry.
The personnel decon area is of sufficient size and consists of showers with smooth tile floor and walls and a door way block to prevent the spread of contamination.
Separate decon facilities are not provided for male and female workers although adequate and separate change rooms have been provided.
Only one feraale now works in potentially contaminated areas.
Offsite personnel decon facilitics are very limited, consisting of janitorial sinks and rest rooms in the QA building.
The Respiratory Protection Program is very small and facilities have not been provided to fit and test respirators or space for proper cleaning and storage. A space has been allocated for equipment decon and the processing of solid waste.
The appraisers noted that local exhaust ventilation is not provided for the dry waste compactor, an operation which could provide an airborne radioactivity problem.
The radiochemistry lab will be moved from its present location in the reactor building to a new building adjoining but separate from the reactor building.
The new lab and storage space alloca-tion appears adequat. ._ . . _ -. -
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- , s a6.' 1. 2 Conclusions Based on the above findings, this portion of the licensee's . program appears adequate, but.;he following matters should tne > considered for improvement.
I ^ ' Relocate' health physics field office and counting equipment 1.
to a near location with adequate space and low potential for contamination.
2.
Relocate contaminated laundry operation to an isolated location , - to reduce potential for spread of contamination.
- 3.
Provide adequate space for respirator fit, testing and cleaning.
! ] 4.
Provide-loca' exhaust ventilation with high effi~ciency filtra- ! tion for'tha solid waste compactor.
5.
Provide adequate facilities near-site for personnel decontamina-tion.
, ~ 6.2 Protective Equipment ' Supplies and _ inventory control sere reviewed for protective equipment, especially. anti-C's.
FSV uses.very few anti-C's, averaging 3 or 4 sets a day. They have a fairly large-inventory, 500 to 700 sets, on < " hand.
Inventory is not formally controlled but is checked before each
i planned outage..All protective equipment is readily available and > - closely controlled by Health _ Physics,
Each'of the two emergency _ kits _contains 75 sets of disposable anti-C's-and are available outside of the reactor _ building.
The supply of lead bricks, supplemental shielding and remote handling tools is very ' limited but probably adequate for the radiation levels currently ex-perienced under the present-operating conditions.
In summary,' the appraisers consider this area.to be acceptable.
I 7.0 Emergency Response /Re-Entry ~ /.n appraiser reviewed the FSV health ~ physics preparedness for responding to 'au accidentLand capabilities.for re-entry.. The review in this area during ,
the appraisal did not duplicate otheriNRC emergency. planning ~ program evalua-tions in progress. The PSCo's. capabilities for management and technical ' support in the event of an equivalent Three Mile-Island type occurrence have been' documented in:the licensee's submittal to the NRC (NRR): dated 0ctober 15, ~ (1979.
Agreements Lwith outside individuals and firms have been made for.
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technical support, environmental and in plant sampling and analyses, and some limited radiation protection assistance. No apparent arrangements have been made for augumenting the onsite health physics staff with properly qualified , health physics assiscance to support a re-entry or recovery phase following an _ accident. Also as noted in' Section 3.1, no apparent provisions for personnel - dosimetry during such events have been made.
The NRC is-conducting a separate nuclear reactor emergency planning evaluation program. Because of thi=, the Health Physics Appraisal Team will not make conclusions of adequacy in this area except to recommend. that a suitable agreement for~ supplemental health physics assistance be entered into and previsions for personnel dosimetry, including rapid processing, during accident situations be provided.
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... , ANNEX A EXIT INTERVIEW The Appraisal Team and Region IV Fuel Facility and Material Safety Branch Chief met with licensee representaives (identified in ' Annex B) onsite at the conclusion of the appraisal c.1 June 27,1980. The appraisers summarized the scope and major findings of the appraisal.
The findings-are classified into three categories: A.
Significant appraisal findings are ' described in Appendix A to the letter forwarding.this report and are summarized at the conclusion of applic- .able sections of this report. Written responses to these findings will be' required to be submitted by the-licensee. Actions taken on these finding's will be reviewed during subsequent inspect ~ ms.
B.
Findings of' lesser significance but which are considered important in implementing a quality health physics program are also summarized at .the end of applicable sections. No written response to these findings c will be required and progress in improvement in these areas will be observed during' routine inspections.
'C.
The apparent noncompliance item identified during the appraisal is specified in Appendix B to the letter forwarding this report.
The licensee is recuired to respond to this finding in writing and the response will oe reviewed and verified during subsequent inspections.
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. . _ _ _ _ _ '. - , ' - ANNEX B PERSONS CONTACTED
- D. Warembourg, Manager of. Nuclear' Production
- F. Mathie,;0perations Manager
- W. Hillyard, Administrative Services Manager
- J. Gahm, Technical Services Supervizor
- F. Jorst,~ Senior Plant' Engineer (RMP)
T. Schleiger, Health Physicist
- W. Woodard, Senior HP Technician Y. Derrer, HP Technician
~E. Stroud, HP Technician
- M. Prochownik, HP Technician
- V. McGaffic, Radiochemist F. Novacheck, Laboratory Assistant S. Willford, Training Instructor
' . T. Chrisler, Training Instructor P. Bollis,-Clerk
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D. Hood, Shift Supervisor J. Maynard;-Equipment Operator B. Freneck, Results Supervisor . C. Olson, Results Technician
- L. Brey, QA Manager.
-*T. Orlin, Superintendent Operations QA R. Reiss, QA Engineer
- Indicates those present at the exit interview.
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. l = , ' ANNEX C' DOCUMENTS REVIEWED-FSV, FSAR FSV, Technical-Specifications 'FSV, Training Programs Administrative Manual FSV, Situation Response. Training Programs FSV, Health Physics Technician Training Check-off List FSV, Personnel Resumes and Training Records FSV, Outline of Job-Related Health Physics Training FSV, Administrative and Health Physics Procedures FSV, Health Physics Procedure HHP-Intervals of Surveys and Use of Survey Maps FSV, Health Physics Procedure HPP-8 Radiation Surveys FSV, Health Physics Procedure HPP-9 Establishing and Posting Controlled Areas FSV, Healtb Physics Procedure HPP-11 Personnel Decontamination FSV, Health Physics Procedure HPP-13 Continuous Air Monitor Instructions FSV, Health Physics Procedure HPP-21 Surface Radioactive Contamination Surveys FSV, Health Physics-Procedure HPP-52 NMC-PCC11 TC Gas Proportional Counter FSV, Health Physics Procedure HPP-16 Full Face Respirator Self. Contained Breathing Apparatus and Half Mask Respirator FSV, Health Physics Procedure HPP-48 Routine Maintenance and Inspection of Respiratory Equipment FSV, Health Physics Procedure HPP-49 Respirator Facepiece Fitting FSV, Health Physics Procedure HPP-50 Breathing Air System Sampling Procedure , _ _ _
a s a . FSV', Health Physics Procedure HPP-53-RTT 7325-1~and RT 73437 Filter and Cart Removal FSV, Health Physics Procedure HPP-55 Reactor Building Ventilation Effluent Activity Determination Using RT 73437 and RT 4801 FSV, Health. Physics Procedure HPP-56 Reactor Building Exhaust Stack Lischarge Activity Calculation FSV, Health Physics Procedure APP-14 Analytical Instrumentation Room FSV, SOP-62 Radioactive Liquid ~ Waste FSV, SOP-63 Radioactive Gas Waste FSV, SOP-73-01 Reactor-Plant Ventilation System FSV, Health Physics Precedure HPP-17 Determining Radioactive Liquid Release Rates FSV, Health Physics Pr :edure KPP-18 Manual Calculation of Radioactive Gaseous Release Rates FSV, Surveillance Procedure SR 5.5.3bc-A Reactor Plant Exhaust Filters, Charcoal Filters, Halogenated Hydrocarbon Removal and HEPA Leak Test FSV, Surveillance Procedure SR 5.5.3a-SA Reactor Building Exhaust Filters and Charcoal Filters Samples FSV, Radiochemistry Procedure RC-9 ' Sample Preparation for Gamma Spectral Analysis FSV, Radiochemistry Procedure'RC-18 Operating Procedure for Northern Scientific Analyzer , T
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_. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ FIGURE 1 FORT ST. VRAIN NUCLEAR GENERATING STATION CONDUCT OF OPERATIONS CilART . HANAGER NUCLEAR PRODUCTION . . OPERATIONS ADMINISTRATIVE -~~ MANAGER ~ ~ ~ ~ ~ ~ ~I SERVICES l MANAGER' t i TECilNICAL g~~~ COMPUTER SERVICES PORC SERVICES SUPERVISOR _ - - -
- -. _ _ _ .. _ _..... IIEALTil SUPERINTENDENT RESULTS SUPERINTENDENT TRAINING SECURITY SCllEDULING CLERICA PilYSICS MAINTENANCE ENGINEERING OPERATIONS SUPERVISOR SUPERVISOR QC/ STORES SUPERVIk SUPERVISOR SUPERVISOR SLO SLO SUPERVISOR SilIFT SUPERVISOR 1/SilIFT SLO l ~ ~
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~~~ EQUIPMENT REACTOR. _. _ _ _.. .....___l AUXILIARY OPERATOR OPERATOR TENDER 2/SilIFT 2/SilIFT LO 2/SilIFT . . -d
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_ - _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ FICURE 2 - FORT ST. VRAIN NUCLEAR. GENERATING STATION RADIATION PROTECTION ORGANIZATION.
MANAGER NUCL i PRODUCTION ,! ' t OPERATIONS TECilNICAL SERVICES MANAGER SUPERVISOR I I IIEALTil PilSYICS RADIATION PROTECTION SUPERVISOR MANAGER' ' l i SENIOR llP.
RADIOCHEMIST-TECHNICIAN I HP RC LABORATORY- .- . TECHNICIAN ASSISTANTS r s
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