IR 05000317/1980009
| ML19350D420 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/14/1980 |
| From: | Donaldson D, Galen Smith, Jason White NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19350D411 | List: |
| References | |
| 50-317-80-09, 50-317-80-9, 50-318-80-07, 50-318-80-7, NUDOCS 8104150465 | |
| Download: ML19350D420 (55) | |
Text
.
.
.
.
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
Health Physics Appraisal Program 50-317/80-09 Report No.
50-318/80-07 50-317 Docket No.
50-318 DPR-53 License No.
DPR-69 Priority
-_
Category c
Licensee:
Baltimore Gas and Electric Comoany P.O. Box 1475
_jal timore. Marvland 21203 Facility Name: Calvert Cliffs. Units 1 and 2 Appraisal At:
Lusbv. Marvland Appraisal Conducted:
May 12-Mav.23. 1980 Team Members:
h.
/8
J R. Wh te', Tea'm Leader, NRC
/ dats ~
~
/d//4 $O D. Donaldson, Radiatio. S cialist, NRC
/ date
/S8'
R. Ryan,Jittelle 77/
d' ate '
l lY
/d
- //O G. W. EndrgM,~ Bat' telly ~ '
d' ate'
date
,
/
n date
////y[d:r Approved by:
we
,
,
Geo ge H/Smi"th, Chief, Fuel Facility and date '
Ma rials Safety Branch Region I Form 12 (Rev. April 1977)
8104150h(ph
.
.
.
TABLE OF CONTENTS 1.0 Radiation Protection Organization 1.1 Description 1. 2 Scope of Responsibility 1.3 Staffing 2.0 Personnel Selection, Qualification and Training i
2.1 Personnel Selection and Qualification 2.2 Personnel Training 3.0 Exposure Control 3.1 External Exposure Control 3.2 Internal Exposure Control 3.3 Surveillance Program 3.4 Respiratory Protection Program 4.0 Radiation Waste Management System 4.1 Program Responsibility 4.2 Waste Processing Systems
.
4.3 Effluent Process Instrumentation 5.0 ALARA Program 6.0 Health Physics Facilities and Equipment 7.0 Administration of the Emergency Plan 8.0 Emergency Organization 8.1 Onsite Organization 8.2 Augmentation of Onsite Emergency Organization 9.0 Emergency Training / Retraining 10.0 Emergency Facilities and Equipment 10.1 Emergency Kits and Instrumentation 10.2 Area and Process Radiation Monitors 10.3 Emergency Operations Center 10.4 Medical Treatment Facilities 10.5 Decontamination Facilities
11.0 Emergency Implementing Procedures
,
l 11.1 Content and Format 11.2 Implementing Instructions l
11.3 Implementing Procedures
'
11.4 Supplementary Procedures l
l
_._ _. -,
_ - _ _ _
_
.
-
.
.... _ _-_
,
-. - - -
- - - - _ - -
,
_ _... _ _.. - -. _ _
.-_ >
-
-
.. _... _ _ _.
. - _ _
. _-. _..-._._._
. -.. _ _ _.._
'
i i
.
.
,
-
.
i i
..
Annex A - Exit Meeting and Licensee Commitments
l l
Annex B - Persons Contacted
!
I I
\\
.
'
Appendix A - Significant-Appraisal Findings Appendix B - Notice of Violation-
-!
'
.
I I
l I
!
!
I
.
,
!
!
!
e k
,
!
.
.
.
1. 0 RADIATION PROTECTION ORGANIZATION 1.1 Description 1.1.1 Current Health Physics Organization The current organization chart in place at Calvert Cliffs Nuclear Power Plant (CCNPP) is described in Baltimore Gas and Electric (BG&E) Company's, Organization Planning Manual.
The structure is as depicted in Figure 1, which indicates
-
that the Radiation Safety and Chemistry Engineer (RSCE), the designated Radiation Protection Manager, reports directly to the Chief Engineer (plant superintendent).
BG&E does not have a corporate level organization or individual assigned to the radiation safety aspects of nuclear power plant operation.
1.1.2 Proposed Health Physics Organization On May 13, 1980, BG&E announced that effective July 1, 1980 that the existing Electric Production Department would become two departments reporting to the Vice President, Supply, i.e., Nuclear Power Department and Fossil Power Department.
In the proposed organizational structure, depicted in Figure 2, the present Radiation Safety and Chemistry Group will be split.
The Radiation Safety Section will report to the Nuclear Power Department Manager, and the Chemistry Section will report to the Plant Superintendent (presently, Chief Engineer), who also reports to the Nuclear Power Department Manager.
At the time of this appraisal, the licensee had not completed the formalization of the reorganization below General Supervisor-Radiation Safety.
1. 2 Scope of Responsibility 1.2.1 Responsibility and Authority The responsibilities assigned to the Radiation Protection organization are implied by several documents, i.e., Quality Assurance Procedures (QAP) and Calvert Cliffs Instruction _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _
_ ____
FIGURE 1 - CURRENT RADIATION PROTECTION
.
ORGANIZATION, CALVERT CLIFFS
PLANT Chairman of the Chief Engineer,
,
Board Calvert Cliffs
.
.
President Radiation Safety
&
Chemistry Engr.
!
Plant Performance Vice President, Health Engineers Supply Physicist Radiation Safety Radiation Safety
&
&
Chemistry Foreman Chemistry Foreman ro Manager Electric Produc-I I
I i
tion Department
'
Dosimetry Radiation Radiation Counting Chemistry Control Support Room Laboratory.
A\\
Respiratory Instrumenta-Protection
-
tion
-
Solid Waste Handling
-
Decontam-ination
_ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _
_ _ _ _ _ _ _ _ _ - _ _ _ _.
_ _ _ _ _ _ _ _ _ _
__ ___
___
_
. _ _ -
FIGURE 2 - PLANNED RE0RGANIZATION
..
.
A
-
..
.
-
Vice-President, Manager, Supply Nuclear Power
<
Department
-
,
I
.
i Manager, General Supy.
Supervisor, General Supv.
Fossil Power A
Radiation Safety Administrative Nuclear Fuel Department Services Management i
,
i General Supv.
Plant Supt, w
Training and Calvert Cliffs
-
Technical Services
-
'
I i
General Supv.'
.
Baltimore cas & Electric Calvert Cliffs Supv.
Chemistry Corporate Office Nuclear Power Operation Plant, Site Office General Supv.
- r Electrical &
rnntynle
.
.
.
.
.
The following documents were reviewed to determine the assignment of responsibility.
QAP-3, Radioactive Waste QAP-18, Radiation Safety and Chemistry CCI400, Radiation Safety Manual QAP-18 is the prime document defining the scope of responsi-bility for radiation safety and indicates that the Radiation Safety and Chemistry Group is responsible for all activities pertaining to radiochemistry and chemistry operations, maintenance of radiation survey instruments and laboratory equipment, and radiation safety.
The Chief Engineer is specified as being the responsible individual.
The general responsibilities of the following personnel are also specif-ied:
Radiation Safety and Chemistry Engineer
--
Radiation Safety and Chemistry Foreman
--
--
Radiation Safety and Chemistry Technicians Radiation Monitoring Personnel
--
Shift Supervisor
--
General Supervisor, Operation Quality Assurance
--
The QAP does not acknowledge the existence of a Plant Health Physicist, and therefore does not provide any description of his responsibilities.
The QAP also does not indicate that RSCE is also responsible for the performance of plant water treatment systems and the operation of the plant's sewage treat =ent facility.
Although statements in the documents representing the RSCE's authority in matters affecting radiation safety are not obvious, there is evidence that the RSCE has sufficient management support to carry out the responsibilities vested in the program.
In the course of this appraisal, it was observed that the preponderance of responsibilities for the radiation protection program were either delegated or passively permitted to fall to the Plant Health Physicist.
The RSCE indicated that in as much as his forte was not health physics, but rather chemistry he assigned the majority of the responsibility for the program to the Plant Health Physicist.
.
-=v-,
.
.
.
.
The job description for the Plant Health Physicist as stated in the "BG&E Organizational Planning Manual", reflects major responsibilities regarding the radiation protection program, in much the same manner as any designated RPM or corporate Health Physicist.
In fact, the majority of the available technical competence in radiation protection is vested in the Plant Health Physicist.
However, the auditors noted t
that while the Plant Health Physicist has the responsibility,
'
he does not, himself, have the necessary authority to carry out or direct the implementation of the program elements for which he is responsible.
He is not a supervisor, nor is he in the line organization.
His position in the organization structure (See Figure 1) is as a technical assistant to the RSCE.
The structure is misleading in this regard.
From interviews with several persons, procedure and records reviews, and direct observation it is evident that CCNPP radiation protection program is centered on the Plant Health Physicist, a situation which results in the formation of an informal organizational (and communication) structure.
Communicating of information appears to be directed to the Plant Health Physicist from both his management and the sub-ordinate organization.
However, except for that information required by the RSCE there is little communication of information from the Plant Health Physicist to any other member of the radiation protection organization, i.e., technicians, foreman, etc.
.
1.2.2 Job Descriptions
Job descriptions (responsibilities and reporting chains) are documented for each position within the Radiation Safety and Chemistry Group (including the Plant Health Physicist).
These job descriptions appear in the BG&E, Organizational Planning Manual.
These job descriptions are generic in nature, particularly in regards to technician and performance engineers.
Observations by the auditor indicate that these two groups are actually more specialized than what is indicated in the job descriptions.
For example, technicians are predominently divided between the specialities of health physics and chemistry.
Within those two groups they are further dedicated to specialty areas such as dosimetry, respiratory protection, radiation control, radioactive solid waste handling, counting, chemistry, etc.
Likewise, at least one of the performance engineers has been assigned sole responsibility for the management and coordination of all radioactive waste.
In these cases, where specialized assignments have been made,
.
.
.
.
.
l there are no job descriptions that detail the responsibili-ties that are associated with the specific job.
The licensee has up to this time relied upon the assigned individuals themselves to determine their associated responsibilities with very little formal guidance or direction from management.
This area is further discussed in Section 2, Personnel Selection, Qualification and Training.
l 1.3 Staffing Currently, the following numbers of personnel comprise the staff for the CCNPP radiation protection and chemistry program:
Number of Persons Number of Directly Supervised Position Title Positions by this Position Radiation Safety and
5 Chemistry (RSC)
Engineer (Radiation Protection Manager)
Plant Health Physicist
0 Performance Engineers
0 RSC Forman, Radiation
s80 RSC Foreman, Chemistry
12 Principle RSC Technicians,
0
,
l Radiation Safety Principle RSC Technicians,
0 Chemistry RSC Technicians,
0 Radiation Safety RSC Technicians,
0 Chemistry Contractor Technicians,
0 Radiation Safety i
.
- -
w
,,-..
.,.--y--+
..-y
, _ _.
-,, _,
.-_%._w.._.--.--.%_p-,,.- - -., ---,_
,
-
, -. - - - -
--c- -, - -_
---e
.
-
.
.
Number of Persons Number of Directly Supervised Position Title Positions by this Position Contractor Personnel,
0 Radiation Safety (non qualified technicians and clerks)
Contractor Personnel,
0 Chemistry (non qualified technicians)
As can be seen the ratio of contractor to utility employees in the radiation safety area is about 6 to 1.
Although the contractor staff is managed by CCNPP personnel and appears to meet the applicable standards of ANSI-18.1-1971, i.e.,
technicians in responsible positions to have at least 2 years previous occupational experience in the speciality, there is a tendency for inordinate reliance being placeo on the contractors, and a long-term commitment to contractor organiza-tions.
The result of such a situation could be an inordinate reliance on personnel who are transient by nature of their employment, who may not bc completely familiar with the facility and its characteristics, and wha are not subject to any formalized training (other than site specific information) pertaining to the health physics speciality.
Such condition would critically impact the effectiveness of the radia-tion safety function in an emergency situation.
The ifcensee indicates that there are currently nine radiation safety technician positions available, but that there is a lack of qualified applicants meeting the licensee's selection criteria, i.e., personnel who have at least three years of previous occupational experience in health physics as well as an acceptable degree of training in the field.
From the above table it is also seen that the span of management control at the first level of supervision (RSC Foremen, Radiation Safety) extends to about 80 people.
Even given the fact that this group contains three principal technicians (non-supervisory personnel)
who essentially act as group leaders, the span of actual control is inordinately large and consequently forces the situation of judgments
,
(assumed risks) affecting radiation safety to be made by non-managemant l
personnel in the name of the company.
While this condition may not be totally objectionable, it does require that the personnel responsible to make such judgments be adequately qualified and trained, a matter which is subject to some question (See Section 2.0, Personnel Selection, Qualification and Training).
I i
l
.
.
.
.
Evaluation:
1.
The reorganization has the potential to improve certain characteristics of the radiation protection program, particularly if the organizational structure and description recognizes the actual significance and import of the current job scope of the Plant Health Physicist.
However, unless either the scope or the structure is amended, the reorganization could compound the already existing responsibility / authority dilemma of the Plant Health
,
l Physicist.
I Based on the above finding, the following improvement in this area is required to achieve an acceptable program:
The position of Plant Health Physicist requires the amendment of functional responsibility to either:
--
include the necessary authority to carry out the assigned respon-sibilities or, relegate the position to one of a technical assistance to the
--
designated RPM.
In this case, it is necessary to assure that the responsibilities assigned to the individual are commensurate with the position.
2.
From observations, it is obvious that in the current program, specialty areas are designated that have not been formally described or acknowledged in any licensee generated document such as Job Description, Quality Assurance Procedures, Organizational Planning Manual, etc.
As a result, the personnel holding these specialty positions (such as in dosimetry, respiratory pro-tection, radioactive solid waste handling, radioactive waste management, (
etc.) have determined their own responsibilities associated with the specialty with no formal review of the intended responsibility or guidance and direc-
'
tion pertaining to authority by management.
Based on the above finding, this portion of the licensee's program appears to be acceptable but the following should be considered for improvement in the program:
--
Job functions that are specialized in nature should be reviewed to determine the associated responsibility and authority vested in the specialized area.
Consideration should be given to formally docu-menting functional responsibility, particularly in the area of radio-active waste management since coordination of this activity requires interfacing with several different organizations.
3.
The span of control, as currently assigned to the RSC Foreman-Radiation Safety is beyond reasonable expectation of affecting adequate management of the program.
The current situation (one supervisor managing 80 persons)
l l
l
-._
-
... _ _.
-..
l
.
.
.
.
will undoubtedly result in the condition that decisions affecting radiation safety that should be subject to professional review, will be made by technical level personnel or lower simply because of the unavailability of management personnel.
Based on the above finding, improvement in the following area is required to achieve an acceptable program:
--
The current span of control, job function, and specialty areas managed '
by the current RSC Foreman are required to be evaluated to determine the need for additional management personnel to support and adequately supervise that portion of the licensee's radiation protection program.
4.
The licensee appears to be tending toward the situation of inordinate reliance on contractor-supplied personnel.
Though this situation appears to be aggrevated in part to increased plant maintenance and upgrading requirements that have been chronic for the last two years, there is evi-dence that the licensee's practice of seeking only fully qualified, fully trained personnel in lieu of providing a Radiation Safety Technician quali-fication and training program sponsored by the utility has created the necessity of relying on contractor health physics technicians.
This item is further discussed in Section 2, Personnel Selection, Qualification and Training.
At the present time there are nine technician positions available at CCNPP but due to a lack of personnel meeting the licensee's selection criteria (fully trained, and fully qualified) the positions have be unfilled for some time and there is no evidence that they will be filled in the near future.
Based on the above finding, this portion of the licensee's program appears to be acceptable, but the following should be considered for improvement.
--
The licensee should evaluate its current reliance on contractor-supplied technicians, particularly those who may fill speciality '
positions.
The evaluation should consider such aspects as the advis-ability of contractual assurance with the contractor services for extended notices prior to action affecting personnel assignment, termination of employment, etc.
The evaluation should also consider means to reduce, mitigate or further decrease inordinate reliance on contractor-supplied technicians.
s
.
.
.
.
.
l 2.0 PERSONNEL SELECTION, QUALIFICATION AND TRAINING 2.1 Personnel Selection and Qualification The licensee selection criteria for Radiation / Chemistry (R/C) tech-nicians are based upon the Nuclear Regulatory Commission's (NRC) Regu-latory Guice 1.8 " Personnel Selection and Training" and the referenced document ANSI 18.1 "American National Standard for Selection and Training of Nuclear Power Plant Personnel".
The Calvert Cliffs Nuclear Power Plant (CCNPP) selection criteria were.
Job related and acceptable j@
descriptions for various positions were available at the time of the Appraisal.
Interviews with both plant R/C technicians and contractor HP technicians, a review of personnel records for both plant and contractor personnel, and discussions with CCNPP R/C supervisory personnel were performed to ascertain whether selection criteria were being used.
The contractor HP personnel, who in fact are transient workers, at CCNPP, outnumber the licensee's R/C personnel by a ratio of about 6 to 1.
Among these contractor HP personnel were a large number of junior technicians and clerks who are apprenticing at the plant to obtain the 2 year experience criteria.
The shortage of plant personnel meeting the selection criteria and the inability of the licensee to maintain an in-house training program for qualification of R/C technicians has made the licensee heavily dependent on these contracter HP technicians.
The CCNPP R/C technicians appear knowledgeable of their duties.
They are rotated through various radiation protection and chemistry positions within the R/C organization, and they are familiar with the promotional criteria.
The experience in Radiation Protection of the CCNPP R/C personnel seems adequate, though, a lack of depth in technical aspects of radiation protection appears evident.
This portion of the licensee's program appeared to be adequate.
2.2 Personnel Training The licensee's Training Program, during the time period of the Appraisal was in the midst of a transition.
The licensee recently became a member of and participant in the "Mid-Atlantic Nuclear Training Group" (MANTG). This MANTG is at present a formation of seven utilities within the Mid-Atlantic region organized to standardize training programs at nuclear power plants.
Training certificates are awarded to participants who successfully complete various programs and will be acceptable to each of i participating utilities.
I
.-- -.
.
.
.
.
The licensee also hired a new Training Director and several new per-sonnel within the Training Department.
At present the Radiation Protection training is given by two contractor personnel who are qualified by experience but who have no formal radiation protection education.
The auditor attended various training lectures, reviewed plant and contractor training records, and had numerous discussions with the Training Director and the Radiation Protection Training personnel.
The Radiation Training Program covered essentially two categories.
The first was General Orientation Training (GOT), and the second was Requalification Training (RT).
The GOT was of 1 to 2 days in length and inciuded video tapes supplemented with lectures by the training personnel.
The RT was of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> duration for individuals who had completed GOT or equivalent.
Both training sessions concluded with written exams which required a passing grade of 70% for GOT and 80%
for RT for successful completion of the training program.
Training personnel and review of records indicate a failure rate of less than 5%.
Those not passing are given an option to retake the training.
The GOT category was further broken into two parts, GOT I and GOT II.
GOT I was taken by all general employees on site.
GOT II, which was a more technical presentation following the MANTG outline was taken by technical support personnel, contractor HP personnel and other radia-tion workers.
It was ascertained by reviewing records that training had been received by essentially all categories of personnel from the Site Manager to clerical personnel.
The R/C technician training responsiblities lie with the Radiation and Chemistry Engineer.
This responsibility and the training program is established and described in CCI 607C " Radiation Safety and Chemistry Personnel Training." The auditor could not find any records of formal i
R/C technician training with the exception of on-the-job (0JT) training.
Based on the above findings, improvements in the following areas are required to achieve an acceptable program:
--
Development of formal performance criteria to be used to evaluate the ability of personnel designated to perform as Radiation Safety and Chemistry technicians.
Such criteria should be documented and should consist of written tests and practical demonstration of ability.
I.
Additionally, the following items should be considered for improvement of the program.
.
s w-
- -
.
.
.
.
Incorporation of plant system training / familiarization into the
--
,
R/C technician training.
3.0 EXPOSURE CONTROL'
3.1 External Exposure Control The external radiation dosimetry system in place at Calvert Cliffs was investigated in considerable detail.
Long term equipment failure in the reader used for the teflon-TLD personnel dosimeters has made that part of the system difficult to administer.
As a consequence, additional time and manpower are needed to maintain this function.
To alleviate this problem, a new reader system is on order.
The dosimeter now in use has no capability to detect fast and thermal neutrons.
It was further observed that the teflon tapes loaded with lithium flouride for neutron detection will not fit in the present reader.
137 A
Cs gamma source was used for all dosimeter calibrations. Most calibration exposures observed were in the range of 200 to 250 mrem, indicating a tendency to calibrate the equipment consistently at one exposure level.
A one point exposure is not an acceptable technique for calibration of personnel dosimetery.
Dosimeters should be calibra-ted at several points in their useful range to assure proper system performance.
ANSI 13.11 " Criteria for Testing Personnel Dosimetry Performance" recommends that several values be selected between 300 and 1000 mrem, as well as values in excess of 1000 mrem.
A few calibration exposures were conducted using a uranium slab for beta response measurements with the teflon-TLD dosimeters.
These beta exposures i.idicated an adequate response of the teflon-TLD to uranium beta radiation, however the auditor noted that teflon-TLD system is usually not used to monitor personnel when significant beta exposure is expected.
Personnel assigned to dosimetry appeared to know what had to be done to make the system work, but their work load was such that they found it difficult to review and coordinate TLD results with the self-reading dosimeters (SRDs) in a timely manner.
There were procedures for correlating SRD and TLD results on a monthly basis and for investigating results when the difference between the SRD and TLD is more than 25 percent.
The program was set up so that unusual exposures would be identified quickly.
Because of problems with the teflon-TLD dosimetry system, the external dosimetry program uses a TLD chip system whenever a worker is planning a job in an area where the exposure rate exceeds 1000 mR/hr.
Special l
- -
_
!
.
.
.
.
dosimeters with an open window and a one centimeter depth dose filter are made up and placed on head, feet, or ankles as needed to evaluate the exposure on the particular job.
Finger rings utilizing teflon-TLDs have been tried but are not presently used because of unsatisfactory results.
Instead, a simple ring made of elastic material with TLD
.
chips placed inside, is used for hand exposure.
Manual Eberline TLD l
reader systems are used for readout of all special dosimetry. Skin exposure is determined through the reading of the TLD chip at the
"open-window" location.
The teflon-TLD personnel dosimeter also has l
an "open-window" area but no nonpenetrating or beta exposures were noted in any of the exposure histories reviewed where the teflon dosimeter was used.
l The auditor noted that the dosimetry system for monitoring skin l
exposure with TLD chips will not respond to low beta radiation due to l
thethicgges@0 f the TLDs unless the energies of the betas are equiva-lent to Sr Y or higher.
In many cases the TLD system could signifi-
!
cantly underestimate beta dose.
There was no evidence that beta l
energy was taken into account in the evaluation of the dosimeters.
At the time of this appraisal neutron dosimetry was not being performed i
according to Regulatory Guide 8.14.
No passive personnel dosimeter l
was in service and the three PNR-4 neutron instruments were not properly
'
calibrated in a known fast neutron field.
Additionally, there was no evidence of techniques or investigatiogg to account for exposure to photons with energies greater than 3 MeV(
N),
even though the heavy metal filter on the fourth region of tgg teflon-TLD dosimeter will provide charged particle equilibrium.or N
photon (6.1 and 7.1 MeV) and provide a reasonably accurate measure of exposure at 6 and 7 MeV.
The TLD chip dosimeters will not properly measure high enepgy photons unless the filtration is equivalent to about 3000 mg/cm Self-reading dosimeters and TLDs are available and required for personnel and visitors who must enter controlled areas even for short duration.
Dedicated exposure record clerks and a security guard record all access to the controlled area.
Exposure records are kept up to date through the use of the computerized system.
Information is available in a timely fashion.
Rad / Chem procedures (RCP3-301) require a review I
of exposure on a quarterly basis.
However this function is not being performed in a timely manner due to a lack of available personnel resources.
The onsite Quality Assurance (QA) department audits the Rad / Chem pro-gram approximately twice a year.
The audit is generally performed to verify training and retraining, up-to-date procedures, accuracy of posting and survey board, and onsite calibration of dosimeters.
There l
l
.
y.y
--- --
.
.
.
.
appeared to be no evidence of a review of offsite calibration by QA.
QA audits have identified problems with out-of-date procedures and at least in one case showed a vigorous follow-up.
The auditing system is prevented from doing an in-depth review because of a lack of personnel experienced in the technical aspects of radiation protection.
Conclusions Based on the above findings, the external exposure control program at CCNPP appears to be acceptable in that the program limits exposure within the specifications of 10 CFR 20 and, in most cases. well below those limits.
However, the following matters should be considered for improvement:
1)
Equipment used to readout the routinely used personnel dosimeter should be made fully operational or replaced.
2)
A wider range of exposures should be used for calibrating the dosimeters.
3)
Beta response of the TLD chip system should be determined to allow more accurate assessment of actual exposure.
The system used now may respond significantly low.
4)
Calibration of neutron instruments should be maintained with a suitable certified neutron source or technique that provides assurance that the instrument will respond to the type of neutron energies experienced at the station.
5)
Personnel having sufficient technical qualification in radiation protection should be utilized for performing periodic reviews of the radiation safety program.
3.2 Internal Exposure Control Dosimetry Program Whenever internal deposition of radioactive material is suspected, nasal and throat smears are taken; and if levels are greater than 1000 cpm, a whole body count is conducted.
Usually the Plant Health Physi-cist is consulted in these cases.
All capability for quantifying internal deposition is placed on the whole body counter (WBC).
All detailed analysis is dene offsite l
through a computer telephone link to Helgeson, Incorporated for data processing.
Models used for determining body burdens of particular radionuclides are based on standard published techniques.
However, there appears to be no evidence that the unit has ever been calibrated
,
l
-
.
L
r
.
.
.
.
properly using a phantom with known amounts of various radionuclides.
This lack of calibration could lead to significant errors in deter-miningboggyburdensfnwrkers.
Presently, source checks with an
internal Am and Co button sources are performed weekly to assure certain electronic gain drifts are controlled.
However, the auditor noted that the WBC needs to be calibrated with a known phantom source system periodically so that actual accuracy in measuring body burdens may be determined.
l All personnel with an annual exposure greater than 500 mrem are given a WBC once a year as a minimum.
There is provision for taking urine and fecal samples at the site but this has never been performed.
If blood samples or other bio-surveil-lance techniques are indicated the plans call for the Plant Health Physicist to call in a physician.
The Plant Health Physicist is also responsible for all follow-up including dose calculations and reports.
No problems were observed in the operation of the internal exposure control system.
The WBC, when used in conjunction with the screening techniques given in the procedures, has worked well to identify and control internal depositions.
The equipment has adequate sensitivity to measure <10% of the Maximum Permissible Body Burden (MPBB) of radionuclides of concern.
Reports are obtained twice a month from Helgeson, Incorporated, and are kept on file at the station.
Exposure Review Any unusual incidents are reported to Operations and documented by the Plant Health Physicist with review by the Plant Operations and Safety Review Committee (POSRC).
The follow-up includes details of any noncompliance with Technical Specifications and other regulatory requirements.
Details cf associated personnel exposure or dose commit-ment are documented and filed by the Plant Health Physicist.
There is no formal ALARA program applied to internal exposure, but
,
there is a subjective ALARA commitment in many areas.
Reviews of l
internal exposure are conducted on a case-by-case basis with corrective l
action required as necessary.
Radiological Investigation Reports go to supervisors and the Chief Engineer.
These investigation "eports were reviewed and appeared thorough and complete.
Exposure Limitations Uptake limits are considered in establishing administrative and physical barrier controls.
Methods and calculations associated with these limits are documente.
.
.
.
Procedures used to authorize access to controlled areas are adequate and carefully adhered to by all personnel.
Previous dosimetery records and training received are required to be documented before access will be granted. Escorted visitors without documented records are limited to 100 mrem. All barriers are checked by technicians during each routine survey. Operators also check barriers each shift.
Spot checks of posted areas are conducted on each shift when work is continuing.
Posting was observed to be in accordance with 10 CFR 20.
Actual
!
exposure is reviewed and compared to the estimates recorded on the SWPs and RWPs routinely.
Proper measures appear to be taken to control local releases and clean up spills.
Clean-up is initiated immediately by maintenance personnel to avoid delays in other work.
Tests of engineering controls are conducted periodically.
Additional ventilation is available when needed.
Conclusion Based on the above findings, the following improvement in this area is required to achieve an acceptable program:
--
The WBC at CCNPP is not being calibrated periodically with a phantom source system as is recommended and endorsed in ANSI N343-1978, " Internal Dosimetry for Mixed Fission and Activation Products".
Such calibration should consist of the measurement of standard radionuclide sources in a phantom or mock-up that will permit reproducible quantitative evaluation that approximates that performed for humans.
Minimal statistical variance should be achieved.
3.3 Surveillance Program Procedures for performing routine and periodic surveys and surveillance activities are established and implemented in accordance with Technical Specification 6.8, " Procedures", and Technical Specification 6.11,
" Radiation Protection Program".
The Rad-Chem Procedure Manual, Chapter 3, contains the principal procedures that apply.
The procedures appear to be sufficiently detailed to allow the user to perform the activity adequately.
The RSCE is charged with the responsibility to maintain and evaluate the HP surveillance program.
There is no off-site or consultant group utilized for program evaluation.
The program, as it exists currently, is completely performed by on-site personnel.
The Radiation Control subsection of the Radiation Protection Organization (See Figure 1)
carries out the majority of responsibilities associated with routine
,
surveillance activities.
-
._.
.
- -
.
-
-
-
.
.
.
.
l The program provides for periodic surveys of radiation, contamination and airborne activity, the results which appear to be adequately dis-l tributed to personnel who have need of the information, i.e., RWP
'
originators, individual workers, RSC-Foreman, etc.
Records of the surveys performed are adequately maintained.
In most cases it would be a relatively simple task to trace the person, instru-ment or other factors relating to the area surveyed if necessary.
Generally the instruments specified by the procedures were adequate with the following exceptions:
a.
Procedure RCP3-401, " Radiological Surveys", specifies the following instruments to be used for beta dose rate determinations:
Instrument Beta Correction Factor E-520/HP-240
PIC-6A
6112 Teletector
Discussions with a RSC Foreman and the principal technician in Radiation Control indicated that none of the instruments specified in the procedure are actually used, but rather R0-2 type instru-ments having a beta correction factor of ~4 are used.
The auditor noted that the PIC-6A in fact does not have a beta-window and is not suited for any type of beta measurement.
The 6112 Teletector which does have a beta-window has extremely limited appeal as a beta measuring instrument due to the high correction factor required.
Upon notification the RSCE indicated that action would be taken to revise the procedure as necessary to indicate the proper instrument selection for beta measurements.
b.
Procedure RCP3-306, " Determination of Neutron Exposure", specifies the technique to be used to calculate neutron exposure to individ-uals. Paragraph 4.2.3 indicates that a neutron dose rate survey, using a PNR-4 neutron c'etecting instrument, is made of the work area.
The neutron dose is then calculated by multiplying the indicated neutron dose rate (mrem /hr) by the individual's stay-time in the neutron field.
Such a technique is described in Regulatory Guide 8.14, " Personnel Neutron Dosimeters".
Procedure RCP-3-301, " Personnel Exposure Control", paragraph 4.2.1.1 identifies an administrative neutron exposure limit of 300 mrem per quarter which may be increased with the specific approval of the RSCE.
.
-
.
S
18 As noted in Section 3.1 of this report the licensee does not currently have a personnel dosimetry system capable of measuring neutron exposure.
Therefore, it is necessary for the licensee to confine personnel neutron exposures as specified in Regulatory Guide 8.14 in order to meet the regulatory requirements of 10 CFR 20.202, " Personnel monitoring", which would ordinarily require the use of appropriate personnel monitoring equipment by each individual who is likely to exceed a dose in excess of 25 percent of the applicable value specified in paragraph (a) of 10 CFR 20.101, " Radiation dose standards for individuals in restricted areas",
i.e., s300 mrem per quarter.
In this regard, Regulatory Guide 8.14 specifies in Section C.1.c,
" Calculated neutron dose equivalent in place of neutron dosimeter",
that neutron dose equivalent may be estimated based on measured neutron dose rates with portable instruments and known personnel occupancy times (as is done by the licensee in accordance with RCP-3-306) in lieu of any personnel neutron dosimeter, provided that the individual is not likely to receive a neutron dose equivalent in excess of 100 mrem per quarter.
Personnel exposure to neutron radiation at CCNPP is such that personnel could exceed 300 mrem neutron dose equivalent per quarter. Procedure RCP3-301, as currently written, could allow such exposure to occur.
However, a review of selected records by the auditor did not reveal any exposure in excess of 300 mrem per quarter; and interviews with licensee representatives indicated
!
j that in actuality 300 mrem is considered the administrative limit which is not normally permitted to be exceeded except in an
'
emergency condition.
l l
The auditor found that the PNR-4 instruments used as the basis l
for calculated neutron dose equivalent are not calibrated against a known neutron source strength, such as a National Bureau of
'
Standards (NBS) traceable neutron source but rather are subjected to electronic alignment, gamma insensitivity verification and a
,
l response check with a 3 curie Am/Be source.
Radiation instrument calibration using known source strengths from an actual radiation source is a common industry-wide practice and is identified as a standard technique in ANSI N323-1978, Radiation Protection j
Instrumentation Test and Calibration.
i l
The auditor noted that without subjecting the PNR-4 instrument to a calibration against a known neutron source, the licensee could not provide assurance that surveys performed with such instruments adequately represented the actual neutron dose rates in the areas occupied by personne.
.
.
.
.
The primary cause of neutron exposure is the frequent entries made into the Reactor Buildings at power to perform various maintenance and surveillance activities. While some of the entires are made because of immediate necessity on an infrequent basis, the majority of the entries have frequent recurrence due to Technical Specification requirements such as,
"4.5.1 Each safety injection tank shall be demonstrated OPERABLE:...
b.
At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each
'
solution volume increase > 1% of tank volume by verifying the boron concentration of the safety injection tank solution."
Due to the chronic leakage involving the safety injection tanks in both units, frequent volume increases greater than 1% of the tank volume occur.
In order to verify that the boron concentra-tion is within specified tolerances following such increases an entry is required to collect a sample from the affected tank for boron analysis in accordance with the Technical Specification.
Currently entries for this purpose occur 2 to 3 times per week.
Note:
Prior to the maintenance performed on the safety injection system during the last outage of Unit 1,
!
entries were required about once per shift.
l Dose per entry per person generally range between 20 and 100 mrem (neutron dose equivalent) depending on the unit involved and the location of the safety injection tank (i.e., there are 4 tanks per unit, each with a different shielding geometry relative to the reactor).
W on notification of this finding the licensee committed to the l
following prior to permitting any further entires in to the
-
Reactor Building during power operations:
l 1.
Portable neutron monitoring instruments, used to estab-lish area neutron dose rates for personnel exposure l
calculations, will be calibrated using an appropriate i
neutron source that is traceable to the National Bureau
'
of Standards (NBS).
Such calibration shall be performed in accordance with an approved procedure which establishes (
the method and frequency at which the calibration shall be performed.
!
2.
Occupational exposure to neutrons will be limited in accordance with Regulatory Guide 8.14, Revision 1, August 1977, until such time as an acceptable personnel
,
.
.
-.
.
.
.
,
neutron monitoring dosimeter is available.
In the case where calculational methods are utilized in lieu of any dosimetry device, an occupational neutron exposure limit of 100 mrem per quarter will be established and such limit shall be reflected in the applicable procedures.
In addition the licensee was required to perform a documented evaluation to establish that ALARA exposure controls have been implemented (considering various options available) by June 30, 1980.
This evaluation has been received by the Directpr, NRC, Region I (Philadelphia) and is currently being reviewed.
The above understandings were documented in a letter to the Vice President-Supply, Baltimore Gas and Electric Company, from the Director, NRC, Region I, dated May 22, 1980.
The auditor noted that 10 CFR 20.201, " Surveys", requires the licensee to make or cause to be made surveys as may be necessary to comply with the regulations contained in 10 CFR 20.
Contrary to this requirement surveys were not performed on the following dates sufficient to assure compliance with 10 CFR 20.101 and 10 CFR 20.202 which prescribe the quarterly limits to which personnel are permitted to be exposed (in the circumstance when appropriate personnel monitoring equipment is not worn),
in that, the portable instruments (PNR-4) used to perform surveys to determine neutron dose equivalent for workers in the Reactor Building were not calibrated to assure that the instru-ments would respond accurately in a neutron field.
Dates on which the PNR-4 instrument was used to determine neutron dose equivalent to personnel in Reactor Buildings include the following:
Date Unit No.
May 17, 1980
May 13, 1980
May 7, 1980
May 4, 1980
,
May 3, 1980
l April 24, 1980
'
April 18, 1980
l April 10, 1980
l April 7, 1980
l May 12, 1980
l May 5, 1980
L
l
.
- - - -
- -.
- - -
- -
.
.
.
Date Unit No.
April 27, 1980
April 16, 1980
April 7, 1980
March 27, 1980
This finding represents noncompliance with 10 CFR 20.201.
(50-317/80-09-01; 50-318/80-09-01)
Evaluation:
The licensee's surveillance program appears sufficient to maintain adequate information, data collection and evaluation process relative to the radiological condition of the plant.
The deficiencies identified (i.e., regarding the select-fon and use cf beta measuring instruments, and the technique and controls applied to neutron dose equivalent measurements) do not appear to be symptoms of a general inadequacy in this portion of the licensee's program, but rather isolated cases of manag6 ment oversight.
Corrective action for both items was immediately initiated by the licensee.
Based on the above findings, while there were deficiencies identified in this portion of the licensee's program, including an item of noncompliance, the overall surveillance activity appears to be acceptable.
3.4 Respiratory Protection Program The appraisal of the Respiratory Protection Program at CCNPP included, but was not limited to, a review of the policy objectives, responsibil-ities of individuals, use of engineering control procedures, purchasing criteria and inventory control, training and testing of personnel,
testing and retesting of respiratory devices, issuance and use of respirators, and preparation for reuse.
The guidance for this review to establish adequacy is provided in NRC Regulatory Guide 8.15 " Accept-able Programs for Respiratory Protection" and detailed in NUREG-0041
" Manual of Respiratory Protection Against Airborne Radioactive Materials."
The auditor participated in the Respiratory Training Program, partici-pated in the fitting and testing for the respiratory devices, visited the various facilities associated with respiratory protection and had numerous discussions with the R/C personnel responsible for the program.
Common practice in the nuclear industry and various regulatory require-ments indicate that engineering controls should be used and are recommend-ed in lieu of respiratory protection.
Areas requiring respiratory protection should be evaluated and documented for the implementation of engineering controls rather than the use of the respirators when
_
.
.
practicable.
Such practice is required in accordance with 10 CFR 20.103 " Exposure of individuals to concentrations of radioactive materials in air in restricted areas."
CC Instruction 4008 "Calvert Cliffs Radiation Safety Manual" recommends the use of respiratory protection when the air sampling program indi-catesairconcentrationsexceedingg.1MPCandalsowhencontamination levels exceed 20,000 dpm per 100 cm beta.
Air samples are taken in areas where the wearing of respirators is mandatory.
The ifcensee's method of sampling appears to depend on a single general air sample in most cases and does not appear to be representative of the individual's breathing zone.
The attached Table 1 is the relationship of whole body counting data related to the 25% of the weekly limit (or 10 MPC hours).
Table 2
" Investigative Levels" is extracted from RCP 3-309 " Bioassays".
From the data of Table 1 which was posted as the Whole Body Counts for May 21, 1980 and Table 2 which is 5% of Maximum Permissible Dose Limit, there does not appear to be a correlation between whole body counting and the Investigative Levels. This was more pronounced when discussions with R/C personnel indicated that investigations are based on the sole judgement of the Plant Health Physicist rather than on procedure.
Review of previous incidents associated with internal contamination indicated the Plant Health Physicist used various ICRP documents in his analysis.
The auditor's review indicated the analysis and associ-ated dose assessments appeared to be correct.
Based on the auditor's findings during this appraisal, this portion of the licensee's program, " Respiratory Protection" appears to be accept-able, however the following matters should be considered for program improvement:
,
No evidence was found by the auditor of a correlation between
--
'
whole body counting data, air sample data, or respiratory protec-tion data that establishes an evaluation of the program's effec-tiveness.
The data seems to exist but no one either has the responsibility nor the inclination to analyze the data sufficient to determine that the respiratory protection program is performing j
as expected.
The auditor could not find documentation of an evaluation of
--
l specific instances where engineering controls were considered or used in lieu of respiratory protection.
Though there is evidence l
that engineering controls are sometimes used, no apparent evalua-tion is performed to justify the use of respirators in lieu of such controls.
.
.
.
The auditor could find no evidence of a corporate policy for
--
respiratory protection.
A policy does exist for the plant however.
The corporate statement should address the program objectives and applications of engineering controls as specified in NUREG 0041.
The air sampling program supporting the Respiratory Protection
--
Program is required to provide air concentration data that is representative of the breathing zone of workers.
Consideration needs to be given to better sampling arrangements to assure that
such samples are sufficient to monitor the exposure of workers.
'
Table 1 (May 21, 1980 Data Posted)
Isotope Counts (25% of Weekly Limit)
Bkg I 131 1120 cts 1110 cts Co 58 5070 cts 761 cts Co 60 1120 cts 413 cts Cs 137 1240 cts
Note:
Conversion to nanocuries 1. For I 131, Co 60 and Cs 137 net counts in peak
Activity in nanocuries
=
i 2. For Co 58 net counts in peak i
Activity in nanocuries
=
.
w-ww--
p 4w+
- - -
,,,,y---
g ge ep-q
, --,.-
e y
--
g p---
-
- - -
,,m w
- - -,,
.
.
.
Table 2 Investigative Levels 5% of Maximum Permissible Annual Dose Limits Organ Dose Equivalence (Rems)
Whole Body 0.25 Red Bone Marrow 0.25 Gonads 0.25 Skin 1.5 Thyroid 1.5 Bone 1.5 All other organs 0.750 Note:
Internal + External = less than Radiation Protection Guide.
4.0 RADI0 ACTIVE WASTE MANAGEMENT SYSTEM 4.1 Program Responsibility QAP-3, " Radioactive Waste", Section 5.1, indicates that "the Chief Engineer has overall responsibility and must review and approve pro-cedures for all activities governing the release and handling of radioactive waste." Sections 5.2 and 5.3 detail the responsibilities of the Radiation Safety and Chemistry Engineer, as well as the Radiation l
Chemistry Foreman in this area.
l
'
The QAP does not recognize that currently responsibility for the implementation of a radioactive waste management program is vested in a single individual, a Radiation Safety and Chemistry Performance Engineer who is designated as Radioactive Waste Engineer (RWE).
As mentioned in Section 1.0 of this report, a functional description of the responsibilities and authorities associated with this position does not exist. The same is true of the Radiation Safety Technician who has been charged with the responsibility for solid radioactive waste handling and shipping.
However, despite this oversight, evidence in the form of observations,
'
interviews with the personnel, and review of records and documents indicates that the activity is being adequately controlled by these individuals.
.
,n
,
,, - - -
.,e--
- -, -,,
-,
'
.
.
4.2 Waste Processing Systems 4.2.1 General The auditor's review of the area indicates that the systems used to process and handle radioactive waste are operating within their design objectives as specified in the Final Safety Analysis Report (FSAR), Section 11, " Waste Processing and Radiation Protection." Note is made of the fact that a cement solidification facility originally designed into the facility was rendered inoperative prior to licensing of Unit 1, and, therefore, there is no solidification capability currently available at CCNPP.
Interviews with the RWE indicate that up to this time CCNPP has not had to rely on solidification due to the ability to adequately process and thereby reuse or discharge radioactive liquid effluent.
The system as designed allows for alternate paths to process and accommodate liquid and gaseous effluents.
The liquid processing system contains three liquid waste evaporators
.
(designed for 20 gpm distillate flow), one of which has
!
never been used but has been maintained for availability if i
required.
The equipment appears to operate within the expected decon-tamination factors, radionuclide concentrations and equip-ment specifications.
However, formalized management of the system does not exist but rather it is left up to the RWE to decide and judge on the acceptability of system performance.
There have been no major changes to the system as designed.
Iri the case where changes either in the design or operation have been made, QAP-15, " Changes, Tests and Experiments",
l appears to be followed.
In this area the auditor reviewed the situation in which the normal procedure for venting the Reactor Containment (RC)
was changed.
l
'
The licensee's test and evaluation of RC Purge and Vent Valves (CV1410, 1411, 1412, 1413) in accordance with NRC letter dated 9/27/79 from D. G. Eisenhut to All Light Water Reactors, and NRC letter dated 10/23/79 from R. W. Reid to
,
A. E. Lundvall, Jr., Subject:
" Containment Purging and
!
Venting During Normal Operations", led the licensee to issue
'
NPE-0 " Standing Order #76-6". This order indicates that purge and vent valves are to be considered disabled and shall remain in the closed position until such time that
_
.
.
.
modification to the system could be made to demonstrate that the valves would close against LOCA pressure and dynamic loadings.
In order to accommodate containment venting, the applicable operating instruction, 0I-36, Section 7 was changed (Change No.79-147) and reviewed by the Plant Opera-tions Review and Safety Committee (PORSC).
The change identified prerequisites such as the issuance of an approved Gaseous Release Permit and installation of a temporary duct extending from the covered pump in the ECCS pump room to the room's HEPA and charcoal ventilation system.
The procedure required placing the ECCS ventilation system in service, and opening Containment Normal Sump Drains EAD-5462-CV, EAD-5463-CV, thereby providing a vent path through the containment sump system to the ECCS pump room.
In this case the ECCS pump room ventilation system provides gaseous effluent clean up prior to release to the main vent.
In-place monitoring of the system exists that the ECCS venti-lation and the main vent.
The PORSC reviewed and recommended approval of the change and determined that the change did not constitute ar. unreviewed safety question in accordance with the criteria presented in QAP-15.
4.2.2 Liquid and Gaseous The auditor verified that HEPA filters and charcoal adsorber systems are subject to tests and laboratory analysis.
The licensee has contracted with a consultant to evaluate the need for liquid waste solidification capabilities in order to accommodate further restrictions that may be placed on the burial of radioactive material (such as dewatered resins) and to assure that the licensee's system is able to handle emergent conditions.
The evaluation will also include radwaste storage capabilities and other waste system features.
!
Specific waste handling capabilities exist for processing contaminated oil.
Procedures exist for the movement and discharge of liquid and gasous effluents. Those procedures address technical specifications, alarm set points, sampling techniques and collection.
.
.
.
The sampling system was reviewed and found to be sufficient for assuring that representative samples are collected.
In assessing operations to determine if Ifquid and gaseous releases are maintained As Low As Reasonable Achievable (ALARA), it was found that though process systems are utilized (such as charcoal /HEPA ventilation system for iodine and particulate activity removal and recycling of reactor coolant), a formalized ALARA program does not exist (See Section 5.0, ALARA Program).
Without adequate documentation of the effectiveness of operations and their impact on effluent releases, no conclusion can be made concerning the ALARA aspects of the program.
In terms of storage capability, the waste processing system contains the following:
Component Quantity Capacity Miscellaneous Waste (1)
4000 gal.
Receiver Tank Miscellaneous Waste (1)
4000 gal.
Monitor Tank Reactor Coolant Waste (2)
90,000 gal.
Receiver Tank (180,000 gal.)
Reactor Coolant Waste (2)
90,000 gal.
Monitor Tank (180,000 gal.)
Waste Gas Decay Tank (3)
610 cu. ft.
(1830 cu. ft.)
According to the RWE, nominally there is 90,000 gallons available in each liquid processing system providing a free
,
l capacity of 180,000 gallons.
The normal worst case condition noted by the RWE would be about 90,000 gallons of free capacity. The RWE noted that in an emergent condition process-ing of filled tanks could be accelerated by the use of the
" spare" liquid waste evaporation previously mentioned in Section 4.2.1 of this report.
In the gaseous processing system, there is normally one
,
empty waste gas decay tank available.
Currently decay times l
'
range from 7 to 45 days (20 days as the average) according to the RWE.
-
-
__
_.__
..
_
_.
_
_
--.
.. _ _
.
.
.
l
4.2.3 Solid Waste Processing Disposition The licensee's program involving the packaging and shipment i
of radioactive waste is in conformance with DOT and NRC regulatory requirements with one exception.
10 CFR 71.12 " General license for shipment in DOT specification containers, in packages approved for use by another person, and in packages approved by a foreign
'
(
national competent authority" grants a general license to the ifcensee to deliver licensed material to a jl carrier for transport provided the licensee has a quality assurance program satisfying the provisions of 10 CFR 71.51, " Establishment and maintenance of a quality assurance program".
10 CFR 71.51 states that the licensee shall establish, maintain and execute a quality assurance program satisfying each of the applic-able criteria specified in Appendix E, " Quality Assurance Criteria for Shipping Packages for Radioactive Material",
l and satisfying any specific provisions which are applic-able to the ifcensee's activities including procurement of packaging.
Contrary to this requirement the licensee has delivered licensed material to a carrier for transport since January 1, 1979 without establishing, maintaining and I
executing a quality assurance program in accordance j
with the criteria specified in Appendix E.
Though there were elements of the radioactive waste shipping
'
activity that could be construed to imply that some aspects of quality assurance were being applied, the licensee did not have an established program nor was I
there execution of any quality assurance functions performed by an authority independent of the Radiation Safety and Chemistry Group.
From discussions with licensee representatives it is apparent that the licensee's intent regarding quality assurance requirements for packages was to accept the quality assurance programs that were implemented by the various suppliers of shipping packages.
This intent is reflected
,
in an April 3, 1980 BG&E response to NRC letter dated March 7, 1980 from Chief, Transportation Certification Branch.
That letter indicated that BG&E would invoke the quality assurance program of a shipping cask supplier in lieu of the licensee's own Appendix B, Part 50 Quality Assurance Program.
,--
-
-, -
F
.
,
A following NRC letter dated April 18, 1980 to BG&E on the subject indicated that it was necessary for the licensee to implement his own QA program; and that invoking a supplier's QA program was not acceptable.
In this latest letter the licensee was directed to submit their intent as regards a quality assurance program meeting the criteria of 10 CFR 71.
l The auditor's review of this area revealed that though there were
,
management approved procedures developed and implemented (i.e., RCP3-506, " Packaging, Labeling and Shipment of Radioactive Materials") that identified some elements required by 10 CFR 71.51, a complete program to meet all of the criteria was not established, nor was there any independent verification being made to assure that the applicable criteria were being satisfied.
Upon notification of this deficiency, the licensee committed to curtail all shipments of radioactive materials until a qualit: assurance program was estabished and implemented meeting the requirements speci-fled in 10 CFR 71.51.
To this end the auditor noted that the licensee has initiated action and intends to submit a description of the quality assurance program to the NRC in accordance with 10 CFR 71 prior to resuming shipping activities.
This finding constitutes an item of noncompliance with the ren,uirements of 10 CFR 71.51.
(50-317/80-09-02; 50-318/80-07-02)
The licensee has contracted a consultant to evaluate the licensee's solid waste processing program to determine what modifications will be needed to cope with anticipated restrictions in burial site requirements.
Up to this time, no changes involving a safety evaluation (in accordance with 10 CFR 50.59) have been made in the current system.
There is evidence that the licensee is actively attempting to reduce the output volume of radioactive solid waste by such efforts as restrict-ing the amount of consumable materials permitted in the controlled area, and using diaphram inserts in the steel drums used for the baling operation in order to increase loading by as much as 30% per drum.
The auditor observed that the licensee's sampling of radioactive material for shipment appears to be efficient and avoids unnecessary build-up and storage of waste material on-site.
4.3 Effluent Process Instrumentation The following references were reviewed to determine the status of the licensee in regards to instrumentation up grading requirements cited in NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations":
.
.
a.
NRC letter dated October 30, 1979 to All Operating Nuclear Power Plants, Subject:
" Discussion of Lessons Learned Short Term Require-ments" b.
BG&E letter dated February 29, 1980, Subject:
" Follow-up Action Resulting from TMI-2 Incident (Lessons Learned)"
c.
NRC letter dated April 7, 1980 regarding the same subject refer-ing BG&E's letter of February 29, 1980 and a previous site visit by Lessons Learned Evaluation Team personnel on February 19, 1980.
Item 2.1.8.b, Increased Range of Radiation Monitors, in NUREG-0578 identified the need to provide an extended range capability in installed equipment, particularly noble gas effluent monitors.
Reference (a)
clarified the recommendation of NUREG-0578 and further specified interim actions that were required until full implementation of the NUREG-0578 items could be accomplished.
These interim items (called Category A items) are items to be accomplished by January 1,1980.
Reference (a) indicated that the interim method for Item 2.1.8.b could include the use of portable instrumentation set in shielded collimators on the sampling line and calculated conversion factors to relate dose rate to concentration.
The method was required to provide for a capability to obtain radiation readings at least every 15 minutes during an accident, and was to include procedures for the calibration of the instrument as well as dissemination of the information provided by the equipment.
The licensee was also required to submit a description of his method including information concerning instrumentation, background correction j
factors, technique to be employed for making measurements, instrument
[
power sources, etc.
Reference (b) in response to the Lessons Learned Requirements stated:
"The noble gas release rate will be determined by means of a portable i
'
dose rate instrument.
Data will be obtained to correlate the expected release rate with dose rate determinations at preselected points.
These instruments will be maintained in the Control Room Emergency Kits (CREK) for implementation."
Appendix A to this letter intended to provide the information required by the Reference (a) was essentially a description of the current Radiation Monitoring System as appears in the FSAR for Calvert Cliffs and did not meet the intent of the information reques.
.
.
.
To implement the licensee's interim plan procedure RCP-1-503, " Post Accident Sampling of Reactor Coolant and Containment Conditions" was issued December 31, 1979.
This procedure entailed the use of a PIC-6A dose rate instrument obtained from the CREK.
The procedure implies that in the event that the Main Vent RMS is off-scale, a person will take readings with a portable instrument at unspecified points on the RMS piping system, or main steam relief piping and report the informa-
tion to the Control Room.
Reference (c) accepted the icensee's method for all Category A items including Item 2.1.8.b.
The auditor's review of this area based on information available from the licensee indicated that in the event of an off-scale condition on the Main Vent RMS it is very likely the Auxiliary Building would be inaccessible due to high concentration of airborne radioactivity.
In such an event it would be necessary to utilize respiratory protection equipment to make entries, particularly to those areas where dose rate measurements are required in order to meet the interim specifications of Reference (a).
In the current plan an individual would have to make an entry every 15 minutes in order to collect the necessary information; a situation that can be easily overcome by utilizing portable remote indicating instruments which are readily available on site.
The auditor also noted that the licensee's information contained in Appendix A to Reference (b) (the FSAR description of the current radiation monitoring system) did not meet the intent of the information requirement.
The action / specification required a description of the equipment, method, procedures and qualifications for whatever mearts the licensee chose to utilize to fulfill interim high range monitoring.
Upon notification, the licensee initiated immediate action to develop
-
a radiation monitoring system having a high range capability using l
i available remote readout instruments, i.e., an RM-16 with appropriate probes located at field locations on the Main Vent System piping.
Upon completion of the installation the licensee committed to inform the ERC in accordance with the specifications of Reference (a), Item 2.1.8.6.
Further review of procedure RCP1-503 indicated that the document was
poorly written and provided little instruction to the user in the
'
event of an actual accident condition.
The procedure was largely a description of the licensee's intent for sampling in an accident
!
situation.
There were no functional instructions to the user pertaining l
to the accomplishment of tasks to be performed, such as reactor coolant i
sampling and analysis.
In addition, the procedure did not provide a means to record pertainent information that would be necessary if the
{
.
.
.
.
licensee utilized such procedure.
For example, the procedure only specifies that the valve lineup for the collection of primary coolant be " appropriate".
There is no reference to a prerequisite procedure, nor any other instruction that assures that the valve line-up is correct. Additionally, the technique for reactor coolant sampling requires diluting the collected coolant in the Miscellaneous Waste Receiver Tank (MWRT) without any information collected as to the prior status of MWRT (concentration and volume of liquid in the tank), and without any means to adequately measure the volume of coolant that is allowed to drain to the tank.
Upon notification, the RSCE indicated that the procedure would be reviewed and amended as necessary to assure that it is able to adequately provide necessary instruction and information to the user personel.
Evaluation:
1.
Responsibilities and authorities of the person currently assigned as responsible for the implementation of the Radioactive Waste management activity is not formalized or documented as previously mentioned in Section 1.2.2 of this
,
report, nor is the position identified in any organizational structure.
l l
The position, as it appears currently, is largely "ad hoc" and is not
!
established as an essential element of the licensee's program.
Currently, the effectiveness of licensee's current effort in this area is due primarily to the self-motivation of the involved personnel.
Based on the above findings, this portion of the licensee's program appears to be acceptable, but the following matter should be considered for improve-ment of the program:
(
Formally establish the position of Radioactive Waste Engineer or
--
equivalent within the organizational structure of the Radiation
,
l Protection Program.
Such establishment should include a description
of the responsibility and authority charged to the position.
l Additionally, it should provide assurance that the designated individual has the necessary qualifications and training to adequately perform and manage this activity.
l l
2.
In the case of an emergent condition at CCNPP, RCP1503 would become a procedure important to the evaluation and analysis of the extent and hazard of an accident.
The procedure as written does not provide the type and quality of instructions that would be necessary to assure safe and adequate performance of the sampling activities that are to be performed. Additionally, the technique utilized to establish release rates from the Main Vent in the event the installed RMS is off-scale is not in accordance with the intent of the " Lessons Learned" recommendations and would cause unnecessary exposure if performed.
t
--_-
-..
- -
.
. -
-
.
.
_ _ _ _
... -.
..
--
.
,
Based on the above findings improvement in the following area is required to achieve an acceptable program:
RCP 1603 is required to be revised so as to adequately provide
--
instructions as necessary to implement the licensee's techniques for performing in accordance with Sections 2.1.8.a. 2.1.8.b and 2.1.8.c of NUREG 0578.
Such revision is to incorporate meaningful and complete instructions sufficient for the user to complete the operation satisfactorily.
Precautions and prerequisites should I
be included.
!
The technique for performing the action to meet 2.1.8.b (Noble
--
Gas Monitoring, High Range Capability) shall be revised to assure
,
that unnecessary personnel exposure is not required. Methods, such as a remote read-out from a fixed probe location shall be considered.
5.0 ALARA PROGRAM An acceptable ALARA program is outlined and described in NRC Regulatory Guide 8.8 "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as Reasonably Achieve-able." Discussions were held with R/C Technicians, Foremen, and the Radia-tion and Chemistry Engineer.
Reviews were made of RWPs and SWPs.
The computer program " Work Classification Routine Maintenance Data Reported for Individual Work Permits" was also reviewed for its content and use. Areas that were addressed included installation of shielding to reduce dose rates in worker environments, use of mobile filtered ventilation systems to decrease e
airborne containments, decontamination of work places to minimize airborne radioactivity and dose rates, and interrogation of workers' supervisors
,
prior to issuance of RWPs and SWPs.
The auditor could not find any evidence of the establishment of an ALARA policy by either corporate or plant management.
Goals for implementation of ALARA and for achievement of ALARA had not been set by the plant manage-ment and individual (s) responsible for an ALARA program had not been identi-fled.
The ALARA program seems to be non-existent in a formal sense although some practices related to ALARA considerations were implemented, such as:
1)
Temporary shielding is used in some work areas.
2)
Blowers are used to ventilate air through absolute filters.
3)
Areas are decontaminated to reduce radiation exposures.
4)
R/C does discuss RWPs and SWPs with work force prior to starting work.
I
_
-g.
qi g.
,. _. - -..-,i
,,,y.-,,.,_,_%,
..,pwe,_a w,,,,,,,,,, -,
,,,,_ _
py.,,. _,,,,,,,, _ _, _,,
,_,,_.,___.,.,_..c_y
,.,
.,,.
..y
-,-9_
r_-.
q
..%
w
.
.
However, it was found that evaluation of ALARA practices and documentation of results achieved is non-existent.
Such evaluation would in fr.t determine whether the instituted practice (s) was in fact reducing radiation exposure as low as reasonably achievable.
Based on the above findings, this portion of the licensee's program requires improvement in order to achieve an acceptable program, in that a formalized ALARA program as described in Regulatory Guide 8.8 is required to be estab-l lished, implemented and maintained.
6.0 HEALTH PHYSICS FACILITIES AND EQUIPMENT The radiation protection and chemistry facilities reviewed included the training facilities, counting rooms, calibration facilities, decontamination facilities, instrument repair, and Radiation Protection Office, change areas and operations control points.
The training facilities include adequate lecture rooms, audible and visual aides and are located in both a warehouse and administration building.
The counting rooms are adequately furnished and sufficient equipment exists.
The in'strument calibration facilities consisted of a "well" type gamma irradiator (Cs 137) located in a room next to the counting room, a neutron l
source (Am 241-Be) for neutron calibrations and beta plaque (depleted uranium) for beta calibrations.
The decontamination facilities include separate areas for personnel decontamination, laundering of respiratory protective equipment, and equipment and tools decontamination.
Station protective clothing is laundered offsite at a vendor's facilities.
Access control to the Auxiliary Building is through the Radiation Control Point. At this location dosimetry checks are made prior to entry, RWP/SWPs are also posted for review by radiation workers.
Access control seems to be adequate during routine operations.
However, it was apparent that during outages the adequate handling of personel entering or exiting would be difficult and would interfere with adequate RWP/SWP indoctrination of l
the radiation worker.
Plant personnel have recognized this problem in that plans have been proposed to alleviate the congestion during outages, but at
'
present only on a short term basis.
The auditor was advised in the exit meeting that a permanent plan was being considered.
The calibration facilities utilized a Cesium 137 source referenced to NBS by an NBS calibrated Victoreen Condenser "R" Chamber.
The range of the gamma calibrator was up to 24 R/ hour.
Beta calibrations were accomplished using a depleted uranium slab; open window (OW) and closed window (CW) responses of the instruments were deter-mined.
A beta factor (BF) was determined as being equal to the l
.
.
.
-
.
OW - CW _
1_
SBDR BF where SBDR is the source beta dose rate.
This BF is set for each instru-ment and is used universally for all source geometries and source to instru-ment distances.
However, a second correction factor should be applied related to source geometry, beta energies and distances.
Also, a corre-lation of the response of the beta personnel dosimeter to that of the survey meter should be conducted and a determination made of the correct beta dose to apply to personnel exposures.
Personnel decontamination is carried out in the First Aid Room, usually under the direction cf R/C and the plant Physician's Assistant.
The equipment available and procedures seem to be adequate.
Laundering of protective clothing is acccmplished offsite in Utica, New York by a licensed vendor.
Laundering of personnel clothing may be done on site in the available laundry facilities.
The equipment available and procedures being used seem adequate.
Decontaminatio' of equipment and tools are accomplished onsite.
The equipment available available and procedures in use appear adequate.
Communications in containment have been deemed troublesome as a result of discussions with R/C personnel.
Adequate communications should be available between R/C monitor control points and the HP Office for adequate radiation protection coverage.
Scenarios can be established where lack of communica-
,
l tion could lead to unnecessary radiation exposures.
Based on the above findings, this portion of the licensee's program appears to be acceptable but the following matters should be considered for program improvement:
l Evaluate the radiation protection facilities deemed necessary for I
--
functionability during abnormal plant operations and provide for those changes deemed necessary.
Provide beta response factors of survey instruments and personnel
--
monitoring devices to beta sources of various energies and geometries.
l l
.
.
.
,
7. 0 ADMINISTRATION OF THE EMERGENCY PLAN The Plant Health Physicist at the Calvert Cliffs site acts in the capacity of an Emergency Planning Coordinator (EPC).
This assignment is not formal.
Emergency planning functions are not included in the individual's job description and there has been no assignment of the emergency planning activities by memorandum or letter from station management.
The auditor determined that emergency planning activities were assumed by the Plant Health Physicist by simple " default".
Recognizing that emergency planning activities were important and that no one was assigned, the Plant Health Physicist does what is necessary to " fill the void".
Discussions with the Plant Health Physicist indicated that several months ago a draft job description was circulated for an Emergency Planning Coordinator position.
The job description was reviewed and commented upon by personnel at the site, but nothing had been heard regarding the position since that time.
The individual who has assumed responsibility for erergency planning also serves in the capacity of Plant Health Physicist and emergency planning represents only a small portion of his duties.
Regarding the performance of emergency planning duties, the individual stated that there is no clear delineation of the authority or responsibility which he or anyone else has been given or may exercise to carry out the emergency planning function.
He was of the opinion, however, that persons at the site per-ceived that he acts with the approval and authority of management in relation to the performance of emergency planning activities.
These discussions also indicate that the informal assignment of planning respon-sibility in conjunction with the general level of staffing at the facility have, to some degree, hampered the ability of the Plant Health Physicist to maintain the emergency planning scheme in as high a state of readiness as he believes is necessary.
The work 1c..
of the Plant Health Physicist and of other elements of the site organization in relation to their primary duties make it difficult for them to accomplish all necessary emergency planning activities in as timely or complete a manner as is desirable.
Discussions with the Plant Health Physicist indicated that there is also no assignment of emergency planning responsibility at the corporate level.
The Plant Health Physicist was of the opinion, however, that the Manager of Quality Assurance essentially is his corporate counterpart for emergency planning.
Based upon the findings in the above area, the following improvement is needed in order to achieve an acceptable program:
Formally assign an individual to perform the Emergency Planning Coordinator function. This assignment should include a clear description of the individual's responsibilities and confer the degree of authority necessary to ensure that the assigned responsibilities can be efficiently and effectively carried out.
-
_ -
-
._
.
o
.
.
8.0 EMERGENCY ORGANIZATION 8.1 Onsite Organization The licensee's onsite emergency organization consists of the following areas of emergency activity with the assignment of individuals to the functional areas as noted:
FUNCTIONAL AREA 0F EMERGENCY ACTIVITY PERSONS ASSIGNED Site Emergency Director (SED)
Shift Supervisor Chief Engineer Plant Engineer (Maintenance)
Plant Engineer (Operations)
Performance Engineer Emergency Radiation Team Leader (ERTL) Radiation Safety & Chemistry Engineer Radiation Safety & Chemistry Foreman RMP Qualified Plant Operator Emergency Radiation Team (ERT)
Radiation Safety & Chemistry Technicians Emergency First Aid & Decontamination Physician's Assistant - Medical Team Leader Services Radiation Safety & Chemistry Foreman RMP Qualified Plant Operator l
Emergency First Aid & Decontamination Radiation Safety & Chemistry Team (EFA&DT)
Technicians Emergency Security Team Leader Assistant General Supervisor, Nuclear Security Foreman Emergency Security Team Security Officers Assembly Area Leaders On-duty Group Supervisors Plant Staff Engineers Fire Brigade Leader Fire Warden Shift Supervisor Senior Control Room Operator Fire Brigade Operations Staff Recovery & Restoration Team Leader Chief Engineer Nuclear Plant Engineer (0ps)
.
.
.
.
.
FUNCTIONAL AREA 0F EMERGENCY ACTIVITY PERSONS ASSIGNED Recovery & Restoration Team Members Persons qualified in:
(ER&RT)
Reactor & Plant Ops Maintenance Radiation Protection Emergency Reentry Teams Members of:
ERT EFT EFA & DT ER & RT The auditor performed a detailed analysis of the licensee's emergency organization to determine if the organizational structure was adequately constituted to effectively respond to the spectrum of emergency condi-tions for which the emergency plan was designed.
There were provisions for an emergency coordinator onsite at all times. (The Calvert Cliffs Site Emergency Plan uses the title Site Emergency Director, SED, in lieu of the term Emergency Coordinator).
,
i The line of succession for the SED position is specified, and while the SED's duties are described in the Implementation Procedures, the extent of the SED's authority to act in behalf of the company during emergencies is not clearly specified.
It was also noted that those responsibilities assigned to the SED which may not be delegated to other element of the emergency organization were not specified.
The concept of operations for site personnel during periods of minimal staffing is reflected in the content of the Implementation Procedures.
This concept indicates that the shift crew would perform initial detection, classification, assessment of radiological consequences, and call-in of personnel to augment the shift crew. More detailed assessment would be performed by augmentation personnel.
Within the scope of the functional areas of emergency activity, the auditor noted that the concept of operations for one area in particular, the Emergency Radiation Team, did not appear to provide an adequate
,
l description of the potential activities that the ECr would be expected l
to perform during a serious emergency.
Discussions with the Plant Health Physicist indicated that the members of the ERT would be expected
,
to perform all of the following activities.
l l
.
-
.
,
Offsite radiation surveys Onsite (out-of plant) radiation surveys In-Plant radiation surveys Personnel monitoring Personnel decontamination Radiochemistry Reentry team HP coverage Emergency First Aid Radiation Protection functions (e.g., dosimetry, access controls, etc.)
Recovery of Environmental TLDs The auditor discussed the scope of these duties with an individual assigned as an ERTL to determine the number of individuals who may be needed to accomplish the aforelisted functions.
This discussion indicated that the composition of the ERT and its operational concept would result in an insufficient number of people to cover all the team's emergency functions in response to a serious emergency.
The auditor determined that a number of the assigned emergency functions were not well-covered within the scope of the emergency plan imple-menting procedures.
(This is discussed in greater detail in Section 11.) This shortcoming was partially a result of the non-specificity of the ERT organizational definition.
There was no clearly designated management structure for each of the various functional areas to be covered by the ERT members.
In this regard, it was noted that Radiation Safety & Chemistry Supervisors and Foremen were only listed as alternates for the ERTL and were not used to provide a supervisory or oversight function for the sub-elements of the ERT's functional activities.
Also, it was noted that the Plant Health Physicist was not integrated into the emergency organization.
l l
The organizational structure of the ERT in conjunction with the assigned i
functional responsibilities is such that during a serious emergency, the ERTL has a span of control, of over 10 functional areas which have interlocking interests.
This span of control would, under normal conditions place a manager under stress.
During emergencies, a span i
of control of 10 presents a substantial potential for poor communication and a degraded response. This situation is complicated even further by the assignment of offsite dose calculation performance and communica-l tion of related information to various offsite agencies to the ERTL.
l In discussing the organizational structure with other members of the
!
emergency organization, the interfaces between and among the onsite functional areas of emergency activity appeared to be understood in relation to short-duration events.
This understanding, however, was less clear in relation to a serious emergency in which functional
,
I needs would arise for which the existing organization is not structured to cope.
This cloudy understanding was especially acute in the area of radiation protection and associated activities which would fall upon the ERT.
-.
.-
-
-
_ _ -.
.
o
O
The auditor determined that the organization was adequately constituted to cope with emergencies up to and including the Plant Emergency category, but would be stressed in response to a serious emergency of the Site or General category involving internal areas of the plant as well as offsite areas.
Based upon the review in the above area, the licensee's emergency organization appears to be adequate to cope with emergency situations of relatively short duration, i.e., 2-4 hours in length.
In order to adequately cope with more serious emergencies and emergencies of longer duration, the following improvements are needed to achieve an acceptable program:
Creation of a management structure for oversight of environmental
--
aspects of response (e.g., offsite and onsite surveys and environ-mental sample collection).
Designation of a management oversight structure for in plant
--
aspects of a response related to radiation protection (e.g.,
radiation protection, chemistry, in plant surveys, personnel monitoring, etc.).
8.2 Augmentation of Onsite Emergency Organization The definition of the integration of corporate resources into the augmentation of the onsite emergency organization is limited to the position of an Emergency Coordinator who performs notifications of various offsite agencies end corporate elements.
There is no clear integration of command and control structure or description of inter-face with the site organization for any corporate element other than the Emergency Coordinator.
It was also noted that there were no provisions for augmentation of the site organization by the various contractors or vendors who may be needed during a serious emergency.
The extension of the organizational capability to be provided by local services support, i.e., ambulance, medical, hospital and fire fighting were clearly integrated into the licensee's augmentation scheme.
Based on the findings in the above area, the licensee's program appears consistent with the existing Emergency Plan, however, the following matters should be censidered for improvement:
The corporate personnel or elements, contractor, vendor and
--
private organizations who will augment the plant staff and their areas of augmentation should be delineated.
_ _ -.
-
-
.
.
.
.-.
.
41 9.0 EMERGENCY TRAINING / RETRAINING The auditor reviewed the licensee's program for training members of the emergency organization to determine the adequacy of the program to prepare response personnel to perform their assigned emergency duties. The basic program is documented in Calvert Cliffs Instruction (CCI) 611C dated 4/21/77.
Training for members of the emergency organization is required on an annual basis and attendance is documented through the use of Calvert Cliffs Train-i l
ing Memoranda (CCTM) maintained by the Training Coordinator.
Instructors for the various emergency training classes are the team leaders of the various teams.
Discussions with the team leaders indicated that there was a wide variance in the method, scope and content of training presented. Generally, instructors use lesson plans which they have prepared.
The lesson plans do not contain student performance objectives or a means to evaluate the ability of the individuals to perform their emergency functions.
In the case of ERT training, the designated team leader, ERTL, (Radiation Safety and Chemistry Engineer) has delegated the training responsibility to two of the alternate team leaders (a Radiation Safety and Chemistry Foreman and a Performance Engineer - Radiation Safety and Chemistry). Discussions with the two individuals indicated that the ERTL had never discussed, reviewed or coordinated the content of the training programs with the individuals involved.
Material presented is limited to the content of the applicable implementation procedures and has not included information on what might be expected under unusual plant conditions, e.g.,
components and I
I areas which would potentially have high radiation levels, expected magnitudes
!
of radiation increases, changes in nuclide composition, etc.
ERT training has also not included information on special surveillance under emergency conditions, e.g., use of equipment, interpretation of results, personnel access controls and special precautions.
In addition, the auditor noted that the current training program did not include two team activities performed by the existing organizatinn (Recovery (
/ Restoration and Reentry).
Training techniques employed vary considerably, from lecture to a combination of lecture and demonstration by the instructor.
In the case of the ERT, training has included tours of the various survey points noted in the procedures.
Training did not, however, include the opportunity for each individual to operate emergency sampling and survey equipment. In general, hands-on use of equipment by class participants with coaching and critique by instructors is limited and inconsistently applied as a training technique.
The current training program, as implemented, has resulted in the training of members of the ERT in changes to procedures and equipment which occur in the period between the routine annual training sessions.
These provisions,
-
____
.
,
however, were not formally established and appear to have only been applied to members of the ERT.
CCI 611C places the responsibility on the Training Coordinator for ensuring that all areas of SEP training are adequate.
It also charges the training coordinator with providing assistance to members of the CCNP staff in preparing training materials, scheduling of training sessions and maintain-ing records.
The method for ensuring that all areas of training are adequate is the conduct of an annual training review, with the results presented in an annual report.
A review of the latest report and discussion with the Training Coordinator indicated that the review is limited to simple verifi-cation that training sessions were conducted.
The review does not include evaluation or review of the adequacy or completeness of the training provided.
Discussions with assigned emergency planning instructors revealed that they had not availed themselves of the' training skills available in the Training Department.
Based on the findings in the above area, the licensee's emergency plan training program appears consistent with the existing Emergency P1&n, however, the following matter should be considered for improvement:
Development of formal lesson plans for each of the functional areas of
--
emergency activity performed by licensee and augmentation groups, to
,
l include clearly defined student performance objectives and provisions for evaluating attainment thereof.
'
10.
EMERGENCY FACILITIES AND EQUIPMENT 10.1 Emergency Kits and Instrumentation The licensee maintains pre positioned stocks of emergency supplies, some of which are in kit form, at various specified locations.
The auditor conducted an independent inventory and operational check of the licensee's emergency equipment and verified that emergency kits and equipment were located as specified in the Emergency Plan and Implementing Procedures, that inventories were correct, and that the
'
equipment was operable.
Equipment designated for emergency environmental air sample counting consists of the MS-2 with SPA-3 detector.
The collection and counting equipment configuration provide a capability to detect and measure radioiodine concentrations in air of at least 5x10E-8 uCi/cc.
Operability checks are routinely performed on all portable emergency instrumentation, and there is an onsite capability for filling self-contained breathing (SCB) devices.
l l
.
,
Based upon the above findings, this portion of the licensee's program appears acceptable.
10.2 Area and Process Radiation Monitors The auditor inspected the area and process radiation monitor readouts in the control room and noted that all required monitors were operable.
,
Inoperable monitors were properly tagged and scheduled for repair.
!
Based upon the above findings, this portion of the licensee's program appears acceptable.
10.3 Emergency Operations Centers
,
The auditor reviewed the licensee's provisions for locations from which emergency activities would be directed.
Initially, emergency activities are directed from the Control Room.
In the event of a Site or General Emergency, a primary Emergency Control Center is estab-lished in the Guardhouse with an alternate center located at the Farm Demonstration Building.
Based on the above findings, this portion of the licensee's program appears consistent with the existing Emergency Plan.
10.4 Medical Treatment Facilities i
The licensee has onsite provisions for treatment of individuals who may be injured and contaminated.
A First-Aid Room is located adjacent to the Controlled Area Washroom.
There are First-Aid kits located at various points throughout the facility.
,
l Based on the above findings, this portion of the licensee's program appears consistent with the existing Emergency Plan.
10.5 Decontaminaction Facilities There were provisions for personnel decontamination in and close to the onsite medical treatment facility.
In addition there are stores of decontamination supplies at the Farm Demonstration Building.
Based on the above findings, this portion of the licensee's program appears consistent with the existing Emergency Plan.
.
.
.
,
.
.
,
11.0 EMERGENCY IMPLEMENTING PROCEDURES 11.1 General Content and Format The general content and format of the procedures which implement the Calvert Cliffs Site Emergency Plan are characterized by vague wording and cumbersome mechanical arrangement.
Emergency Action levels and
.
Protective Action Guides (PAGs) are not clearly highlighted within the l
sections of the procedure, and while action steps are displayed in a step-by-step sequential fashion, the nature and content of the informa-tion contained in the steps is such that the procedures are of question-able value in guiding the user.
There is a scarcity of guidelines for areas in which the procedure user is permitted to exercise judgment and procedural steps which require other functions or jobs to be performed do not reference the interfacing procedures.
Procedures do have signoff sheets, checklists and data sheets to document details of actions taken.
Details of specific procedures are discussed in subsequent paragraphs.
11.2 Implementing Instructions The licensee utilizes a procedural approach in which there are not separate procedures for each class of emergency specified in the plan.
Rather there is a single, large Emergency Action Procedure (SEPIP A)
with an " Initial Action" section (Section I), which is followed by a
" Classification of Emergency and Evacuation Criteria" section (Section II).
The classification tables contained in this section do not orchestrate or reference the implementation of more specific pro-I cedures.
Section IV of SEPIP A consists of a single checklist with 23 i
steps for use by the Site Emergency Director.
This checklist is utilized for all emergencies regardless of classification.
Based or, the findings in the above area, the following improvement is needed in order to achieve an acceptable program:
Development of a separate procedure (implementing instruction)
--
l for each class of emergency specified in the Emergency Plan which
!
specifies the EALs and pre planned response actions required to
!
be considered or implemented in response to each class of emergency.
11.3 Implementing Procedurcs a.
Notifications
!
The sequence of notification to alert or mobilize the emergency organization and supporting agencies is specified in Figures IA l
l t
.
.
,
and 18 and phone numbers are in SEPIP C, Communications.
There is one notification sequence for Local or Plant Emergencies (Figure 18) and another for Site or General Emergencies (Figure 1A).
In general, notification provisions do not specify the action levels for notification of various participating organizations.
Figures 1A and 18 of the licensee's Emergency Plan Implementing Procedures include vague statements such as "if required" or
"according to conditions" regarding notifications to the NRC, Maryland Health Department, and Calvert County Commissioners.
This vagueness could result in the failure to notify supporting groups in a timely manner or create a "second guessing" atmos-phere of who should be notified and when.
The auditor noted that the principal method for notifying the NRC reflected in the plan was via commercial telephone by the Emergency Coordinator.
This notification sequence is not consistent with existing NRC regulations which requires notification of the NRC via the Emergency Notification System (ENS).
The licensee does not use pre planned messages in performing initial notification of supporting agencies.
The notification procedures also do not include an authentication scheme for offsite agencies.
Based on the findings in the above area, the following improve-ments are necessary to achieve an acceptable program:
Clarification of the action levels which will result in
--
notification of the elements of the response organization, thereby eliminating "if required" or "according to con-ditions" statements.
Updating of notification procedures to reflect use of the
--
ENS for notification of the NRC.
l b.
Radiological Surveys - Offsite, Onsite and In-Plant The licensee's provisions for conducting offsite and onsite environmental surveys and in plant radiation surveys under emergency conditions are inadequate.
The methods and equipment to be used are not specified.
Discussions with licensee management indicated that reliance is placed on the individual's experience and training to know the proper methods and equipment.
Discussions with emergency team members indicated that the individuals them-selves had the same understanding.
The auditor determined that I
the combination of procedure content and experience level of the
.
g---
,,- - -,,.
-, -, -
-.-,--g
,,
e
-
w+
n
-.e-
,
m,-
, -
.
'
.
,
individual who would perform radiation surveys would be adequate for short duration emergencies in which the radiological profile of the plant and environment do not change considerably.
For more serious accidents involving dramatic increases of radiation levels and radionuclide compositions different than the norm, the procedures fail to provide a reasonable degree of assurance that the techniques employed or data collected would be adequate to serve as a basis for assessment and protective action activities.
I i
Data sheets available for recording survey results do not contain key elements such as the date and time of the surveys, survey location, instrument mode (e.g., window open/ closed), background levels, etc.
There are no provisons for labeling environmental camples or for a central collection point for samples and original field data.
Based upon the findings in the above area, the following improve-ments are necessary to achieve an acceptable program:
Development of specific procedures governing the performance
--
of offsite, onsite and in plant radiation surveys to include:
specification of the methods and equipment to be used.
-
data sheets
-
'
provisions for labeling samples
-
-
cautions / precautions to be observed c.
Personnel Monitoring and Decontamination Procedures related to personnel monitoring and decontamination
were equally vague.
SEPIP A, Page A17, Paragraph 3.e, states
" Man site evacuation checkpoint to monitor and assist." As in the case of radiological surveys, the methods and equipment to be used are not clearly specified, provisions for data recording are not delineated and contamination action levels are not specified.
Based on the above findings, the following improvement is necessary to achieve an acceptable program:
Development of procedures governing the performance of
--
personnel monitoring and decontamination.
_
'
.
,
d.
Evacuation of Onsite Areas Section A-III, Pages A-5 and A-6 list evacuation criteria. Initial assembly areas are specified, however, reassembly areas for post-evacuation are not.
Based on the above findings, this portion of the licensee's program appears acceptable, however, the pre-selection and designa-tion of post-evacaution assembly areas should be considered for improvement.
e.
Assessment Actions The licensee's system for gathering information upon which to base decisions regarding corrective or protective actions is described in SEPIP G and Section V of SEPIP A.
There are provisions for calculating initial offsite exposures and magnitudes of radioactive releases. Within the assessment scheme there were no provisions for trend analysis of assessment data or for the comparison of results with action levels adopted by the state of Maryland.
While such comparisons appear to be made during the classification process (SEPIP A,Section II), the continued evaluation beyond the initial assessment appears weak.
Based on the above findings, this portion of the licensee's program appears acceptable, however, the following item should be considered for improvement:
Development of an overall assessment procedure which would
--
integrate and coordinate all the sources of assessment data
,
in a cohesive manner.
f.
Radiological and Environmental Monitoring Program (REMP)
The licensee does not have well-developed plans for implementing an accelerated or special environmental monitoring program during emergencies, e.g., environmental TLDs, soil and water samples, etc.
Current provisions for environmental monitoring are limited to air samples and area radiation readings taken by the Emergency Radiation Team and a one line statement, " collect TLDs from environmental monitoring stations as directed."
Based on the above findings, the following improvement is needed in order to achieve an acceptable program:
Development of provisions for an emergency REMP.
--
.
-.
_ _ _.
_
, __
.
,
g.
Onsite First Aid / Rescue Procedures governing the conduct of onsite first aid or rescue are described in SEPIP A,Section V.C, Emergency First Aid and Decontamination, and SEPIP A,Section VI, Personnel Injury.
The general methods for receiving, recovering or transporting and handling an injured person who may also be contaminated are described in general terms.
The interface and action levels for use of the offsite medical treatment facility are also covered.
Based upon findings in the above area, this portion of the licensee's program appears acceptable.
h.
Security Durina Emergencies SEPIP A,Section IV.0, Emergency Security Team, details actions to be carried out by members of the security force to implement security measures during emergency conditions.
Duties assigned to the Emergency Security Team include such activities as manning evacuation checkpoints, verifying personnel or vehicles are monitored prior to departing the site, logging persons on and offsite, and overall coordination of personnel accountability and status reports.
Based upon the findings in the above area, this portion of the licensee's program appears acceptable, however, the following item should be considered for improvement.
Emergency security procedures should be reviewed to assure
--
they have been developed in accordance with the requirements of Appendix C to 10 CFR 73 and that they complement the radiation emergency plan.
1.
Radiation Protection During Emergencies Section 8 of this report discusses an organizational shortcoming
'
concerning the performance of radiation protection functions during emergencies.
This organizational omission is also reflected within the scope of Emergency Plan Implementing Procedures.
There were no Emergency Plan Implementing Procedures directly relating to such radiation protection functions as personnel dosimetry, exposure records, the need for positive access controls, instructions to emergency workers, dose assessment of individuals who may be exposed or clear provisfors for preventing re-exposure of individuals or limiting further exposure.
In this regard, t5e auditor also reviewed the licensee's routine radiation protection procedures and noted that these procedures did not reflect their applicability during emergency situations.
Discussions with
.
.
,
Ifceasee personnel indicated that special controls would be implemented for emergency conditions but that these conditions and considerations had not been fully developed.
Based upon the findings in the above areas, the following improve-ments are needed in order to achieve an acceptable program:
Development of procedures governing the continuity of the
--
radiation protection program (e.g., access control, dosimetry, etc.) under emergency conditions to include considerations for changing and unusual conditions such as higher doses, different energies or changing radionuclide compositions.
j.
Repair / Corrective Actions The licensee does not appear to have a procedure governing repair or corrective actions.
SEPIP A,Section I.E, Emergency Reentry Team, discussed general concepts to be followed by any teams which may reenter the facility. Within the scope of Emergency Plan Implementing Procedures, there was not a clear description of the concept of operation or composition of repair or corrective action teams, the individuals to whom the team will report, and the steps to assure that individuals are properly briefed as to the radiological conditions, stay times, etc.
Based upon the findings in the abcve area, the licensee should develop a procedure governing the actions of repair or corrective action teams in order to achieve an acceptable program.
11.4 Supplementary Procedures a.
Inventory Operational Check and Calibration of Emergency Equipment, Facilities and Supplies SEPIP 8 of the licensee's Site Emergency Plan contains listings of emergency equipment to be maintained at various locations throughout the facility.
This document is essentially not a procedure in that all it contains is an inventory listing.
There were no instructions governing the method by which the various types of checks and inventories are to be conducted.
Based upon the above findings, in order to achieve an acceptable program the licensee should develop a procedure which details the method by which the various items of emergency equipment are operation-ally checked and inventoried to assure that all emergency equipment is maintained in a high state of readiness.
)
,
-
)
.
,
i b.
Drills The licensee's provisions for conducting drills are detailed in Procedure CCI 611C.
The auditor noted that this procedure is also the training procedure for the emergency plan training program.
Overall responsibility for scheduling and conducting drills is assigned to the Calvert Cliffs Training Coordinator. The auditor noted that the drill procedure did not contain provisions for the preparation of a scenario in advance of the drill.
In addition, while an attachment to the procedure contained an emergency plan drill critique sheet, the procedure itself did not describe the provisions to be used for documenting and evaluating observer and participant comments as part of the drill.
There were also no provisions in the procedure for ensuring that management controls are implemented to assign responsibility for corrective action and completion dates, and for review to assure that assigned corrective actions are completed in accordance with established schedules and are adequate to resolve any noted deficiencies.
Based upon the findings in the above area, the auditor determined that the licensee's present drill procedure was inadequate and the following improvements are needed in order to achieve an acceptable program:
Provisions should be clearly delineated for the use of observer
--
staff for all drills.
--
Provisions should be detailed for evaluating and assuring ade-quate resolution of items highlighted during the drill as needing improvement.
l
_
_
_..
.
_
-.
.
.
. _ _.
"
~
.
,
ANNEX A Exit Meeting and Licensee Commitments 1.0 On May 23, 1980, the Health Physics Appraisal Team met with the licensee representatives, denoted in ANNEX B, to discuss the scope and findings of this appraisal.
In the course of that discussion, the following plans for program improvement were identified by the licensee.
1.1 Prior to resuming shipments of radioactive material, a quality assurance program sufficient to satisfy the criteria of 10 CFR 71.51 shall be developed and implemented.
1. 2 Instrumentation used for providing surveys to determine neutron dose equivalent will be subject to calibration utilizing a neutron cource adequate for the purpose of providing such calibration.
The calibration procedure shall be revised as necessary to assure that adequate cali-bration is performed on a periodic basis.
Such action will be com-pleted prior to permitting personnel to be subjected to exposure from neutron radiation.
Other actions pertaining to this item are documents.
1.1 In accordance with the letter from Mr. B. H. Grier, Director, NRC Region I (Philadelphia), to Mr. A. E. Lundvall, Jr., Vice President, Baltimore Gas and Electric Company, dated May 22, 1980, the licensee will take the necessary corrective action to assure that appropriate controls are applied regarding personnel exposure to neutron radiation.
1. 2 RCP-1-503 will be revised as necessary to assure that it provides sufficient direction to personnel performing sampling activities in post-accident conditions.
1.3 The licensee's approach to meeting the requirement of Lesson Learned Item 2.1.8.b will be amended by providing for a remote monitoring technique to assure high range capability to the existing monitoring system.
.
'
.
,
j ANNEX B Persons Contacted The following personnel were contacted in the course of the Health Physics Appraisal:
- Mr. L. Russell, Plant Superintendent
- Mr. E. Riemer, Plant Health Physicist
- Mr. A. Kaupa, Radiation Safety and Chemistry Engineer
- Mr. T. Sydnor, Supervisor, Quality Assurance
- Mr. J. Carlson, Foreman, Radiation Safety
- Mr. J. Speciale, Foreman, Chemistry
- Mr. P. Crinigan, Performance Engineer Mr. J. Lenhart Technician, Radiation Safety and Chemistry (RSC)
Mr. R. Cauldwell, Technician, RSC Mr. S. Cherry, Technician, RSC Mr. E. Roach, Technician, RSC Mr. R. Sprecher, Technician, RSC Mr. L. Hopson, Technician, RSC Mr. S. Hutson, Tecnnician, RSC Mr. G. Phair, Technician, RSC Mr. T. Guf f, Technician, RSC Mr. R. Denton, Nuclea.- Plant Engineer Mr. E. Earts, Performance Engineer
- Denotes those persons attending the exit meeting on May 23, 1980.
The appraisal team interviewed several other personnel, including tech-nicians, members of the plant operations, maintenance and quality assurance staffs, and personnel assigned to the plant personnel training department.
.
-*