ML20030B328
| ML20030B328 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 07/31/1981 |
| From: | Davidson D CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8108060335 | |
| Download: ML20030B328 (57) | |
Text
"
I THE CLEVELAND ELECTRIC ILLUMIN ATING COMPANY P o box 5000 m CLEVEL AND. OHIO 44101 m TELEPHONE (216) 622-980s a ILLUM:NATING BLDG e 55 PUBLIC SQUARE Serving The Best Location in the Nation Dalwyn R. Davidson VICE FRE stDE NT 5 f STE M E NGINE ER NG AND CONST AUC TION e
July 31, 1981 J'
A
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(1.
6 AUG 0 5198W t.9 1 Mr. Robert L. Tedesco u.qj.
Assistant Director for Licensing f
Division of Licensing
/h U. S. Nuclear Regulatory Co= mission Washington, D. C.
P0555 Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Response to Request for Add..tional Infor=ation -
Padiation Protection
Dear I/x. Tedesco:
This letter and its attachment is submitted to respond to your letter dated June ll, 1981 concerning radiation pro-tection and radiological assessment. The requested infor-
=ation is provided as draft responses for discussion with your staff in the meeting scheduled for August 19, 1981.
It is our intention to incorporate these responses in a subsequent amendment to our Final Safety Analysis Report.
Very truly yours, 2
Dalwyn.. Davidson Vice President System Engineering and Construction DRD:dlp Attachment cc:
G. Charnoff NRC Resident Inspector ool p$%
8108060335 810731 L)
PDR ADOCK 05000440 A
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r 471.06 In the Perry Nuclear Power Plant (PNPP) organization, Figure 13.1-3 (13.1) the Supervisor, Health Physics Unit (Radiation Protection Manager (RPM) in Regulatory Guide 8.8), reports to the Radiation Protection Section General Supervising Engineer, who reports to the Plant Manager.
The RPM should have direct access to responsible management personnel and be independent of operating divisions as specified in Regulatory Guide 8.8, Section C.1.b(3) and in NUREG-0731 " Criteria for Utility Management and Technical Competence." Section 13.1 or Section 12.5 of the FSAR should be revised to show that in health physics matters, the Health Physics Supervisor has direct access to the Plant Manager.
Response
The Radiation Protection Section General Supervising Engineer is the Radiation Protection Manager (RPM) in Regulatory Guide 8.8.
See revised Section 13.1.2.2.
General Supervising Engineer, Technical Section The General Supervising Engineer, Technical Section is responsible for directing all activities associated with providing technical support and services related to monitoring plant performance, equipment and system testing, instrument maintenance, calibration and repair and reactor technology. He is also respons?ble for the prograaming, operation and maintenance cf the process computer and related software developmect. The General Supervising Engineer, Technical Section is a member of the Plant Operations Review Committee and reports to the Superintendent, Plant Operations.
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G General Supervising Engineer, Radiation Protection Section g
s a
The General Supervising Engineer, Radiation Protection Section is designated as 33 O
the Radiation Protection Manager (RPM) and is, therefore, responsible for ALARA program coordination and directing all activities associated with the chemical, h
radiochemical, radwaste and other radiological control services required to support plant operation and maintenance activities. This includes conducting laboratory and plant survey activities required to ensure that personnel exposure to radiation and radioactive materials is within regulatory guidelines and that such exposure is kept as low as reasonably achievable (ALARA). The General Supervising Engineer Radiation Protection Section is a member of the Plant Ope.at'ons Review Committee, ALARA Committee, and reports to the Plant Manager.
General Supervisor, Nuclear Services Section The General Supervisor, Nuclear Services Section is responsible for directing all training activities required to develop and maintain a qualified workforce, and for insuring that all provisions of the PNPP Security Plan are implemented. He is also responsible for providing the necessary general maintenance and administrative services required to effectively support plant activities. He reports to the Plant Manager and, via the Security Supervisor, maintains a communications link with the corporate Security Advisor.
i 13.1-14
471.07 Based on information contained in NUREG-0731 " Criteria for Utility
)
(13.1)
Management and Technical Competence," it is our pcsition that your organization chain contain a qualified health physicist to provide backup in the event of the absence of the Supervisor, Health Physics.
The December 1979 revision of ANSI 3.1 specifies that individuals temporarily filling the RPM position should have a B.S. degree in science or engineering, two years experience in radiation protection, one year of which should be nuclear power plant experience, six months of which should be onsite.
It is our position that such experience be professional experience.
Identify and provide an outline of the qualifications of the individual who will act as the backup for the RPM in his absence.
Response
i The Health Physics Supervisor will act as the backup for the RPM in his absence. The following is his resume of experience and training:
Education Miami Dade Community College, A.A., PreEngineering, 1977 U.S. Army Engineer Reactors Group, Health Physics Speciality, 1965.
Experience I
l January, 1980 to Present: Health Physics Superviscr, Cleveland Electric Illuminating Co., Perry Nuclear Power Plant (2, 1205 MWe l
BWRs). Assist the General Supervising Engineer, Radiation Protection Se
'on in the development of the Radiation Protection Program for the Perry Nuclear Power Plant. This includes facility l
and equipment reviews, procedure preparation, and developing Health Physics Technician staffing and training. Assigned to the Edison Electric Institute's Health Physics Committee, 1981.
July 1977 - January 1979: Station Health Physicist, Houston Lighting & Power Company, South Texas Project (2-1250 MWe PWRs) and Allen's Creek Project (1-1200 MWe BWR). Assist in the development of a Radiation Protection Program for the company nuclear projects. This included writing of health physics procedures, developing health physics technician training courses, and performing ALARA reviews of facility and equipment design.
June, 1976 - June, 1977: Health Physics Shift Supervisor, Florida Power & Light Company, Turkey Point Plant (2-760 MWe PWRs). Responsible for Radiation Protection Men activities for radiological surveys.
Initiate, review, and authorize Radiation Work Permits; ship and receive radioactive materials; control of personnel exposure; direct health physics coverage for the following functions: refueling operations, steam generator eddy current testing and tube plugging, equipment maintenance, calibration and repair of systems instrumentation, radioactive laundry and waste processing; write and review health physics operating procedures.
May, 1974 - June, 1976: Health Physics Administration Assistant, Florida Power & Light Company, Turkey Point Plant. Responsible for all phases of personnel exposure records; write and maintain computer programs and associated files (Fortran IV); research and order health physics supplies and equipment; draft various health physics reports; conduct basic and refresher training in health physics fundamentals for vendor, maintenance, and operations personnel.
October, 1968 - May, 1974: Radiatioa Safety Office, The Ohio State University. Provide ra('iological health services to operations staff of the Nuclear Reactor Laboratory.
Perform analycis on water and air samples, conduct area and irradiated materials surveys.
Initiate and maintain environmental surveillance program for the
Nuclear Reactor Laboratory. Review and recommend radiological practicen for medical and reasearch applications for use of radioisotopes. Generally oversee Broad License requirements and ensure compliance with Nuclear Regulatory Commission regulations at The Ohio State Uciversity.
December, 1967 - October, 1968: Health Physics - Plant Chemistry Specialists, Plant Operator, SM-1 Nuclear Power Plant, Fort Belvoir, Virginia. Provide health physics services and advice to operating crew, plant management, and maintence personnel. Perform water analysis and treatment on potable, reactor associated, and secondary water systems. Operate equipment and plant components, participate in core loading, control red maintenance, and reactor physics testing.
e June, 1966 - December, 1967: Health Physics - Plant Chemistry Specialist, Plant Operator, PM-3A Naval Nuclear Power Plant, McMurdo Sound, Antarctica. Provide health physics services and advice to operating crews, plant management, and maintenance personnel.
Perform water analyis and' treatment on potable, reactor disposal system for liquid and gaseous wastes. Operate plant systems and components. Operate desalinization unit for producing fresh water from sea water.
Feb rua ry, 1965 - June, 1966: Health Physics - Plant Chemistry Instructor, Nuclear Power Plant Operator School, Fort Belvoir, Virginia. Give instruction in health physics theory and practices; counting and monitoring equipment theory and operation; radiochemistry, inorganic, qualitative, and quantitative chemistry; and gamma spectroscopy. Write lesson plans and student handouts for portions of these courses.
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471.08 Regulatory Guide 1.8, states "The RPM should have a bachelor's (13.1.3.2) degree or the equivalent in a science or engineering subject including some formal training in radiation protection" and at least 5 years of professional experience in applied radiation i
protection.
It is our position that equivalent as used in Regulatory Guide 1.8 for the bachelor's degree means (a) four years of formal schooling in science or engineering (b) four years of applied radiation protection experience at a nuclear facility, (c) four years of operational or technical experience or training in nuclear power, or (d) any combination of the above totaling four years.
From th( information submitted in Resume Number 27, we are unable to determine that the Supervisor, Health Physics Unit, has training and experience equivalent to that specified in Regulatory Guide 1.8.
Therefore, justify the selection -of the individual delineated for this postion based on his training and experience and specify, as required, how he wi,ll achieve the aforementioned experience or training, prior to the plant being licensed, to qualify as the RPM.
Response
l The RPM is the Radiation Protection Section General Supervising l
Engineer (resume 19 of Table 13.1-3).
The resume indicates he possesses a Bachelor of Science in Chemical Engineering and t
i therefore meets the requirement of. Regulatory Guide 1.8.
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471.09 Provide an outline of the function, responsibility, and (13.1.2.2) authority of the Perry Nuclear Power Plant's Health Physics Supervisor (RPM).
(See Regulatory Guide 8.8, Section C.1.b(3) for examples of some duties)
Response
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The General Supervising Engineer, Radiation Protection Section j
is the Radiation Protection Manager. An outline of his function, i
responsibilities, and authority is provided in revised Sections 12.5.1 and 13.1.2.2.
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12.5 HEALTH PHYSICS PROGRAM 12.5.1 ORGANIZATION The Radiation Protection Section General Supervising Engineer (RPSGSE) is responsible for implementing the plant (site) radiation protection program, which 9
encompasses the handling and monitoring of radioactive materials, including 7
special nuclear, source, and by product materials. He has responsibility for o
ALARA and assures that plant operation meets the radiation protection requirements of federal and state regulations which are applicable to the h:
radiation protection program. He reports to the Plant Manager, Perry Nuclear Power Plant (See Figure 13.1-3 and Section 13.1).
Reporting to the RPSGSE are the Health Physics, Radwaste and Chemistry Unit Supervisors. The Health Physics Unit Supervisor is responsible for the health physics portion of the radiation protection program.
The Health Physics Technicians who report to the Health Physics Unit Supervisor, perform the various surveys and analyses for health physics protection. One health physics personnel is provided for each shif t.
For a more detailed discussion of the responsibilities and authority of the supervisory positions mentioned above, and of the training and qualifications of the personnel holding these positions, refer to Sections 13.1.2.2 and 13.2.3 and Appendix 13A.
12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 12.5.2.1 Facilities All health physics and radiochemistry facilities are located on elevation 599'0" (B1) of the Control Complex. These facilities include the following rooms and areas:
a.
Personnel Decontamination Room b.
Medical Aid Room 12.5-1
General Supervising Engineer, Technical Section 1
i The General Supervising Engineer, Technical Section is responsible for directing all activities associated with providing technical support and services related to monitoring plant pe r fo rma n ce, equipment anti system testing, instrument maintenance, calibration and repair and reactor technology.
lie is also responsible for the programming, operation and maintenance of the process computer and related sof tware development. The General Supervising Engineer, Technical Section is a member of the Plant Operations Review Committee and reports to the Superintendent, Plant Operations.
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General Supervising Engineer, Radiation Protection Section c{
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The General Supervising Engineer, Radiation Protection Section is designated as es O
the Radiation Protection Manager (RPM) and is, therefore, responsible for ALARA program coordination and directing all activities associated with the chemical, h
radiochemical, radwaste and other radiological control services required to support plant operation and maintenance activities. This includes conducting laboratory and plant survey activities required to ensure that personnel exposure to radiation and radioactive materials is within regulatory guidelines and that such exposure is kept as low as reasonably achievable (ALARA). The General Supervising Engineer Radiation Protection Section is a member of the Plant Operations Review Committee, ALARA Committee, and reports to the Plant Manager.
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General Supervisor, Nuclear Services Section The General Supervisor, Nuclear Services Section is responsible for directing all l
training activities required to develop and maintain a qualified workforce, and for insuring that all provisions of the PNPP Security Plan are implemented. He is also responsible for providing the necessary general maintenance and administrative services required to effectively support plant activities. He reports to the Plant Manager and, via the Security Supervisor, maintains a communications link with the corporate Security Advisor.
13.1-14
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a 471.10 As recommended in Regulatory Guide 8.8, Section C.1.b(3),
(12.1.2) the responsibility and authority for implementing the plant's program for maintaining occupational radiation exposures ALARA should be assigned to an individual (or committee) with organizational freedom to ensure dec-lopment and implementation.
l Identfy by title the individual (s) responsible for the ALARA program coordination and describe how he (they) are placed in the organization, particularly the mechanism for communication I
with plant management.
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Response
f The response to this question is provided in revised Sections 12.5.1 and 13.1.2.2.
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c 12.5 HEALTH PHYSICS PROGRAM 12.5.1 ORGANIZATION The Radiation Protection Section General Supervising Engineer (RPSGSE) is responsible for implementing the plant (site) radiation protection program, which
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encompasses the handling and monitoring of radioactiva materials, including 7
special nuclear, source, and by product materials. He has responsibility for e2 o
ALARA and assures that plant operation meets the radiation protection requirements of federal and state regulations which are applicable to the
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radiation protection program. He reports to the Plant Manager, Perry Nuclear Power Plant (See Figure 13.1-3 and Section 13.1).
Reporting to the RPSGSE are the Health Physics, Radwaste and Chemistry Unit Supervisors. The Health Physics Unit Supervisor is responsible for the health physics portion of the radiation protection program.
The Health Physics Technicians who report to the Health Physics Unit Supervisor, perform the various survt ys and analyses for health physics protection. One health physics personnel is provided for each shift.
For a more detailed discussion of the responsibilities and authority cf'the supervisory positions mentioned above, and of the training and qualifications of the personnel holding these positions, refer to Sections 13.1.2.2 and 13.2.3 and Appendix 13A.
12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 12.5.2.1 Facilities All health physics and radiochemistry facilities are located on elevation 599'0" (B1) of the Control Complex. These facilities include the following rooms and areas:
a.
Personnel Decontamination Room b.
Medical Aid Room 12.5-1
l General Supervising Engineer, Technical Section The General Supervising Engineer, Technical Section is responsible for directing all activities associated with providing technical support and services related to moni t oring plant l>erformance, ciani nnent anl system testing, ins t rument i
maintenance, calibration and repair and reactor technology. lie is also I
responsible for the programming, operation and maintenance of the p'rocess j
computer and related sof tware development. The General Supervising Engineer, i
Technical Section is a member of the Plant Operations Review Committee and reports to the Superintendent, Plant Operations.
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General Supervising Engineer, Radiation Protection Section g
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The General Supervising Engineer, Radiation Protection Section is designated as es 0
the Radiation Protection Manager (RPM) and is, therefore, responsible for ALARA s
ss program coordination and directing all activities associated with the chemical, rs radiochemical, radwaste and other radiological control services required to support plant operation and maintenance activities. This includes conducting labocatory and plant survey activities required to ensure that personnel exposure i
l to radiation and radioactive materials is within regula* ry guidelines and that such exposure is kept as low as reasonably achievable (ALARA). The General I
Supervising Engineer Radiation Protection Section is a member of the Plant Operations Review Committee, ALARA Committee, and reports to the Plant Manager.
General Supervicor, Nuclear Services Section The General Supervisor, Nuclear Services Section is responsible for directing all training activities required to develop and maintain a qualified workforce, and for insuring that all provisions of the PNPP Security Plan are implemented. He is also responsible for providing the necessary general maintenance and f
administrative services required to effectively support plant activities. He reports to the Plant Manager and, via the Security Supervisor, ma cains a communications link with the corporate Security Advisor.
1 13.1-14
471.11 Based on information contained in Regulatory Guide 8.10, (12.1.2)
Section C.1.b.
It is our position that the plant's management staff should periodically review operating procedures and i
exposure information to determine major changes ta problem 1
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areas, and areas in which worker groups are accumulating the highest exposures. The staff at Perry Nuclear Power Plant should use the information obtained by management review to I
cecommend equipment modification or changes in plant procedures. Outline your methods for implementing this position.
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Response
The response to this question is provided in revised Section 12.1.3.
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is compatible with maintaining occupational radiation doses ALARA. All designs influencing radiological centrol in the plant are reviewed by competent professionals in the area of radiation protection.
As indicated in Section 12.1.2.3, all personnel associated with design aspects influencing radiation protection have been given the basic ALARA principles.
After the preliminary design and layout of the system, a shielding engineer analyzes the various radiation sources and specifies shielding as necessary to conform to the appropriate radiation zone requirements. Radiation protection personnel review the system in terms of the total plant operation and specify necessary changes to keep occupational radiation doses ALARA. Both snielding engineers and radiation protection personnel are part of the overall design team and report their findings directly to the design project management for the PNPP.
These findings are considered in conjunction with design requirements from disciplines not directly associated with radiation control to determine what modifications, if any, should be esde to promote ALARA radiation doses.
12.1.3 OPERATIONAL CONSIDERATIONS The ALARA Committee is comprised of senior plant staff members chaired by the Plant Manager and including the Radiation Protection Manager. One of the functions of the committee is to periodically review procedures and the report of 7
Personnel and Man-Rem by Work and Job Function to ensure exposures are maintained ALARA and to recommend methods of reducing Man-Rem waere applicable through Ts
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procedural changes or equipment and system modifications.
12.1.3.1 ALARA Training Program The Radiological Trainine Program at PNPP will help implement the Company's ALARA policy. The Training Pi am will assure that workers understand how radiation protection relates to le i, jobs and all workers will have frequent opportunities to discuss radiation safety with the Health Physics Unit personnel when the need arises.
12.1-6
4 All vork at PNPP involving systems that contain, collect, store or transport radioactive materials and may cause radiation exposure will require a Radiation Work Permit (RWP). The RWP will help implement the Company's ALARA policy by defining the radiological hazards and requiring specific radiological precautions. The RWP also becomes a record of how various jobs were performed I
and the radiological problems associated with specific jobs. By reviewing expired RWP's, recommendations can be made to change procedures or equipment that will result in lower radiation exposures in the future.
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471.12 As requested in Regulatory Guide 1.70, Section 12.2.1, provide (12.2) maximum neutron and gamma dose equivalent levels, in routinely visited areas in the containment, in the vicinity of major drywell shield penetrations. Areas of interest are, i.e.,-Reactor Water Cleanup and Standby Liquid Control System (drywell purge penetrations), TIP Station (personnel and equipment lock drywell penetrations), etc.
Describe location, dimensions and shielding for the drywell shield penetrations. Provide average neutron and gamma exposure levels at the CRD hydraulic control units, at the CRD master control and at the containment personnel lock area and provide an estimate of average daily personnel exposure time in these areas.
Response
The response to this question is provided in revised Section 12.3
'.2.1. and Table 12.4-7.
12.3.2.2 Design Description 12.3.2.2.1 Plant Shielding Description Detailed layout drawings showing all plant structures are shown in Figures 1.2-3 through 1.2-17.
A general description of the major shielding in the buildings housing radioactive process equipment and fluids is outlined as follows:
a.
Reactor building complex The reactor building complex shielding includes the biological shield wall, drywell shield walls and the shield building wall.
The purpose of the biological shield is te minimize gamma heating in the drywell shield wall, to provide access to the drywell during shutdown and to reduce activation of drywell equipment and materials. The design dose rate used in sizing this shield is to maintain a radiation level in the drywell below 100 Rem /hr at full power operation.
(Zone V).
l The drywell shield wall maintains the area outside the drywell at Zone II level except for some individual cubicles housing rsdioactive process equipment and piping, such as cubicles for the reactor water cleanup system and the chase for the main steamline pipes. The shielding for these is sized to maintain a Zone II level outside of each respective cubicle.
Areas in containment routinely visited during power operation include the following sy stems: SLC(C41), RWCU(G33), CRD(C11), and TIP(CSI). The expected occupancy requirements to these areas and the average personnel exposures is provided in Table 12.4-7.
Their respective locations are shown N
on Figures 12.3-2, 12.3-3, 12.3-4 and 12.3-6.
Routine surveillance and control functions are all accomplished from Zone II areas where the design
[
D basis gamma dose rate is < 2.5 mr/hr. The neutron dose rate in these areas 7
is negligible due to shielding provided by the five foot thick concrete drywell shield wall.
12.3-8 1
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l There are several major penetrations through the drywell shield wall. These as well as all other plant penetrations are located and/or designed to preclude the possibility of streaming from high to low radiation areas, or otherwise will be adequately shielded. Details of the personnel access lock 1
and shield door and the equipment hatch are shown on Figures 3.8-31 and J
3.8-32, respectively. Figure 12.3-2 provides their orientation in the plant layout.
The shield building completely surrounds the steel containment vessel and ensures that lecels outside the building are less than 0.5 mrem /hr (Zone I) during normal plant operation.
In addition, the building serves to attenuate radiation to plant personnel and the general public in the event of an accident.
I 12.3-8a
TABLE 12.4-7 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE OPERATIONS AND SURVEILLANCE Average Exposure Number Dose Dose Rate Time of (man-Rems /
Activity (mrem /hr)
(hr)
Workers Frequency year)
Control room 0.1 6,000 2
1.2 Walking ano checking:
Turbine and feedwater 100 0.1 1/ shift 10.0 heat exchanger 1.0 1.0 1
1/ shift 1.0 Containment cooling system 1.0 1.0 1
1/ day 0.36 Standby liquid control system 1.0 1.0 1
1/ day 0.36 ECCS and process equip 1.0 1.0 1
1/ shift 1.0 C&I panels and equip in containment 1.0 1.0 1
1/ shift 1.0 Fuel pool system 1.0 0.4 1
1/ day 0.1 RWCU 1.0 0.5 1
1/ shift 0.6 CRD system 1.0 0.5 1
1/ shift 0.6 Recirc flow control 1.0 0.6 1
1/ day 0.22 Misc auxiliary building 1.0 1.0 1
1/ shift 1.0 100 0.1 1
1/ shift 10.0 Traversing incore probe system 10 0.1 1
1/ shift 1.2 Misc. in containment 1.0 1.0 1
1/ day 0.4 Instrument calibration in containment 1.0 0.6 1
1/ week 0.03 Radiochemistry 1.0 1.0 2
1/ day 0.73 Health physics surveys 1.0 4.0 1
1/ day 1.46 a
15 1.0 1
1/ day 5.48 7
100 0.5 1
1/ week 2.6 3
Sample stations in reactor building 5.0 0.5 1
1/ shift 2.7 12.4-16
l TABLE 12.4-7 (Continued)
Average Exposure Number Dose Dose Rate Time of (man-Rems /
Activity (mrem /hr)
(hr)
Workers Frequency year)
Other local samples 5.0 0.1 1
1/ day 0.15 cd Remote sampling 1.0 0.3 1
1/ day 0.1
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4 Containment personnel 1.0 0.05 3
1/ shift 0.16 r-lock 7
Total 43 12.4-17
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471.13 Because the steam dryer and steam separator must be transferred (12.2.2) partially out of water during refueling, there is p.'tential for high concentrations of airborne radioactive material during the transfer. You should outline our proposed methods (other than maintaining wet) to reduce airborne radioactive material during these transfers. Provide an estimate of expected airborne concentration, on the separator transfer from the reactor vessel to the storage area. Consider equipment contamination buildup after at least 10 years of operation and address particulates as well as iodines.
(See Regulatory Guide 1.70, Section 12.2.2)
Response
The primary method of minimizing exposure to airborne radioactive material during the transfer of the steam dryer and steam separator is through administrative controls. These include direct Health Physics surveillance, the use of respiratory protection equipment and excluding'the entry to cottainment of all personnel except those directly involved in the transfer operations. Predicted radiation levels after ten (10) years operation are estimated to be 7R/hr for the steam dryer and l
25R/hr for the steam separator. Radiation levels caused by
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contamination will be predominantly due to the following nuclides:
Mn-54,Zn-65, Fe-59, Co-58, Co-60, Cr-51, I-131 and I-133.
After removal and storage of the reactor pressure vessel (RPV) head, the steam dryer assembly is transferred dry to its storage space.
The steam line plugs are installed to permit flooding of the RPV and simultaneous removal ofthe safety / relief valves for testing.
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The steam separator bolts are untensioned and unlatched from the RPV flange.
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The upper pools are flooded and the steam separator is transferred with the shroud head under water to its storage space.
471.14 Describe the Perry Nuclear Power Plant's accident radiation (12.3) monitoring system.
Include the location of installe.f instruments that have emergency power supplies (including LOCA) and the location of portable instruments placed to be readily accessible to personnel responding to an emergency. Regalatory Guide 1.97 (Revision 2) specifies the area radiation monitors in areas requiring access after an accident and portable survey meters should have a range up to 10' R/hr.
Response
Accident conditions are to be monitored by redundant high range gamma monitors to be located in the reactor building and in the drywell. The four monitors are to have a range of 1 R/hr. to 7
10 R/hr. and will be powered by the diesel backed 120 AC bus.
These monitors shall be procured in accordance with Regulatory Guide 1.97 and NUREG 0737 Table II.F.1-3.
Revised Table 12.5-2 and Table 7-5 of the Emergency Plan (Appendix 13A) list the location and number of portable monitoring equipment available to personnel responding to an I
emergency.
High range noble gas monitors are to be added to the effluent flow paths, i.e. :
a) main plant unit vent b) heater bay / turbine building vent c) off-gas vent These monitors will provide range extension to include the high level noble gas concentration in accordance with Regulatory Guide 1.97 and NUREG 0737. Power is to be derived from the diesel backed 120 VAC bus.
I TABLE 12.5-2
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e PORTABLE SfJRVEY INSTRUMENTS a
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Quantity Range Sensitivity / Accuracy Remarks m
G-M Pancake 6
0-50K CPM 5,000 CPM /mR/hr G-M Hand 6
0-50K CPM 2,500 CPM /mR/hr mR/hr and CPM l
f G-M mR/hr 20 0.1-2,000 110% of Full Scale Wide Range l
2 mR/hr l
Ion Chamber mR/hr 15 1-1,000 110% From 60 kev 1
1 mR/hr to 1.3 MeV I
mR/hr Telescoping 7
0.01 R/hr -
115% From 70 kev 999 R/hr to 1.3 MEV
,t e-l Ion Chamber R/hr 4
1 mR/hr -
115% of Full Scale 19.99 KR/: r p
J-i i
mR/hr Neutron 3
0-5K mR/hr
.025 eV (Thermal) l to approx. 10 MeV For calibration R-Chamber 1
0-200R
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Alpha-Proportional 2
0-50,000 CPM 46% of 2 n Counter l
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12.5-16 L
TABLE 7-5 PORTABLE RADIATION MONITORING EQUIPMENT A. PORTABLE RADIATION SURVEY INSTRUMENTS
(
TYPE RANGE DETECTOR QUANTITY LOCATION High Range 1 mR/hr-19.99 KR/hr lon Chamber 2
OSC Survey Instrument p
High Range 1 mR/hr-50 R/hr Ion Chamber 4
EOF Survey Instrument 1
TSC 4
OSC Low Range 0-200 mR/hr G.M.
4 EOF Survey Instrument 1
TSC 4
OSC 5
Alpha Survey 0-10 Proportional 1
RCA y
Instrument U
5 Neutron 0-10 CPM BF3 1
RCA Survey Instrument ImRem/hr-SRem/hr BF3 1
RCA B. PERSONNEL MONITORING DEVICES TYPE RANCE QUANTITY LOCATION Direct Reading 0-1003 10 EOF Pocket Dosimeter 5
10 TSC 70 RCA Thermoluminescent 0-100R 20 EOF Dosimeter (TLD) 50 RCA
E 471.15 Outline the Perry Nuclear Power Plant's program for implementing (12.3.1.1) a chemistry control program to reduce radiocobalt production and crud buildup in normally radioactive systems (See Regulatory Guide 8.8, Section C.2.e(3)).
f
Response
N The response to this question is provided in revised Section 12.3.1.1.
To further eliminate corrosion and crud transport, the General Electric oxygen control program will be evaluated to control oxygen in the condensate and feedwater to 50 1 20 PPB. The condensate cleanup system is further described in Section 10.4.6.
The Company has participated in the BWR Radiation Assessment and Control (BRAC) Program since its inception in 1973. BRAC is a joint venture between utilities, General Electric and EPRI formed to investigate and research the problem of radiocobalt production and associated radiation field buildup on primary t
l system piping. The Perry Plant design and operation of
{
cleanup systems is based on results and recommendations that I
the BRAC program has developed througt its research, l
In addition to the installed water purification systems, consideration has been given to reduce radiocobalt production and crud buildup in normally radioactiv: systems through material selection and equipment deer,n as discussed in Section 12.3.1.1.
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-..--,n-
,., ~, -.,.,,,, -. -. -., -. - -,... -, -,,,,,
h.
Counting room.
Figures 12.3-1 through 12.3-11 also illustrate traffic patterns during normal ooeration and locations of airborne radioactivity and area monitors.
The counting room is located so that the background radiation levels will be low enough to allow for continuous occupancy and to provide an accurate analytical environment under t.ormal operating conditions and anticipated operational occurrences. The counting room is sized to provide adequate space for the required instrumentation. See Section 12.5 for a discussion of instrumentation in the counting room.
Nonradioactive equipment that may require maintenance is located, when possible, in either Zone I or Zone II.
Adjacent areas containing potentially radioactive systems are designed to maintain a radiation level less than the Zone IV maximum (ICJ mrem /hr) during required maintenance.
Equipment located in Zone IV or Zone V is designed to minimize required maintenance and to be operated remotely. Shield wall penetrations for remote operating devices, electrical equipment, pipes and ventilation ducts are designed and located at positions that prevent a direct line of sight to any significant source, thereby minimizing radiation streaming.
The primary defense against corrosion product buildup and associated neutron activation in the reactor vessel followed by crud transport is to minimize the j
l input of impurities (i.e. iron, cobalt) in the feedwater. The Perry Plant design t
includes both full flow condensate filters and deep bed demineralizers. This i
I design provides maximum removal of both suspended and dissolved impurities.
In addition an extensive condenser sampling and analyses system is provided to
()
ensure prompt detection of small condenser leaks. The condensate demineralizers s
are provided with conductivity cells to measure water quality in the bed effluent r;
and at two thirds resin depth. The second conductivity cell will provide l
indication of resin depletion and allow for regeneration prior to total bed breakthrough.
i l
l 12.3-2 i
I
The following design consideration has been given to reduce radiocobalt production and crud buildup in normally radioactive systems:
System materials are specified for low corrosion and erosion rates and for a.
low neutron activation source characteristics. Hardfacing materials which have high cobalt content, such as Stellite, are used only where substitute materials cannot satisfy performance requirements.
b.
Packless valves are specified for systems which normally handle radioactive fluids. Where packed valves are specified, they are provided.
12.3-2a
471.16 The doses to plant personnel in the reactor building as they exist 0
(12.4.3) after a type 2 safety / relief valve isolation scram is estimated in Table 12.4-14 of the FSAR. However, it appears that the doses provided are average values and may not reflect actual doses to workers. Accordingly, explain all of the assumptions used to calculate the whole body, skin, and thyroid dose to plant personnel following a safety / relief valve discharge listed in Table 12.4-14.
Provide an estimate of personnel exposure (similar to Table 12.4-14),
resulting from actuation of safety relief valves, based on the following considerations:
1.
Use design basis radiation sources:
a)
Noble gas concentrations corresponding to an off-gas release rate of 0.1 C1/sec after 30 minutes decay:
b)
Halogen concentration in reactor water FSAR, Table 11.1-3, i
2.
Operator working at TIP drive floor at a location closest to the low-set safety relief valve discharge; 3.
Asseme that all safety relief valves open; low set relief valves remain open following closure of others (TYPE 2 occurrence);
4.
Operator exposure 4 minutes; 5.
Normal ventilation in containment (do not assume homogeneous mixing of airborne contaminants in the entire containment volume within the first four minutes);
6.
Dose reduction factor can be applied if a clear-air shower is provided in the vicinity of the containment personnel lock;
7.
Containment airborne concentrations should not be corrected for plate out on walls, as it will be negligible in the first few minutes.
Specify the noble gas and halogen pool retention factors, the average radiohalogen (reactor water to steam) carry-over factor, and other significant, dose-effecting factors employed in the calculations.
Response
The response to this question is provided in revised Sections 12.2.2.1, 12.4.3, and Table 12.4-14.
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During refueling it is anticipated that the only major contribution to airborne activity in the reactor building will be from radioiodides. The fuel pool cooling and cleanup system is designed to clean and purify the water in the spent fuel pool.aad the upper fuel pools in the containment. The iodine activity in the pools will be reduced by passing the water through a 1,000 gpm filter /demineralizer. The resultant airborne concentrations of iodine in the reactor building are expected to be less than 2 percent of the equilibrium values during normal operation. For the purposes of calculating operating exposures in Section 12.4, a value of 2 percent of the normal operation thyroid dose rate is assumed.
Another source of potential airborne contamination in the reacter building is the activity release through relief valve discharge to the suppression pool.
These are classified as Type 1 and Type 2 events. Type 1 events are of minor consequences because of the relatively short duration of the blowdown (<15 seconds). Type 2 events are of more concern because they involve isolation and depressurization of the system. The expected frequency of the Type 2 events is 2.5 times per year. Reference 2 provides the source terms used to determine the containment airborne concentrations following a Type 2 event and d
the methodology used to determine operational doses following the event.
Secti.on 12.4.3 presents the anticipated operator exposures per event.
+
l 12.2.2.2 Radwaste Building Leakage to the radwaste building is assumed to be 2,000 gallons per day at 10 l
percent of the primary coolant iodine activity. The airborne noble gas activity in the radwaste building is negligible. A partition factor of.001 0
3 is assumed for iodine. The radwaste building free volume is 1.1 x 10 ft and the purge rate is 30,000 cfm.
i l
Table 12.2-14 presents the calculated airborne concentrations in the radwaste building.
12.2-11
12.2.2.3 Turbine Building Leakage to the turbine building atmosphere is assumed to be 1,700 lb/hr of steam at main steam activity. A partition factor of 1 is assumed for both noble gases and halogens. The turbine building free volume is assumed to be 6
3 5
3.2 x 10 gg and the purge rate is 1.8 x 10 cf,,
Table 12.2-15 presents the calculated airborne concentrations in the turbine building.
12.2.2.4 Fuel Handling Area of the Intermediate Building Leakage to the fuel handling area atmosphere is based on evaporation from the spent fuel pool. The evaporation rate is calculated to be 320 lb/hr assuming the buiding is at 90 F and 50 percent relative humidity, and the pool is at 120 F.
The equilibrium I-131 concentration in the pool was conservatively s3 N
-6 taken at 1x10 pCi/cc based on information given in Reference 3.
The
-2 6
3 P-building free volume is assumed to be 1.5x10 ft and the purge rate is a.
30,000 cfm.
Table 12.2-16 presents the calculated airborne concentration for I-131 and proportional values for I-133 and I-135.
12.2.2.5 Other Buildings Other plant buildings are expected to have negligible noble gas and iodine airborne activity concentrations.
12.
2.3 REFERENCES
FOR SECTION 12.2 1.
Smith, J.M., " Noble Gas Experience in Boiling Water Reactors," Paper No.
A-54, presented at Noble Gases Symposium, Las Vegas, Nevada, September 24, 1974.
12.2-12
2.
General Electric Co., " Mark III Containment Dose Reduction Study", 22A5718 Rev. 2, Jan. 29, 1980.
3.
Johnson, A.B., " Behavior of Spent Nuclear Fuel in Water Pool Storage,"
BNWL-2256, Bate 11e, Pacific Northwest Laboratories, September, 1977.
12.2-12a
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s
,_o-,
-u-N O
TABLE 12.2-13 (Deleted)
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12.2-32
_..- _ -. _ _-. _. _ _ _____ _. _ _. -... _ _ _ _____...._._._.. -.._ _.., _ _ _ - _. _ _. - _ - _ _ _ _ _ - _. _... -.. -. _... _ -, _. _. _ _ _ _. _, ~. _ _
TABLE 12.2-13 (Continued)
(Deleted) l l
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1 12.2-33
12.4.3 ESTIMATED IN11ALATION DOSES i
Radiation doses associated with airborne radioactivity have not been analyzed in terms of tasks due to the lack of sufficient data.
Inhalation doses sere estimated using the airborne sources of radioactivity described in Section 12.2.2.
The ventilation system has been designed to move air from areas of unlimited occupancy (no radiation sources) to areas of limited occupancy (potential airbcrne radioactivity sources). Appropriate health physics procedures will be established to measure the radiological conditions in areas with potential airborne contamination, ensuring that radiation doses are maintained ALARA. Where required, respirators will be used to further reduce inhalation doses.
Table 12.4-13 lists specific areas of the plant where airborne activity may be present in quantities that would result in a measurable dose to the whole body and, as a result of gaseous iodines, a dose to the thyroid. The areas included are the reactor building (during normal operation and during refueling), the fuel handling building, the radwaste building and the turbine building. For aach area man-hours per work function have leen estimated using References 4 and 6.
The exposure rates have been determined from specific activities listed in Section 12.2.2.
Sources of airborne activity will be primarily from valve and pump leakage. During refueling operations the refueling and spent fuel pools will release small amounts of airborne activity l
to the reactor building and the fuel handling building, respectively.
tiowever, as a result of pool cleanup systems, it is anticipated that contributions from these sources will be minimal. The resulting annual man-Rem doses per unit are listed in Table 12.4-13.
An additional airborne activity source will come from the actuation of the safety / relief valves after an isolation scram (Type 2 event). Type I and Type 2 events of steam discharges to the suppression pool are discussed in Section 12.2.
Table 12.4-14 gives the resulting doses to personnel in the reactor building as they exit af ter a Type 2 event.
It has been assumed that at the initiation of the isolation scram, an operator is located at the TIP drive d
floor. The operator egress is at the personnel air lock 180' from the TIP Q
N 12.4-3
i i
i floor area at the same elevation. Operator egress is conservatively assumed I
to take four minutes.
The dose rates used are those calculated for the immediate area above the
,9 l
suppression pool. Normal ventilation is assumed and airborne concentration.
[{
l are not corrected for plate out on walls. The dose assessment methodology I'
.%~
including pool retention factors and average radiohalogen carry-over factors l
are provided in Reference 2 of Section 12.2.3.
i No dose calculations regarding Type 1 events are presented since resulting personnel doses would be negligible.
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12.4-3a
P TABLE 12.4-14 SAFETY / RELIEF VALVE DISCHARGERS DOSE FOR TYPE 2 EVENT Organ Dose Whole body + eye (y) 150 mrem / event G
Skin ( )
440 mrem / event so Thyroid
.87 mrem / event 12.4-24
471.17 Verify that your portablc radiation detection instruments (12.5.2) calibration progra.n meets Regulatory Guide 8.25 or describes an equivalent alternative.
Response
The response to this question is provided in revised Sections l
12.5.2.3.4 and 12.5.3.2.8.
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1 Personnel contamination survey instruments shall include Geiger-Mueller friskers,
\\
portable monitors, and hand and foot counters. These instruments will be calibrated according to Fealth Physics Instructions semiannually when in use, or prior to use after repair.
Personrel internal exposures will be evaluated by i
whole body counting as described in Section 12.5.3.6.2.
Typical personnel monitoring instruments are listed on Table 12.5-3.
12.5.2.3.4 Heal *.h Physics Equipment l
Portable air samples are used to determine airborne radioactive material N
concentrations. Air samplers will be celibrated for flow canually in accordance s
with Regulatory Guide 8.25 and Health Physics Instructions. Typical surveys will
- k J'
be performed fer particulate and radioiodine airborne concentrations.
Iortable continuous air monitors will normally be located in the common refueling area, solid waste drumming area (radwaste), the turbine operating floor, and in the heater bay. Local information and trend indicatica is provided. Ala rm setpoints are variable in accordance with health physics procedures. Audible and visual alarms are provided to warn local personnel of airborne radioactive concentrations in excess of specified limits.
Respiratory equipment will be provided and stored in the clothing storage area or any remote controlled access point in the plant, as required. Emergency respiratory equipment shall be stored at strategic locations within the plant.
The equipment will be maintained and used in accordance with applicable Eealth Physics Instructions. These instructions are prepared in conformance with Reg Guide 8.15.
Protective clothing will be provided for personnel working in radiologically controlled areas. Specific requirements for clothing will be prescribed by Health Physics personnel based on actual or anticipated radiological conditions.
An adequate int ^ntory of protective clothing will be maintained in the Clothing Storage Area, anc at any plant controlled access point necessary to support plant 12.5-5
12.5.3.2.7 Sampling Periodic sampling of process streams will help keep radiation exposures ALARA.
Analysis of samples will help verify that process stream monitors are accurate and are providing reliable information.
Most of the sampling of radioactive systems will be performed in chemical fume huods. The fume hoods provide negative air pressure and minimize the possible spread of contamination. Appropriate protective clothing and equipment will be specified in the sampl:24 procedures. Where applicable, survey meters will be used to monitor radiation levels at the fume hood and on the sample container.
The possibility of radioactive spills and radiation exposure will be maintained ALARA during sample transport by the use of special shielded or remote handling and transportation devices.
12.5.3.2.8 Calibration Periodic calibration of radiatio,,n detection instruments will help keep r diation exposures ALARA by assuring that the instruments are accurate and are providing reliable information. Portable radiation detection instruments are calibrated in accordance with manufacturers recommendations and Fealth Physics instructions.
5f Portable survey meters will be calibrated using an enclosed and shielded calibrator. Although the calibrator can calibrate instruments up to approximately 500 R/hr, it is shielded so that the external radiation level is about 0.1 mR/hr at one meter. Portable scurces used to calibrata fixed instruments like the Area Radiation Monitoring system are in shielded containers i
that slip over the detectors, keeping radiation exposure to personnel ALARA.
12.5.3.2.9 Plant Cleanliness Plant cleanliness is maintained in accordance with Regulatory Guide 1.39 as discussed in Sections 1.8 and 17.2.
O 12.5-10
471.18 Identify the iodine counter and gamma spectrometer marked (12.5.2.3.1)
"later" in Table 12.5-1 or indicata when this information will be,.covided.
Response
The iodine counter and gamma spectrometer information will be provided in an amendment submitted on or about January, 1982.
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8
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T 471.19 Table 12.5-2, portable survey instruments show the number of (12.5.2.3.2) radiation detection instruments which will be available for both units.
It is our position that the number of portable radiation sutvey instruments (especially those which are most frequently used by radiation protection personnel, 0 - 5000 mr/hr) be increased to reflect the following considerations (a) both units t
can be shut down for repair at the same time, (b) a number of instruments out of service (in need of calibration or repairs),
i and (c) a number of spare, operational instruments, should be always available for use in unusual occurrences. You should evaluate the number of portable survey instruments required by the plant using the above considerations and amend the FSAR accordingly.
(See Regulatory Guide 8.8, Section C.4)
Response
The response to this question is provided in revised Section 12.5.2.3.2 and Table 12.5-2.
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- - -... ~ - -.
I Each portable instrument will be. calibrated according to Health Physics Inscructions when in use, or prior to use after repair. Sufficient quantities of portable instrumentation will be available to permit calibration, caintenance, or repair to instruments without causing a shortage of operational equipment.
Typical portable equipment is listed in Table 12.5-2 An additional 30 percent i[
of each type of instrument listed in Table 12.5-2 is available for replacement of instruments that are out of service for repair or calibration.
[2 A large, heavily shielded, self-contained, malti-source calibrator will be provided for calibrating gamma dose rate instrumentation. Other sources will be provided as required. Instruments may also be calibrated by a qualified consultant. All sources used for calibration will be traceable to NBS.
(
12.5.2.3.3 Personnel Monitoring Instruments Personnel monitoring shall be provided by use of the thermoluminescent dosimeters (TLDs), direct reading pocket ionization dosimeters, and some electronic instruments. All persons entering the Radiological Control Area will be issued a TLD which will be used to measure exposure to beta gamma radiation. This badge will contain thermoluminescent chips with suitable energy filters. TDLs will be analyzed at least monthly for personnel records, and anytime personnel doses need to be ascertained as prescribed by Health Physics Instructions. The exposure history established by the TLD readings shall constitute the official record of personnel exposure at PNPP.
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It is not expected that an individual's Neutron dose will exceed 300 mrem per quarter; therefore, a calculated Neutron dose equivalent measured with portable j
monitoring instruments and known occupancy times will be issued in place of t
Neutron dosimeters.
In the event that such operating data indicates the Neutron t
dose equivalent does not exceed 300 mrem per quarter, the calculated Neutron dose equivalent will be assumed equal to zero.
Direct reading dosimeters shall be issued to personnel as necessary for indication of exposure. These dosimeters provide "up-to-the-minute" indication of radiation exposure. These dosimeters may also be used to monitor the 12.5-4
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TABLE 12.5-2 PORTABLE SURVEY INSTRUMENTS
'rype Quantity Range Sens it ivi ty/ Accu racy Remarks G-M Pancake 6
0-50K CPM 5,000 CPM /mR/hr G-M Hand 6
0-50K CPM 2,500 CPM /mR/hr mR/hr and CPM l'
{
I i
G-M mRIhr 20 0.1-2,000 110% of Full Scale Wide Range l
l mR/hr r
l Ion Chamber mR/hr 15 1-1,000 10% From 60 kev mR/hr to 1.3 MeV 2I I
mR/hr Telescoping 7
0.01 R/hr -
115% From 70 kev 999 R/hr to 1.3 MEV s
o-',
Ion Chamber R/hr 4
1 mR/hr -
15% of Full Scale 19.99 KR/hr p
7 l
mR/hr Neutron 3
0-5K mR/hr
.025 eV (Thermal) l to approx. 10 MeV l
R-Chamber 1
0-200R For calibration l
Alpha-Proportional 2
0-50,000 CPM 46% of 2 n Counter j
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12.5-16
TAlllI 7-5 PORTABII RADIATION MONITORING EQUIPMENT A. PORTABLE RADIATION SURVEY INSTRUMENTS TYPE RANCE DETECTOR QUANTITY LOCATION High Range 1 mR/hr-19.99 KR/hr Ion Chamber 2
,p Survey Instrument High Range 1 mR/hr-50 R/hr Ion Chamber 4
EOF 1
TSC Survey Instrument 4
OSC Low Range 0-200 mR/hr G.M.
4 EOF i
TSC Survey Instrument 4
OSC Alpha Survey 0-10 Proportional I
RCA y.
Instrument U
5 Neutron 0-10 CPM BF3 1
RCA Survey Instrument ImRem/hr-SRem/hr BF3 I
RCA B. PERSONNEL MONITORING DEVICES TYPE RANGE QUANTITY LOCATION j
l I
Direct Reading 0-100R 10 EOF Pocket Dosimeter 5
TSC 10 RCA 0-500 mR 20 EOF 10 TSC 70 RCA i
Thermoluminescent 0-100R 20 EOF Dosimeter (TLD) 50 RCA
471.20 Based on information contained in Regulatory Guide 8.8, (12.5.2.3.3)
Section C.3.b(2), it is our position that all personnel assigned TLD or film badgss also wear direct reading pocket dosimeters when entering controlled areas. The readings from these dosimeters should be used to keep a running total of the dose to individuals prior to TLD or film badge processing and to allow analysis of doses by tasks. Outline your methods for implementing this position.
Response
The response to this question is provided in revised Section 12.5.2.3.3.
1 1
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j extremities. The dosimeters will be calibrar.ed semi-annually, or anytime damage is suspected.
O The individual records the pocket dosimeter. reading on the RWP Exposure Record i
Card while exiting the Radiological Control Area. Daily, the readings are l
transferred to the ir.dividual's occupational exposure records by Health Physics Personnel. This provides a running total of the dose to the individual.
1 The TLD results are compared to the pocket dosimeter readings upon receipt.
g d
Discrepancies between the pocket dosimeter readings and TLD results of greater N
than 25 percent are investigated by the Health Physics Unit to determine the b
}
probable cause.
Annually or as required by the Health Physics Supervisor, an analysis of dose by j
task is performed to determine waich operations or tasks can be modified to j
reduce exposure.
t i
12.5-4a
. - _ _,, _ _ _ _ _.. _ _ _. _ _ ~._,._.,__,___,_-...,_ _ __ _
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471.21 In Table 12.5.4 you indicate that 50 full-face masks will be (12.5.2.3.4) available at the plant. This number appears low for routine (nonemergency) use compared to the needs of currently operating
)
nuclear power plants. You should revise this inventory or justify this number by evaluating the number of full-face masks expected to be needed during normal and expected i
operational occurrences, during emergency situations and number of masks not available to plant personnel due to J
routine maintenance.
(See Regulatory Guide 8.8, Section C.4).
Response
The response to this question is provided in revised i
Table 12.5-4.
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I TABLE 12.5-4 HEALTH PHYSICS EQUIPMENT T
Quantity Range Remarks m
Large Calibrator 1
75mR/hr-50 curies; Cs-137 500H/hr traceable to NBS High Vol. Air Sampler 3
Approx. 30 CFM i
Low Vol. Air Sampler 3
Approx. 3 CFM Self-Contained 10
Breathing Apparatus 4
Full Face Masks 200
N Full Face Filters 300
< 50 PF 7
l Respirator Hoses 20 N/A Respirator Junction 3
N/A To 6 Respirators Box Per Junction Box Waterproof Suits 30 N/A Clothing Sets 3,000 N/A (Coveralls, Hoods, and Booties) 2 Rubbers (Pr.)
3,000 N/A Rubber Gloves (Pr.)
5,000 N/A i
Cotton Gloves (Pr.)
2,000 N/A Doz.
Wet / Dry-55 Gal.
Vacuum Cleaners 2
N/A Laundry Monitor 1
160-7000 CPM i
12.5-18
_, - _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. ~
- i 471.22 Section 12.5.3.6.2, states all personnel who wear respirators 1
( 12. 5. 3. 6. 2.'
will be whole body counted or have a bioassay at least once i
]
per year. " Personnel who freqenti ' use respirators, or are l
suspected of having an accidental exposure to airborne j
radioactivity may be bioassayed or whole body counted more J
often." You should specify that your bioassay program will be l
implemented in accordance with Regulatory Guide 8.26
" Applications of Bioassay for Fission and Activation Products,"
j or you should describe your equivalent bioassay program.
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Response
i j
The response to this question is provded in revised Section i
12.5.3.6.2.
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Exposure data for all personnel will be recorded on Form NRC-5, or the equivalent. Occupational exposures incurred by individuals who are expected to exceed 1.25 Rem in any qu. rter will be summarized on Form NRC-4, or the equivalent. These records will be maintained by CEI and will be preserved indefinitely or until the NRC authorizes their disposal. Current exposure status will be made available to each supervisor and individual, as required, to assist in keeping individual radiation exposures ALARA. Each worker shall receive exposure reports in accordance with 10 CFR 19.13.
Personnel monitoring, and equipment for personnel monitoring and surveys, will be in conformance with the requirements of 10 CFR 20 and Regulatory Guides 8.4 and 8.9, as discussed in Section 1.8.
12.5.3.6.2 Internal Exposures g
N The bioassay program is implemented in accordance with Regulatory Guide 8.26.
All personnel who take part in the Respiratoty Protection Program (wear or may wear respirators) will be whole body counted or have a bioassay at least once per year. Personnel who frequently use respirators, or are surpected of having an c
l accidental exposure to airborne radioactivity, may be bioassayed or whole body counted more often. The results of the whole body counting or bioassays will be I
maintained with, and become part of, an individual's dosimetry file.
12.5.3.7 Evaluation and Control of Potential Airborne Radioactivity i
Portable air samplers, air monitors, and fixed air monitors are ised to determine the concentrations of airborne radioactivity in the plant. Particulate fil*.ers and charcoal cartridges from the samplers and monitors will be analyzed in the health physics service room or the counting room using equipment described in Sections 12.5.2.1 and 12.5.2.3.1.
Samples from the continuous air monitors will l
be changed and analyzed at least weekly. Where applicable, air samples will be taken as a routine part of radiological surveys as described in Section 12.5.3.1.
For additional details, see Sections 12.5.2.3.4 and 12.5.2.3.5.
12.5-13 L
e 471.23 Section 12.5.3.8 of the PNPP FSAR " sealed radionuclides (12.5.3.8) having activities greater than the amounts listed in Appendix C of 10 CFR Part 20 will be subject to controls for radiological protection." Since the radionuclides and activities listed in Appendix C are associated with allowable sewerage relear.1 limits authorized in 10 CFR 20.303 and not intended as de mininus quantities, you should revise your procedures to require all licensed sources to be subject to material cantrols. However, pursuant to Section 30.18 of 10 CFR Part 30 sealed sources obtained from a manufacturer licensed to distribute exempt quantities in accordance with Section 32.18 of 10 CFR Part 32 may be exempted from your material control system.
Response
Section 12.5.3.8, c.s modified in Amendment 2 (5-22-81),
provides a response to this question.
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respirators and to the users' supervisors. All personnel who are expected to continue using respirators will be retrained at least annually to retain a high degree of proficiency and help maintain radiation exposures ALARA.
12. 'i..l. 3 R.nli o.ir t i ve,- flatei ia l llanill Mg anil Storage Methods Handling of radioactive samples is described in Section 12.5.3.2.7.
Various other types and quantities of radioactive sources are used to calibrate equipment.
Recognized methods for the safe handling of radioactive materials, such as those recommended by the National Council on Radiation Protection and Measurements, will be used to maintaici potential external and internal radiation exposures ALARA.
Radioactive sources will be used or handled by, or under the direction of, Health Physics personnel.
Individuals using these sources will be familiar with the radiological restrictions, regulations, and limitations placed on their use.
These limitations will help protect be*.h the user and the source integrity, and other personnel in the work vicinity.
Special Nuclear Material is maintained, handled, and stored in accordance with 10 CFR 70.
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12.5-14 Am. 2 (5-22-81) i
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471.28 Please provide the information requested in II.B.2, II.F.1(3)
(Appendix 1A) and III.D.3 of NUREG-0737, " Clarification of TMI Action Plan Requirements."
Response
Item II.B.2 addresses the requirements for post-LOCA vital area l
access.
i Vital Area Access A review of the plant identified systems which were likely to contain highly ridioactive fluid following a design basis LOCA.
The radioactive esterial was assumed to be instantaneously mixed in those systems, connected either to the reactor coolant system or to the containment atmosphere, that are not isolated at the start of the accident.
Non-essential systems that are isolated and have no post-accident function were not considered in the review.
After determining the systems and post-accident source distribution to be used for the shielding review, the SDC point kernel shielding code was used to calculate the associated post-accident radiation doses.
Areas which may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident are vital l
areas. The evaluation to determine the necessary vital areas included the control room, technical support center, post-LOCA I
hydrogen control system, containment isolation system, sampling I
and sample analysis areas, ECCS alignment functions, motor control center, instrument panels, emergency power supplies, security center and radwaste control panels. Of these it was determined that for the Perry plant, the control room and l
l technical support center will require continuous occupancy and i
l the post-accident sample area will require infrequent occupancy.
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II.F.1-3 High range gamma monitors are to be added to the reactor building and to the drywell to provide conformance with 7
NUREG 0737 Table II.F.1-3 with a range of 1 R/hr to 10 R/hr and to respond to the requirements of Regulatory Guide 1.97, Revision 2.
III.D.3 Iodine samplers are being designed to be added to the main plant vent, heater bay / turbine building vent, and the off gas vent to be in accordance with NUREG 0737, bection III.D.3 and II.F.1-2.
These are to be powered by a diesel backed bus.
Iodine absorbers will include the use of silver zeolite cartridges, ihesamplerdesignshalltakeintoconsiderationthe accessibility of the collection filters, the flushing requirements, the dose rate potential for handling and transporting the samples, and the requirements of emergency power for the sample pumps along with auto start of sampling upon detection of high activity by the radiation monitoring instrumentation.
A counting room and/or portable instruments will be available for analysis of these samples under post-accident conditions.
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f CONNECTICUT YA N K EE ATO MIC POWER COMPANY BERLIN, CO N N ECTIC U T P. O BOX 270 H ARTFORD CONNECTICUT 06101 TrLeessoas 203-666 6911 July 22, 1981 Docket No. 50-213 B10235 f
k Director of Nuclear Reactor Regulation f(
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g Attn:
Mr. Dennis M. Crutchfield, Chief J U(r]g' Operating Reacters Branch #5 r
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U. S. Nuclear Regulatory Commission i
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y Gentlemen:
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N W Haddam Neck Plant Steam Line Break Detection Logic The NRC Staff has verbally expressed concerns on the steam line break detection logic for the Haddam Neck Plant and how the detection logic meets the single failure criterion with one reactor coolant system loop isolated. To assist the Staff in their review of this concern, Connecticut Yankee Atomic Power Company (CYAPCO) is providing drawing No. 16103-32001, Sheet llc. CYAPCO has determined that this drawing provides sufficient information to demonstrate that the steam line break detection logic does meet the single failure criterion with one reactor coolant loop isolated.
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We trust the Staff will tind this adequate to resolve this concern for the Haddam Neck Plant.
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Very truly yours, 1
CONNECTICUT YANKEE ATOMIC POWER COMPANY
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W.'G. Counsil Senior Vice President l
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