ML18078A835
| ML18078A835 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 09/25/2017 |
| From: | Duke Energy Progress |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17298A847 | List: |
| References | |
| RNP-RA/17-0063 | |
| Download: ML18078A835 (52) | |
Text
HBR 2 UPDATED FSAR CHAPTER 12 12.0 RADIATION PROTECTION
HBR 2 UPDATED FSAR 12-i Revision No. 21 CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS
SECTION TITLE PAGE
12.0 RADIATION PROTECTION 12.1.1-1
12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 12.1.1-1
12.1.1 POLICY CONSIDERATIONS 12.1.1-1 12.1.1.1 CORPORATE HEALTH PHYSICS POLICY 12.1.1-1 12.1.1.2 FACILITY MANAGEMENT POLICY 12.1.1-2 12.1.1.3 FACILITY MANAGEMENT RESPONSIBILITIES 12.1.1-3
12.1.1.4 POLICY IMPLEMENTATION 12.1.1-4 12.1.1.5 ALARA PROGRAM IMPLEMENTATION COMPONENTS 12.1.1-4 12.1.2 DESIGN CONSIDERATIONS 12.1.2-1 12.1.3 OPERATIONAL CONSIDERATIONS 12.1.3-1 12.1.3.1 PLANT ORGANIZATION 12.1.3-1 12.1.3.2 OPERATING EXPERIENCE 12.1.3-1 12.1.3.3 EXPOSURE REDUCTION 12.1.3-2 12.2 RADIATION SOURCES 12.2.0-1
12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1-1
12.3.1 SHIELDING 12.3.1-1 12.3.1.1 DESIGN BASIS 12.3.1-1 12.3.1.2 PRIMARY SHIELD 12.3.1-1 12.3.1.3 SECONDARY SHIELD 12.3.1-2 12.3.1.4 ACCIDENT SHIELD 12.3.1-2 12.3.1.5 FUEL HANDLING SHIELD 12.3.1-3 12.3.1.6 AUXILIARY SHIELDING 12.3.1-3 HBR 2 UPDATED FSAR 12-ii Revision No. 21 CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS (continued)
SECTION TITLE PAGE 12.3.2 VENTILATION 12.3.2-1 12.3.3 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION 12.3.3-1 12.3.3.1 AREA RADIATION MONITORING SYSTEM 12.3.3-1 12.3.3.1.1 DESIGN BASIS 12.3.3-1 12.3.3.1.2 GENERAL SYSTEM DESCRIPTION 12.3.3-2 12.3.3.1.2.1 CHANNEL R1, 2, 3, 4, 5, 6, 7, 8, 9 & 33 12.3.3-3 12.3.3.1.2.2 CHANNEL 32A AND 32B 12.3.3-3 12.3.4 CRITICALITY MONITORS 12.3.3-5 12.3.4.1 CRITERION 1 12.3.3-5 12.3.4.2 CRITERION 2 12.3.3-6 12.3.4.3 CRITERION 3 12.3.3-6 12.3.4.4 CRITERION 4 12.3.3-7 12.3.4.5 CRITERION 5 12.3.3-7 12.3.4.6 CRITERION 6 12.3.3-8 12.3.4.7 CRITERION 7 12.3.3-8 12.4 DOSE ASSESSMENT 12.4.0-1 12.5 HEALTH PHYSICS PROGRAM 12.5.1-1 12.5.1 ORGANIZATION 12.5.1-1 12.5.
1.1 INTRODUCTION
12.5.1-1 12.5.1.2 RESPONSIBILITIES 12.5.1-1 12.5.1.3 AUTHORITY 12.5.1-3 12.5.1.4 EXPERIENCE AND QUALIFICATION 12.5.1-3 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 12.5.2-1 12.5.2.1 PERSONNEL PROTECTIVE EQUIPMENT 12.5.2-1 12.5.2.2 RADIATION INSTRUMENTATION 12.5.2-1 12.5.2.3 PERSONNEL MONITORING 12.5.2-1 HBR 2 UPDATED FSAR 12-iii Revision No. 16 CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS (continued)
SECTION TITLE PAGE
12.5.2.4 FACILITIES AND ACCESS PROVISIONS 12.5.2-2 12.5.3 PROCEDURES 12.5.3-1 12.5.3.1 CONTROL OF ACCESS AND STAY TIME IN RADIATION AREAS 12.5.3-1
12.5.3.1.1 PHYSICAL CONTROLS 12.5.3-1 12.5.3.1.1.1 SECURITY CHECK POINT AND ACCESS CONTROL 12.5.3-1 12.5.3.1.1.2 DOOR AND AREA POSTING AND LOCKING 12.5.3-1 12.5.3.1.1.3 HEALTH PHYSICS SURVEILLANCE 12.5.3-1 12.5.3.1.2 ADMINISTRATIVE CONTROLS 12.5.3-2 12.5.3.1.2.1 TRAINING 12.5.3-2 12.5.3.1.2.2 RADIATION WORK PERMITS 12.5.3-2 12.5.3.2 ASSURING OCCUPATIONAL RADIATION EXPOSURE ARE ALARA 12.5.3-2
12.5.3.2.1 ALARA PROCEDURES COMMON TO EXTERNAL AND INTERNAL EXPOSURE 12.5.3-2 12.5.3.2.1.1 TRAINING 12.5.3-2 12.5.3.2.1.2 RADIATION WORK PERMIT (RWP) 12.5.3-2 12.5.3.2.1.3 WORK SCHEDULING 12.5.3-3 12.5.3.2.1.4 JOB PRE-PLANNING AND EXPOSURE GOALS 12.5.3-3 12.5.3.2.1.5 ALARA PROGRAM REVIEWS 12.5.3-3 12.5.3.2.2 EXTERNAL ALARA 12.5.3-3 12.5.3.2.2.1 ADMINISTRATIVE LIMITS 12.5.3-3 12.5.3.2.2.2 TEMPORARY SHIELDING AND SPECIAL TOOLS 12.5.3-4 12.5.3.2.3 INTERNAL ALARA 12.5.3-4 12.5.3.2.3.1 CONTROL OF ABSORPTION 12.5.3-4 HBR 2 UPDATED FSAR 12-iv Revision No. 25 CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS (continued)
SECTION TITLE PAGE
12.5.3.2.3.2 CONTROL OF AREA AND EQUIPMENT CONTAMINATION LEVELS 12.5.3-4 12.5.3.2.3.3 AIRBORNE EXPOSURE EVALUATION 12.5.3-4 12.5.3.3 RADIATION SURVEYS 12.5.3-4 12.5.3.4 CONTAMINATION SURVEY PROCEDURES 12.5.3-5 12.5.3.4.1 PERSONNEL CONTAMINATION SURVEYS 12.5.3-5 12.5.3.4.2 EQUIPMENT CONTAMINATION SURVEYS 12.5.3-5 12.5.3.4.3 SURFACE CONTAMINATION SURVEYS 12.5.3-6 12.5.3.5 AIRBORNE RADIOACTIVE MATERIAL 12.5.3-6 12.5.3.5.1 AIRBORNE CONCENTRATION SAMPLING 12.5.3-6 12.5.3.5.2 RESPIRATORY PROTECTION 12.5.3-6 12.5.3.5.3 HANDLING OF RADIOACTIVE MATERIAL 12.5.3-7 12.5.3.6 PERSONNEL MONITORING 12.5.3-7 12.5.3.6.1 EXTERNAL RADIATION EXPOSURE ASSESSMENT 12.5.3-7 12.5.3.6.2 INTERNAL RADIATION EXPOSURE ASSESSMENT 12.5.3-8 12.5.3.6.3 METHODS OF RECORDING AND REPORTING 12.5.3-8 12.5.3.7 HEALTH PHYSICS TRAINING PROGRAMS 12.5.3-9 12.5.3.7.1 HEALTH PHYSICS TRAINING 12.5.3-9 12.5.3.7.2 RESPIRATORY PROTECTION TRAINING PROGRAM 12.5.3-10
HBR 2 UPDATED FSAR 12-v Amendment No. 10 CHAPTER 12 RADIATION PROTECTION LIST OF TABLES
TABLE TITLE PAGE
12.3.1-1 PRIMARY SHIELD NEUTRON FLUXES AND DESIGN PARAMETERS 12.3.1-5 12.3.1-2 SECONDARY SHIELD DESIGN PARAMETERS 12.3.1-6 12.3.1-3 ACCIDENT SHIELD DESIGN PARAMETERS 12.3.1-7 12.3.1-4 REFUELING SHIELD DESIGN PARAMETERS 12.3.1-8 12.3.1-5 PRINCIPAL AUXILIARY SHIELDING 12.3.1-9 12.3.3-1 AREA RADIATION MONITORING SYSTEM DETECTING MEDIUM CONDITIONS 12.3.3-4
HBR 2 UPDATED FSAR 12-vi Revision No. 19 CHAPTER 12 RADIATION PROTECTION LIST OF FIGURES
FIGURE TITLE
12.5.2-1 RADIATION CONTROL AREA SITE
12.5.2-2 RADIATION CONTROL AREA REACTOR AUXILIARY
HBR 2 UPDATED FSAR 12.1.1-1 Revision No. 27 12.0 RADIATION PROTECTION 12.1 Ensuring that Occupational Radiation Exposures are as Low as Reasonably Achievable (ALARA) 12.1.1 Policy Considerations
12.1.1.1 Corporate Health Physics Policy
It is the policy of the Duke Energy Progress, LLC to develop, implement, and maintain sound
health physics programs at each Company facility where radiation-producing equipment and/or
radioactive materials are used or stored. The health physics programs shall be structured to
ensure that radiation doses to Company personnel, contractor personnel, and the general public
are maintained at levels which are as low as reasonably achievable (ALARA) and consistent
with the United States Nuclear Regulatory Commission Regulations in Title 10 of the United
States Code of Federal Regulations and with applicable state regulations.
The line management at each of these Company facilities is responsible and accountable for
implementing and enforcing the facility's health physics program. Radiation control personnel shall be assigned to these Company facilities to assist line management in carrying out their
responsibility to protect workers and the general public. Radiation control personnel shall have
sufficient independence from the line management which they assist to assure that proper
health physics practices are not compromised by operational pressures. Furthermore, radiation
control personnel shall have access to higher levels of management for the resolution of health
physics concerns which cannot be resolved at a lower management level.
All Company and contractor employees working in a facility where exposure to radiation might
occur are personally responsible for maintaining radiation doses and releases of radioactive
materials to unrestricted areas as far below specified limits as reasonably achievable, to
minimize the creation of radwaste, and to support the requirements of the health physics
programs consistent with the proper discharge of their duties. Personnel who habitually or
willfully disregard or violate health physics procedures and practices will be subject to
disciplinary action.
A manual shall be developed and maintained as the controlling document for the Company's
health physics programs and shall set forth policies and standards for the Company's health
physics programs. The health physics programs at the Company's nuclear generating plants
shall comply with the manual and meet the intent of the Institute of Nuclear Power Operations' "Guidelines for Radiological Protection at Nuclear Power Stations."
The goal of the Company is to maintain the annual integrated occupational dose at nuclear
plants and dose to members of the public from Company activities among the lowest in the
country. The design, maintenance, and operation of nuclear facilities shall be consistent with
this goal. Modifications to existing nuclear facilities shall be designed and implemented in
compliance with the health physics programs to meet the ALARA objective.
HBR 2 UPDATED FSAR 12.1.1-2 Revision No. 27 To support this goal and to allow management to conduct effective health physics programs, the Company will commit sufficient resources in the form of facilities, equipment, and personnel to
the programs. Personnel involved in the conduct of the health physics programs, including
general employees and contractors, shall be given adequate training and instruction to allow
them to contribute to the programs.
Nuclear Oversight (NOS) evaluates the performance and effectiveness of the Company's health
physics programs on a continuous basis to assure they are being carried out in an effective
manner. These activities are to detect deficiencies in the desired levels of performance and
quality, reporting these conditions to the Vice President - Robinson Nuclear Plant, and ensuring
adequate action is taken to correct and eliminate these conditions.
12.1.1.2 Facility Management Policy
Duke Energy Progress, LLC is committed to a program of keeping occupational radiation exposure as low as reasonably achievable (ALARA). The Operating License, issued by the
Nuclear Regulatory Commission, carries with it an obligation to both workers and the general
public to maintain exposures as low as is reasonably achievable, considering costs and
expected benefits. Duke Energy Progress, LLC follows the general guidance of Regulatory
Guides 1.8., 8.8, 8.10, and publications which deal with ALARA concepts and practices, including Title 10, Code of Federal Regulations, Part 20. As discussed in Section 12.1.1.1, corporate management has formally committed itself to this concept by issuing and endorsing
the Corporate Health Physics Policy, which ensures compliance with all state and federal
regulations that pertain to the safe operation of nuclear power plants.
The implementation of the Corporate Health Physics Policy is accomplished through a number
of mechanisms and procedures in all phases of plant operation. The Radiation Control and
Protection Manual provides the direction necessary for implementing corporate policy.
The Radiation Control and Protection Manual sets forth the basic philosophy and general
radiation protection standards and procedures that are essential to the safe operation of DEP's
nuclear facilities. The Site Vice President of each nuclear facility is responsible for ensuring the
requirements of this manual are included in the Radiation Control and Protection Program at
that facility. Requirements of the Radiation Control and Protection Manual and/or applicable
Nuclear Generation Group procedures.
The primary purpose of the Radiation Control and Protection Program is to provide personnel
with a safe environment in which to work, to protect the general public and the off-site environs, and to establish procedures and a system of records to meet all the requirements of applicable
regulations.
Effective control of radiation exposure involves the following major considerations:
- 1. Management commitment to, and support of, the Radiation Control and Protection Program
- 2. Careful design of facilities and equipment to minimize radiation exposure during operation and maintenance.
HBR 2 UPDATED FSAR 12.1.1-3 Revision No. 26 3. Good radiation protection practices, including good planning and the proper use of appropriate equipment by qualified, well-trained personnel.
The management of DEP is firmly committed to performing all reasonable actions for ensuring that radiation exposures are maintained ALARA.
12.1.1.3 Facility Management Responsibilities
Management's commitment to the Corporate Health Physics Policy is reflected in the careful
preparation of plant operating and maintenance procedures, the provision for review of these
procedures and for review of equipment design to consider the results of operating experience, and most importantly, the establishment of an on-going training program. Training is provided
for personnel, so that each individual will be capable of carrying out his responsibility for
maintaining his own radiation exposure, as well as that of others, ALARA consistent with
discharging his duties. The development of a proper attitude and an awareness of the potential
problems in the area of health physics is accomplished through proper training of all plant
personnel.
The responsibility for implementation of the ALARA program resides with the Site Vice
President, with primary support from all facility Section and Unit heads. The Superintendent -
Radiation Control reports to the General Manager - Robinson Plant and makes recommendations to plant management concerning the most effective radiation exposure
reduction methods. He is assisted in this task by the Specialist - ALARA who includes, as a
major portion of his assignment, an analysis of plant operations and maintenance with respect
to maintaining an ALARA approach to personnel radiation exposure.
The success of the program depends upon cooperation between many plant operating groups.
The Specialist - ALARA acts as liaison between the groups while maintaining a high degree of
organizational freedom. An ALARA Committee, composed of all major plant operating groups
and chaired by the Site Vice President - Robinson Plant or designee, ensures radiation
exposure reduction initiatives are identified and implemented, ensures interface with various
plant groups, and provides a mechanism for the review of outages and maintenance activities.
Data and experience compiled from the operation of HBR 2 and similar nuclear plants are used
as a basis for the review of plant operations, features, and proposed modifications.
The Superintendent - Radiation Control reports to the Plant General Manager, assuring an open
line of communication with appropriate management. The overall effectiveness of the program
is reviewed periodically by appropriate plant and corporate personnel. Formal guidance is
provided in the Radiation Control and Protection Manual (discussed in Section 12.1.1.2) and a
written ALARA program. Formal support for the plant's ALARA program may be requested from
company resources, outside agencies or contact personnel, through the efforts of the
Superintendent - Radiation Control, who is responsible for ensuring that the direction provided
in the Radiation Control and Protection Manual is implemented and that the health physics
programs comply with the Corporate Health Physics Policy.
HBR 2 UPDATED FSAR 12.1.1-4 Revision No. 18 12.1.1.4 Policy Implementation The management's ALARA policy is implemented at HBR 2 by the Radiation Control Staff under
the direction of the Superintendent - Radiation Control, Radiation Control Supervisors, and
Specialist - ALARA. It is formalized by incorporating ALARA philosophy and considerations into
permanent plant procedures which deal specifically with ALARA concerns. The operational
ALARA considerations identified in Sections 12.1.3. and 12.5.3.2 have been implemented
according to plant procedures.
A training program has been established to give appropriate plant personnel the knowledge
necessary to understand why and how they should maintain their occupational radiation
exposure ALARA.
12.1.1.5 ALARA Program Implementation Components
The Plant Management's responsibilities for implementation of corporate policy include:
- 1. Ensuring that an effective measurement system is established and used to determine the degree of success achieved by plant operations with regard to the ALARA goals and
objectives.
- 2. Ensuring that the measurement system results are reviewed on a periodic basis, and that corrective action is taken when attainment of the specific objectives appears to be jeopardized.
- 3. Ensuring that the authority for providing procedures and practices, by which the specific goals and objectives are achieved, is delegated.
- 4. Ensuring that the resources needed to achieve ALARA goals and objectives are made available.
The health physics program is based on regulations and experience including or considering the
following:
- 1. Detailed procedures have been prepared and approved for radiation protection and are a part of HBR's health physics program.
- 2. Radiological incidents are thoroughly investigated and documented in order to minimize the potential for recurrence. Reports are made to the NRC, in accordance with 10CFR20.2202 and 10CFR20.2203, 10CFR20.2204 and the ODCM.
- 3. Periodic radiation, contamination, and airborne activity surveys are performed and recorded to document radiological conditions. Records of surveys are maintained in
accordance with 10CFR20.2103.
- 4. Radiation areas, high radiation areas, and very high radiation areas are segregated and identified in accordance with 10CFR20.1003, 10CRF20.1501, 10CFR20.1601, 10CFR20.1602, 10CFR20.1901, 10CFR20.1902, and the Technical Specifications.
Airborne radioactivity is determined and posted in accordance with 10CFR20.1003, 10CFR20.1501, 10CFR20.1901, and 10CFR20.1902.
HBR 2 UPDATED FSAR 12.1.2-1 12.1.2 DESIGN CONSIDERATIONS
HBR 2 was designed and constructed in accordance with the proposed General Design Criteria (proposed Appendix A to 10CFR50, published in July, 1967) and other appropriate design
guidance, including experience with other plants, that was available at the time. The general
design criteria are discussed in Section 3.1.
HBR 2 UPDATED FSAR 12.1.3-1 Revision No. 18 12.1.3 Operational Considerations Operational considerations at HBR that promote the ALARA philosophy include the
determination of the origins of radiation exposures, the proper training of personnel, the
preparation of radiation protection procedures, the development of conditions for implementing
these procedures, and the formation of a review system to assess the effectiveness of the
ALARA philosophy.
Operational radiation protection objectives deal with access to radiation areas, exposure to
personnel, and decontamination. Working at or near highly radioactive components requires
planning, special methods, and criteria directed toward keeping occupational radiation exposure
ALARA. Job training and debriefing following selected high exposure jobs contribute toward
reduced exposures. Decontamination also helps to reduce exposure. Procedures and
techniques are based upon operational criteria and experience that have worked to keep
radiation exposure ALARA.
12.1.3.1 Plant Organization
As described in Section 12.5.1, the plant organization provides both the Radiation Control
Supervisors and Superintendent - Radiation Control access to the Plant General Manager. This
organization allows the Plant General Manager to be involved in the review and approval of
specific ALARA goals and objectives as well as review of data and dissemination of information
related to the ALARA program.
The organization also provides the Specialist - ALARA through the Superintendent - Radiation
Control, who is normally free from routine health physics activities, to implement the plant's
ALARA program. This individual is primarily responsible for coordination of plant ALARA
activities and routinely interfaces with first line supervision in radiation work planning and post-
job review.
12.1.3.2 Operating Experience
The Radiation Work Permit process described in Section 12.5.3 provides a mechanism for
collection and evaluation of data relating to personnel radiation exposure. Information, collated
by systems and/or components and job function, assists in evaluating design or procedure
changes intended to minimize future radiation exposures.
The Radiation Control Supervisors and the Specialist - ALARA, are responsible for the review of
radiation exposure records, investigating not only the individual exposures, but the exposures
as classified by job description and job location.
HBR 2 UPDATED FSAR 12.1.3-2 Revision No. 17 12.1.3.3 Exposure Reduction Specific radiation exposure reduction techniques that are used at HBR are described in Section
12.5.3. Procedures assure that: applicable plant activities are completed with adequate
preparation and planning; work is performed with appropriate health physics recommendations
and support; and, results of post-job data evaluation are applied to implement improvements.
In addition, the radiation control staff is, at all times, vigilant for ways to reduce radiation exposures by soliciting employee suggestions, evaluating origins of plant exposures, investigating unusual exposures, and assuring that adequate supplies and instrumentation are
available.
HBR 2 UPDATED FSAR
12.2.0-1 Revision no. 19 12.2 RADIATION SOURCES
Radiation shielding for normal plant operation is designed for operation at the maximum
licensed thermal power level, and for safe operation with one percent fuel element defects.
Reactor coolant inventories of fission products are discussed in Chapter 11 and remain
bounding for operation at 2339 MWt within the RCS activity limits contained in the Technical
Specifications. .
HBR 2 UPDATED FSAR 12.3.1-1 Revision No. 21 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 SHIELDING
12.3.1.1 Design Basis
Radiation shielding is designed for operation at maximum calculated thermal power and to limit
the normal operation levels at the site boundary to below those levels allowed for continuous
non-occupational exposure. The plant is capable of safe operation with the RCS inventory of
fission products discussed in Chapter 11.
In addition, the shielding provided ensures that in the event of a design basis accident the
contained activity does not result in offsite radiation exposure in excess of 10 CFR 50.67.
The shielding is divided into five categories according to function. These functions include the
primary shielding, the secondary shielding, the accident shielding, the fuel transfer shielding, and the auxiliary shielding.
12.3.1.2 Primary Shield
The primary shield is designed to:
a) Reduce the neutron fluxes incident on the reactor vessel to limit the radiation induced increase in transition temperature.
b) Attenuate the neutron flux sufficiently to prevent excessive activation of plant components.
c) Limit the gamma flux in the reactor vessel and the primary concrete shield to avoid excessive temperature gradients or dehydration of the primary shield.
d) Reduce the residual radiation from the core, reactor internals and reactor vessel to levels which will permit access to the region between the primary and secondary shields after plant shutdown.
e) Reduce the induced secondary radiation leakage to obtain optimum division of the shielding between the primary and secondary shields.
The primary shield consists of the core baffle, water annuli, barrel-thermal shield (all of which
are within the reactor vessel), the reactor vessel wall, and a concrete structure surrounding the
reactor vessel.
The primary shield immediately surrounding the reactor vessel consists of an annular reinforced
concrete structure extending from the base of the containment to Elevation 275.0 ft. The lower
portion of the shield is a minimum thickness of 6.5 ft of regular concrete (density of 2.3 g/cc)
and is an integral part of the main structural concrete support for the reactor vessel; it extends
upward to the operating floor, with vertical walls 6 ft thick, to form an integral portion of the
refueling cavity. The primary shield has been designed for a uniform temperature of 50EF and
for thermal gradients as shown on Figures 3.8.3-1 through 3.8.3-4. All thermal stress
calculations assumed 70°F as a starting temperature.
HBR 2 UPDATED FSAR 12.3.1-2 Revision No. 16 The primary concrete shield is air cooled to prevent overheating and dehydration from the heat generated by radiation absorption in the concrete. Eight "windows" have been provided in the primary shield for insertion of the out-of-core nuclear instrumentation.
Cooling for the primary shield concrete and the nuclear instrumentation is provided by
12,000 cfm cooling air.
The primary shield neutron fluxes and design parameters are listed in Table 12.3.1-1.
12.3.1.3 Secondary Shield
The main function of the secondary shielding is to attenuate the radiation originating in the
reactor and the reactor coolant. The major source in the reactor coolant is the Nitrogen-16
activity, which is produced by neutron activation of oxygen during passage of the coolant
through the core. The secondary shield will limit the full power dose rate outside the
Containment Building to less than 1 mR/hr.
The secondary shield surrounds the reactor coolant loops and the primary shield. It consists of
interior walls in the containment structure, the operating floor, and the reactor containment
structure. The containment structure also serves as the accident shield.
The main portion of the secondary shield above grade consists of the
3 ft -6 in. cylindrical portion of the reactor containment, and a concrete crane wall, 3 ft thick, surrounding the reactor coolant loops. The secondary shield will attenuate the radiation levels
in the primary loop compartment from a value of 25 rem/hr to a level of less than 1 mR/hr
outside the Reactor Containment Building.
The crane wall has been designed for a uniform temperature of 50ºF and 120ºF.
The secondary shield design parameters are listed in Table 12.3.1-2.
12.3.1.4 Accident Shield
The accident shield ensures safe radiation levels outside the Containment Building following a
maximum hypothetical accident.
The accident shield consists of the 3 ft -6 in. reinforced concrete cylinder capped by a
hemispherical reinforced concrete dome 2 ft -6 in. thick. Supplemental shielding has been
provided for containment penetrations. This includes a 3 ft -6 in. concrete shadow shield for the
equipment access hatch, and a 2 ft concrete shield for the personnel lock.
Smaller penetrations associated with piping, cables, and ventilation are directed into pipe
tunnels which are shielded with 24 in. of concrete. The Control Room is protected with concrete
sidewalls 18 in. thick, and a concrete roof 22 in. thick.
The accident shield design parameters are listed in Table 12.3.1-3.
A design review of plant shielding was performed in response to NUREG-0737 Item II.B.2.2 to
ensure adequate post accident access to vital areas.
HBR 2 UPDATED FSAR 12.3.1-3 Revision No. 20 12.3.1.5 Fuel Handling Shield The fuel handling shield permits the safe removal and transfer of spent fuel assemblies and
control rod clusters from the reactor vessel to the spent fuel pit. It is designed to attenuate
radiation from spent fuel, control clusters, and reactor vessel internals to less than 2.5 mR/hr at
the refueling cavity water surface and less than 1.0 mR/hr in the Fuel Handling Building.
The refueling cavity is irregularly shaped, formed by the upper portions of the primary shield
concrete, and other sidewalls of varying thicknesses. A portion of the cavity is used for storing
the upper and lower internals packages; these are shielded with concrete walls, 3 and 4-ft thick, respectively. The remaining walls are a minimum thickness of 6 ft of concrete, and provide the
shielding required during transfer of spent fuel.
The refueling cavity, which is flooded with borated water to between approximate elevations 274
ft and 273 ft (consistent with the normal SFP level of between 37 ft 5/8 in. and 36 ft 2 1/2 in.)
during refueling operations, provides a temporary water shield above the components being
withdrawn from the reactor vessel. The water height during refueling is approximately 25 ft
above the reactor vessel flange. This height ensures that a minimum of 10 ft 6 in. of water will
be above the active fuel of a withdrawn fuel assembly. Under these conditions, the dose rate is
less than 2.5 mR/hr at the water surface.
The spent fuel assemblies and control rod clusters are remotely removed from the reactor
containment through the horizontal spent fuel transfer tube and placed in the spent fuel pit.
Concrete, 6 ft thick, shields the spent fuel transfer tube. This shielding is designed to protect
personnel from radiation during the time a spent fuel assembly is passing through the main
concrete support of the reactor containment and the transfer tube.
Radial shielding during fuel transfer is provided by the water and concrete walls of the fuel
transfer pit. An equivalent of 6 ft of concrete is provided to ensure a maximum dose value of 1.0
mR/hr in the areas adjacent to the spent fuel pit.
Fuel is stored in the spent fuel pit of the Fuel Handling Building which is located adjacent to the
Containment Building. Shielding for the spent fuel storage pit is provided by 6 ft thick concrete
walls, and the pit is flooded to a level such that the normal water level is at least 21 ft above the
stored spent assemblies. During spent fuel handling a minimum of 7 ft 6 in. is maintained above
the fuel.
The refueling shield design parameters are listed in Table 12.3.1-4.
12.3.1.6 Auxiliary Shielding
The function of the auxiliary shielding is to protect personnel working near various system
components in the Chemical and Volume Control System, the Residual Heat Removal System, the Waste Disposal System and the Sampling System. The shielding provided for the Auxiliary
Building is designed to limit the dose rate to less than 1 mR/hr in normally occupied areas, and
at or below 2.5 mR/hr in periodically-occupied areas.
HBR 2 UPDATED FSAR 12.3.1-4 Revision No. 14 The auxiliary shield consists of concrete walls around certain components and piping which process reactor coolant. In some cases, the concrete block walls are removable to allow
personnel access to equipment during maintenance periods. Periodic access to the Auxiliary
Building is allowed during reactor operation. Each equipment compartment is individually
shielded so that compartments may be entered without having to shut down and, possibly, to
decontaminate the adjacent system.
The shield material provided throughout the Auxiliary Building is concrete (density of 2.3 g/cc).
The principal auxiliary shielding provided is tabulated in Table 12.3.1-5.
HBR 2 UPDATED FSAR 12.3.1-5 Revision No. 19 TABLE 12.3.1-1 PRIMARY SHIELD NEUTRON FLUXES AND DESIGN PARAMETERS CALCULATED NEUTRON FLUXES ENERGY GROUP INCIDENT FLUXES
- LEAKAGE FLUXES *
(n/cm 2 - sec) (n/cm 2 - sec) E > 1 Mev 1.2 x 10 9 1.4 x 10 2 5.3 kev < E <
1 Mev 2.7 x 10 10 3.3 x 10 2 .625 ev < E < 5.3 kev 1.6 x 10 10 6.0 x 10 2 E < .625 ev 4.1 x 10 9 7.0 x 10 4 DESIGN PARAMETERS Core thermal power 2300 MW
- Active core height 144 in. Effective core diameter 119.7 in.
Baffle wall thickness 1.125 in.
Barrel wall thickness 2.03 in. Thermal shield wall thickness 2.69 in. Reactor vessel ID 155.5 in. Reactor vessel wall thickness 9.31 in. Reactor coolant cold leg temperature 555ºF Reactor coolant hot leg temperature 611ºF Maximum thermal neutron flux exiting primary concrete7.0 x 10 4 n/cm 2-sec
- Reactor shutdown dose exiting primary concrete
<15 mR/hr
- Although this analysis was performed for a power level of 2300 MWt, evaluations have shown that either the analysis remains bounding for operation at 2339 MWt or the impact on the analysis is negligible.
HBR 2 UPDATED FSAR 12.3.1-6 Revision No. 19 TABLE 12.3.1-2 SECONDARY SHIELD DESIGN PARAMETERS Core power density 86.4 w/cc
- Reactor coolant liquid volume 9400 ft 3 Reactor coolant transit times:
Core 0.9 sec Core exit to steam generator inlet 2.0 sec Steam generator inlet channel 0.6 sec Steam generator tubes 3.2 sec Steam generator tubes to vessel inlet 2.7 sec Vessel inlet to core 2.1 sec Total Out of Core 10.6 sec Total power dose rate outside secondary shield
<1 mR/hr
- Although this analysis was performed for a power level of 2300 MWt, evaluations have shown that either the analysis remains bounding for operation at 2339 MWt or the impact on the
analysis is negligible.
HBR 2 UPDATED FSAR 12.3.1-7 Revision No. 19 TABLE 12.3.1-3 ACCIDENT SHIELD DESIGN PARAMETERS Core thermal power 2300 MW
- Minimum full power operating time 1000 days Equivalent fraction of core melting 1.0 Fission product fractional releases:
Noble gases 1.0 Halogens 0.5 Remaining fission product inventory 0.01 Clean-up rate following accident 0
Maximum integrated dose (infinite exposure) in the Control Room
<2rem
- Although this analysis was performed for a power level of 2300 MWt, evaluations have shown that either the analysis remains bounding for operation at 2339 MWt or the impact on the
analysis is negligible.
HBR 2 UPDATED FSAR 12.3.1-8 TABLE 12.3.1-4 REFUELING SHIELD DESIGN PARAMETERS Total number of fuel assemblies 157 Minimum full power exposure 1000 days Minimum time between shutdown and fuel handling 96 hr Maximum dose rate adjacent to
spent fuel pit 1.0 mR/hr Maximum dose rate at water surface 2.5 mR/hr
HBR 2 UPDATED FSAR 12.3.1-9 Revision No. 19 TABLE 12.3.1-5 PRINCIPAL AUXILIARY SHIELDING
COMPONENT CONCRETE SHIELD THICKNESS (ft - in.)
Demineralizers 4 - 0 Charging pumps 2 - 4 Liquid holdup tanks 2 - 6 Volume control tank 3 - 7 Reactor coolant filter 2 - 9 Gas stripper 2 - 6 Gas decay tanks 6 - 0 Gas compressor 2 - 8 Waste evaporator 2 - 0 Liquid waste holdup tank 2 - 0 Design parameters for the auxiliary shielding include:
Core thermal power 2300 Mwt
- Fraction of fuel rods containing small clad defects 0.01 Reactor coolant liquid volume 9400 ft 3 Letdown flow (normal purification) 60 gpm Cesium purification flow (intermittent) 60 gpm Cut-in concentration deborating demineralizer 160 ppm Dose rate outside Auxiliary Building
<1 mR/hr Dose rate in the building outside shield walls
<2.5 mR/hr
- Although this analysis was performed for a power level of 2300 MWt, evaluations have shown that either the analysis remains bounding for operation at 2339 MWt or the impact on the analysis is negligible.
HBR 2 UPDATED FSAR 12.3.2-1 12.3.2 VENTILATION
Ventilation is discussed in Section 9.4.
HBR 2 UPDATED FSAR 12.3.3-1 Revision No. 26 12.3.3 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION
The Radiation Monitoring System (RMS) consists of the following:
a) Area Radiation Monitoring System
b) Process and Effluent Radiological Monitoring and Sampling System
The Area Radiation Monitoring System is described below, while the Process and Effluent
Monitoring and Sampling System (which includes Airborne Radiation Monitoring) is described in
Section 11.5.
12.3.3.1 Area Radiation Monitoring System
12.3.3.1.1 Design Basis
The Radiation Monitoring System is designed to perform two basic functions:
a) Warn of any radiation health hazard which might develop
b) Give early warning of a plant malfunction which might lead to a health hazard or plant damage Instruments are located at selected points in and around the plant to detect, compute, and record the
radiation levels. In the event the radiation level should rise above a desired setpoint, an alarm is initiated in the Control Room. The automatic Radiation Monitoring System operates in conjunction
with regular and special radiation surveys and with chemical and radiochemical analyses performed
by the plant staff. Adequate information and warning are thereby provided for the continued safe
operation of the plant and assurance that personnel exposure does not exceed 10CFR20 limits.
The only components of these systems which are located in the containment are the detectors for
certain area monitoring channels. The low range channels would not be expected to operate
following a major loss-of-coolant accident and are not designed for this purpose. Components of all
other area monitoring channels are designed for post-accident operation. Area monitors are of a
nonsaturating design so that they "peg" full scale if exposed to radiation levels above full scale
indications. The Time-to-Count technique employed in the GM tube detectors eliminate the dead
time and the saturation effects. The Ion chamber detector operates in 'current mode' such that
increasing dose rate beyond full scale continues to increase or maintain current output of the
detector tube.
The components of the Radiation Monitoring System in the control room are designed according to
the following normal environmental conditions:
a) Temperature - an ambient temperature range of 70º to 77ºF
b) Humidity - 20 to 80 percent relative humidity
c) Pressure - normal atmospheric pressure (components inside the containment are designed to withstand test pressure).
d) Radiation - Negligible HBR 2 UPDATED FSAR 12.3.3-2 Revision No. 15 12.3.3.1.2 General system description
The Area Radiation Monitoring System consists of twelve channels which monitor radiation
levels in various areas of the plant. These areas are as follows:
CHANNEL AREA MONITOR R-1 Control Room R-2 Containment R-3 Pass Panel Area R-4 Charging Pump Room R-5 Spent Fuel Building R-6 Sampling Room R-7 In-Core Instrumentation Cubicle R-8 Drumming Station R-9 Failed Fuel R-32A Containment R-32B Containment R-33 Monitor Building
All of the Control Room system equipment is centralized in three cabinets. High reliability and
ease of maintenance are emphasized in the design of this system. Sliding channel drawers are
used for rapid replacement of units, assemblies, and entire channels. It is possible to
completely remove the various chassis from the cabinet, after disconnecting the cables from the
rear of these units.
A Multipoint strip chart recorder is provided in the Radiation Monitoring System cabinets in the
Control Room. Each monitoring channel can be recorded.
Table 12.3.3-1 indicates the detector medium and temperature conditions during normal
operation.
The relation of the radiation monitoring channels to the systems with which they are associated
is given in the sections describing those systems. Routine test and recalibrations will ensure
that the channels operate properly.
HBR 2 UPDATED FSAR 12.3.3-3 Revision No. 26 12.3.3.1.2.1 Channel R1, 2, 3, 4, 5, 6, 7, 8, 9 and 33
Each channel has a fixed position gamma sensitive GM tube detector (except for channel R-9, which has an Ion Chamber detector) assembly located remotely in the plant. Each detector
contains a source check mechanism and signal processing electronics. Each detector is
connected to a field located ratemeter which provides power to operate the detector and
provides outputs to a ratemeter in the control room. The control room ratemeter provides
outputs to the control room recorder, the local readout, the ERFIS computer, and upon alarm, to
the RTGB annunciator panel. The field located ratemeter indicates alarm conditions and
provides audible warning on a high radiation alarm. R-1 remote buzzer is disabled since the
control room is manned all the time. R-1 also provides a signal to operate control room
ventilation equipment on an alarm signal.
The pulse rate from each detector is averaged by the ratemeter, converted to a reading in Mr/hr
and displayed digitally at the ratemeter. Local readouts are on digital ratemeters. The range of
R-1 through R-8 is 0.1 mR/hr to 10,000 mR/hr. The range of R-9 and R-33 is 1.0 mR/hr to
100,000 mR/hr.
The control room ratemeter front panel consists of the following:
A vacuum fluorescent display where channel activity and error messages are displayed;
A three position key switch for selecting ratemeter operating mode;
A six button keypad for accessing and modifying ratemeter data, and performing ratemeter
functions;
Four indicator lamps representing High Alarm, Alert Alarm (not used), Normal Operation, and
Fail Alarm.
12.3.3.1.2.2 Channel 32A and 32B
The Post Accident Containment Radiation Monitoring System is designed to provide
measurement of containment radiation exposure levels during an accident to help assess the
severity of the accident.
The system consists of two, independent monitors with a measurement range of 1R/hr to
10 7 R/hr. The monitors and associated equipment are qualified for continuous post-accident operation. The monitors are mounted on opposite sides of the containment building and are
powered by independent vital power supplies. Readout and recording for each monitor are
provided in the control room.
HBR 2 UPDATED FSAR 12.3.3-4 Revision No. 15 TABLE 12.3.3-1 AREA RADIATION MONITORING SYSTEM DETECTING MEDIUM CONDITIONS
CHANNEL MEDIUM TEMPERATURE RANGE (ºC)
R-1 Air 10-50 R-2 Air 10-50 R-3 Air 10-50 R-4 Air 10-50 R-5 Air 10-50 R-6 Air 10-50 R-7 Air 10-50 R-8 Air 10-50 R-9 Air 10-50 R-32A Air 0-177 R-32B Air 0-177 R-33 Air 10-50
HBR 2 UPDATED FSAR 12.3.3-5 Revision No. 25 12.3.4 Criticality Monitors
H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2 has chosen to comply with the
provisions of 10 CFR 50.68(b) "Criticality accident requirements," in lieu of providing the
criticality monitoring and alarm system required by 10 CFR 70.24. A discussion of the
compliance with each of the criteria of 10 CFR 50.68(b) is provided below.
12.3.4.1 Criterion 1
"Plant procedures shall prohibit the handling and storage at any one time of more fuel
assemblies than have been determined to be safely subcritical under the most adverse
conditions feasible by unborated water."
Response:
Fuel handling procedures are written to instruct the operators to move one fuel assembly from
the shipping cask, allow for fuel inspection, and insert the fuel assembly into the new fuel
storage rack, before moving a subsequent assembly. Additionally, HBRSEP possesses only
one new fuel crane; therefore, handling of more than one new fuel assembly at a time in
accordance with procedural controls is prevented by this material limitation. Criticality analyses
which incorporate administrative controls show that subcriticality will be maintained with k eff less than 0.95 for unborated water and less than 0.98 in an optimum moderation condition in the new
fuel storage areas.
Fuel handling activities in other areas are procedurally controlled to preclude accidental
criticality. In the spent fuel pool during the storage and handling of the fuel, analysis requires a
minimum boron concentration of 800 ppm to be safely subcritical for an assumed mishandling
event. Technical Specifications 3.7.13 mandates a minimum boron concentration of 1500 ppm
to protect against a dilution event. This provides adequate boron concentration to meet the
mishandling event requirements.
In the cavity, the analysis requires a minimum boron concentration during refueling operations
which involve fuel handling, and is bounded by Technical Specifications 3.9.1 and the Core
Operating Limits Report. Use of refueling equipment and movement of fuel assemblies is
governed by plant procedures. The design of the fuel handling equipment does not allow for
handling of more than one fuel assembly in a local vicinity. For example, there can be one
assembly in the new fuel elevator, but at the same time an assembly could be in transition via
the spent fuel tool to its final location in the spent fuel pool.
Criticality may be possible under certain circumstances described in the Carolina Power & Light (CP&L) Company letter to the NRC dated May 10, 1999, that requested an interpretation of the
criteria of 10 CFR 50.68(b). However, the NRC response to the May 10, 1999, letter dated May
19, 1999, stated that there is no need for an interpretation and "that the staff concludes that H.
B. Robinson 2 meets 10 CFR 50.68." In light of this correspondence from the NRC, criticality
considerations
HBR 2 UPDATED FSAR 12.3.3-6 Revision No. 16 are adequately considered in the reactor vessel and in the spent fuel pool by existing analyses and by the administrative controls in place to control fuel movement and storage in these areas.
12.3.4.2 Criterion 2
"The estimated ratio of neutron production to neutron absorption and leakage (k eff) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of
the maximum fuel assembly reactivity and flooded with unborated water and must not exceed
0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be
performed if administrative controls and/or design features prevent such flooding or if fresh fuel
storage racks are not used."
Response:
The current criticality analysis of record for HBRSEP is documented in EMF-94-113, "H. B.
Robinson New and Spent Fuel Criticality Analysis." This document was submitted to the NRC
by CP&L letter dated July 28, 1994, in support of a request for a license amendment to increase
the allowable fuel enrichment to 4.95 + 0.05 (nominal 4.95) weight percent of U-235. The
amendment was approved by the NRC by letter dated January 5, 1995.
Documentation that the k eff of the fresh fuel storage racks filled with fuel of the maximum permissible U-235 enrichment and flooded with unborated water does not exceed 0.95, at a
95% probability, 95% confidence level, is provided in EMF-94-113, section 5.1. The successful
conclusions of this analysis assume that certain new fuel rack locations are "locked out" and
unavailable for storage of fuel. The locking out is accomplished by the installation of chains and
locks on designated fuel rack cells that prohibit the insertion of a new fuel assembly.
CP&L letter dated May 10, 1999, requested an interpretation regarding whether the phrase "racks are. . .loaded with fuel," requires that rack cells that are "locked out" by administrative
controls be assumed to contain fuel for the purposes of the analysis required by 10 CFR
50.68(b)(2). The "locking out" and administrative controls are described in fuel handling
procedures. The NRC response to the May 10, 1999, letter dated May 19, 1999, stated that
there is no need for an interpretation and "that the staff concludes that H. B. Robinson 2 meets
12.3.4.3 Criterion 3
"If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are
assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-
density hydrogenous fluid, the k eff corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be
performed if administrative controls and/or design features prevent such moderation or if fresh
fuel storage racks are not used."
HBR 2 UPDATED FSAR 12.3.3-7 Revision No. 19 Response:
The optimum moderation condition occurs at about 5% interspersed water volume.
Documentation that the k eff corresponding to this optimum moderation does not exceed 0.98, at a 95% probability, 95% confidence level, is provided in EMF-94-113, section 5.1. The
successful conclusions of this analysis assume that certain new fuel rack locations are "locked
out" and unavailable for storage of fuel. Administrative controls are adequate to assure that "locked out" fuel storage locations do not need to be included in the assumed fuel storage
locations for the criticality analyses.
12.3.4.4 Criterion 4
"If no credit for soluble boron is taken, the k eff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95
percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the
k eff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with
borated water, and the k eff must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."
Response:
The HBRSEP, Unit No. 2 analysis for the low density spent fuel racks does not take credit for
soluble boron. Documentation that the k eff of low density spent fuel storage racks filled with fuel of the maximum permissible U-235 enrichment, and flooded with pure water, is less than or
equal to 0.95, at a 95% probability, 95% confidence level, is provided in EMF-94-113, section
5.2. The analysis assumes fuel with an enrichment of 5.0 weight percent of U-235.
The HBRSEP, Unit No. 2 analysis for the high density spent fuel racks does take credit for
soluble boron. Documentation that the k eff of high density spent fuel racks filled with fuel of the maximum permissible U-235 enrichment, and flooded with pure water, remains less than 1.0 at
a 95% probability, 95% confidence level, is provided in report HI-992350. The analysis
assumes fuel with an enrichment of 5.0 weight percent of U-235. If flooded with borated water (minimum 200 ppm soluble boron concentration), the k eff will remain less than or equal to 0.95, at a 95% probability, 95% confidence level.
12.3.4.5 Criterion 5
"The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary
for a critical mass."
Response:
The quantity of SNM other than nuclear fuel stored on site is less than the quantity necessary
for a critical mass. Procedures will contain a precaution that non-fuel quantities of SNM on site
shall be limited to less than the amounts necessary for a critical mass.
HBR 2 UPDATED FSAR 12.3.3-8 Revision No. 20 12.3.4.6 Criterion 6 "Radiation monitors are provided in storage and associated handling areas when fuel is present
to detect excessive radiation levels and to initiate appropriate safety actions."
Response:
The spent fuel pool area has an existing monitor, R-5, that will meet the criterion. This monitor
provides an audible alarm.
Monitoring of fuel in the reactor vessel during fuel handling will be accomplished by the source
range monitors which are required by Technical Specifications section 3.9.2. The audible count
rate which can be heard in containment will warn personnel with an increased count rate. High
rate alarms are provided in the control room and operators will dictate appropriate safety actions
if unexpected readings are noted.
A portable (temporary) radiation monitor will be provided in the new fuel storage and handling
areas when fuel is in storage or being handled in those areas.
Monitors will not be provided in areas where fuel is contained in shipping containers including
the IF-300 cask, since these activities are performed in accordance with 10 CFR 71 or 10 CFR
72 and are, therefore, outside the scope of 10 CFR 50.68.
12.3.4.7 Criterion 7
"The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0)
percent by weight."
Response:
The limitation of U-235 enrichment in fresh fuel is provided in Technical Specifications Section 4.3.1.2, which states: "The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent."
HBR 2 UPDATED FSAR
12.4.0-1 Revision No. 27 12.4 DOSE ASSESSMENT Historic data on annual man-rem doses resulting from operation and maintenance of HBR 2 is
maintained by Duke Energy Progress, LLC.
HBR 2 UPDATED FSAR 12.5.1-1 Revision No. 19 12.5 Health Physics Program 12.5.1 Organization
12.5.1.1 Introduction
The HBR health physics program has been established to provide an effective means of
radiation protection for plant personnel, visitors, and the general public. To provide this
radiation protection, the health physics program incorporates a dedicated philosophy from
management, qualified personnel to direct and to implement the health physics program, the
appropriate equipment and facilities, and written procedures based upon acceptable radiation
protection practices and guidance.
The health physics program at HBR is developed and implemented to evaluate and document
plant radiological conditions and to ensure that every reasonable effort is made to maintain
occupational radiation exposure (ORE) as low as reasonably achievable (ALARA). The
organization of the health physics program provides a flexible, responsive, and comprehensive
structure for attaining these goals. The qualifications of all plant personnel are provided in
Sections 13.1.3 and 1.8.
12.5.1.2 Responsibilities
The Radiation Control Supervisors are under the supervision of the Superintendent - Radiation
Control, and are responsible for providing the information necessary to establish compliance
with regulations pertaining to radiation safety, for uniformly enforcing plant health physics
requirements, and for ensuring every reasonable effort to minimize personnel exposures. In
addition, they are responsible for ensuring that the staff members who implement the health physics program are trained and retrained in operational health physics principles. They are
assisted by a staff which includes a number of RC Technicians. The ALARA program is
implemented and evaluated under the technical direction of an Specialist - ALARA and the
management direction of an RC Supervisor who reports to the Superintendent - Radiation
Control. The Specialist - ALARA may report to an RC Supervisor that provides the
management direction for the ALARA Program, if deemed necessary by the Superintendent -
Radiation Control. The Specialist - ALARA provides technical direction and expertise to the RC
subunit and ensures compliance with the corporate commitment to ALARA expressed in the
Corporate Health Physics Policy.
The RC unit coordinates with the operations, maintenance, and engineering units and provides
health physics coverage for all activities that involve radiation or radioactive material. In
addition, RC provides various other services, including the following:
- 1. Preparing health physics procedures for routine and non-routine activities that may be encountered in operating, maintaining, inspecting, and testing the plant.
HBR 2 UPDATED FSAR 12.5.1-2 Revision No. 18 2. Ensuring that the provisions and standards of 10CFR20 for permissible dose limits and potential release levels are not exceeded.
- 3. Providing a personnel radiation dosimetry program and maintaining dosimetry records
- 4. Providing radiation surveys of plant areas and maintaining survey records
- 5. Assisting in plant training programs by consulting with the plant training organization
- 6. Providing and calibrating radiation detection instruments and equipment for assessing the radiation environment at HBR
- 7. Providing, maintaining, and issuing protective clothing and equipment
- 8. Assisting in the shipping and receiving of all radioactive material to ensure compliance with regulatory requirements
- 9. Preparing, maintaining, and issuing reports of the required regulatory, plant, and personnel records that involve radiological aspects at HBR
- 10. Assisting in the decontamination of personnel, equipment, and facilities at HBR
- 11. Administration of the HBR respiratory protection program
The responsibilities of the Radiation Control Supervisors are to provide the day-to-day execution
of the health physics program through supervision of the routine and special surveys and the
programs required by applicable regulations and procedures. The Radiation Control
Technicians implement the health physics program by performing routine and special surveys and by providing health physics surveillance in accordance with plant health physics
procedures.
It is the responsibility of each individual to obey all radiation control procedures and to report to
either the Radiation Control Supervisors, or Superintendent-Radiation Control or their designees
any circumstances where procedures may be incorrect or unsafe activities may be occurring.
The Supervisor - Radiological Services is available to provide technical direction, guidance, support, and assistance for all radiation control programs and related activities. The
Supervisor-Radiological Services reports to the Manager Material Services. The Manager
Material Services has the organizational freedom to communicate directly with the Chief
Executive Office to resolve any concern in the area of health physics should the concern not be
resolved satisfactorily at a lower management level.
HBR 2 UPDATED FSAR 12.5.1-3 Revision No. 22 Nuclear Oversight (NOS) evaluates the performance and effectiveness of the Company's health physics programs on a continuous basis to assure they are being carried out in an effective
manner. These assessments are to detect deficiencies in the desired levels of performance
and quality, reporting these conditions to the Vice President - Robinson Nuclear Plant, and
ensuring adequate action is taken to correct and eliminate these conditions.
12.5.1.3 Authority
The plant General Manager, who is ultimately responsible for all plant activities including
radiation safety, receives direct reports from the Superintendent - Radiation Control concerning
the status of the health physics program. To ensure uniform enforcement of health physics
requirements, the Plant General Manager delegates his authority to the Superintendent-
Radiation Control and the Radiation Control Supervisors. The Superintendent-Radiation
Control and the Radiation Control Supervisors have the authority to cease any work activity
when, in their judgment, worker safety is jeopardized, or in the event of unnecessary personnel
radiation exposures.
The Radiation Control Supervisors delegate authority to responsible RC Technicians to cease
any work activity which is not being performed in accordance with health physics requirements.
The Radiation Control Technicians have the authority to ensure that jobs are conducted in
accordance with health physics requirements.
In the absence of Radiation Control supervision, the authority associated with the above
positions may be delegated, in accordance with the plant's health physics procedures, to the
most senior health physics individual on shift.
12.5.1.4 Experience and Qualification
The radiation control staff, which is responsible for the health physics program at HBR, meet
minimum experience and qualification requirements as described in Section 1.8. Each
Radiation Control Technician must be qualified in the use and application of radiation protection
control procedures.
HBR 2 UPDATED FSAR 12.5.2-1 Revision No. 18 12.5.2 Equipment, Instrumentation, and Facilities 12.5.2.1 Personnel Protective Equipment
Personnel entering the Radiation Control Area may be required to wear protective clothing. The
nature of the work to be done is the governing factor in the selection of protective clothing to be
worn by individuals. Some of the protective apparel available are shoe covers, head covers, gloves, coveralls, etc. Additional items of specialized apparel such as plastic or rubber suits, and respirators are available for operations involving wet surfaces or airborne radioactivity. In
all cases, health physics-trained personnel shall evaluate the radiological conditions and specify
the required items of protective clothing to be worn.
Respiratory protective devices may be required in situations arising from plant operations in
which an airborne radioactive area exists or is expected. Airborne concentrations are monitored
by health physics personnel and the necessary personnel protective devices are specified
according to the concentration and type of airborne contaminants present while maintaining the
total effective dose equivalent ALARA.
Respiratory devices which may be available for use include:
- 1. Full-face air purifying respirator
- 2. Air supplied full-face respirators or hoods
- 3. Self-contained breathing apparatus.
Respirators are maintained by checking for mechanical defects, contamination, and cleanliness
by qualified personnel.
12.5.2.2 Radiation Instrumentation
Facilities are provided for the Environmental & Chemistry and Radiation Control Group. These
facilities include both laboratory and health physics work areas. These facilities are equipped to
analyze routine air samples and contamination smear surveys. These facilities also serve as a
central location for portable radiation survey instruments, respiratory protection equipment, and
contamination control supplies.
Survey instruments are calibrated periodically.
12.5.2.3 Personnel Monitoring
All individuals, except casual visitors, who enter the Radiation Control Area are monitored for
external whole body radiation exposure using an appropriate individual monitoring device.
Individual monitoring devices used at HBR include (but are not limited to) thermoluminescent
dosimeters (TLDs), electronic pocket dosimeters, and self-reading pocket ionization chambers.
Individuals subject to monitoring for occupational radiation exposure are issued individual
monitoring devices and are required to wear them at all times while in the Radiation Control
Area.
Special or additional individual monitoring devices are issued as may be required at the
discretion of health physics personnel.
HBR 2 UPDATED FSAR 12.5.2-2 Revision No. 21 Individual monitoring devices capable of measuring neutron dose equivalent are issued as required. The neutron dose equivalent may be determined initially by using survey data and
stay times to calculate exposure.
Casual visitors are not subject to monitoring, recordkeeping, and reporting requirements of 10
CFR 19 and 10 CFR 20. However, they are issued a self-reading dosimeter (e.g., pocket
ionization chamber or electronic pocket dosimeter) for verification purposes.
12.5.2.4 Facilities and Access Provisions
The plant site is divided into two categories, the Clean Area and the Radiation Control Area for
radiation protection purposes as shown on Figures 12.5.2-1 and 12.5.2-2.
Normal entry to and exit from the primary Radiation Control Area is through a single access
control point. Additional access control points may be established to support plant operations.
Numerous locations for emergency Radiation Control Area entry/exit are available.
Satellite Radiation Control Areas exist in other areas on plant site and access to these areas is
administratively controlled by the Environmental and Radiation Control Organization.
Radiation Control Areas are surveyed, classified, and conspicuously posted in accordance with
10 CFR 20 regulations.
The general arrangement of the service facilities is designed to provide personnel
decontamination and change areas (see Figure 12.5.2-1). Clean locker rooms in the RCA
Access Facility are used to store individuals' clothing when they must dress out to perform work
in the Radiation Control Area. A dress out area is provided in the Radiation Control Area. A
supply of clean protective clothing for personnel is maintained in this area. Appropriate
personnel contamination survey devices are located at the exit point of the Radiation Control
Area so that personnel can survey themselves upon leaving the Radiation Control Area. A
decontamination shower and washroom is located in the Radiation Control Area.
Decontamination areas are also provided in the Radwaste Building for the decontamination of
hand tools and small equipment. Decontamination areas are established or removed in other
plant areas based on plant configuration or need. In addition, satellite decontamination areas
may be established for special conditions or projects.
HBR 2 UPDATED FSAR 12.5.2-3 Revision No. 14 Administrative and physical security measures are employed to prevent unauthorized entry of personnel to any High Radiation Area. These measures include the following:
- 1. Any areas accessible to individuals in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 100 mRem in any one hour at 30
centimeters from the radiation source or from the surface the radiation penetrates, are
barricaded and conspicuously posted as "High Radiation Areas".
- 2. Locations where the above value exceeds 1 Rem in any one hour at 30 centimeters from the radiation source or from any surface the radiation penetrates are conspicuously
posted and, in addition, locked doors are provided to prevent unauthorized entry.
- 3. Areas, accessible to personnel, in which radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in any one hour at one meter from a radiation source or from any surface that radiation penetrates, are designated as Very
High Radiation Areas. Accesses to these areas are administratively controlled. Entry
into Very High Radiation Areas is forbidden unless there is a sound operational or safety
reason for entering.
- 4. Any individual or group of individuals entering "high radiation areas" is provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously monitors the radiation dose in the area.
- b. A radiation monitoring device provided for each individual which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels
in the area have been established and personnel have been made knowledgeable of
them. c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing
positive control over the activities within the area and shall perform periodic radiation
surveillance at frequency specified by the Radiation Control Supervisor in the Radiation
Work Permit.
HBR 2 UPDATED FSAR 12.5.3-1 Revision No. 25 12.5.3 Procedures The health physics procedures developed for HBR are an integral part of the ALARA policy, as
discussed in Section 12.1. These procedures, developed through careful planning and
preparation and utilized by well-trained and qualified personnel, contribute significantly to the
overall reduction of the occupational radiation exposures. The health physics procedures cover
the appropriate administrative, operating, and ALARA-related operations and conditions at HBR.
As indicated in Section 12.1, ALARA considerations have been embodied in applicable
procedures. In addition, procedures for personnel radiation protection shall be prepared
consistent with the requirements of 10CFR20 and shall be approved, maintained and adhered
to for all operations involving personnel radiation exposure.
12.5.3.1 Control of Access and Stay Time in Radiation Areas
As specified in Section 12.1, physical and administrative controls are instituted at HBR to
ensure that the philosophy of maintaining personnel exposures as low as reasonably achievable (ALARA) is implemented.
12.5.3.1.1 Physical controls
12.5.3.1.1.1 Security check point and access control
The plant's security checkpoints are continuously manned. TLD badges issued to individuals
will remain the custody of the individual. All personnel will maintain their TLD with their security
badge. Each individual will be responsible for bringing their security badge to work each day
and ensuring the TLD is available to be exchanged during TLD changeout period. A restricted
area access list is maintained at the security entrance. Any individual not authorized access
must be accompanied by a person who is an authorized escort for restricted areas. The
training, retraining, and testing requirements for unescorted access are described in the HBR
Plant Access Training Procedures.
12.5.3.1.1.2 Door and area posting and locking
Physical control is provided by the posting and locking, as appropriate, of Radiation Areas, High
Radiation Areas, and very High Radiation Areas. These areas, as defined in 10CFR20.1003, are posted in accordance with 10CFR20.1902. Plant areas that are routinely accessible are
surveyed in accordance with plant procedures to determine radiation and contamination levels.
The surveys are performed on a frequency that is appropriate with the potential hazard present.
In addition to recording the results of these surveys in accordance with 10CFR20.2103, the
radiological postings are updated to reflect operating conditions.
12.5.3.1.1.3 Health physics surveillance
When appropriate, health physics surveillance of work activities is provided to assure a positive
control of access and stay time in radiation areas. Surveillance may also be provided for tasks
in areas where conditions may warrant timely instructions to workers (e.g., on jobs where the
radiological conditions can fluctuate greatly).
HBR 2 UPDATED FSAR 12.5.3-2 Revision No. 19 12.5.3.1.2 Administrative controls
12.5.3.1.2.1 Training
As specified in Section 12.5.3.7, personnel allowed unescorted restricted area access receive
health physics and related training in accordance with 10CFR19.12.
12.5.3.1.2.2 Radiation Work Permits
The Radiation Work Permit (RWP) system described in Section 12.5.3.2 is implemented to
administratively control access and personnel dose in radiation areas. For personnel or groups
who must enter these areas as a necessary part of work duties, a Radiation Work Permit (RWP)
will be issued in accordance with plant procedures.
12.5.3.2 Assuring Occupational Radiation Exposures are ALARA
To effectively implement the corporate ALARA commitment as discussed in Section 12.1.1, an
ALARA program is utilized at HBR to assure that activities are performed with ALARA personnel
exposure. Progress Energy considers it necessary to apply the basic concepts of ALARA to
both internal and external exposure to assure proper emphasis on both modes of potential radiation exposure. Procedures employed to implement the program described are subject to
review and revision to ensure that the ALARA program is responsive to plant needs and
conditions.
12.5.3.2.1 ALARA procedures common to external and internal exposure
12.5.3.2.1.1 Training
Individuals allowed unescorted restricted area access receive health physics training as
described in Section 12.5.3.7. The individual's responsibility to avoid unnecessary exposure is
emphasized during health physics training sessions.
As appropriate, individuals involved in potentially high occupational radiation exposure jobs
receive pre-job instruction in exposure reduction techniques and controls applicable to the
specific job. Post-job reviews are held, as appropriate, to provide a positive feedback on
improved job performance.
12.5.3.2.1.2 Radiation Work Permit (RWP)
For entry into the Radiation Control Area (RCA), an RWP is initiated and approved prior to
commencement of scheduled work. Plant procedures specify that an RWP will be completed
for work that is performed involving the following situations:
- 1. Entry into any area where whole body radiation levels are in excess of 100 mrem/hr.
- 2. Any maintenance work which involves opening of any system which contains, or could potentially contain, radioactive material in excess of limits established by plant
procedures
HBR 2 UPDATED FSAR 12.5.3-3 Revision No. 19 3. Any maintenance of contaminated or potentially contaminated equipment using methods involving abrasion, cutting, machining, or welding Radiation control personnel evaluate the radiological conditions associated with the work to be
performed and specify appropriate protective clothing/devices, respiratory protective equipment, dosimetry, special samples, surveys, and precautions to be taken.
The RWP is evaluated to ensure that the work will be performed utilizing good health physics
practices and an ALARA approach.
The RWP is approved by a Radiation Control Supervisor or designee. The RWP
implementation process is detailed in plant procedures.
12.5.3.2.1.3 Work scheduling
Use of the work management system establishes a data base from which supervisory staff are
able to efficiently schedule workers.
12.5.3.2.1.4 Job pre-planning and exposure goals
Preparing an RWP is a form of job pre-planning. The responsible supervisor ensures that
individuals selected to perform the task are familiar with the appropriate procedures to be
employed.
On major dose accumulating job functions, total person-rem (also called man-rem) exposure
goals are established prior to commencement of scheduled work. Significant deviations above
established goals are investigated by ALARA staff. Methods to improve performance on future
jobs will be investigated and implemented, if appropriate.
12.5.3.2.1.5 ALARA program reviews
In an effort to continually improve the ALARA program, the Specialist - ALARA and radiation
protection personnel utilize benchmarking, operational experience, self-assessment, and the corrective action program to identify deficiencies and improvement opportunities in the ALARA
program. Results of these efforts are documented and reported to the appropriate levels of
plant management.
12.5.3.2.2 External ALARA
12.5.3.2.2.1 Administrative limits
Administrative limits are implemented by plant procedures to maintain personnel exposures
ALARA with respect to federal regulations. Plant radiation exposure limits may be exceeded
only after approval of plant management. Unapproved radiation exposures exceeding plant
limits are investigated by radiation control personnel to identify causes and establish methods to
prevent recurrence.
HBR 2 UPDATED FSAR 12.5.3-4 Revision No. 18 12.5.3.2.2.2 Temporary shielding and special tools
During the planning phase of RWP work, qualified radiation control personnel and the ALARA
staff evaluate the use of temporary shielding. Care is taken to ensure that installation and
removal of shielding does not cause larger man-rem total exposures than expected without its
use.
Every reasonable effort is expended to ensure that any necessary, special, or modified tools are
available for specific tasks.
12.5.3.2.3 Internal ALARA
To minimize potential intake of radioactive material in excess of federal limits, plant limits are
established. Airborne radioactivity concentrations in excess of these limits may require work
restriction, engineering controls, use of respiratory protection, and/or special in-vivo or bioassay
studies.
When work orders indicate that work is required in areas containing potential airborne
radioactive material, appropriate air samples are taken. These data samples are normally of
short-term, low-volume nature in order to obtain representative of normal breathing rates. Any
area that is posted as an airborne radioactivity area is sampled and analyzed prior to
commencement of scheduled work.
12.5.3.2.3.1 Control of absorption
When work is scheduled on equipment or systems that contained or may contain radioactive
liquids, every reasonable effort to prevent skin contact with radioactive solutions is made.
12.5.3.2.3.2 Control of area and equipment contamination levels
Contaminated areas and equipment are decontaminated to as low a level as reasonably
achievable in accordance with plant procedures.
12.5.3.2.3.3 Airborne exposure evaluation
Exposure to airborne radioactive material is evaluated in accordance with 10CFR20.1201(d)
and Subpart H - Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas to aid in work planning and to demonstrate the effectiveness of the internal ALARA program.
12.5.3.3 Radiation Surveys
The health physics program utilizes a comprehensive system of radiation surveys to document
plant radiological conditions and identify sources of radiation that contribute to occupational radiation exposure. Radiation control personnel normally perform radiation and contamination
surveys of all accessible areas in the plant. The surveys are performed on an appropriate
frequency, depending on the probability of radiation and contamination levels changing, and the
frequency with which the areas are visited. Surveys related to specific operations and
maintenance activities may be performed prior to, during, and/or after the activity, based on
information required to keep radiation exposures ALARA.
HBR 2 UPDATED FSAR 12.5.3-5 Revision No. 15 Radiation level surveys may be performed for alpha, gamma, beta, and/or neutron exposure rates. Contamination surveys are normally performed to establish gross beta-gamma
contamination level, but may be processed for specific types of radiation (beta-alpha-gamma) or specific radionuclides (via gamma spectroscopy). Availability of current survey information aids
in keeping exposures ALARA.
The radiation survey program is subject to evaluation by radiation control supervision to ensure
that necessary data are collected while exposures to surveyors are ALARA.
12.5.3.4 Contamination Survey Procedures
A system of periodic contamination evaluations is utilized to minimize the spread of radioactive
material. Evaluation of personnel, equipment and surface contamination is also made to
demonstrate the effectiveness of engineering and procedural controls. In addition, the
contamination survey programs are evaluated to assure that survey personnel exposures are
12.5.3.4.1 Personnel contamination surveys
Evaluation of exposures due to personnel contamination is conducted in accordance with
Section 12.5.3.6.
Instrumentation designed for contamination monitoring is strategically located within the
Restricted Area. Every effort is made to locate these instruments in as low a radiation
background area as possible in order to maximize sensitivity. Personnel are trained in the use
of the instruments and interpretation of the readings.
Personnel contamination causing an alarm requires notification of radiation control personnel.
Radiation control personnel take appropriate actions to minimize further spread of
contamination, and direct appropriate decontamination of affected areas and personnel.
12.5.3.4.2 Equipment contamination surveys
Movement of equipment from the Radiation Control Area requires the assistance of RC
personnel. Fixed and removable contamination levels are evaluated as appropriate and a
clearance for removal is issued in accordance with plant procedures.
Reusable protective clothing and shoe covers used in contamination zones
are collected in receptacles at access areas and sent for laundering/decontamination.
Change-out procedures require that individuals leaving the Radiation Control Area have to
perform surveys of personal items that may have become contaminated during work.
Radioactive material is shipped in accordance with U. S. Department of Transportation and
NRC Regulations. Plant procedures implement the applicable regulations with regard to proper
packaging and labeling requirements. Appropriate removable contamination and dose rate
surveys are taken, records completed, and shipment labeled accordingly.
HBR 2 UPDATED FSAR 12.5.3-6 Revision No. 15 12.5.3.4.3 Surface contamination surveys
A smear survey program is utilized to assure that a representative number of routinely
accessible surface areas within the Radiation Control Area are checked for removable
contamination. Emphasis is placed on survey of the clean side of established contamination
area access areas.
Occupied plant areas outside the Radiation Control Area are surveyed to assure that a
representative number of floor surfaces are checked for removable contamination. The exit
areas from the Radiation Control Area receive emphasis to minimize the spread of
contamination.
Lunch room facilities and vending machine areas frequented by Radiation Control Area workers
are checked for removable contamination. Stoves, benches, table tops, and floor surfaces are
representatively smeared to assure minimal contamination in eating areas.
12.5.3.5 Airborne Radioactive Material
Accessible areas containing concentrations of airborne radioactive materials as specified in
10CFR20.1003 are posted in accordance with 10CFR20.1902(d) and applicable plant
procedures.
12.5.3.5.1 Airborne concentration sampling
Routine sampling in selected areas of potential airborne radioactivity is accomplished with air
monitors. Special air samples are taken, as required. The majority of special air samples are
taken as a result of RWP requests and pertinent results are recorded thereon.
12.5.3.5.2 Respiratory protection
The respiratory protection program assures that personnel intake of radioactive material is
minimized. The respiratory protection program is not used in place of practical engineering
controls and prudent radiation control practices. Every reasonable effort is made to prevent
potential, and minimize existing, airborne concentrations. When controls are not practicable, or
conditions unpredictable, respiratory protective devices may be utilized to minimize potential
intake of airborne radioactive material.
The RNPD Respiratory Protection Program ensures that the following minimum criteria are met:
written standard operating procedures; proper selection of equipment, based on the hazard;
proper training and instruction of users; proper fitting, use, cleaning, storage, inspection, quality
assurance, and maintenance of equipment; appropriate surveillance of work area conditions;
consideration of the degree of employee exposure to stress; regular inspection and evaluation
to determine the continued program effectiveness; program responsibility vested in one qualified
individual and an adequate medical surveillance program for respirator users.
HBR 2 UPDATED FSAR 12.5.3-7 Revision No. 26 12.5.3.5.3 Handling of radioactive material
Recognized methods for the safe handling of radioactive materials are incorporated into
procedures to ensure proper usage. Procedures specify handling techniques, storage, and
other safety considerations, as listed below:
- 1. Minimizing distances that large radioactive sources are transported
- 2. Use of shielded transporters
- 3. Storage of sources in appropriately shielded containers
- 4. Proper labeling of radioactive material containers in accordance with 10CFR20
- 5. Inventorying of all radioactive sources in accordance with plant procedures
- 6. Leak testing of sources at six-month intervals in accordance with license conditions
- 7. Monitoring of all packages received containing radioactive material in accordance with 10CFR20.1906.
- 8. Radioactive material is stored outside the primary radiological control area. These storage locations are typically plant components being stored and under a radiological
surveillance program. In addition, HBR operates 2 ISFSIs; 7P ISFSI under a site
specific license and a 24P ISFSI under a general license. The ISFSIs are used to store
15 x 15 used fuel assemblies. These locations are:
a) Inside Protected Area Dose rates are established and maintained at less than or equal to
1 mrem/hr on the building exterior.
- Building 458: Reactor Head Storage
- Building 385: Steam Generator Storage
- Building 457: Reactor Services ISFSI Dose rates are established and maintained at less than or equal to
2 mrem/hr at the building RCA fence line.
- Building 380: Reactor Services Building 7P ISFSI
- 24P ISFSI b) Outside Protected Area Dose rates are established and maintained at less than or equal to 0.05 mrem/hr on the building exterior.
- Building 425: Spare Reactor Coolant Pump Motor
12.5.3.6 Personnel Monitoring
12.5.3.6.1 External radiation exposure assessment
Individual monitoring devices are used at HBR to evaluate external occupational exposure to
radiation sources. The appropriate individual monitoring devices are issued in accordance with
plant procedures implementing 10CFR20.1502(a).
HBR 2 UPDATED FSAR 12.5.3-8 Revision No. 26 Individuals who are issued individual monitoring devices are instructed in the purpose and use of the devices, plant administrative exposure limits, and interpretation of the self-reading
monitoring device data.
Administrative exposure limits are established and implemented by health physics procedures
to ensure the limits of 10CFR20.1201 are not exceeded and personnel occupational exposures
are maintained ALARA.
12.5.3.6.2 Internal radiation exposure assessment
To demonstrate the effectiveness of engineering controls and the respiratory protection
program, personnel are periodically monitored for internal radioactivity. The following methods
for assessing internal exposure are used to ensure compliance with occupational dose
equivalent limits:
- 1. Air sampling to determine the concentrations of airborne radioactive materials in the work area; or
- 2. Whole body counting techniques to determine the quantities of radioactive material in the body; or 3. Bioassay techniques to determine the quantities of radioactive material excreted from the body; or 4. Any combination of the above listed methods.
12.5.3.6.3 Methods of recording and reporting
Updates of exposure totals are compiled from self-reading pocket dosimeter readings.
Unapproved exposures exceeding plant limits will be reported to the Plant General Manager
and appropriate supervision, and investigated by radiation control personnel to identify causes
and establish methods to prevent recurrence.
Occupational radiation exposure received during previous employment is used in preparation of
individuals' Forms NRC-4, or equivalent in accordance with 10CFR20.2104. Records used in
preparing Form NRC-4, or equivalent, are retained and preserved until the NRC authorizes
disposition.
Records of the radiation exposure of personnel who are issued individual monitoring devices in
accordance with 10CFR20.1502 are maintained on Form NRC-5, or equivalent. Records of
radiation exposures of individuals receiving exposure under the provisions of 10CFR20.1206
are documented in accordance with 10CFR20.2105 and maintained on Form NRC-5 or
equivalent.
Reports of exposure to radiation or radioactive materials are made to individuals as specified in
10CFR19.13. When reports of individual exposure to radiation or radioactive material are made
to the NRC, the individual concerned is also notified. This notice is forwarded to the individual
at a time no later than the transmittal to the NRC and complies with 10CFR19.13.
HBR 2 UPDATED FSAR 12.5.3-9 Revision No. 26 Reports of individual monitoring are submitted in accordance with 10CFR20.2206 on or before
April 30 of each calendar year. As part of a routine annual operating report, personnel
exposure information is submitted within the first quarter of each calendar year. It includes a
tabulation of the number of plant, utility, and other personnel (including contractors) for whom
monitoring was required or provided, the exposures then received, and associated man-rem
exposure according to work and job functions.
Reports of overexposures at HBR are submitted to the NRC and the individual(s) involved in
accordance with 10CFR19.13 and 10CFR20.2203. Reports are also forwarded to appropriate
committees for review and recommendation for follow-up action
12.5.3.7 Health Physics Training Programs
Health physics training programs assure that personnel, who have unescorted access to the
restricted area, possess an adequate understanding of radiation protection to maintain
occupational radiation exposures ALARA. Special training/retraining is administered upon
recommendation of the Training Supervisor or Radiation Control Supervisor. Record-keeping
and training scheduling is performed by the Manager - Training or designated alternate. This
program covers the following:
- 1. General employee health physics
- 2. General employee respiratory protection
- 3. Contractor health physics
- 4. Contractor respiratory protection
- 5. General employee retraining
- 6. Radiation control technician training
- 7. Radiation control technician retraining
12.5.3.7.1 Health physics training
Persons allowed unescorted access to Radiologically Controlled areas must demonstrate
proficiency in the following areas as evidenced by passing a computer based or written
examination:
- 1. Requirements of 10 CFR 19.12
- 2. Radiation/Contamination (examples and control)
- 3. ALARA (Corporate commitments, meaning, and individual responsibility)
- 4. Personnel Monitoring and Self-Survey Requirements
- 5. Radiological Control Signs and Posting Requirements HBR 2 UPDATED FSAR 12.5.3-10 Revision No. 26
- 6. Radiation Exposure Control and Limits
- 7. Radiation Emergency Plan and Applicable Procedures
- 8. Prenatal Radiation Exposure
Training is administered to provide radiation workers with an adequate knowledge to effectively cope with job situations while maintaining radiation exposures as low as reasonably achievable.
To assure individual proficiency in radiation protection practices, retraining/retesting should be
performed annually 25% (maximum 15 months). Scheduling, records and test results are
maintained by the Manager - Training or designated alternate. Individuals changing job
classification receive training of the level required by their new job classification.
12.5.3.7.2 Respiratory protection training program
Individuals requiring access to areas where respiratory protection is utilized are required to
complete the Respiratory Protection Training Program. The instructor is a qualified individual
with a thorough knowledge regarding the application and use of respiratory protective
equipment and the hazards associated with radioactive airborne contaminants.
Training may include lectures, computer based training (CBT), demonstrations, discussions of
pertinent plant procedures, and actual wearing of respirators to become familiar with the various
devices utilized at HBR.