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==SUMMARY==
OF THE TESTIMONY OF' GERALD P. LAHTI                                                  -
i                                        REGARDING CONTENTIONS 42, 111 AND 112 Gerald P. Lahti is the Assistant Division Head.                                          ;
of the Nuclear Safeguards and Licensing. Division in charge i
of Shielding and Radiological Safety at Sargent &.Lundy. As such Mr. Lahti is responsible for the design of radiation J
i shielding and other radiation protection features at the Byron Station. 'His testimony' addresses those portions of                                          .
Contentions 42, 111 and 112 which. assert that plant design for reducing occupational radiation exposure was.not con-sidered in the Byron design.
.L Sargent & Lundy lis responsible for the design of 4
the so-called " balance of plant", i.e. all structures and a
components other than nuclear steam supply system, the steam 4
i                generators and ancillary equipment. The design of radiation
;                protection features for Byron assures that operation and j
maintenance of the plant do-not result in doses of radiation in excess of those allowed by NRC regulations and are in com-pliance with regulatory ALARA principles.                                Mr..Lahti states that there are . 4 well-known design principles which reduce I,
radiation exposure: shielding, distance, reduction of the. time j                spent by the worker in a radiation area and removal of the radioactive source. Each of these principles has been applied
,i as appropriate to the Byron Station design. Lahti Exhibit                                                    1 contains extracts from the FSAR for Byron which detail the manner in which the Byron design reduces occupational radiation exposure.
                                          -,r 8302220325 830215
'      PDR ADOCK 05000454 T                    PDR
 
  . o TESTIMONY OF GERALD P. LAHTI REGARDING CONTENTIONS 42, 111 AND 112 0.1.        State your name and present occupation.
A.l.        My name is Gerald P. Lahti. I am Assistant Division Head of the Nuclear Safeguards and Licensing Division in charge of Shielding and Radiological Safety at Sargent &
Lundy in Chicago, Illinois.
Q.2.        Briefly state your educational and professional qualifications.
A.2.        I received a BSCE in Civil Engineering from Wayne State University in 1959. I received a MSE (Nuclear Engineering) from the University of Michigan in 1960, and completed additional part time course work in Mechanical and Nuclear Engineering at the University of Delaware, Case Western Reserve University, and the University of Toledo. From 1960 to 1963 I was employed by E. I. duPont deNemours & Co., Inc., and mathematically analyzed and designed polymer transfer systems and extrusion dies.
From 1963 to 1973 I was a member of the National Aero-nautics and Space Administration (NASA) staff at Lewis Research Center, Cleveland, Ohio. There, I evaluated radiation hazards and designed radiation shields for nuclear reactors considered for power or propulsion
 
systems in space vehicles. In 1968 I assumed super-visory responsibilities in'this area. In 1973 I joined Sargent & Lundy and have been employed in the i
Shielding and Radiological Safety Section continuously since that time. The Shielding and Radiological Safety Section, which is under my supervision, designs and
;            evaluates all radiation shielding and other radiation protection features incorporated in nuclear power plant design. I am also responsible for assessing the radiological impact of radionuclides released during normal and abnormal power plant operations. I am a Registered Professional Engineer in the State of Illinois and a member of the American Nuclear Society and Health
;            Physics Society. I am a past Chairman of the ANS's Radiation Protection and Shielding Division.
l l
0.3.      What is the scope of your testimony?
A.3.      My testimony addresses those portions of Contentions 42, 111 and 112 which assert that plant design for re-ducing occupational radiation exposure were
;            not considered in the design of the Byron Station.
I Q.4.      What have been your duties and responsibilities with respect to the design of radiation protection features for Commonwealth Edison Company's Byron Nuclear Station?
l      A.4.      To supervise the activities of engineers assigned to the Shielding and Radiological Safety Section to make L
 
certain that the design of Byron Station is such'that operation and expected maintenance does not result in radiation doses in excess of the levels specified in 10 CFR S 20.101 and takes account of other NRC regulations regarding occupational radiation exposure, specifically the ALARA (as-low-as-reasonably-achievable) provision of 10 CFR S 20.1.
Q.5.        What is the scope of the Sargent & Lundy design effort in connection with compliance with NRC regulations regarding occupational radiation exposure?
A.5.        Westinghouse is responsible for the design of the nuclear steam supply system, that is the nuclear reactor, the steam generators and ancillary equipment. _ Generally, Sargent and Lundy is responsible for the design of what is designated " balance of plant". The balance of plant includes the structure to house the NSSS and the design of mechanical systems other than the NSSS.      Some of these systems, such as the radwaste system, will routinely contain radioactive materials over the operating life of the plant. Others, although not themselves handling radioactive materials, are in proximity to radioactivity.
Q.6.      Are there certain design principles which reduce radiation exposure and which have been taken into account in the design of Byron Station?
 
    . o A .' 6 .        Yes. There are four means of reducing radiation ex-posure. They are: (1) provide shielding between the radioactive source and accessed area; (2) provide A
distance between the radioactive source and workers; (3) reduce the time spent by the worker in the radiation j                    field; and (4) removal of the radioactive source material.
l Q.7.            Please describo generally the application of these principles to the design of a nuclear power plant such as Byron.
A.7.          Example of the use of each of the above is as follows:
: 1. Shielding can be interposed between radioactive areas and those areas containing systems which do not have radioactivity permitting general access in these latter areas; and shielding can be interposed between radioactive components to permit maintenance on one t
component without undue hazard from the other component.
l                  But in some cases it is not practical to interpose shielding; for example, consider work on a pump containing radioactive material. Other principles are then called upon.
: 2. Perhaps distance can be provided by remote handling tools or remote operations.
: 3. Or in our pump example, the ALARA guideline can bs met more effectively by the reduction of the time spent by the worker in the radiation field.      This can be done by providing, in design, enough laydown space for
 
removed mechanical parts, maintenance tools and supplies, J
permanent galleries for ready access to radioactive areas when necessary, cranes for efficient handling of materials in~a radioactive area and removable shield walls to improve access to radioactive components when maintenance is required; or
: 4. Providing draining and backflushing pro-visions to remove the source from the work' area.
Generally, it is a combination of all of the above design features which contribute to ALARA. The plant procedures as described in the testimony of Mr. Van Laere further help keep doses ALARA.
Q.8.      Is the manner in which specific features of Byron
:          Station have been designed to take account of ALARA f
principles described in the FSAR?
A.8.      Yes. Attached to my testimony as Lahti Exhibit 1 are section 12.3 through section 12.3.2.1.8. These sections were prepared under my supervision and control and describe the detailed application of the design principles described above to the Byron Station.
 
l
      ,  .                                B/B-FS AR 12.3 s-RADIATION PP.OTECTION DESIGN FEATUREO
    \
Radiation protection design features are provided to reduce direct radiationradiation, areas,  control airborne radioactivity, identify decontaminate personnel and equipment, calibrate radiation monitors, and maintain personnel raciation exposure as low as is reasonably achievable (ALARA) .
12.3.1    Descriotion of Facility Desian Considerations 12.3.1.1 Equioment Selection, Lavout, and Secrecation In selecting and shielding equipment and compor ents containing radioactive materials, prime consideration is given to protecting i
the operating and maintenance personnel from radiation, and to maintain personnel exposures ALARA.
Equipment containing rooms or cubicles, radioactive materials is located in separate where practicable, to protect operating and maintenance personnel from radiation associated with other equipment.
Components serviced whenever            are remotely operated and/or remotely practicable.
Items which require frequent maintenance and which are
;        radioactive or potentially radioactive, such as pumps, valves, and instrumentation are to the extent practicable, separated from passive radioactive components such as tanks, filters, demineralizers, etc.
{
l        Areas containing more than one piece of radioactive equipment are, where practicaole, designed and provided with shielding such that othermaintenance      of one item is not restricted by radiation from preces of equipment. Where it is not practicable to provide  permanent    shielding, 12.3.2) ror temporary shielding  provisions (discussed in Subsection provided.                            to minimize maintenance doses are i
Components which are not radioactive or potentially radioactive are physically separated, to the extend practicable, from components which are radioactive or potentially radioactive.
Radiation detector proDe access holes are provided in shield walls (e.g.,
shield hatches) for all isolated equipment cubicles where access is only by means of removable shield walls.
Partially shielded configurations are reviewed for radiation scatt erin g.
12.3.1.2    Cubicle Access i
Access to radioactive or potentially radioactive cubicles or            l compartments is through entrances designed, where practicable,      to f
(      permit access to an area of the room which has the lowest or relatively lowest radiation level.
Entrances are designed to    l 12.3-1 1
 
B/B-FSAR n
prevent source openings    and intoradiation occupiedfrom  passing directly through entrance areas.
(,                                              This is done, where practicable, and  potentiallyby providing radioactivelabyrinthine cubicles. entrances to radioactive Typical 12.3-2.
labyrinthine entrances are seen in Figures 12.3-1 and Radiation traveling through such labyrinthine entrances collides with the shield walls and consequently can be attenuated to some small fraction of the incident quantity.
Cubicle access for Byron /Braidwood is either through a labyrinthine entrance with an overlap of 1 to 1 1/2 times the passageway width as seen in Figure 12. 3-1 or through a double labyrinth arrangement as seen in Figure 12.3-2.
Not  all entrances to radioactive areas are designed with labyrinthine entrances.
Where labyrinthine entrances are not feasible, other alternatives include:
a.
shield doors installed at personnel entrances,
: b. removable concrete block walls, and c.
wall and floor removable shield hatches and plugs (such as for the radwaste filter and demineralizer compartments).
The following considerations govern the design of labyrinths:
a.
A labyrinth is located and sized to cause unscattered radiation to be attenuated by the required amount of shielding, as shown in Figure 12.3-1. Normally, the labyrinth overlap is designed so that (with worst-case sources) the streaming leaving the laoyrinthine entrance due to scattered radiation gives a dose rate which is less than three to five times the design dose rate of the surrounding area. Where ctrong sources of low energy gamma radiation are encountered, a double laoyrinthine entrance such as depicted in Figure 12.3-2 is used in order to meet this criterion.
b.
When the design of the labyrinth is determined by other design considerations, a shield door, isolation of the entrance (e.g., rope of f area) , e xtended labyrinth specified.
overlap, or a removable labyrinth is also c.
If the labyrinth height is shorter than the ceiling height, as is often the case, a roof is provided above the labyrinth section.
k-12.3-2
 
      .  .                                  B/B-FSAR
: d. Galleries and other elevated occupied areas are
(-                    protected  from radiation passing through the roof of the labyrinth.
The dose maintain the design    roofsrate haveofathese thickness  which will elevated areas.
e.
Labyrinths inside source cubicles require roofs if any part of a source is higher than the top of the labyrinth.
The roof thickness is dependent upon the location of the source, and the thickness is eclculated on a cubicle-by-cubicle basis.
12.3.1.3    Draining and Flushina Cacability of Equicment consideration is given in the radiation protection design to identifyingofthe capability        need for adequate draining and flushing equipment radioactive service.          designed for radioactive or potentially The potentially high activity radaaste storage tanks were selected and their designs reviewed to assure adequate draining capability    to minimize activity buildup and excessive radiation levels over the plant lif etime.        Tanks containing radioactive material have sloped bottoms wherever practicable so that sludge accumulation is minimized and ease of drainage is enhanced.
Where practicaole, equipment is selected and the design reviewed
(-    to assure that there are no obvious ledges or pockets where radioactivity may be trapped or accumulated.
Io the extent practicable, drain piping is of welded construction and is welded in a manner, e. g. , using consumable inserts , to minimize crevices which might collect radioactive material. (Use of backing rings in the welds or use of socket welds may be acceptable if the weld is embedded in concrete.)
The design of the spent resin storage and exchange systems is reviewed to assure that the layout and components are such as to prevent the retention bends, horizontal          of resin beads or fragments in connections, sections,  reducers, etc.
All equipment drains which are considered to be radioactive are direc ted to appropriate liquid radwaste storage tanks. Sumps are used as intermediate collection points. Such sumps and tanks            are appropriately areas.
shielded  or appropriately  loca  ted within radiation l
The design of the radwaste filters was checked to assure that the filters can be drained and flushed prior to filter element replacement.
Elushing capability of radioactive service equipment is important f
L      to assure the          a minimum of radioactive crud or sludge retention in equipment l                          prior to maintenance or removal of the equipment.
12.3-3 l
 
I
    .    -                                  B/B-FSAR All potentially high activity source storage vessels were C'      selected and capability. their designs checked to assure adequate draining These tanks include the volume control tank, the spent resin storage tank, the concentrates holding tank, the regeneration waste drain tank, the auxiliary building floor and equipment drain tanks, and the recycle holdup tanks.
Draining capability is assured:
a.
to minimize personnel exposure during testing, surveillance, and maintenance activities and b.
to minimize activity (crud) buildup and avoid excessive plant        radiation levels to accessible areas during lifetime.
Adequate draining capability is assured wherever practicable by selecting be          tanks provided  with,which have sloped bottoms and which have, or can the tanks. Drainage    drain lines connected to the lowest level of of the above listed high activity source storage tanks is via remotely operaten valves or by valves which are located remotely from the tank cubicle in lower radiation areas.
(For location of valves with respect to shielded areas, refer to subsection 12.3.1. 8. )
Flushing of radwaste tanks is accomplished by washing down the tank interiors with demineralized water and/or cleaning agents.
(      Where practicable, provisions are made to remove crud sedimentation by remote mechanical means with hoses.
Where practicable, flushing of radwaste tank interior is accom-plished by an installed sparger (where justified) a recirculation line for the pump servicing the tank to theor by providing bottom of the tank so that a spraying effect can be utilized to get drainedsettled outdeposits    in suspension, so that they may be pumped or of the tank. For manual flushing, adequate capability is provided in the form of water connections located l
i near the tank cubicles.
Flushing is required when major maintenance and/or removal of the tank is necessary and also when necessary to reduce radiation levels in adjacent areas due to sources within the tank. Flushed water is directed to tanks having sufficient capacity and shielding necessary to'contain and shield the flushed water.
When items suchpracticable, as pumps.theWhere above applies to other high activity source adequate draining and flushing capability is not practicable, shielding is designed to account for worst-case radioactive crud buildup.
L 12.3-4
 
B/B-FSAR 12.3.1.4 Floor and Sink Drains _
(      Adequate floor drainage is provided for each room or cubicle housing liquids. components which contain, or may contain, radioactive Floors are properly sloped to the floor drain to facilitate floor drainage and prevent water puddles.
All  floor drains which are considered to be radioactive are directed to appropriate liquid radwaste storage tanks. Sumps are used as intermediate collection points. Such sumps and tanks are aappropria rea s.      tely shielded or appropriately located within radiation -
Shielding Subsection 12.3.1.6.      of radwaste drain piping is discussed in To the extent practicable, greater potential radiation area floor drains to protect against backflow of radioactive liquids lowerspill large    potential    radiation occurs.        areas, if drainage is blocked or if a system is prevented by the use of water-filled sealsAir            circulation throug or by sealing individual floor drains. The use of such          (loop seals) seals alsofloor the      prevents drainbackflow system. of radioactive gases into the room from Sink drains which are expected to contain radioactive fluids are reviewed for appropriate shielding and routing requirements.
(  Loop radioactiveseals are    present on sir.k drain lines which may handle fluids.
All floor drains in the auxiliary, containment, fuel handling, and below,  radwaste/ service buildings, except for those areas listed either the auxiliary building floor drare considered to be radioactive and shall drain tank through various sump pumps.ain requirement are:
tanks or the Exceptions      chemical to this a.
diesel generator oil storage tank rooms, b.
auxiliary feedwater tunnel, c.
main steam / steam generator feedwater tunnel,
: d. tendon tunnel, e.
tendon tunnel access area,
: f. diesel-generator rooms,
: g. cable spreading rooms,
: h. switchgear rooms, L
12.3-5 i
I
 
1 l
B/B-FSAR            AMENDMENT 22 SEPTEMBER 1979 ;
l
[)            i. office Treas in service building,
: j. storage rooms in service building,
: k. auxiliary electrical equipment room,
: 1. battery rooms,
: m. auxiliary building HVAC equipment area (elevation 451 feet), and
: n. essential service water pump.
12.3.1.4.1  Desian or Drain System
: a. Equipment drains in the turbine building dicenarge to the two turbine building equipment drain sumps, one por unit, from which they are piped to the turDine cuilding equipment drain rank. The drainc are then treated and recycled to the condensate storage tanks.
: b. Equipment drains in the auxiliary, containment, and fuel handling buildings discharge to the two auxiliary building equipment drain collection tanks.
Pumps are provided to pump the drains to the
(                  auxiliary building equipment drain tanks.
: c. Floor drains that are expected to handle chemical waste solutions from potentially radioactive areas are kept separate from other floor drains and are routed to the chemical drain tank, unless otherwise specified.
: d. Leak detection sumps are provided for various areas in the auxiliary building that contain safety-related equipment required for long-term operation.
: e. A storm drain system, complete with oil separators, is provided to remove all roof and storm drainage.
: f. Borated equipment drains are recycled to the recycle holdup tanks.
: g. High radiation area floor drains are routed separately from low radiation area floor drains to prevent backtlow of nigh contamination into low radiation areas.
: h. Tne top elevation of rioor drains are set below nominal elevations of the rloor area to be drained.
k,I              i. Floors are sloped to the drain to facilitate floor drainage and prevent water puddles.
12.3-6
 
        .  .                                    B/B-FSAR j.
Slotted cover plates are used to prevent solids from C-entering floor drain sumps. These cover plates are removable to provide full access to the sump.
k.
The arrangement of drains from cubicles containing radioactive equipment is such that air from a zone of high airborne radioactivity potential does not circulate through the drain system to normally accessible areas. The prevention of air circulation is done through the use of loop seals.
1.
Drain piping of equipment and systems which carry caustics or acids is the same material as the equipment or system they~are draining.
m.
Drain lines are positively sloped 1/8-inch per foot to assure complete drainage of piping.
: n. Drain piping is of welded construction and is welded ir, a manner to avoid crevices (except where embedded in concrete) , which might collect radioactive solids.
All potentially high radioactive drain piping from the equipment to the loop seal is welded using a consumable insert.
o.
Equipment drains which interconnect pieces of
(.                      equipment are designed so as not to inadvertently                '
transfer fluid from one piece of equipment to another.
: p. Shielding of radwaste drain piping is provided as necessary. Radwaste drain piping not specifically shielded is routed so that is is not exposed to normally high access areas and general access routes.
Vertical runs of radwaste drain piping not specifically shielded is run against walls and sufficiently isolated so as to facilitate j
compensatory shielding, if required.
q.
All floor drain piping to the sumps (unless otherwise l
noted) is carbon steel unles required to be otherwise 1
by design due to flow of corrosive liquids.
r.
i l
Primary sample drains are routed to the chemical l                          drain tank and from these processed in the radwaste e vaporators.
l 12.3.1.5    Ventina of Eauiement l
Where equipment practicaole, all radioactive or potentially radioactive
!        is vented to a filtered vent (such as filters, demineralizers, and radwaste tanks) l
(-                                          header to minimize the possibility L'      of  airborne due to equipmentradioactivity  in occupied    areas or equipment cubicles venting.
12.3-7
 
B/B-FSAR Radwaste sumps (i.e. , sumps designed to handle drains from radio-C    acitve service equipment or from floor areas of potentially radioactive components) are normally either vented to a high radia tion area, such as to within the cubicle the sump is located, if it is high radiation cubicle, or to a filtered vent header. Venting of radwaste sumps is important to control the concentrations of radioactive contaminants normally released to the air from potentially contaminated water held in the sumps.
Subsection 12.3.1.5.1 lists the sumps with venting. For sumps which are in shielded cubicles and which vent to the cubicle, cubicle ventilation rates are such as to assure adequate control over expected airborne concentrations of radiciodine. If venting to other areas is required, the sump covers have air inleakage and have no special provisions for sealing since the sump can maintain a slightly negative pressure with respect to the area in which the sump is located. A small amount of air inleakage      to the sump is desiracle to maintain air flow through the vent line.
12.3.1.5.1      Sumos Reauirina Ventina Venting is provided for the following sumps:
a.
auxiliary building equipment drain collection sumps, b.
fuel handling building floor drain sumps, c.
fuel handling building decontamination sump, k                  d. drum storage area sump, and
: e. radwaste drumming station sump.
Venting of these sumps minimizes the possibliity of potential airborne radioactivity in the sump areas. Venting is via a small vent line connected to the sump cover plates. This line is routed to a filtered vent header. Slightly negative pressure is i
maintained sump is located.
in the vent line with respect to the area in which the 12.3.1.6 Routina and Shieldina of Lines and Ventilation Ducts
!        12.3.1.6.1    Routing and Shielding of Lines All potentially radioactive process lines are evaluated to deter-mine proper routing and shielding requirements, based on minimizing    radiation exposures to station operating and main-tenance personnel.
Radioactive process piping is routed in shielding pipe tunnels, trenches, or chases, or in areas wnere the radiation field due to the pipe is consistent with the radiation zone for that area.
12.3-8
 
  , -                                B/B-FSAR To aid in preventing crud buildup in process piping, sharp bends, C. dead ends, and other obvious crud traps are minimized. In general, socket welds and welds employing backing rings are avoided to the extent practicable; these welds contribute to radioactive crud accumulation which results in increased radiation fields near the weld. Where practicable, welds employing consumable inserts are used instead of socket welds or welds using backing rings because the consumable insert weld makes the inside-of pipe surface smoother and minimizes crevices which may trip crud at the weld. Socket welds and welds employing backing rings are used, however, if the weld is to be embedded in concrete (such as in concrete floor slabs) ; for these cases, radiation fields due to radioactive crud accumulation are attenuated by the concrete around the weld.
Shielding of radwaste drain piping (including floor and sink drain piping) is provided as necessary. Radwaste drain piping not specifically shielded is routed so that it is not exposed to normally occupied areas and general access routes. Vertical runs of radwaste drain piping not specifically shielded are run against walls and sufficiently isolated so as to facilitate compensatory shielding, if required.
To the extent practicable, radioactive or potentially radioactive sample lines used for grab samples are routed so that grab samples can be taken in low radiation areas.
(    Radioactive lines are process system piping, drain lines, sample lines, and other lines which normally do, or may contain, radioactive fluids. Special attention is given to the routing and shielding of radioactive lines.
The f ollowing guidelines are followed for routing and shielding of radioactive lines:
: a. Routing of radioactive lines in low radiation zones is avoided to the extent practicable.
: b. Lines that require shielding are routed in shielded pipe tunnels or in radiation areas to the extent practicable.
: c. Penetrations through shielded pipe tunnels are not maae    by lines which do not, themselves, run through the pipe tunnels.
: d. Lines that carry radwaste cemineralizer resins, filter backwash, filter /demineralizer sludges, or otner particulates have large radius bends ano are continuously sloped.
On radwaste demineralizer resin lines, welded piping is used but the use of socket welds or welds employing backing rings is avoided to the extent practicable; also, the use of loop seals on these lines is avoided to the extent practicable.
12.3-9
 
                      ,                B/B-FSAR e.
Slightly radioactive lines are routed in a manner
(                  which and    minimizespersonnel.
maintenance  radiation exposure to plant operating Slightly radioactive lines in low radiation zones are, to the extent practicable, routed at a minimum elevation above .the finished floor of 10 feet 0 inch, or as high above the floor as is practicable.      To the extent practicaole, slightly radioactive lines are not routed near normally travelled passageways, nor near galleries or other elevated work areas.
f.
For field routing of 2-inch and under nonseismic radioactive followed.      piping, the guidelines listed below are 1.
Piping is installed at as high an elevation as is practicable but, in no case, below 10 feet 0 inch from the finished floor level.
2.
Piping is routed as close as possible to existing walls or structures to take advantage of their shielding effect.
3.
Radioactive piping is not routed near groups of nonradioactive piping thereby not limiting accessibility components.
to nonradioactive system 4.
Radioactive piping is not routed near an area radiation monitor thereby causing abnormally high radiation readings which are nonrepresentative of the general area in which the radiation monitor is located.
5.
To aid in preventing radioactive crud buildup in the piping,  sharp bends, dead ends, and other obvious  crud practicable.
traps are avoided to the extent The use of socket welds or welds employing backing rings on the piping is avoided to the extent practicable.
12.3.1.6.2    Routina and Shielding of Ventilation Ducts HVAC duct routing are reviewed to assure that air flow is from areas of lower potential radiation contamination to areas of higher potential radiation contamination.
Ventilation duct penetrations of shield walls, floors, and ceilings are evaluated to determine if parapet and labyrinthine shielding around the ducts is necessary. Penetrations in shield walls for hVAC ducts is discussed further in Subsection 12.3.2.3.
L 12.3-10
 
i B/B-FSAR              AMENDMENT 40    l NOVEMBER 1982 12.3.1.7    Waste Filters and Demineralizers
(      The waste filters and demineralizers which accumulate radioactivity and which, if unshielded, could cause the area design dose rate to be exceeded, are located, to the extent practicable, in separately shielded cubicles.      Shielding is provided between such adjacent filters and demineralizers to minimize personnel exposure during removal or maintenance opera tion s.
A radiation detector probe access hole is provided in all filter and demineralizer removaale shield hatches so that radiation levels of the contained equipment may be measured without removing the shield hatches.      The probe access hole is shown in Figure 12.3-4.
The waste tilters are designed where practicable to permit removal cy a remote handling device.      Draining and flushing of radwaste filters is discussed in Subsection 12.3.1.3.
Waste filters also includes HVAC filters which may accumulate aircorne raaicactive materials. These tilters are located in areas of the station where access is controlled. Shielding is provided as necessary around HVAC filters (e.g., charcoal filte rs) to ensure tnat resultant dose rates from the filter areas are less than the design dose rates for the areas, and to minimize radiation exposure to maintenance personnel during
(      filter removal or maintenance.
HVAC filters are designed for easy removal and sized to allow proper disposal as per Regulatory Guide 1.52, " Design, Testing and Mainenance Criteria for ESF Atmosphere Cleanup System Air Filtration and Adsorption Units of LWR's ," Revision 2.
l For charcoal air filters, charcoal filtration capacities are such as to assure that radiciodine loadings meet criteria for ESF atmospheric cleanup system air riltration and adsorption units.      l l
l        12.3.1.8    Valves and Instruments t
Where practicable, valves are located and shielded from adjacent radiation sources so that they can be operated or serviced without causing excessive exposure to operating or maintenance personnel.
Shielded valve aisles are provided where necessary to allow greater accessibility to frequently operated or maintained valves. The valves are installed in the valve aisle shielaed from the equipment they serve.      The valves and associated piping are shielded from the valve operating area (where the valve handwheels are generally installed) , located outside of the valve aisle.
L 12.3-11
 
O
* B/B-FSAR                AMENDMENT 21 JULY 1979 12.3.1.8.1      Valves I                                                                              .
a.
To extent practicable, all valves servicing radioactive or potentially radioactive equipment are located in shielded valve aisles, apart from the (adjacent) equipment being serviced. Walk-in valve aisles are used where practicable (see figure 12.3-3). Locating valves in pipe tunnels cannot be avoided entirely, tevever.
b.
All radioactive or potentially radioactive manually operated valves (and associated piping) are shielded from tne valve operating area, to the extent practicaole. Where practicable, use is made of remote manual valve operators (valve extensions or reacn rods) connected to the manual operated handwheels or geared handsheels and passing througn the shielding to allow valve operation in the valve operating area (see Figure 12. 3- 3) . This protects valve operating personnel from radiation due to radioactivity in the valves and associatea rluid piping in tne valve aisle.
c.
Radioactive pipe runs to and from valves located in valve aisles are minimized to reduce the amount of radioactive material in valve aisles. This is done by maximizing the amount of radioactive runs behind
(
shielding (e. g. , running as much as or the.
radioactive pipe behind the shield wall which separates the valve aisle from the [ad 3acent]
equipment compartment of the component whicn the valve services) .
d.
To the extent practicable, all motor-operated valves and pneumatic operated valves (air-operated valves) which are in radioactive or potentially radioactive service are located in areas which are shielded from the (adjacent) component or item of equipment which the valves service. Locating these valves--which are typically higher maintenance items than manual operated valves--in shielded areas minimizes potential personnel radiation exposures due to cther nearby radiation sources during valve maintenance and inservice inspection.
e.
Valves servicing radioactive or potentially radioactive equipment are installed and positioned
,!                    with respect to other valves so that (1) service or maintenance time is minimized, and (2) compensatory sheilding (e. g. , lead blankets) is used, where pra ct icable , to protect workers f rom adjacent radioactive valves and piping.
()
s-i 12.3-12
 
e-
          . .                                  B/B-FSAR f.
For valve maintenance, provision is made for draining
(~                  or flushing the valve and associated connecting lines of radioactive fluids so ' that radiation exposures are minimized.
12.3.1.8.2      Instruments a.
Output devices such as instrument readouts, pressure switches, electrical bistable devices, electric converters, control devices, etc. , are instal 2 ad and located in such a manner as to minimize plant personnel exposure to radiation.
b.
Instrumentation displays and controls are installed in the lowest practicable radiation zone. (Use of transducers is maximized in high radiation areas.)
Instrument readouts are located and positioned in areas ( e. g. , at valve operating stations) which will result in the lowest personnel exposures, if consistent with other requirements such as instrument accuracy and precision.
c.
Instrument readouts are designed, located, and positioned to minimize the time and exposure required to take a reading.      The following is considered in the location and positioning of the instriu.ent readout devices to assure ALAP exposures.
: 1. Locate in readily accessiola areas.
2.
Position at convenient elevation for ocservation and application of parallax corrective devices.
3.
Face readout toward direction convenient for reading.
: 4. Provide easily readable numbers and easily observable pointers and needles.
: 5. Preclude or minimize application of scale multipliers on readout.
6.
l                                Locate to take advantage of amount of lighting available.
d.
Location of instruments and instrument readouts considers the possibility of local hot spots due      to streaming radiation or from the accumulation of radioactivity equipment.
in lines, ducts, filters, and e.
Wherever practicable, radiation monitoring equipment
    /'-                  with remote readout is located in areas to which personnel normally have access.
l 12.3-13 l
 
B/B-FSAR t
Figure  12.3-3 shows a typical walk-in valve aisle arrangement for
(~    Byron /Braidwood.
12.3.1.9    Contamination Control and Decontamination In addition to the safety design features discussed above, the following saf ety design features specifically relating to decontamination and contamination control are incorporated into the radiation protection design of the station.
: a. Curbs Where practicable and where failure of radioactive storage tanks, vessels, or associated piping is postulated, either the floor of the cubicle is situated at an elevation lower than the entrance to the cubicle, or curb walls are provided to restrict radioactive material to the cubicle.
Curbs are provided for equipment decontamination pads to restrict washdown water to the pad and avoid contamination of adjacent areas.
: b. Protective Surface Coatings Wherever these exists a potential for leakage or f                    spillage of radioactive material onto concrete
(                    surfaces (e. g. , shield wa lls, ficors, or ceilings),
such surf aces are coated with a nonporous coating to enhance decontamination.
The f ollowing guidelines and criteria are used for the application of coating systems to potentially contaminated concrete surf aces in the station to enable them to be effectively decontamina ted.
l          The f unction of the protective coating system is to facilitate i
decontamination of surfaces by providing a clean, smooth, and hard  finish that is minimally free of cracks, is nonabsorbent, and is wa ter-repellent.
Surface contaminants can then be removed by means of washing, sweeping, scruboing, or wiping in one or more applications.
: a. The coating systems are capable of performing their surface protective functions throughout the 40-year plant lifetime (including reasonable maintenance and touch-up activities) and under the variable radiation I                        source and environmental conditions anticipated for the plant.
b.
l                        The coating systems applied to floors, curbs, dado, I                      and wainscot are capable of maintaining their
  ''-                    integrity in protecting these surraces under conditions of water immersion. The coating syst ems t
12.3-14 L
 
B/B-FS AR used on floors, curbs, and dado is therefore solvent-(                  based.
based. The wainscot can be either solvent or water-c.
To enable the coating systems to perform their intended function, a surface preparation system appropriate  to the surface as well as to each coating system, is first applied. The surface preparation system includes surface cleaning, the filling of holes and the application of primer coating.
d.
The coating systems used on floors and ramps are capable of maintaining their integrity in protecting these surfaces under the traffic patterns (people, lift trucks, etc.) anticipated in the various areas.
The thickest of the field coating systems should be specified for such areas that involve continuous use to  avoid deteriorating and thereby compromising the coating.
Protective coating systems are applied to concrete surfaces on the following basis:
: a. Where no other requirements are necessitated, all walls are coated to 1 foot-0 inch dado height to protect this lowest section during sweeping and washing of the floor.
: b. Walls that require only partial height coverage are coated to one of several standard wainscot heights (usually 5 feet-0 inen or 8 feet-0 inch) . General examples include the walls of potentially radioactive heat exchangers, and certain access area locations, c.
Cubicles containing radioactive equipment that reach above the highest wainscot level are coated to full height and in most cases, the ceiling. The coating of rooms utilizing monorail or crane systems for l
handling radioactive materials are based on the elevated height of the materials.
d.
Cubicles containing radioactive processing equipment i
l
!                      such as pumps or pressurized pipe with valving are coated to full height, where necessary. Potentially
                      ' radioactive contaminated water f rom the room or area on the floor above, through penetrations in the ceiling.
e.
Walls are coated to full height if the potential exists for leakage of radioactive contaminated water from the room or area on the floor above, through penetrations in the ceiling.
(-
12.3-15 l
 
B/B-FSAR' f.
Cubicles, rooms, and areas that require complete wall
(~)                          coverage have their ceiling fully coated as well.
This includes the underside of removable shield hatches and plugs as well as fixed ceilings.
g.
Pipe tunnels that are accessible and contain radioactive pipes are fully coated.
Areas that require partial wall coverage (although complete wall coverage following:
may be dictated by equipment size) include the a.
areas around sampling stations or panels receiving radioactive process streams f or monitoring; b.
areas through which heavy traffic patterns are expected; and c.
clothing change areas, personnel monitoring points, and counting room.
12.3.1.9.1        Equioment Decontamination Facilities Equipment decontamination facilities are provided in the station as tools, required etc. for the decontamination of contaminated equipment, The design of these facilities includes adequate shielding, and ventilation and filtration of the room air.
A separate area, the auxiliary machine shop and equipment decon-tamination facility is provided on elevation 426 feet 0 inch in the auxiliary building for cleaning, repairing, and decontamina-ting tools and small pieces of equipment. Decontamination and repair is done under radiation protection personnel supervision.
The room is equipped with shelves, tables, hooded sinks, welding equipment, a lathe, a saw, a drill press, and tanks. For larger pieces of equipment, cleaning and decontamination takes place in the spent fuel shipping cask decontamination pit, in the fuel handling building.
12.3.1.9.2 Personnel Decontamination Facilities A personnel decontamination facility is supplied on elevation 426 feet 0 inch in theofauxiliary decontamination                        building to provide for prompt plant personnel, l                                                          if the need should arise.
12.3.1.9.3          Station Decontamination Radiation decontamination of the station is currently expected to be required at least once during the life of the station. The radiation protection saf ety design features discussed above assure is required. less complicated station radiation decontamination when it L.
12.3-16
 
      .    -                                  B/B-FS AR 12.3.1.10    Traffic Patterns and Access Control Points Traffic patterns are established to maintain occupational radia-tion exposures ALARA. Anticipated traffic patterns have been used to determine design dose rates in the various areas, and thus have affected the determination of radiation zones.
The majority of normal personnel traffic occurs between the service building and the auxiliary building. The remainder of the traffic occurs in operating areas (where panels and motor-control centers are located), hallways, elevators, and stairwells.
Access control points (i . e. , check points for p'ersonnel) are flexible and are determined on a day-to-day basis, depending on contamination levels and maintenance activities.
12.3.1.11    Radiation Zones Five main radiation zones have been defined as a means of classifying the occupancy restrictions on various areas within the plant boundary.      The design criteria tor each zone are described    in the subsections which follow and are tabulated in Table 12.3-1. The radiation zones assigned to the areas of the plant, and upon which the shielding has been designed, are shown in the radiation zone maps in Figures 12.3-25 through 12.3-44.
(,        Zone I-A Zone I-A has no restriction on occupancy. A I-A area would repre sent, for example, plant site where radiation due to occupancy on a 40 hr/ week, 50 week /yr basis, will not exceed the whole body dose of 0.5 rem /yr, as specified in paragraph 20.105 of 10 CFR 20. The environs around the plant such as the pump house,    electrical switchyards, and turbine hall, are examples of a Zone I- A area.
It is expected that nonplant personnel or vistors to the site will receive considerably less than 0.5 rem /yr because of the relatively small time interval during which t, hey are on the site.
Zones I-B, I-C, and I-D Zones I-B, I-C, and I-D are radiation areas which individuals can occupy on a 40 hr/ week, 50 week /yr casis, and not exceed the allowaole whole body dose of 1.25 rems per calender quarter.          The design dose rates are from 0.5 to 2 mrem /hr in these zones.        The area will be posted and will remain accessible.          Corridors in the l          auxiliary building and areas outside radioactive enclosures where l
l personnel can walk freely are included in this zone.
l l
12.3-17 l
\
 
B/B-FSAR AMENDMENT 28 OCTOBER 1980 Zone II-A
(          Zone II-A is a radiation area that plant personnel can occupy periodically.
This zone has a design dose rate of 4 mrem /hr.
The radiation level in a Zone II-A area will be posted, but the area will remain accessiole to the plant personnel.
Zone II-B, II-c, and II-D Zones II-B,'II-C, and II-D are radiation areas which will be posted with " Caution-Radiation Area" signs.
are from greater than 4                                The radiation levels mR/hr to 100 mR/hr.
The time a worker determined          withfactors:
by four    a permit can stay in this room isOccupancy is limited. l a.
the actual radiation level in the room; b.
the nature of the radioactivity (airborne, etc. ) ;                                            gamma, c.
the past radiation history of the worker; and d.
nature of the required job.
The " nature of the required job" means that the necessity of tne job being done to ensure the safe operation of the station will ce nned.
pla considered when work in these radiation areas is being Auxiliary equipment which requires manual operation or inspection or  maintenance zones.                during  unit operation will not ce located in these All equipment in areas designated as Lone I-A,            I-B, I-C, I-D: or II-A will not contain radioactive materials, or if it does, the activity will be sucn that the dose rate outside the equipment is censistent with the design dose rate in the area. Such equipment could include fluid system, monitor tanks, and monitor pumps.
Zone III Zone III is a "high radiation area" (10 CFR 20.203) which will be secured oy locked doors or other suitable means. Radiation in this    area is in excess of limited.                          100 mR/hr and occupancy periods are very l
This is a restricted area and entry into the area is prevented by a locxed door, or entry will energize a conspicuous visicle and/or audible alarm in such a manner that the individual entering, and the control room, are maae aware of the entry.
Entry is permissible only with the knowledge of the control room (shif t operating engineer or a designated alternate) . A Zone III area    will be time permissoble      monitored limit. prior to occupancy to determine the 12.3-18
 
B/B-FS AR Zone IV
(..        Zone IV has no restriction on occupancy. Zone IV covers the lowlevel counting room and personnel dosimeter storage area where a low dose rate is desirable. The dose rate in this zone will be less than or equal to 0.01 mrem /hr above background.
Zone V Zone V is the main control room area. Zone V normal dose rate will be less than or equal to 0.2 mR/hr during normal operations.
During an accident the integrated whole oody dose will not exceed 5 rem.
12.3.1.12    Laboratory comclex The station laboratory complex is located in a controlled access area on the mez=anine floor of the auxiliary building. The facilities located in this complex are:        a high level laboratory, a low level laboratory, a counting room, a chemical storage room, a radiation chemistry management room, a hot instrument room, a personnel decontamination room, a supply room, a radiation office, and supervisor offices. This complex will serve as a center for the health physics and radiochemistry activities at the station.
12.3.1.12.1    Hich_ Level Laboratcry                                  '
The high level laboratory is designed to provide for the safe and efficient radioactiveprocessing samples. and analysis of radioactive and potentially Such samples may be expected for such systems as the:      primary coolant loop, chemical and volume control, radwaste.
f uel handling and storage, steam generator blowdown, and l          The major facilities provided in this laboratory are: fume hoods (with HEPA filtered exhausts) , sinks (with drains routed to the liquid radwaste system), sufficient workbench space to allow frequently used equipment to be left in place, sufficient built-in storage space to assure a safe, uncluttered work environment, l
'          computer grade regulated electrical circuits, and a close tolerance HVAC system (temperature and humidity) to assure optimal performance of sensitive laboratory equipment.
To minimize the accumulation and spread of surf ace contamination;
'          floor coa tings, surface coatings, workbench surfaces, fume hood interiors, and sink and drain pipe materials are chosen to l
minimize the adherance and ease the removal of contamination. To minimize the spread of airborne radioactivity, fume hoods are provided for the storage and processing of volatile radioactive samples, and the high level laboratory is kept at a negative pressure with respect to all adjacent areas. All air exhaust
{        from this laboratory is filtered prior to its release to the environment via the station vent stack.
l l
12.3-19
 
    *    "                                    B/B-FSAR AMENDMENT 21 JULY 1979 12.3.1.12.2      Low Level Laboratory
(*      The low level lacoratory is located adjacent to the high level laboratory on the mezzanine floor of the auxiliary building. It is designed to provide a radiation and contamination free environment for the chemical preparation and analysis of nonradioactive samples (i.e., those samples whien could not pose a radiological danger to the laboratory worxers) . The major equipment hood s , sinks, provided in the low level laboratory includes: tume workbenches, and storage facilities.
12.3.1.'12.3      countina Room The counting room is located near tne high and the low level lacoratories on the mezzanine rloor of the auxiliary building.
This room is provided with computer grade regulated electric circuits and nonfluorescent lighting to assure the optimal              >
perf ormance of the counting equipment. The desired radiation level in the counting room should be nelow background (0.1 mrem /hr) ,      To assure that the counting room will not be af f ected oy any in-plant airborne radioactivity the room is maintained at            l a positive pressure with respect to all surrounding areas and is ventilated with fresh filtered and conoitioned air.          The room HVAC is designed to maintain the temperature and humidity tolerances required by the detectors ana their associated electronics ana computer equipment. '1he use of thick concrete
(,    walls f or shielding the counting room was precluded due to considerations of natural radiation emanating from the concrete itself.
The original equipment to be provided in the counting room includes:
: a. gamma-ray spectrometer suosystem,
: b. multichannel analyzer subsystem,
: c. data analysis subsystem,
: d. standard alpha, beta counting subsystem, e.
low-background alpha, beta counting subsystem, f.
automatic sample changer (f or d and e) , and g.
liquid scintillation counting system.
Localized        radiation snielding will ce providea for the counting equipment as is needed.
12.3-20 L
 
B/B-FSAR            AMENDMENT 40 NOVEMBER 1932 12.3.1.12.4  Chemistry storace
(        A chemistry storage room is located between the low level chemistry room and the counting room on the mezzanine floor of the auxiliary building.      It provides storage space for the radiochemistry supplies used in the laboratory complex.
12.3.1.12.5    survev Instrument storace Room The survey instrument storage room is designed to provide a location where radiation proctection instrumentation can be stored, serviced, and decontaminated when necessary. This room and its facilities are designed to minimize the accumulation and spread of airborne and surface contamination.
12.3.1.12.6    Decontaminatiora Room The decontamination room associated with the laboratory complex is designed to facilitate the decontamination of station personnel.
12.3.1.12.7    Sunoly Ro-The supply room in the laboratory complex is designed to provide for the storage of health physics related supplies.
12.3.1.12.8    office soace J
The radiation chemical management office, the radiation office, and the supervisor of fices in the laboratory complex are provided to assure adequate, local office space for the laboratory complex workers.
12.3.1.13    Laundry Facility The station laundry facility is located on the mezzanine floor of the auxiliary building.      It is designed to receive, decontaminate, store, and distribute the radiation protection apparel used in-plant.      The ma]or equipment provided in this facility includes:
a 25- and a 50-pound washer, two 50-pound    l dryers, a contaminated clothes hamper (where incoming contaminated apparel will be dropped off) , a sorting table (with local ventilation), sinks, a folding and monitoring table, and clean clothing storage racks. The floor and surface coatings and the equipment in the laundry have been chosen to minimize the buildup of and ease the removal of surface contamination. The laundry is kept at a negative pressure with respect to all surrounding areas, to minimize the spread of airoorne contamina tion.
The laundry facility also includes a mask faciliti .        The mask area    includes inspecting,  andspace  and storing    equipment for collecting, cle aning, masks.
12.3-21
 
        *
* B/B-FSAR              AMENDMENT 40 NOVEMBER 1982 12.3.2      shielding The design of the station shielding is based on the design dose rates and the established design criteria.          Using the sources given in Section 12.2 and the shielding design criteria, the shielding design is determined.
12.3.2.1      General Shieldina Design Criteria Every component that handles radioactive fluids may require              l shielding; the thickness of which is based on the operational cycle of the component, the design dose rate, and the shielding ma terial.
12.3.2.1.1      Regulatorv Recuirements The shielding design dose rates for Byron /Braidwood meet 10 CFR 20 and 10 CFR 50, which are concerned with allowable radiation to individuals in restricted and unrestricted areas. The only shielding required to be safety-related is the control room anc the primary containment shielding; this shielding satisfies the requirements stated in Criterion 19 of 10 CFR 50, Appendix A,        and 10 CFR 20.
12.3.2.1.2      Shieldina R ecuirements Radiation protection of personnel, equipment, and materials is
('          largely dependent upon the adequacy of the design of the station shielding system.        Radiation shielding has the passive protection function of radiation attenuation and consists of material placed between radiation sources and personnel and/or equipment and materials needing protection from radiation.
The shielding system is designed and constructed to assure that the station can ce operated and maintained such that the resultan t radiation level and doses are within the limitations of applicable regulations and are as low as is reasonably achievable          l
( AL ARA) . Specific design dose rate limits recommended to achieve this objective are discussed in Subsection 12.3.1 and listed in l              Table 12.3-2.
Shielding must be capable of performing its protective function i
throughout the plant lifetime and under the variable source and i
environmental conditions associated with all normal, anticipated abnormal operational, and design-basis accident conditions identified in the saf ety analysis reports and as noted below.
: a. Normal Goeratino Conditions For the purposes of shielding design, normal station operating conditions are considered to include conditions generally known as anticipated abnormal
(                            operational occurrences. Two modes of normal station
(                            operation are:
12.3-22 t
 
B/B-FSAR
: 1. normal power operation of the reactor, including s
anticipated operational occurrences, and
: 2. normal shutdown of the reactor.
Shielding is designed to provide the required protective function under such conditions.
: b. Accident Conditions Station snielding provides protection to plant operating personnel and the general public under postulated design-basis accident conditions as defined in the Chapter 15.0.
: 1. Control Room Habitability The main control room and associated areas is shielded such that, after a postulated design-basis accident (LOCA) , the whole body dose in the control room for the duration of the accident will not exceed 5 rem or its equivalent to any part of the body, including ingress and egress, as.per requirements 10 CFR 50, Appendix A, Criterion 19. Subsection 6.4.2.5 descrices control room shielding.
(              The radiation shielding protecting the main control room (and associated areas) is designed based on the anticipated radiation environment resulting from the postulated LOCA. Figure 6.4-2 shows an isometric view of the main control room shielding.
: 2. Direct Of fsite Doses All sources in the plant are adequately shielded to assure that radiation levels at the restricted area boundary are in compliance with 10 CFR 20 limits. Adequate station shielding is provided to limit site boundary doses, due to direct and scattered radiation from contained sources within the plant, to as low as practicable limits during normal operation in conformation with 10 CFR 20 and to within the limits specified in 10 CFR 100 during accident conditions.
: 3. Seismic and Safety Classification Structural walls of the station are designed, as required, to meet Seismic Category I requirements. Walls which are shielding walls f
L                may be designed Seismic Category I, depending upon the particular design requirements other 12.3-23
 
B/B-FSAR than radiation protection requirements (e.g.,
C                      structural integrity, load bearing capacity, et c. ) that the walls must meet.
The primary shield, the shield walls for the main control room, and the shield walls for the spent fuel pool are examples of shield walls which are designed Seismic category I.
: c. Protection of Equipment Appropriate shielding is provided, where needed:
: 1. to limit radiation heating of build structural concrete,
: 2. to reduce neutron activation of equipment, and
: 3. to limit radiation to equipment and materials.
Protection from neutrons and from neutron-induced gamma rays is important around neutron sources such as the nuclear reactor core. The primary shield around the reactor vessel and the neutron shield plugs at the reactor vessel nozzles and examples of station shielding designed to protect personnel and equipment against neutron radiation and neutron-      '
C                  induced gamma rays.
: d. Additional Recuirements In addition to the radiation protection f unctions discussed above, the shielding system have other f unctional requirements. These depend, however, generally upon the location of the shield and the access requirements to equipment or areas beyond the shield. Thus, access to an area may be through the shield itself; e.g., through removable shield walls.
;                      Removable shield walls, portable shields, and l                      compensatory shielding are discussed in Subsection l                      12.3.2.1.6.
l 12.3.2.1.3  Desian Recuirements The station shielding system must be capable of performing its protective functions throughout the plant lifetime and under the variable source and environmental conditions which are anticipated and/or postulated for the plant.
The radiation attenuating materials which comprise the station shielding system are selected to assure no significant loss in l
radiation attenuation characteristics for at least 40 years of
(,    plant operation.
12.3-24 l
 
1 B/B-ESAR                              '
12.3.2.1.4    General Description and Design Parameters The shielding system includes all concrete walls and associated radiation attenuating materials (e.g. , lead, steel, and water) which are used to protect the public, plant personnel, equipment, and materials from radiation emitted from radioactive sources contained or generated within the plant. The radiation exposure of individuals, equipment, and materials is a fcnction of the following basic parameters, which are given due consideration in the shielding design:
: a. source strength (type, intensity, energy) ;
: b. number of sources, source geometry, and self-absorption factors;
: c. shielding material, geometery, and mass between source (s) and receptor;
: d. distance between source (s) and receptor;
: e. time that receptor is exposed; and
: f. allowed dose rate or dose.
Where radioactive crud buildup sources are known, the source C    strength parameter is appropriately adjusted and the shielding      '
designed to accommodate the effects of crud buildup for at least 10 years of reactor operation. Where radioactive crud buildup sources are not known, but expected, the shielding design reflects appropriate conservatism to accommodate the expected effects of crud buildup for at least 10 years of reactor operation, and/or protective measures are used, where practicable , e.g. , tnos e discussed in Subsection 12.3.1.
12.3.2.1.5 Shielding Materials and construction Methods Bulk shielding structures such as cubicle shielding walls, floors, and ceilings are mainly designed of ordinary concrete, either of (solid) block or poured-in place construction. Where space concrete limitations (e.g.,
are encountered, a special high density Hematite concrete) is employed to assure adequate radiation protection.      Concrete is a mixture of materials, the exact  proportions of which may differ from application to application. Concrete for radiation shielding is classified as oridinary or high density according to the unit weight or the aggregate. The design or concrete mixtures and forms, the construction of concrete radiation shielding structures, and the l
quality assurance provisions needed to verif y that the desired quality of construction has oeen achieved is in accordance with accepted design criteria for concrete radiation shields.
L 12.3-25
 
    . .                                B/B-ESAR            AMENDMENT 40 NOVEMBER 1982 Poured-in-place concrete construction is normally used for
(      shielding structures which are load-bearing structural walls.
Concrete block walls are provided where necessary to accommodate eqaipment installation, removal, and construction. Concrete block wall installation is controlled to assure as-built radiation attenuation characteristics similar to those expected f rom equivalent poured concrete.
In the case of the primary shield around the reactor vessel, nuclear  heating is not severe enough to warrant special designs (e.g., wa ter cooling coils) for cooling the primary shield.
A special hinged door design is used at the reactor vessel nozzle inspection openings. Inese doors are designed to allow pressure relief in the event of a reactor coolant pipe break. The shield consists of special neutron shielding material, Heat Resistant Cement 277, supplied by Rea'ctor Experiments Inc.
Where a potential of leakage or spillage of radioactive material exists, effective features are provided in the design of the shielding to prevent the _ spread of contamination by seepage through walls. As discussed in Subsection 12.3.1.9, wall surfaces are coated with a nonporous coating to permit effective decontamina tion.
12.3.2.1.6  Removable Shield Walls, Portable Shieldino, and Comoensatory Shielding Shielding is designed to be removable, where required, to provide personnel access for inspection, servicing, maintenance, or replacement of plant equipment.
Removable shield panels are provided in shield walls, floors, or ceilings as necessary where frequent access for maintenance or removal of equipment is required and if radiation levels in the area can cause excessive exposure.      Such shielding is designed to minimize exposure to operating and maintenance personnel.
Compensatory, portable, or temporary shielding is considered in station design only as required where other more permanent shielding is not practicable.      Where compensatory shielding is necessary, provisions are made to accommodate such shielding in terms of space, structural loading, clearances, and equipment accessibility.
The station snielaing system uses three types of removaole shield walls:    stacked unmortared block, shield hatches and plugs, and shield doors. The primary functions of a removable wall are equipment installation, inspection, ma intenance, and removal.
12.3-26
 
B/B-FSAR              AMENDMENT 40 NOVEMBER 1982 The following are guidelines for the design of removable shield (S      walls. The term major requires the removal of a removable shield wall in addition to repairing and maintaining equipment.
12.3.2.1.6.1    Stacked (Unmortared) Block Removable stacked block walls that are provided to accommodate removal of equipment are constructed such that the top of the removable unmortared block sections are offset and provided with a lintel arrangement. The blocks are held in place by special metal frames to resist lateral pressure and seismic loads. Use of stacked unmortared block avoids unnecessary exposure associated with disassembly or mortared blocks.
Removable stacked block shield walls are used in the shield design when a room contains equipment that seldom requires replacement or ma jor maintenance. Seldom is defined in the section as once a year.      The type of shielded equipment which fits into this category are heat excnangers, pumps, and radwaste tanks.
12.3.2.1.6.2    Removaole shield Hatenes and Plucs Removable shield hatches (or removable floor slabs) and plugs are used in the shield' design when a room contains equipment which often requires replacement or maintenance.        Often is defined in this section to mean more frequent than once a year.                  -
In addition to all equipment that falls into "the of ten category," shield hatches or plugs are used, whenever practicable, for access to equipment and piping which have, or are in radiation areas that have, a dose rate grcater than 3 R/hr.
The use of removable shield hatches or plugs minimizes the main-tenance exposure to station personnel; shield hatch and plug design and construction shall be in accordance with ANSI 101.6-1972.
A radiation detector probe access hole is provided in all filter and demineralizer removable shield hatches so that radiation levels of the contained equipment may be measured without
,        removing the shield hatches. This is provided by boring a l
vertical stepped hole in the top of the shield hatch for insertion of a radiation detector.      The arrangement is pictured in Figure 12.3-4.
The types of equipment that require removable shield hatches are        l demineralizers, filters , and pumps and motors, which are radio-active or are in radioactive areas.      Removaole shield plugs are also used for shielding around the reactor nozzle inspection openings.
12.3-27
 
  ~. .
B/B-FSAR 12.3.2.1.6.3    Shield Doors Shield doors are used when access requirements, maintenance re-quirements, or design consideratiun make it undesirable to adequately employ the removable shield walls mentioned previously. Shield doors can also be used with labyrinthine entrances where the dose rate at the entrance due to scattered radiation is greater than the design dose rate.
12.3.2.1.7    Inspection (Ins ervice) and Maintenance Reauirements Shielding is designed to permit access for required inspections, testing, and maintenance of plant systems and components which require these functions.
During construction, shield walls are visually inspected for cracks and separations that might compromise the shield. There are initial preoperational radiation surveys :aken as well as periodic routine radiation surveys during power operation. Thes e surveys serve as a check on the radiation buildup within auxiliary equipment and the adequacy of shielding design.
Installed radiation monitoring systems survey continuously the radiation condition at certain areas of the plant and also serve as a check on the adequacy of shield wall design and construction.
As discussed in Subsection 12. 3.1, biological protection    of
{-    personnel during anticipated inspection and maintenance activities are considered in shielding design in the effort to maintain exposures ALARA.
12.3.2.1.8    shield Thicknesses Shield thicknesses are designed to reduce the average area dose rate to or below the assigned area dose rate level for worst-case conditions of normal plant operation or, where applicable, for accident conditions. Worst-case conditions include source terms appropriate to maximum power level and 1% failed fuel fraction as discussed in Section 11.1.
Shielding thickness are designed with consideration given to all sources    in the area including localized hot spots or penetra tion s. Design parameters are listed in Subsection 12.3.2.1.4. Byron /Braidwood's shielding design is pictured in Figures 12. 3-5 through 12. 3-2 4. Computer codes used in shielding design account for energy spectra and source strengths for each nuclide (including daughter products) , material cross sections or attenuation coefficients for each material or element comprising the shield, dose buildup factors, and other relevent parameters.
L 12.3-28}}

Revision as of 06:17, 31 March 2020

Testimony of GP Lahti Re Contentions 42,111 & 112 on Plant Design for Reducing Occupational Radiation Exposure.Each Design Principle Which Can Reduce Radiation Exposure Used Appropriately
ML20065A929
Person / Time
Site: Byron  Constellation icon.png
Issue date: 02/15/1983
From: Lahti G
COMMONWEALTH EDISON CO., SARGENT & LUNDY, INC.
To:
Shared Package
ML20065A894 List:
References
NUDOCS 8302220325
Download: ML20065A929 (34)
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