ML20035C301
| ML20035C301 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/16/1993 |
| From: | Byrne J, Marshall W GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20035C124 | List: |
| References | |
| NUDOCS 9304070025 | |
| Download: ML20035C301 (24) | |
Text
NQslear TERlU801c03_,04 nEv 9-i, ISSUE DATE March 1993 C ITS *
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O wits DIVISION l
i TECHNICAL EVALUATION REPORT i
FOR CONTAINMENT AIR l
CONTROL ENVELOPE (CACE) bAb DATE ~356bS COG ENG DATE 3'Ib'91 RTR 3~Ib~
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DOCUMENT PAGE OF i
9304070025 930327 PDR ADOCK 05000320 i
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ENuclear Do-m2 15737-2-G03-104 l
Tstle Page of Technical Evaluation Report for Containment Air Loatrol Envelope 2
23 Rev.
SUMMARY
OF CHANGE Approval Date i
0 Issued f,or Use 1
Revised and Issued for Use 2
Revised and Issued for Use 3
Revised to reflect changes in the operation of. the containment purge system, to clarify the radiation monitoring equipment and to reflect revised isotopic distribution.
4 Revised to allow operational controls over surface radioactivity limits, and airborne radioactivity limits prior to opening the roll-up door.
5 Revised to include a detailed description of moni-toring radiation levels when the roll-up doors are open. Also, to correct a typographical error in the X/Q used for fire analyses.
6 Revised to reflect new isotopic distribution for packages of radwaste.
7 Revised to update receptor locations for offsite releases, recalculated offsite dotes for normal oper-ations.
8 Revised to reflect Mode 3 conditions, reduce allowable I
5/92 curies to be staged, incorporate recalculated fire inhalation dose and recalculated annual release doses, delete redundant HEPA Ventilation Unit from the design, and replaced Figure 1 for clarity. Due to format change, all pages become Revision 8.
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Revised to reflect the Facility Mode 3 requirement
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3/93 for containment isolation.
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' TABLE OF CONTENTS SUBJECT PAGE l
1.0 INTRODUCTION
4.0 t
i 1.1 General 4.0 1.2 Organization of Repon 4.0
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i 1.3 Conclusions
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. i 2.0 FACILITY DESCRIPTION 4.0 l
2.1 Purpose of the Facility
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2.2 Summag Description 5.0 i
t 2.2.1 Location 5.0 t
2.2.2 Design Basis 5.0 2.2.3 Building Description
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t 2.3 Major Systems
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2.3.1 HVAC 7.0 2.3.2 Other Major Systems 9.0 2.4 Equipment Haten Removr1 9.0 t
3.0 TECHNICAL EVALUATION
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f 3.1 Dose Assessment and Accident Analysis 10.0' 3.1.1 On-site Dose Assessment 10.0 3.1.2 Offsite Dose Assessment 10.0 3.2 Occupational Exposure 12.0 i
3.3 Design Conditions 13.0 3.3.1 Normal Operations 13.0
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t 3.3.2 Incidents of Moderate Frequency 13.0 l
3.3.3 Infrequent Incidents 14.0 4.0 SAFETY EVALUATION 15.0
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4.1 Technical Specifications / Recovery Operations Plan 15.0 4.2 Safety Questions 15.0 3.0 Revision 9 a
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SUBJECT PAGE-1 1
- l-TABE 3-1 17.0 t
TABE 3-2 18.0 f;
TABE 3-3 19.0 TABE 3-4 20.0
-r FIGURE 1 21.0 FIGURS 2 22.0
'I FIGURE 3 23.0 V
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1.0 INTRODUCTION
1 1.1 General-i The Containment' Air Control Envelope (CACE) pmvides space to mobilize equipment and materials needed to suppon the in-containment activities in i
preparation for the Post Defueling Monitored Storage (PDMS) period. Location of the CACE at the equipment hatch allows equipment and materials to be moved into j
and out of the containment building with a minimum of difficulty through the j
equipment hatch airlock doors. The CACE will serve as an aid in the contml of the i
spread of contamination and airbome radioactivity during those times when the f
airlock doors are opened in accordance with approved procedures.
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This repon does not apply to removal of the equipment hatch, since there is no f
known intention of removing the hatch in the foreseeable future.
I 1.2 Organization of Repon This repon is organized as follows:
l Following this introduction are a description of the design and operational considerations and a discussion of the safety issues associated with the freility.. The repon concludes with the safety evaluation required by 10 CFR 50, paragraph 3
50.59, " Changes, Tests and Experiments."
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1.3 Conclusions l
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The evaluation of the safety concerns detailed in this repon results in the following 1-conclusions:
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The CACE fulfills the need for a facility which allows a large entryway into -
and out of the containment while acting as an aid in the control of the spread j
of contamination and airborne activity.
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The constmetion and operation of the facility is not an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.
I 2.0 FACILITY DESCRIPTION I
2.1 Purpose of the Facility j
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The CACE is used as a staging / packing area for materials and equipment mquiring transfer into or out of the reactor building, while helping to control airborne releases from the reactor building. Contaminated material may be wiped down, wrapped, f
or otherwise protected prior to being brought into the CACE to ensure surface radioactivity does not exceed limits established by the Radiological Controls l
Depanment.
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This building is temporary and is not designed to satisfy the criteria for a pennanent TMI Unit 2 facility. The CACE is not designed to function as a storage area for radioactive waste, but will be used to tempomrily stage radioactive material.
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.2.2 _
Summary Description 2.2.1 Iecation As shown on Figure 1, the CACE is located southwest of the Unit 2 reactor building at the cquipment hatch. The building is built on top of the existing contml building area roof slab at the 305' elevation.
Access through the CACE personnel or roll-up door or through the M-20 area (El. 280'-6" of the control building area) is controlled by j
Radiological Control Procedures.
2.2.2 Design Basis The facility helps to control the releases of contamination and airborne l
mdioactivity from the reactor building when both the equipment hatch airlock doors are open in accordance with approved procedures. It also l
controls paniculate releases from contaminated materials brought into the CACE from the reactor building. - The CACE is a temporary i
facility which will be removed or upgraded to satisfy the criteria for a j
permanent TMI Unit 2 facility prior to plant restan.
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The CACE structure is classified as Not Important to Safety.
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The CACE is designed for the probable natural phenomena as required by the local building codes. It does not have as part of its design basis the severe natural phenomena used for permanent nuclear power plant l
l structures. These severe natural phenomena, such as tornadoes, safe shutdown earthquakes (SSE), and probable maximum floods, are not i
postulated to occur during the shon-tenn design life of the CACE.
. The CACE is designed to conform with 10 CFR Part 20'.1 (c). This
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ensures that personal exposures associated with the CACE are ALARA..
Transit and short-term staging of contaminated material in the CACE contributes to keeping exposures ALARA. In addition, access to the I
building will be controlled in accordance with the Radiological Control l
Procedures in effect at TMI Unit 2.
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2.2.3 Building Description The CACE, shown in Figures 2 and 3, is located adjacent to the reactor building. It is attached to the missile shield door stmeture and the contml building area roof slab which are seismically separated from the reactor building. The missile shield door has been rolled bd as shown in Figure 3 and the joints between the door structure and the -
adjacent stmetures are sealed. The missile shield door will remain '
rolled back for the duration of CACE use.
The building has a-stmetural steel frame with 2-hour fire rated metal siding for a fire from outside the CACE.
The roof is non-fire rated ' galvanized metal-decking. Access to the CACE is through a personnel door located on the nonh side of the building and a 27' mil-up tmck door on the west side of the building. Personnel can also gain entry through the M-20 area and through the equipment hatch airlock from the reactor building.
Interior surfaces of the CACE are covered with sheet metal for ease of decontamination.
2.3 Major Systems L
2.3.1 HVAC I
2.3.1.1 Design Bases The CACE HVAC design assures the following:
a.
Minimize the exfiltration of airborne contaminants to the outside -
1 environment b.
Maintain the concentration of airborne paniculate.in the CACE below the limits defined in 10 CFR 20, Appendix B for 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> occupancy c.
Direct air flow fmm the outside, thmugh the CACE and into the reactor build.ing d.
Maintain a negative pressure inside the CACE with respect to the outside environment e.
Limit differential pressure across the CACE walls to a maximum -
of 2 inches w.g.
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Operate in a manner not to reduce the reactor building average air temperature below 50*F g.
Provide ventilation for the CACE 6.0 Revision 9 -
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2.3.1.2
System Description
2.3.1.2.1 General Description The CACE HVAC System. consists of one filtered exhaust unit, its associated ductwork, - dampers, controls and three pressure relief.
intakes. The reactor building purge system maintains the reactor building atmosphere at a slightly higher negative pressure than the CACE to induce air flow from the outside, through the CACE and into the reactor building. There is no physical connection between the two ventilation systems. There is the possibility if both systems are mn at -
-i the same time to compete against each other during the few times that both equipment hatch doors are simultaneously opened. This mode of operation is prohibited by administrative controls.
The filtered exhaust system has three functions--to provide internal i
cleanup of the CACE atmosphere; to inhibit exfiltration by exhausting the building inleakage; and to reduce the amount of airborne particulate released to the environment. This system will be operated only when one or both equipment hatch airlock doors are closed. The exhaust unit takes iniet air from the CACE atmosphere, processes the exhaust through a prefilter and HEPA filter, and discharges to the outside' i
environment.
The ventilation exhaust will be monitored for i
particulates. The HEPA filters will not be shop tested, but will be DOP tested in place.
Counterweighted pressure relief dampers are provided to limit differential pressure across the CACE walls to a maximum of 2 inches l
w.g. (negadve pressure inside the CACE with respect to the outside.
environment). The dampers provide pressure relief to protect the structure.
Supplemental heating and cooling equipment will be provided for use, if required, to maintain a suitable environment for pe.sonnel and equipment.
2.3.1.2.2 System Operation EXHAUST SYSTEM - The filtered exhaust system may be operated, when one or both equipment hatch airlock doors are closed, when it is necessary to remove any airborne particulate contamination that may be present in the CACE. The system is started by a local handswitch.
l The exhaust system need not be operated when the roll-up tmck door i
is open.
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-t' The exhaust system unit takes inlet air from the CACE atmosphere, processes the' exhaust through a high efficiency prefilter and HEPA I
filter before discharging to the outside environment. An isoladon damper -is provided in' the exhaust duct. The exhaust duct ' is
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isokinetically sampled for particulate activity. A adiation monitor is l
provided with local readout and alarm. Actuation of the alarm trips the i
filter unit and closes the isolation damper on high radiation. The-handswitch provided on the filter unit may also be used to initiate
.j building isolation.
PRESSURE RETJEF - Actuadon of the relief mode'is initiated by.
counterweighted pressure relief dampers set to open when differential l
pressure acmss the CACE walls exceeds 0.75 inches w.g.
The dampers are provided for pressure relief only to protect the structure.
The dampers will normally be closed.
SUPPLEMENTAL HEATING AND COOIlNG - Supplemental heating and cooling units will be operated, if required, to maintain a suitable environment for personnel and equipment. The heating and cooling units will be portable and will be operated independently of the CACE ventilation system. When operating, the units will also be used to circulate air in the CACE when one or both equipment hatch airlock doors are closed.
Heating equipment will maintain a minimum e
temperature of 50*F in the CACE during winter. This equipment will be configured to ensure no release pathway is created from the CACE to the environment.
REACTOR BUILDING PURGE SYSTEM - Whenever both equipment hatch personnel airlock' doors are open, the reactor building purge system will be opemted in one of two modes.
l In the first mode the reactor building purge exhaust system is operated to maintain a minimum capture velocity (200 fpm) through the open airlock in order to prevent uncontrolled outflow from the reactor i
building. This mode is to be used only when the CACE doors are to be maintained open.
In the second mode the reactor building purge exhaust system is operated in a manner that maintains a negative pressure in the reactor building and CACE. To maintain the CACE at a negative pressure slight enough to avoid opening the CACE relief dampers, one or both purge supply train isolation dampers may also be opened.
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2.3.2 Other Major Systems -
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Electrical - Electrical service is pmvided to supply power for lighting, receptacles and electrically operated equipment. LAll electrical systems and the metal structure are gmunded.
Communications - Communications systems consist of surveillance cameras, a sound powered phone and dedicated P.A. -system linked directly to the command center and/or the control room.
Radiation Monitoring - A mobile airborne paniculate monitor with local :
alarm, readout and recorder is. provided for monitoring local air activities. The exhaust fan has a constant air monitor with local madout-and alarm. No permanent area radiation monitors are planned to be installed in the CACE since the radiation levels inside the CACE are expected to be low. Ponable area radiation monitors will be added if required.
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2.4 Equipment Hatch Removal The CACE provides an area large enough to allow for removal and reinstallation i
of the equipment hatch. The roll-up tmek door has been provided to allow removal -
of the equipment hatch from the CACE. The foreseen use of the CACE will only i
utilize opening of the airlock doors and will not utilize removal of the equipment l
hatch.
3.0 TECHNICAL EVALUATION
This section summarizes the licensing issues which were considered in the design of the.
CACE. These issues deal with the expected performance of the facility during normal
'I operations and various design basis events.
The licensing issues associated with the operation of the CACE are:
l Demonstrating compliance with 10 CFR Part 20 with respect to on-site dose limits.
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Demonstrating compliance with 10 CFR Part 50, Appendix I, with respect to offsite radiation doses due to normal operations within the CACE.
o Asse.ssing the consequences of potential accidents in the CACE that could lead to radioactive releases to the envimnment.
Demonstrating compliance with the principles of ALARA.
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Demonstrating that the design conditions specified in the TMI-2 General Project Design Criteria (GPDC) are satisfied.
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j Each of these issues is addressed in the following sections.
3.1 Dose Assessment and Accident Analysis
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3.1.1 On-Site Dose Assessment The CACE is designed for material handling activities that transfer j
material into and out of the reactor building. Measurements of the dose rate outside of the equipment hatch have been taken with the equipment j
hatch airlock doors closed and open. At a point approximately 6 ft.
i from the outer airlock door a dose rate of approximately.3 mrem /hr has been measured with the doors closed, while with both airlock dwrs open a dose rate of approximately.6 mrem /hr has been measured.
Access to the CACE will be controlled by Radiological Control Procedures.
The staging of contaminated material in the CACE may temporarily i
increase the inside and outside area dose rates. Therefore, any staging of contaminated material within the CACE will be controlled, j
monitored, and the operation reviewed prior to implementation in 1
accordance with Radiological Control Procedures on a case-by-case basis. This does not preclude the establishment of procedures or limits for *. asks which are generic in nature, such as staging of contaminated trash from the reactor building.
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3.1.2 Offsite Dose Assessment 3.1.2.1 Normal Operations It is postulated that material handling in the CACE 'may genemte airborne radioactivity which then may be released to the environment.
It is expected that these releases will be extremely low. However, to i
demonstrate the small offsite dose consequences, conservative assumptions were made to obtain the maximum credible calculated doses.
These conservative assumptions included the quantity of radioactivity expected to be processed in the CACE and conservative
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release fractions used for normal operations. Although the CACE ventilation is equipped with a HEPA filter, which normally removes greater than 99.9% of particulate radioactivity, it was assumed that there was no filtration of the CACE exhaust. Thus the actual releases during normal operations in the CACE with the ventilation system in i
operation would be expected to be at least a factor of 1000 lower than calculated releases.
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The handling of contaminated material in the CACE was evaluated to determine the bounding offsite doses. The primary source for airborne 10.0 Revision 9 1
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l mdioactivity in the CACE will be the result of activities related to handling contaminated material from the containment. To assess this dose the following assumptions were made:
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The maximum exposed contaminated surface area that will be-staged through the CACE annual' f is equivalent to the surface area of 10,000 drums (22,400 square meters).-
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The surface contamination is 50,000 dpm/100 cm, and IE-3 of.
2 the surface contamination is released due to material handling.
c.
The total annual release is therefore 5.06 E-5 curies.
d.
No credit is taken for the CACE building, the CACE ventilation l
system with its HEPA filter, or the fact that all material is wiped down or bagged prior to passage into the CACE further reducing or removing extemal contamination.
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Isotopic distribution of surface contamination.is derived from
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radwaste conversion factors -for defueling waste which are
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developed in accordance with TMI-2 procedures. The assumed distribution-for defueling waste is given in Table 3-1.
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s The dose to the public was calculated for these releases using the methodology specified in Regulatory Guide 1.109 and based on 1985 meteorological data for x/Q and D/Q, which has been shown to conservatively represent off-site dose. Receptor locations for residents, gardens, beef and dairy famis were taken from the 1990 land use l
census for TMINS.
The resulting annual dose to the maximally exposed individual is summarized in Table 3-2.
i 3.1.2.2 Contaminated Material Fire For the purpose of evaluating the consequences of a potential fire in the CACE the following assumptions were made:
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The maximum number of curies staged in the CACE at any one time is limited to 250 millicuries. The isotopic distribution of the I
contamination is derived from radwaste conversion factors for defueling waste which are developed in accordance with TMI f procedures. See Table 3-1 for tne assumed distribution.
b.
A release fraction of IE-2 was used to estimate the airborne i
release based on the Atomic Energy Commission report, l
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" Environmental - Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," December 1972.
c.
No credit was taken for HEPA filtration or the CACE building.
- The resulting inhalation dose was calculated at the exclusion ' area boundary distance of 610 meters. The 1-hour meteorological dispersion 3
parameter (x/Q) of 7.67E-4 sec/m for a ground level release was used.
The resulting doses are tabulated in Table 3-3.
3.1.2.3 High Velocity Winds From the TMI-2 FSAR, the design wind velocity, based on the 100-year recurrence interval, is 80 miles per hour at 30 feet above grade. The CACE is designed to withstand this condition.
An evaluation was conducted to assess the radiological consequences of a wind condition at the design wind velocity. Assumptions used in this analysis include the following:
i a.
The maximum number of curies staged in the CACE at any one
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time is limited to 250 millicuries. rhe isotopic distribution of the -
contamination is derived from mdwaste conversion factors for defueling waste which are developed in accordance with TMI-2 procedures. See Table 3-1 for the assumed distribution.
b.
A conservative release fraction of IE-3 was used to estimate the airborne releases.
c.
No credit was taken for the ventilation or the CACE building.
The resulting inhalation dose was calculated at the exclusion area boundary distance of 610 meters.
The meteorological dispersion 8
parameter (x/Q) of 5.5E-6 sec/m for an 80 mile per hour wind was used. The resulting doses are tabulated in Table 3-4.
3.2 Occupational Exposure It is expected that the general area dose mtes that will be experienced in the CACE i
will be less than 1 mrem /hr. This dose rate will result in a small, but uncalculated,
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occupational exposure. However, use of the CACE reduces personnel exposure below that which would occur if the CACE were not used. Worker exposure is reduced because the CACE provides a lower background area to stage and assemble large pieces of equipment which would otherwise have to be transported into the containment and assembled there. This same area will allow contaminated material from the reactor building to be placed in dmms or LSA boxes in a low radiation f
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area. ' This will result in lower occupational exposure for activities associated with j
the reactor building. The minimum number of persons needed to perform activities j
are assigned to ensure that total exposure is ALARA. Access and operations withm j
the CACE are controlled by Radiological Control procedures.
3.3 Design Conditions l
The design conditions which must be satisfied are specified in the TMI-2 GFDC.
l These fall into three categories: ' nonnal operation, incidents of moderate frequency, l
and infrequent incidents. Each of these categories is addressed below.
3.3.1 Normal Opemtions j
t Normal operation conditions are discussed in the previous sections.
These operations will be carried out without unplanned or uncontrolled releases of radioactive materials to the environment because i
o any mdioactive material transferred into the CACE will be wiped i
down, wrapped,. or otherwise protected to meet the limits established by the Radiological Controls Department.
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o When both equipment hatch airlock doors are open, the air flow will be from the CACE into the containment, and exhausted to l
the environment through the containment purge exhaust system.
o Under normal operating conditions, the CACE rollup doors will j
not be opened if a significant potential for airborne contamination j
exists. A continuous air monitor will be operating in the CACE l
when a contaminated area is posted and the rollup doors aru open. The alarm set point for the continuous air monitor is established to meet the conditions of Paragraph 2.3.1.1.b and will be set to alarm at or below 2 MPC-hours (occupational) in i
accordance with Radiological Controls procedures. If the monitor alarms, the CACE rollup doors will be closed, and the CACE ventilation system will be operated in accordance with Section
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Under these conditions, the airborne radioactivity concentrations released to the environment should not exceed 10CFR20 Appendix B Table II values pursuant to 10CFR20 Paragraph 20.106. If an unanticipated airborne release does I
occur, the resulting off-site doses would not be expected to-i exceed a small fraction of 10CFR50 Appendix I values.
3.3.2 Incidents of Moderate Frequency l
The CACE and the equipment provided with the CACE serve no safety related functions and since there is no interface with any safety system, j
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it will not interfere with the performance of any safety related feature.
Therefore, loss of electrical power in the CACE, inadvertent actuation of a component provided ~ with the CACE, single operator error-1 associated with the operation of the CACE, or a single failure of an active component in the CACE will not endanger the health and safety -
of the public.
j Failure of the reactor building purge system will not result in an uncontmiled release of radioactivity to the environment. Pmssure relief i
dampers are pmvided to protect the CACS from too great a negative pressure.
t Normal operations in the CACE will involve the handling of.
l contaminated radioactive material. During the course of handling the packages there is the possibility that a package could be broken open.
This would not result in an uncontrolled release of radioactivity to the environment because of the design of the HVAC system, discussed in '
Section 2.3.1. ' Releases of radioactivity to the environment would be'-
l minimized by the filters in either the containment purge exhaust system.
or filtered exhaus system provided with the CACE. The result of a package breaking vn is enveloped by the normal release calculation.
Packages exceeding t. S 250 millicurie limit for staged materials may be transferTed through ths CACE. These packages will comply with the -
requirements of 49 CFl Part 173 for strong-tight containers. Since the -
materials will not be s iged in the CACE and will.bi contained in.
strong-tight containers, releases due to a postulated drop of'these containers need not be et sluated.
i 3.3.3 Infrequent Incidents Rupture of tanks and pipe breaks are not considered because no tanks j
or liquid lines will be installed in the CACE. A fuel handling incident l
occurring in the CACE is not considered because fuel handling in the -
CACE is not planned. (The handling of waste from the reactor building that may contain small quantities of transuranic material:is not.
considered to be a fuel handling activity). During fuel handling in the containment, the airlock doors will be shut. This will ensure that releases of radioactivity to the environment will be within acceptable limits.
The effect of fire and an operating basis canhquake are considered below.
3.3.3.1 Operating Basis Eanhquake i
In the event of an OBE, the CACE will not cause any damage to the j
reactor building because of the seismic expansionjoint that separates the 14.0 Revision 9 l
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o reactor building from the control building and the missile shield door
-l stmeture to which the CACE is attached. The_ consequences of the j
collapse of the CACE on the control building roof slab are considered bounded by an aircraft impact, described in the TMI-2 FSAR.-
q 3.3.3.2 Fire Protection As noted in Section 2.2.2, opening both of the personnel airlock doors is accomplished by approved procedures. The existing pmcedum.
requires that when both airlock doors am open,. someone is to.be standing by to close the doors expeditiously. in the event of an emergency. Should a fire occur in the containment when both of the -
airlock doors are open, one of the doors will be closed as required by procedure, and the control room will be notified, thereby reestablishing the containment boundary and preventing an uncontrolled release of radioactivity to the environment. The addition of the CACE will not change the reactor building fire boundary since the equipment hatch will' i
remain installed.
4.0 SAFETY EVALUATIOE 4.1 Technical Specifications / Recovery Operations Plan l
The operation of the CAC2 with respect to staging contaminated material does not impact any existing Technical Specification / Recovery Operations Plan requirement i
nor does it require any additional Tecimical Specification / Recovery Operation Plan l
requi ement. The opening of 11+ airlock doors will continue in accordance with existing procedures.
4.2 Safety Questions (10CFR50.59)-
10CFR50, Pamgmph 50.59, permits the holder.of an operating license to make changes to the facility or perform a test or experiment, provided the change, test, or experiment if it is determined not to be an unreviewed safety question and does not involve a modification. of the plant technical specifications.
r A proposed change involves an unreviewed safety question if:
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The possibility of occurrence or the consequences of an accident or a.
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malfunction of equipment imponant to safety previously evaluated in the safety analysis repon may be inemased; or i
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The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or 15.0 Revision 9 i
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The margin of safety, as defm' ed in the basis for any technical specincation.
is reduced.
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The CACE does not increase the probability of occunence of an accident or-malfunction of equipment imponant to safety previously evaluated in the safety-analysis repon. Containment isolation will be maintained with the CACE installed j
in accordance with existing technical specifications. As can be seen from Figures i
2 and 3, the CACE is supponed by the existing imid. Aield area (i.e., missile shield suppon stmeture). The CACE is attached to the missile shield door structure and the control building area roof slab which are seismically separated from the' j
reactor building itself. There is no interface between systems provided in the l
CACE and any safety related systems. Therefore, the CACE will not impact j
existing safety related structures or systems and there will be no increase in the j
probability of an accident or malfunction of equipment imponant to safety.
i The CACE does not increase the consequences of an accident beyond acceptable l
criteria established by the NRC.
Section 10.4.1.2 of NUREG-0683, " Final 4
Programmatic EnvironmentalImpact Statement Related to the Decontamination and 4
Disposal of Radioactive Wastes Resulting from March 28,1979 ' Accident Thme.
Mile Island Nuclear Station, Unit 2," evaluated a fire in a low level waste stomge l
area. The results of this evaluation were compared to the offsite dose limits presented in 10 CFR Pan 20. Based on 10 CFR 20.105(a), the NRC concluded that since the offsite dose due to a fire in a low level waste stomge area did not exceed the limits of 10 CFR Pan 20 for normal operation, this type of accident did not pose a significant risk to the heahh and safety of the public.
i Using the same criteria as above, a postulated fire in the CACE can be demonstrated not to pose a signincant risk to the health and safety of the public.
l Since specific organ dose limits are not given in 10 CFR Pan 20, the methodology i
and weighting factors contained in ICRP Publication 26, " Recommendations of the Intemational Commission on Radiological Protection," adopted January 17,197'i, were used to determine a whole body dose equivalent in risk to the organ doses reponed in Table 3-3. This equivalent whole body dose was calculated to be S.1 mrem. Since 8.1 mrem is a very small fmetion of the 10 CFR Pan 20 limit of 500 mrem per year for normal opemtions, it can be concluded that a Gre in the CACE l
does not pose a signincant risk to the health and safety of the public.
j 4
The possibility of an accident or malfunction of a different type than any previously j
evaluated in the safety analysis repon is not created by the existence of the CACE.
i l
This is due primarily to the passive nature of the facility and the ability to quickly reestablish containment isolation in the event of an emergency. Also, the operation of the CACE does not result in a reduction in the margin of safety as defined in the technical specifications since the CACE does not impact any systems covered in the technical specifications and any release of radioactivity from the CACE will be monitored for compliance with environmental technical specifications.
Based on the above, the CACE is deemed not to be an unreviewed safety question as defined in 10 CFR 50.59.
16.0 Revision 9 i
~
TABLE 3-1 ASSUMED ISOTOPIC DISTRIBUTION FOR DEFUELING WASTE.
t
.I
-I Radionuclide Fraction Sr-90
' O.509-i Cs-134 0.100 I
Cs-137 0.340 Pu-238 2.96E-4 Pu-239 3.44E-3 Pu-240 9.07E-4 i
Pu-241-4.12E-2
?
i Am-241 5.59E-3.
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17.0 Revision 9
i TABLE 3-2 i
CALCULATED AhWUAL DOSE TO THE MAXIMALLY EXPOSED INDIVIDUAL FROM ROUTINE RELEASES FROM THE CACE l
'I Annual Dose from Inhalation, Vegetable, Intake, Meat Consumption, Cow Milk and Ground i
Plane Dose (mrem /vr)
Ace Group Bone Total Body 4
Adult 5.3E-1 9.l E-2 Teen 6 lE-1
- 1. lE-1 Child 7.9E-1 1.7E-1 i
Infant 1.2E-1 1.9E-2 i
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18.0 Revision 9 e
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TABLE 3-3 t
CALCULATED INHALATION DOSE AT THE EXCLUSION AREA BOUNDARY FOR A FIRE IN THE CACE l
l
's Contmlling Organ Weighting Equiv. Whole l
Organ Ace Group Dose (mrem)
Factor
- Body Dose (mrem)
-Bone Adult 45 0.12 *
- 5.4 i
i Total Adult 2.7 1.0 2.7 Body Total 8.1
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r r
ICRP Publication 26,' " Recommendations of the International Commission on Radiological Protection," adopted January 17,1977
"" Waighting factor for red bone marrow is used for all bone dose..This overestimates 'the
[
equivalent whole body dose since some radionuclides tend to remain deposited on the bone.
1 surfaces, for which a lower weighting factor may be used.
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19.0 Revision 9 l
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TABLE 3-4 ii CALCULATED INHALATION DOSE AT THE EXCLUSION AREA BOUNDARY FOR. A HIGH VELOCITY WIND I
Orcan Controlling Age Gmun Dose (mrem)
Bone Adult 3.0E-2 Total Body Adult 1.9E l i
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