ML20207E580
| ML20207E580 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/04/1986 |
| From: | Smith E GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20207E575 | List: |
| References | |
| 15737-2-G03-104, 15737-2-G03-104-R07, 15737-2-G3-104, 15737-2-G3-104-R7, NUDOCS 8701020215 | |
| Download: ML20207E580 (24) | |
Text
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3170-011 Nuclear TER 15737-2-G03-104 REV.
7" ISSUE DATE March 8, 1985 0 ITS O NSR O NITS TMI-2 DIVISION TECHNICAL EVALUATION REPORT FOR Containment Air Control Envelope (CACE)
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(\\lLJC3lEDElf 3170-011 15737-2-G03-104 Title PAGE 2
OF 23 Technical Evaluation Report for Containment Air Control Envelope Rev.
SUMMARY
OF CHANGE O
Issued for 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 monitoring 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 doses for normal operations.
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REVISION STATUS SHEET JOB 15737 REV.
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Containment Air Control Envelope 7
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15737-2-G03-104 3170-011 Containment Air Control Envelope Technical Evaluation Report Table of Contents E*S*.
1.0 Introduction 5
1.1 General 5
4 1.2 Organization of Report 5
1.3 Conclusions 5
2.0 Facility Description 5
2.1 Purpose of Facility 5
2.2 Summary Description 6
2.2.1 Location 6
2.2.2 Design Basis 6
2.2.3 Building Description 6
2.3 Major Systems 7
2.3.1 HVAC 7
2.3.2 Other Major Systems 9
2.4 Equipment Hatch Removal 9
3.0 Technical Evaluations 10 3.1 Dose Assessment and Accident Analysis 10 3.1.1 On-site Dose Assessment 10 3.1.2 Offsite Dose Assessment 11 3.2 Occupational Exposure 13 3.3 Design Conditions 13 3.3.1 Normal Operation 13 3.3.2 Incidents of Moderate Frequency 14 3.3.3 Infrequent Incidents 15 3.3.3.1 Operating Basis Earthquake 15 3.3.3.2 Fire Protection 15 4.0 Safety Evaluation 15 4.1 Technical Specifications / Recovery Operations Plan 15 4.2 Safety Questions 16 Revision 3 0073Y
15737-2-G03-104 3170-011
1.0 INTRODUCTION
l.1 General The Containment Air Control Envelope (CACE) provides space to mobilize equipment and materials needed to support the in-containment activities through defueling. Location of the CACE l
at the equipment hatch allows equipment and materials to be moved into and out of the containment building with a minimum of difficulty through the equipment hatch airlock doors. The CACE will serve as an aid in the control of the spread of contamination and airborne radioactivity during those times when the airlock doors are opened in accordance with procedures approved by the NRC.
This report does not apply to removal of the equipment hatch. A separate report will be prepared for removal of the equipment hatch.
1.2 Organization of Report This report is organized as follows:
After this introduction, a description of the design and operational considerations is presented. This is followed by a discussion of the safety issues associated with the facility.
The report concludes with the safety evaluation required by 10 CFR 50, paragraph 50.59, " Changes, Tests and Experiments."
4 1.3 Conclusion The evaluation of the safety concerns detailed in this report results in the following conclusions:
o 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 of contamination and airborne
- activity, o
The construction and operation of the facility is not an unreviewed safety question as defined in 10 CFR 50 paragraph 50.59.
1 2.0 FACILITY DESCRIPTION 2.1 Purpose of the Facility The CACE is used as a staging / packing area for materials and equipment requiring 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, or otherwise protected prior to being brought into the CACE to ensure surface radioactivity does not exceed limits established by the Radiological Controls Department.
i.
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3170-011 This building is temporary and is not designed to satisfy the criteria for a permanent TM1 Unit 2 facility. The CACE is not designed to function as a storage area for radioactive waste, but will be used to temporarily stage radioactive material.
2.2 Summary Description 2.2.1 Location As shown on Figure 1, the CACE is located southwest of the Unit 2 reactor building at the equipment hatch. The building is built on top of the existing control 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 Radiological Control Procedures.
2.2.2 Design Basis The facility helps to control the releases of contamination and airborne radioactivity from the reactor building when both the equipment hatch airlock doors are open in accordance with procedures approved by the NRC.
It also controls particulate releases from contaminated materials brought into the CACE from the reactor building. The CACE is a temporary facility which will be removed or upgraded to satisfy the criteria for a permanent TMI Unit 2 facility prior to plant restart.
The CACE structure is classified as Important to Safety for fire protection only. The HVAC equipment is classified as Not Important to Safety; however, the ventilation exhaust monitors are classified g
Important to Safety as the exhaust path is a release point.
I 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 phenemena used for permanent nuclear power plant structures. These severe natural phenomena, such as tornadoes, safe shutdown earthquakes (SSE), and probable maximum floods, are not postulated to occur during the short-term design life of the CACE, The CACE is designed to conform with 10 CFR Part 20.1 (c). This 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 che building will be controlled in accordance with the Radiological Control Procedures in effect at TMI Unit 2.
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
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.....15737-2-G03-104 3170-011 structure and the control building area roof slab which are seismically separated from the reactor building. The missile shield door will be rolled back as shown in Figure 3 and the joints between the door structure and the adjacent structures will be sealed. The missile shield door will remain rolled back for the duration of CACE use. The building has a structural 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 north side of the building and a 27' roll-up truck 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 2.3.1 HVAC 2.3.1.1 Design Bases The CACE HVAC design assures the following:
a.
Minimize the exfiltration of airborne contaminants to the outside environment b.
Maintain the concentration of airborne particulate 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 from the outside, through the CACE and into the reactor building 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.
f.
Operate in a manner not to reduce the reactor building average air temperature below 500F 3
g.
Provide ventilation for the CACE 2.2.1.3
System Description
2.3.1.3.1 General Description The CACE HVAC System consists of two filtered exhaust units, their associated ductwork, dampers, controls and three pressure relief intakes. The system interfaces with the reactor building purge system on an operational basis when both equipment hatch airlock doors are open. The purge system maintains the reactor building i
atmosphere at a slightly higher negative pressure than the CACE to induce air flow f rom the outside, through the CACE and into the reactor building. There is no physical connection between the two ventilation systems. Revision 3 0073Y
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3170-011 e.
The filtered exhaust system has three functions--to provide internal 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 units take inlet air from the CACE atmosphere,-
process the exhaust through a prefilter and HEPA filters, and discharge to the outside environment. The ventilation exhaust will be monitored for particulates in compliance with the Recovery Operations Plan. 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 w.g. (negative 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 personnel and equipment.
2.3.1.3.2 System Operation EXHAUST SYSTEM - The filtered exhaust system will 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. Normally one unit will be operated at a time, but both filter units may be operated simultaneously if required. The exhaust system need not be operated when the roll-up truck door is open.
The exhaust system unit takes inlet air from the CACE atmosphere, processes the exhaust through a high efficiency prefilter and HEPA filters before discharging to the outside environment. An isolation damper is provided in each exhaust duct and each exhaust
+
duct is isokinetically sampled for particulate activity. The radiation monitor is provided with local readout and alarm.
Actuation of the alarm trips the filter unit and closes the isolation damper on high radiation. The handswitches provided on each filter unit may also be used to initiate building isolation.
PRESSURE RELIEF - Actuation of the relief mode is initiated by counterweighted pressure relief dampers set to open when differential pressure across 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 COOLING - Supplemental heating and cooling units will be operated, if required, to maintain a suitable environment for personnel and equipment. The heating and cooling Revision 3 0073Y
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H5737-8c20E-B@4 3A70-011 1
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 temperature of 500F 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 operated in one of two modes.
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 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, I
one or both purge supply train isolation dampers may also be opened.
2.3.2 Other Major Systems Electrical Electrical service is provided to supply power for lighting, receptacles and electrically operated equipment. All electrical systems and the metal structure are grounded.
Coamunications 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 particulate monitor with local alarm, readout and recorder is provided for monitoring local air activities. Each exhaust fan has a constant air monitor with local readout 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.
Portable area radiation monitors will be added if required.
2.4 Equipment Hatch Removal The CACE provides an area large enough to allow for removal and reinstallation of the equipment hatch. The roll-up truck door has Revision 3 0073Y
15737-2-G03-104 3170-011 4
been provided to allow removal of the equipment hatch from the CACE.
The initial use of the CACE will only utilize opening of the airlock doors and will not utilize removal of the equipment hatch.
3.0 TECHNICAL EVAMATION 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 operations and various design basis events.
The licensing issues associated with the operation of the CACE are:
o Demonstrating compliance with 10 CFR Part 20 with respect to on-site dose limits.
j o
Demonstrating compliance with 10 CFR Part 50, Appendix I, with respect to offsite radiation doses due to normal operations within the CACE.
o Assessing the consequences of potential accidents in the CACE that could lead to radioactive releases to the environment.
1 o
Demonstrating compliance with the principles of ALARA.
o Demonstrating that the design conditions specified in the E4I-2 I
General Project Design Criteria (GPDC) are satisfied.
Each of these issues is addressed in the following sections.
3.1 Dose Assessment and Accident Analysis 3.1.1 On-Site Dose Assessment I
The CACE is designed for material handling activities that transfer t
material into and out of the reactor building. Measurements of the dose rate outside of the equipment hatch have been taken with the equipment hatch airlock doors closed and open.
At a point approximately 6 ft. from the outer airlock door a dose rate of approximately.3 ares /hr has been measured with the doors closed, while with both airlock open a dose rate of approximately
.6 ares /hr has been measured.
Access to the CACE will be controlled by Radiological Control Procedures.
The staging of contaminated material in the CACE will temporarily increase the inside and possibly the outside area dose rates.
Therefore, any staging of contaminated material within the CACE will be controlled, monitored, and the operation reviewed prior to
}
implementation in accordance with Radiological Control Procedures l
on a case-by-case basis. This does not preclude the establishment i
of procedures or limits for tasks which are generic in nature, such j
as staging of contaminated trash from the reactor building.
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15737-2-G03-104 3170-011 3.1.2 Offsite Dose Assessment 3.1.2.1 Normal Operations It is postulated that material handling in the CACE may generate airborne radioactivity which chen may be released to the environment. It is expected that these releases will be extremely low. However, to 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 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 would be expected to be at least a factor of 1000 lower than calculated releases.
The handling of contaminated material in the CACE was evaluated to determine the bounding offsite doses.
The primary source for airborne radioactivity 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:
a.
The maximum exposed contaminated surface area that will be staged through the CACE annually is equivalent to the surface area of 10,000 drums (22,400 square meters).
b.
The surface contamination is 50,000 dpm/100 cm2, and 10-3 of the surface contamination is released due to material handling.
c.
The total annual release is therefore 5.06 x 10-5 curies.
d.
No credit is taken for the CACE building or the CACE ventilation system or HEPA filters.
I e.
Isotopic distribution of surface contamination is derived from rudwaste conversion factors for defueling waste which are developed in accordance with THI-2 procedures. The assumed distribution for defueling waste is given in Table 3-1.
The dose to the public was calculetad for these releases based on the following parameters:
a.
A milk cow is located 1.1 miles east of the release point. The corresponding meteorological dispersion and deposition parameters at this location are 6.91 x 10-6 sec/m3 and 2.05 x 10-8/m2 for X/Q and D/Q, respectively. Although this may I
not be the nearest milk cow, it is the receptor location with l
the highest X/Q and D/Q for this pathway. Revision 7 0073Y
15737-2-G03-104 3170-011 b.
The nearest residence is 0.4 mile east southeast of the release Thecorregpondingmeteorologicalparametersare point.
sec/m and 1.80 x 10-7/m2 for X/Q and D/Q, 3.96 x 10-5 respectively, at this location.
c.
The nearest milk goat is 1.2 miles north of the release point.
The corresponding meteorological parameters are 7.83 x 10-6 sec/m3 and 1.71 x 10-8/m2 for X/Q and D/Q, respectively, at this location.
Intake of radioactive materials by the meat pathway is also assumed for this location.
d.
The nearest vegetable garden is located 0.5 miles east southeast of the release point. The corresponding meteorological parameters are 2.51 x 10-5 sec/m3 and 1.06 x 10-7/m2 for X/Q and D/Q, respectively, at this location.
e.
The dose rate from the ground plane source was calculated based on the location of the nearest residence as described in b.
above.
The resulting annual dose to the maximally exposed individual is summarized in Table 3-2.
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:
a.
The maximum number of curies staged in the CACE at any one time is limited to 10 curies. The isotopic distribution of the contamination is derived from radwaste conversion factors for defueling waste which are developed in accordance with THI-2 procedures. See Table 3-1 for the assumed distribution.
j b.
A release fraction of 10 2 was used to estimate the airborne release based on the Atomic Energy Commission report,
" Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," December 1972.
No credit was taken for HEPA filtration or the CACE building.
c.
The resulting inhalation dose was calculated at the exclusion area boundary distance of 610 meters. The 1-hour meteorological dispersion parameter (X/Q) of 6.1 x 10-4 sec/m3 for a ground level release was used as discussed in Appendix 2D of the Three Mile Island Unit 2 Final Safety Analysis Report (FSAR).
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. Revision 7 0073Y
3170-011 1
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:
4 a.
The maximum number of curies staged in the CACE at any one time is limited to 10 curies. The isotopic distribution of the contamination is derived from radwaste conversion factors for defueling waste which are developed in accordance with THI-2 procedures. See Table 3-1 for the assumed distribution.
i b.
A conservative release fraction of 10-3 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 parameter of 6.8 x 10-6 sec/m3 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 rates that will be experienced in the CACE will be less than 1 ares /hr. This dose rate will result in a small, but uncalculated, 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 I
l area to stage and assemble large pieces of equipment which would otherwise have to be transported into the containment and assembled i
i there.
This same area will allow contaminated material from the j
reactor building to be placed in drums or LSA boxes in a low radiation area.
This will result in lower occupational exposure for activities associated with the reactor building.
The minimum number of persons needed to perform activities are assigned to ensure that total exposure is ALARA. Access and operations within the CACE are controlled by Radiological Control procedures.
3.3 Design Conditions The design conditions which must he satisfied are specified in the IMI-2 GPDC. These fall into three categories: normal operation, j
incidents of moderate frequency, and infrequent incidents.
Each of j
these categories is addressed below.
3.3.1 Normal Operations l
Normal operation conditions are discussed in the previous i
sections. These operations will be carried out without unplanned l
or uncontrolled releases of radioactive materials to the environment as a result of:
any radioactive material transferred into the CACE will be o
l wiped down, wrapped, or otherwise protected to meet the limits established by the Radiological Controls Department.
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15737-2-C03-104 3170-011 o
When both equipment hatch airlock doors are open, the air flow will be from the CACE into the containment, and exhausted to the cavironment through the containment purge exhaust system.
o Under normal operating conditions, the CACE rollup doors will not be opened if a significant potential for airborne contamination exists. A continuous air monitor will be operating in the CACE when a contaminated area is posted and the rollup doors are 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 2 MPC-hours (occupational) in 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 2.3.1.
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 occur the resulting off-site doses would not be expected to exceed a small fraction i
of 10CFR50 Appendix I values.
3.3.2 Incidents of Moderate Frequency The CACE and the equipment provided with the CACE serve no safety related functions and since there is no interface with any safety system, 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 associated with the operation cf the CACE, or a single failure of an active component in the CACE will not endanger the health and safety of the public.
i Failure of the reactor building purge system will not result in an uncontrolled release of radioactivity to the environment.
Pressure relief dampers are provided to protect the CACE from too great a l
negative pressure.
i Normal operations in the CACE will involve the handling of 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 l
the environment would be minimized by the filters in either the containment purge exhaust system or filtered exhaust system provided with the CACE.
The result of a package breaking open is enveloped by the normal release calculation.
Packages exceeding the 10 curie limit for staged materials may be transferred through the CACE.
These packages will comply with the requirements of 49 CFR Part 173 for strong-tight containers. Since
]
the materials will not be staged in the CACE and will be contained in strong-tight containers releases due to a postulated drop of these containers need not be evaluated.
' Revision 7 0073Y A m
LQ67RV-7FMF-WJo 3170-011 3.3.3 Infrequent Incidents Rupture of tanks and pipe breaks are not considered because no tanks or liquid lines will be installed in the CACE. A fuel handling incident 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 earthquake are considered below.
3.3.3.1 Operating Basis Earthquake In the event of an OBE, the CACE will not cause any damage to the reactor, building because of the seismic expansion joint that separates the reactor building from the control building and the missile shield door structure to which the CACE is attached.
The consequences of the collapse of the CACE on the control building roof slab are considered bounded by an aircraft impact, described in the TMI-2 FSAR.
3.3.3.2 Fire Protection As noted in Section 2.2.2, opening both of the personnel airlock doors is accomplished by procedures approved by the NRC. The existing procedure requires that when both airlock doors are 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 by the individual required by the procedure, and the control room 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 remain installed.
4.0 Safety Evaluation 4.1 Technical Specifications / Recovery Operations Plan The operation of the CACE with respect to staging contaminated material does not impact any existing technical specifications nor does it require any additional technical specifications. The ventilation exhaust radiation monitor is presently included in Recovery Operations Plan and no other modifications to the Recovery Operations Plan are required.
Since the opening of the airlock doors will continue in accordance with existing procedures, the operation of the CACE does not require changes to the Recovery Operation Plan.
l l
i Revision 7 0073Y n
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15737-2-G03-104 3170-011 O*
4.2 Safety - Questions (10CFR50.59) j 10CFR50,' Paragraph 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 is determined not to be an unreviewed safety question and does not involve a modification of the plant technical specifications.
a i
A proposed change involves an unreviewed safety question if:
a) The possibility of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the nafety analysis report may be increased; or b) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or c) The margin of safety, as defined in the basis for any technical specification, is reduced.
t j -
The CACE does not increase the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
The containment integrity a
{
will be maintained with the CACE installed in accordance with existing technical specifications. As can be seen from Figures 2 and 3, the CACE is supported by the existing hatch shield area (i.e., missile shield support structure). The CACE is attached to the missile shield door structure and the control building area roof slab which are seismically separated from the reactor building itself.
There is no interface between systems provided in the CACE and any safety related systems. Therefore, the CACE will not impact existing safety related structures or systems and there will be no increase in the probability of an accident or malfunction of l
equipment-important to safety.
4 s
The CACE does not increase the consequences of an accident beyond I
acceptable criteria established by the NRC. Section 10.4.1.2 of
)
NUREG-0683, " Final Programmatic Environmental Impact Statement Related to the Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident Three Mile Island Nuclear Station, Unit 2," evaluated a fire in a low level waste storage area.
The results of this evaluation were compared to the offsite dose limits presented in 10 CFR Part 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 storage area did not exceed the limits of l
10 CFR Part 20 for normal cperation, this type of accident did not pose a significant risk to the health and safety of the public.
Using the same criteria as above, a postulated fire in the CACE can be demonstrated not to pose a significant risk to the health and I
safety of the public. Since specific organ dose limits are not l
given in 10 CFR Part 20, the methodology and weighting factors I Revision 7 0073Y
15737-2-Gd3-1 4 3170-011 contained in ICRP Publication 26, " Recommendations of the International Commission on Radiological Protection," adopted January 17, 1977, were used to determine a whole body dose equivalent in risk to the organ doses reported in Table 3-3.
This equivalent whole body dose was calculated to be 119 area. Since 119 area is within the 10 CFR Part 20 limit of 500 mrem per year for normal operations, it can be concluded that a fire in the CACE does not pose a significant risk to the health and safety of the public.
The possibility of an accident or malfunction of a different type than any previously evaluated in the safety analysis report is not created by the existence of the CACE. This is due primarily to the passive nature of the facility and the ability to quickly reestablish containment integrity 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. Revision 7 0073Y
. - - - =
15737-2-G03-104 3170-011 TABLE 3-1 ASSUMED ISOTOPIC DISTRIBUTION FOR DEFUELING WASTE Radionuclide Fraction Sr-90 0.509 Cs-134 0.100 Cs-137 0.340 Pu-238 2.96E-4 Pu-239 3.44E-3 Pu-240 9.07E-4 Pu-241 4.12E-2 Am-241 5.59E-3 i Revision 7 0073Y
15737-2-G03-104 3170-011 TABLE 3-2 CALCULATED ANNUAL D03E TO THE MAXIMALLY l
EXPOSED INDIVIDUAL FROM RELEASES FROM THE CACE I.
Annual Dose from Inhalation, Vegetable Intake, Cow Milk, Meat, and l
Ground Plane Dose to Organ (arem/yr)
Age Group bone Total Body Iaing Liver Adult 2.3E-2 2.9E-3 1.1E-3 2.5E-3 Teen 2.2E-2 2.5E.?
1.8E-3 2.8E-3 child 2.2E-2 3.0E-3 1.6E-3 2.5E-3 Infant 7.3E-3 8.5E-4 1.2E-3 1.6E-3 II.
Annual Dose from Inhalation, Vegetable Intake, Goat Milk, Meat, and Ground Flane Dose to Organ (ares /yr)
Age Group Bone Total Body Lung Liver Adult 2.4E-2 3.1E-3 1.1E-3 2.7E-3 Teen 2.3E-2 2.8E-3 1.8E-3 3.1E-3 Child 2.4E-2 3.4E-3 1.7E-3 3.0E-3 Infant 9.2E-3 1.2E-3 1.3E-3 2.6E-3 i
- i i Revision 7 0073Y
15737-2-G03-104 3170-011 9
Table 3-3 CALCULATED INHALATION DOSE AT THE EXCLUSION AREA BOUNDARY FOR A FIRE IN THE CACE Controlling Organ Weighting Equiv. Whole Organ Age Group Dose (arem)
Factor
- Body Dose (area)
Bone Teenager 656 0.12**
78.7 Total Body Teenager 27 1.0 27.0 Lung Teenager 64 0.12 7.7 Ilver Teenager 96 0.06 5.9 Total 119.3
- ICRP Publication 26, " Recommendations of the International Commission on Radiological Protection," adopted January 17, 1977 C* Weighting 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 surfaces, for which a lower weighting factor may be used.
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15737-2-G03-104 3170-011 TABLE 3-4 CALCULATED INHALATION DOSE AT THE EXCLUSION l
AREA BOUNDARY FOR A HIGH VEIDCITY WIND Organ Controlling Age Group Dose (mrem)
Bone Teenager 7.3E-1 Total Body Teenager 3.0E-2 Imag Teenager 7.1E-2 Liver Teenager 1.1E-1 l Revision 6 0073Y jj
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