ML20217E081
| ML20217E081 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/24/1998 |
| From: | Bennett A, Boughton K, Irani A GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20217E040 | List: |
| References | |
| TR-121, TR-121-R, TR-121-R00, NUDOCS 9803300376 | |
| Download: ML20217E081 (29) | |
Text
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PU NUCLEAR TMI-I CONTROL ROOM HABITABILITY FOR MAXIMUM HYPOTHETICAL ACCIDENT 1
TOPICAL REPORT #121 REV.0 s,.
AUTHOR:
OLah-p sluju Ardesar A trani DATE (Contributors: K.G. Boughton.P. A. Bennett. T.Y. Byoun. DJ. Distel. B A. Parfitt)
APPROVALS:
SECTION MANAGER DATE AR dA
</u/n DEPARTMENTDIRECTOR DATE
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DIRECTOR.ENGINEF/MG DATE (SIGNIFICANT!MPACT REVIEW) 4 9803300376 980324 7
PDR ADOCK 05000289!
P PDR
.6 Topical Rcport #121 Rev.O Page 2 of 29 ABSTRACT By previous agreement with the NRC staff, GPUN was required to only address the whole body and beta skin doses for Control Room Habitability. However, by letter dated September 24,1997, the NRC requested GPUN to provide an analysis of the thyroid dose within six months. This Topical Report describes the methodology and results of this re-analysis. The ARCON96 computer program and an altemate methodology based on data sources from wind tunnel experiments are used for evaluating atmospheric dispersion. The STARDOSE computer program is used to calculate control room dose.
The analysis shows that the worst radiological consequences for the control room occur with an assumed failure of dampers AHD-37/39, and result in a dose of 4.0 Rom skin, 0.3 Rem whole body and 29 Rem thyroid. These values are below the 10 CFR Appendix A, GDC 19 limits and the NRC Standard Review Plan 6.4 guideline of 30 Rem for the j
control room. Further, an evaluation was perforgled to assure that the Maximum Hypothetical Ace; dent (MHA)is the most severe design basis acedent for Control Room
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operator dose. This revised methodology significantly enhances the Control Room dose assessment by utilizing updated radiological methodologies and including thyroid dose.
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Topical Report #121 Rev.O Page 3 of 29 TABLE OF CONTENTS TMl-1 CONTROL ROOM HABITABILITY FOR MAXIMUM HYPOTHETICAL ACCIDENT PAGE
1.0 INTRODUCTION AND BACKGROUND
4 2.0 CONTROL BUILDING EMERGENCY ENVELOPE DESCRIPTION 5
2.1 Control Building Design 5
2 . Control Building Ventilation System 6
3.0 ATMOSPHERIC DISPERSION MODELLING 9
3.1 Yard intake X/Q 11 3.2 Ventilation Exhaust X/O 12 4.0 CONTROL ROOM DOSE ANALYSIS 13 4.1 STARDOSE Code and Model 13 4.2 Analysis Assumptions 15 4.3 Cases Analyzed and Results 17 4.4 Evaluation of Other Design Basis Accidents 18 5.0
SUMMARY
AND CONCLUSIONS 20
6.0 REFERENCES
20 4
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Topical Riport #121 Rev. O Page 4 of 29
1.0 INTRODUCTION AND BACKGROUND
As a result of NUREG-0737, Item Ill.D.3.4, GPU Nuclear had previously performed testing and analysis to demonstrate that for TMI-1, the Control Room (CR)
Ventilation System was able to adequately protect the CR operators against the effects of accidental release of radioactive gases and provide verification that the plant can be safely operated from the CR under design bas:s accident conditions.
The NRC staff reviewed and approved the GPU Nuclear submittal and issued a supplemental Safety Evaluation Report (SER) on August 14,1986 (Reference 1).
This letter recognized that, by previous agryment with the NRC staff, GPUN was required to address only the whole body And beta skin doses. Consideration of thyroid doses from iodine releases was deferred until completion of the source term re-evaluation effort (revised source term). It was recognized by the NRC that in the interim, adequate protection is provided because the control building ventilation system is able to maintain a positive pressure in the main control room, charcoal filters are provided in the control building ventilation. system, and self-contained
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breathing apparatus and potassium iodide are available.
By letter dated September 24,1997, the NRC acknowledged that the staff position at the time of the above SER was an interim approach with the expectation that the source term study would be completed in a few years (Reference 2). The NRC letter further stated that since the aforementioned study is yet to be completed GPU Nuclear should provide an analysis of the control room operator thyroid dose within six months.
In order to address this NRC request, GPUN reviewed the previous control room habitability analyses to identify the most limiting cases for re-analysis. The previous CR Habitability study was performed for a Maximum Hypothetical Accident (MHA),
which is a large break loss-of-coolant accident (LOCA) with a postulated gross release of fission products based on TID-14844. Extemal pathways have been analyzed with postulated single failures in the Control Building Ventilation System.
' Additional Control Building Ventilation System testing has been performed to measure differential pressure from the Control Building Emergency Envelope (CBEE) to the Patio and stairwell areas.
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Topic:t Rrport #121 Rev.0 Page 5 of 29 This Topical Report desenbes the methodology and results of this re-analysis The -
ARCON96 computer program and an alternate methodology based on data sources from wind tunnel experiments are used for evaluating atmospheric dispersion. The STARDOSE computer program is used to calculate control room dose. A proposed Technical Specification limit of 15 gph ECCS leakage through mechanical joints into the Auxiliary Building is accounted for. Also modeled is an assumed 180 gph ECCS leakage through boundary valves to tanks outside the Auxiliary Building and vented to the atmosphere. This refined methodology is used to establish revised control room operator whole body and skin dose a(well as thyroid dose. The radiological consequences for the control room are c6mpared to the dose acceptance limits as established by 10 CFR 50, Appendix A General Design Criterion (GDC) Number 19 and NRC Standard Review Plan 6.4. Finally, an evaluation of other design basis accidents is performed to confirm that the MHA is the most limiting with respect to control room operator dose.
' 2.0 CONTROL BUILDING EMERGENCY ENVELOPE DESCRIPTION 2.1 Control Building Design The Control Building Emergency Envelope (CBEE) has a minimum free air volume of approximately 192,000 cubic feet, and includes the following elevations / rooms:
Elevation 355'-0" Main Control Room, Offices and Tech Support Center Elevation 338' - 6" Relay Room,4160V ES Switchgear Rooms ES Relay Cabinet Rooms Elevation 322'- 0" Battery Rooms, inverter Rooms,480V ES Switchgear Rooms, Remote Shutdown Cabinet Room The CBEE excludes the stairwell and Control Building Hallway (Patio). Also excluded are the first floor of the Control Building (Chemistry Lab and HP Access), and the fifth floor (Ventilation Equipment Room). All areas are shown on Figure 1.
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7 Topicci RIport C121 -
Rev.0 Page 6 of 29
' 2.2 Control Building Ventilation System During normal operating conditions, the TMI-1 Control Building Ventilation System (CBVS) serves the first through fourth floors of the Control Building. The fifth floor ventilation equipment rooms have a separate ventilation system.
During emergency conditions that reqIiire the Control Building Ventilation System to be placed on emergency recirculation, filtered, conditioned air is recirculsted within the CBEE only. A schematic of the Control Building Ventilation System during recirculation operations is shown in Figure 2. Emergency Recirculation operation is initiated by a high radiation condition in the Main Control Room, an Engineered Safeguards signal, or by an Air intake Tunnel device signal (smoke or combustible vapor detector). If the initiating signal is the Main Control Room radiation monitor, then the CBEE ventilation is recirculated throug% s HEPA filter
/ carbon adsorber filter unit (one of two redundant strings). Otherwise, normal supply and return fans are used to recirculate through a roughing filter bank. All fans, filter banks and cooling coils are redundant.
, in the emergency radiation recirculation mode of operation, the CBEE isolation dampers AH-D28 (first floor isolation), AH-D37 (exhaust isolation) and AH-D39 (supply isolation) are pneumatically controlled, and all fail in the closed position.
The recirculation damper, AH-D36, fails in the open position to ensure the recirculation path is open. Operating Procedure OP-1104-19 and Abnormal Procedure EP-1203-34, " Control Building Ventilation System," requhe the operator to modulate Damper AH-D-39 to gradual dial position 10.5 when entering the emergency mode of operation. With AH-D-39 gradual dial adjusted.
AH D-39, AH-D-37 and AH-D-36 will be in the intermediate position and AH-D-28 L will be closed.
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Topical Rtport #121 Rev. O Page 7 of 29 While on emergency recirculation, and with the gradual dial position 10.5, the Main Control Room is maintained at a positive pressure of at least 0.1 inches W.g. with respect to the Patio and stairwell areas, with or without single failure of isolation dampers AH-D28, AH-D37 and AH-D39. This was demonstrated by a Control Building Differential Pressure and Air Flow Measurements Test conducted in 1998. The test also demonstrated that no areas of the CBEE experienced negative pressures (Reference 3).
The CBVS is monitored via the Heating and Ventilation control panel in the Main Control Room. Each of the Normal Supply Fans (AH-E17A/B), Emergency Supply Fans (AH-E18A/B) and ReturriAir Fans (AH-E19A/B) has an indicator o'f its operational status. Supply Damper (AH-D39), Exhaust Damper (AH-D37) and Recirculation Damper (AH-D36) have a combined indication of Normal or Recirculation Mode indicated on the ESAS panel in the Main Control Room. The first floor isolation damper (AH-D28) is separately monitored for status (open/ closed).
The CBVS dampers are opposed blade dampers with pneumatic piston operators, and have been proven reliable over many years at TMI. A Failure Modes and Effects Analysis of the TMI CBVS was performed in 1984 by the Impell Corporation (Reference 13) to support the previous Control Room Habitability submittal to the NRC. This analysis listed the failure modes of system dampers, which include:
damper operatorlinkage jammed e
damper operator piston jammed e
damper operator failure e
air not vented from damper operator e
broken damper operator return spring.
e Corrective Maintenance Procedure 1410-AH-2, " Air Handling Damper Maintenance directs personnel on how to identify damage to a damper / operator, I
along with specific guidance on repair. For the hypothetical single active failures assumed in this report (AH-D28,37,39), any of the dampers are assumed to be W commoinTORiCR Habnabiiny coc
Topical Report #121 Rev. O Page 8 of 29 repairable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Reference 8) in accordance with the above referenced procedure, due to their location inside the Control Building and simplicity of design. Actual repairs would be expected to take much less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This procedure provides instruction for inspection, adjustment and repair of HVAC dampers (blades, frame, operator and adjustment) as described below:
AH-D28 (30* x 30" First Floor Isolation Damper)
Required Position:
closed Location:
Within CBEE,322 ft. elevation.
Maintenance issues: Both damper andqperator can be worked from floor level.
Operator mounted on outside of duct. Damper proven to always exfiltrate filtered air from envelope. Readily available materials can be used to ensure damper will close (Warehouse stocked parts or wire).
AH-D37 (60" x 60" CBEE Exhaust Damper)
Required Position:
throttled open with gradual dial control Location:
Control Building Patio, 365-ft, elevation of I&C Offices.
Maintenance issues: Damper Access door requires short stepladder from I&C J
Office Library. Operator mounted inside duct, so access inside duct would be necessary to repair damper. Readily available materials can be used to ensure damper will throttle open (Warehouse stocked parts, wire, compressed air or nitrogen bottle to locally operate operator).
AH-D39 (72" x 72" CBEE Supply Damper) l Required Position:
throttled open with gradual dial control Location:
Control Building Patio, 322-ft. elevation.
Maintenance issues: Damper Operator mounted on outside of duct, and is easily.
worked with a stepladder (stored on Patio elevation). An access door is available for manipulating blades (accessed by stepladder). Readily available materials can be used to ensure damper will throttle open (Warehouse stocked 1
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Topicci Rtport #121 Rev.O Page 9 of 29 parts, wire, compressed air or nitrogen bottle to locally operate damper).
Operator dose exposure for the repair action will be evaluated to ensure 10 CFR Part 20 exposure limits are not exceeded.
The CBVS ducting and active components are designed to Seismic Category i criteria except for an approximate 50 foot section of ducting upstream of the CBVS air intake isolation damper AH-D-39 which is designed to Seismic 11 anti-falldown criteria. This section of ductinens from the end of the Seismic I j
concrete air intake tunnel at Control Building Elevation 306' to Control Building Elevation 330' and contains branchoffs to non-essential system fans. This passive section of ducting is contained within the Auxiliary & Fuel Handling Building and the Control Building which are Seismic Category I structures that are tornado missile and flood protected. Thus, the essential portions of the TMI-1 CBVS are adequately protected from the effects of floods, hurricanes, tomadoes and intemally or extemally generated missiles as required by Standard Review Plan Section 9.4.1, " Control Room Area Ventilation Systems." The TMI-1 control room habitability evaluation assumes air intake flow via the Seismic l air intake tunnel and the Seismic ll portion of ducting upstream of intake damper AH-D-39 to preclude negative pressures in the CBEE. This assumption is consistent with the CBVS design basis as described in TMI-1 FSAR Section 9.8.1.4 and the previous TMl-1 control room habitability evaluation approved by NRC in the Supplemental Safety Evaluation Report dated August 14,1986.
3.0 ATMOSPHERIC DISPERSION MODELLING Primary parameters affecting the ovaluation of a radiological release from containment upon the control room and the capability to maintain its habitability depend upon the transport and settling processes involved in a direct release of radioactive material to the environment. In this regard atmospheric dispersion of the contaminant resulting from the combined effects of site meteorology and site Weommon\\TDR\\CR Hab:tabity.c!cc
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Topical Report #121 Rev.0 Page 10 of 29 geometry greatly influence the concentration of the contaminant at points of air intake into the control room.
ForTMI-1, the evaluation of a Maximum Hypothetical Accident assumes a diffuse containment leak at Technical Specification leak rate limits directly to the environment. The modeling used assesses the reduced concentration affects of the contaminant at various intake points to the control room by atmospheric dispersion only with consideration of building wake effects. Depoition of particulate material by gravitational settling, dry deposition, or the affects of precipitation scavenging are not include (as mechanisms for concentration
/ eduction consistent with industry practice for these types of assessments. The oispersion models presented herein consider the contaminant released at ground level (Height = 1 meter) and at various release heights on the containment surface for representative points of interest. No credit is considered for momentum or buoyancy effects upon the released contaminant to the environment.
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1 The dispersion modeling for TMI-1 considers two principal geometries for the entrance of the contaminant into the control room. These are shown on Figure 3.
The first is the normalintake point known as the " yard intake", which is located at approximately 91 meters from the containment surface at azimuth 230 from containment centerline reference and elevation 312 ft. True North is at 0
- The second is known as the " exhaust intake". As discussed in Section 2.2, the operator would modulate Damper AH-D-39 to gradual dial position 10.5 when entering the emergency mode of operation. With AH D-39 dial adjusted, AH-D-39, AH-D-37 and AH-D-36 will be in the intermediate position and AH-D-28 will be closed. In the event that AH-D-37 and AH-D-39 fail to open, the exhaust is evaluated as an entry point since it has been proven by test to draw outside air into the vent llation system under this failure mode. The ventilation exhaust is located at 1.5 meters from the containment surface at an azimuth of 176
- and elevation 392 ft. Grade elevation for the site is approximately 303 ft.
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1 Topical R: port #121 Rev. O Page 11 of 29 Unfiltered infeakage directly to the control room has been considered in the mathematical model evaluating dose. This leakage is associated with ingress and egress activities to the control recm and is considered a conservativa assumption since field testing demonstrated the existence of an adequate positive pressure in the Main Control Room. To assess this leakage, environmental transport was assumed as if the geometry of a virtualinleakage point was equivalent to that of the ventilation exhaust opening entrance point.
This assumption has been used for the purpose of determining relative concentration values associated with unfiltered inleakage. The assumption represents an attempt to model the entrance of unfiltered air from the environment since the actual physicaIstructure of the control room itself has no direct physical interface points which could introduce contaminant from the environment.
Relative concentrations (X/Q) have been determined for the above geometries by two methods. The yard intake X/Q 's have been determined utilizing the ARCON96 computer code. ARCON96 was developed by the Pacific Northwest National Laboratory to calculate relative concentrations in plumes from nuclear power plants at control room air intakes in the vicinity of the release point. The exhaust point X/Q's have been developed by utilizing data sources from wind tunnel experiments for simple building shapes with surface releases.
4 3.1 Yard intake X/Q The yard intake location is sufficiently downwind of the containment structure and within the boundaries of the building wake effect, making ARCON96 a suitable method for evaluating this geometry.
For this analysis a ground level diffuse release is assumed from the containment. The model assesses the building wake effect for the simple geometry of the containment structure alone', neglecting the wake effects of adjacent and adjoining structures. The diffuse leak approach utilized in ARCON96 is transformed into a continuous virtual point source upstream of the containment structure to apply the corrected Gaussian model W.commoniTDRiCR Habitability.duc
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1 Topical Rtport #121 Rev.O Page 12 of 29 described by the code. To establish the virtual point distance the user must estimate initial diffusion coefficients representing the 12 e distribution of the contaminant at the diffuse release point (containment center line). This has been determined from the projected cross sectional area of the containment structure, Sitt meteorological data for 1976,1978,1980,1981,1982, and 1996 have been utilized as input to the code for determining the 95* percentile X/O values.
The resulting X/Q values at thar'espective intervals are as follows (Reference 4):
Interval MQ 0-2 Hrs.
1.70E-04 sec / m^3 2-8 Hrs.
1.14E-04 sec / m^3 8 24 Hrs.
4.60E-05 sec / m^3 1-4 Days 3.65E-05 sec / m^3 4-30 Days 3.03E-05 sec / m^3 3.2 Ventilation Exhaust X/Q The ventilation exhaust point geometry is sufficiently close to the containment surface that it is questionable as to whether or not ARCON96. a corrected Gaussian model, is suitable for use.
As an altemative to ARCON96, the ventilation exhaust geometry has been evaluated using data sources from wind tunnel model experiments.
The use of experimental models provides a conservative estimate of X/Q values since the flow field near building surfaces differ significantly from the uniform flow fields utilized in Gaussian models.
The TMI-1 model evaluates selective release points on the containment surfaces exposed to the environment. A fractional release is determined for each release point representative of a release area associated with w....mmmunar.a u,,4.x
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1 Topical Report 0129 Rev.O Page 13 of 29 the point. Experimental surface concentration coefficients for buildings are used to determine X/Q. The coefficient, K, is equal to (X/Q)Au, where A is blockage area and u is wind speed. From a known blockage area associated with wind direction and speed, the sector (X/Q)u value is determined. Using hourly site meteorological data for 1976,1978,1980, 1981,1982, and 1996, hourly X/O 's are calculated and the applicable 95*,90*, 80*, and 60* percentile values are determined for use in the
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i dose assessment model.
The resulting X/O values at the respective intervals are as follows (References 5 & 6):
interval X/O 0-8 Hrs.
2.13E-03 sec / m^3 8-24 Hrs.
1.44E-03 sec / m^3 1-4 Days 9.42E-04 sec / m^3 4-30 Days 5.09E-04 sec / m^3 4.0 CONTROL ROOM DOSE ANALYSIS 4.1 STARDOSE Code and Model Control Room Habitability for an MHA postulates a large release of fission products to the reactor building. The release is not mechanistic and a means for it to occur is not postulated. The intent of the analysis is to demonstrate the safety of the plant to large releases of radioactivity. As a result, the acceptance criteria for this event are based on dose. For the CR, radiological consequences must be consistent with the requirements of 10CFR 50, Appendix A. GDC 19 limits.
The STARDOSE computer code is designed to calculate Control Room and offsite doses for nuclear power plant Design Basis Accidents. It is proprietary to Polestar Applied Technology. The code modeling is flexible in terms of the l
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Topic:1 Rtport #121 Rev.O Page 14 of 29 numbers and arrangements of control volumes and junctions. A finite differen'ce numerical scheme has been used to maximize flexibility. The STARDOSE approach to calculating offsite and Control Room doses is fully consistent with applicable Code of Federal Regulations and Regu!atory Guides (Reference 7).
The magnitude of the radioactive release is based on the fission product buildup _
from fuel burnup. One hundred percent of the noble gases and fifty percent of the iodine's in the core are assumed to be released into the Reactor Building.
Further, only fifty percent of the elemental iodine's released to the reactor building is assumed to plate out. As a result, as much as twenty-six percent of the core lodine activity and one hundred percent of the core noble gas activity is assumed to be released to the Reactor Building atmosphere. Decay of fission products are assumed to occur while they are confined to the Reactor Building (RB), but are not assumed to occur once they pass to the environment. Reactor Building pressure is assumed to be at design pressure and therefore Reactor Building leakage is assumed at its design leak rate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The RB leakrate is then assumed to be at one half its design value for the next 29 days. This is conservative, since peak pressure lasts only for the initial 100 seconds after a LOCA. Within the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, building pressure is less than 5 psig.
todine removal from the Reactor Building atmosphere is performed assuming two spray pumps and two cooling fans. The use of two sprays was found to give a higher dose consequence than one spray. This can be explained by considering that the elemental iodine spray removal coefficient is limited to a maximum of 10 hr ", and ESF leakage would begin earlier with 2 spray pumps in operation.
STARDOSE uses a two-compartment model to calculate spray iodine removalin the containment since the droplets do not reach certain areas. The volumetric flow rate between the sprayed and unsprayed areas is assumed to be 58,000 cfm based on the capacity of two cooling fans.
. Engineered Safety Features (ESF) Systems leakage through mechanical joints into the auxiliary building is assumed to be 30 gph, which is twice the proposed Technical Specification limit of 15 gph, as required by SRP 6.4. In STARDOSE, the iodine activity of the coolant leaking into the Auxiliary Building is based on Wcommon\\TDR\\CR Habitability doc
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Tomcal Rsport 0121 Rev.0 Page 15 of 29 50% of the core iodine activity. This is modeled in STARDOSE by assuming that the leakage goes directly to the environment at a rate of 0.067 c.'m (30 gph) To account for the evolution of 10% of the iodine from the liquid into the atmosphere of the auxiliary building, an artificial filter with a 90% efficiency for iodine has been placed in this junction. The effect is that 10% of the iodine in the leakage of reactor coolant from ESF systems is released to the environment, in accordance with the guidelines of SRP 6.4.
It is assumed that Auxiliary Building Ventilation System is not opew.y and charcoal filtration is not available. This i(conservative, as the char; Nd filters would reduce the dose consequences by a factor of 10. Also modeled is an assumed 180 gph ECCS leakage through boundary valves to tanks outside the Auxiliary Building and vented to the atmosphere.
The STARDOSE model consists of a single control volume representation of the three floors that comprise the Control Building Envelope. Model schematics of the three cases analyzed in Section 4.3 are shown on Figure 4.
4.2 Analysis Assumptions Conservatism's and assumptions utilized in the dose analysis are:
The isotopic Core inventory is the same as in the current FSAR for a power level of 110% of 2568 (Mwt).
The activity released from the core is based on ANSI /ANS 56.5-1979, PWR and BWR Containment Spray System Design Criteria". The fractional activities released into the reactor building are as follows:
One hundred percent of the noble gases Elementaliodine:- 47.75% of the totaliodine in the core. 50% of the elemental iodine released is assumed to immediately plate out.
Organic iodine :- 1% of the total iodine in the core.
Particulate iodino:- 1.25% of the total iodine in the core.
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Rev. O Page 16 of 29 Containment leakage rate to the atmosphere of 0.1 percent is assumed for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then 0.05 percent for the remaining 29 days. 0.1 percent leakage is based on peak Reactor Building pressure during a LOCA l
lodine removal from the Reactor Building atmosphere is performed assuming two containment spray trains and only two out of three RB emergency cooling units (fans) are in operation Containment spray iodine removal efficiencies for two spray trains are:
Elemental 10 hr, Particulate 6.06 hr, Organic 0.0144 hr" d
d Credit is taken for containment spray only untill the time of sump recirculation (29.1 minutes)
A volumetric flow rate of 58,000 cfm was used between the sprayed and unsprayed regions The breathing rate for the ' control room operator is 3.47 x E-4 cu m/sec Exposure is based on occupancy of the Control Room by an operator for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following release, followed by 60 % occupancy for the next 3 days, and 40 % occupancy for the remaining 26 days The Auxiliary Building Ventilation System is assumed not to be operating and charcoal filtration is unavailable ECCS leakage through mechanicaljoints into the Auxiliary Building was -
assumed to be 30 gph. This is twice the proposed Technical Specification limit of 15 gph ECCS leakage through boundary valves to tanks vented to the atmosphere is assumed to be 180 gph for the first hour, then reduced to 96 gph until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and then reduced to 30 geh for the remaining 29 days 10% of the iodine in the ECCS coolant goes airbome in the Auxiliary Building and is immediately released to the environment Unfiltered inleakage to the control room is assumed to be 10 cfm to account for ingress and egress per SRP 6.4 Hypothetical failures of dampers AH-D28,37 and 39 can be repaired in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W: common \\ TOR \\CR Habitabsty. doc
Topied R port #121 Rev.O Page 17 of 29 4.3 Cases Analyzed and Results The radiological accidents assume that ES actuation results in the closing of the intake dampers while the detectors in the control room cause automatic shutdown of normal ventilation in the control building envelope, followed by activation of the ventilation system in the emergency mode. Previously evaluated operator dose due to single active failures in the CB HVAC system were reviewed to determine the most limiting cases for re-analysis (Reference 9).
Internal pathways are not considered credible based on testing which shows a lack of negative pressure in the CBEE with respect to the Patio and stairwell areas. With the assumption of loss of Patio and Auxiliary & Fuel Handling Building ventilation systems, there is no driving force for transport of contaminated air to the CBEE. In addition, the operator will take action to repair and manipulate failed dampers, and to restore offsite power to return Auxiliary &
Fuel Handling Building ventilation to operation. This will result in extemal dose consequences discussed below.
4.3.1 Base Case As discussed in Section 2.2, in the event of a high radiation signal, the control building ventilation system emergency recirculation mode sequence takes place automatically. The isolation dampers AH-D-39, AH-D-37 and AH D 28 close; the recirculation damper AH-D-36 opens.
However, since the operator would modulate Damper AH-D-39 to gradual dial position 10.5, AH-D-39, AH-D-37 and AH-D-36 will be in the intermediate position and AH-D-28 will be closed. The STARDOSE model for analyzing this case is shown on Figure 4. The resulting 30 day CR
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doses are 23.9 Rem thyroid. 0.1 Rem whole body and 1.4 Rem skin (Reference 10) 4.3.2 Case 1 - Failure of AH-D-28 This case is similar to the Base Case, except that AH-D-28 is assumed to fail open. This is conservative, as it results in a larger amount of air entering the CBEE from the outside as shown in Figure 4. Since Abnormal Procedure 1203-34, " Control Building Ventilation System" wccmmcmTcR',ca nad,:ad,; ty h
e Topical Raport #121 Rev.0 Page 18 of 29 requires the operator to visually inspect AH-D-28,37 and 39 dampers for j
proper position and correct as necessary, it is conservatively assumed j
this condition exists for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Manual manipulation and/or repair of the damper will revert this case to the base case after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The resulting 30 day CR doses are 24.6 Rem thyroid 0.1 Rem whole body and 1.5 Rem skin (Reference 10).
4.3.3 Case 2 - Failure of AH-D-37 and AH-D-39 l
This case assumes that when the operator goes to gradual dial position i
10.5, Dampers AH-D-39 and AH-D-37 fail to open to their intermediate i
positions. This results in back-leakage rates through the dampe'rs as shown on Figure 4. Since Abnormal Procedure 1203-34 " Control Building Ventilation System" requires the operator to visually inspect AH-D-28,37 and 39 dampers for proper position and correct as necessary, it is conservatively assumed this condition exists for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Manual manipulation and/or repair of the dampers will revert this case to the base case after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The resulting 30 day CR doses are 29 Rem thyroid, 0.3 Rem whole body and 4.0 Rem skin (Reference 10).
4,4 Evaluation of Other Design Basis Accidents The MHA control-room habitability dose is compared with those of various accidents analyzed in Chapter 14 of the TMI-1 FSAR. Review of all the accidents in Chapter 14 of the FSAR indicate that several accidents do not cause radiation release. These are: uncompensated reactivity changes, startup accident, rod withdrawal accident, moderator dilution accident, loss of coolant flow, and stuck-out/-in control rod. Consequently, these accidents are not considered in this evaluation. The accidents considered in this evaluation are:
Maximum Hypothetical Accident (MHA)
Fuel Handling Accident (FHA) in reactor bu!l ding and fuel handling building Steam Generator Tube Failure (SGTF)
Loss of Electric Power (Loss of Load transient and Station Blackout)
Main Steam Line Break (MSLB)
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9 Topics! R: port #121 Rev.O Page 19 of 29 Rod Ejection Accident (REA)
Waste Gas Decay Tank Rupture (WGDTR)
Spent Fuel Cask Drop Accident (SFCDA)
The description of radiation releases during the above accidents are given in Table 1 and source term data comparisons are given in Table 2. Except for the MHA and REA, all the accident releases are short term, i.e., the release is terminated within approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the accident. The source term comparison is performed based on the 24-hour release of I-131, to ensure that the comparison is consistent and cons'ervative. This is because the long-term X/Q-value reduction fectors (wind direction factors, wind speed factors, and occupancy factors) are applied to the MHA dose analysis, which can not be applied to other accidents.
The MSLB source term was based on the reactor coolant activity limit of 0.35
)
C/ gram, reflecting the current Technical Specifications. As shown in Table 2, the MSLB source term (4741 Ci) is the only one to challenge the MHA (8728 Ci),
and source terms of all other accidents are orders of magnitude lower than MHA.
In Figure 5, the radiation release points are identified for each accident against the control room air intake, which is located 91 meters away from the reactor containment in the southwest direction. The source-to-receptor configuration for each accident in Figure 5 is assumed to be under the influence of building wake effect of the reactor building and surrounding plant structures and, therefore, a diffused source. Internal paths of the activity release to the control room were not considered in this evaluation as discussed in Section 4.3.
The comparison of the X/Q-values was performed by using the guidelines given in Reference 12. Integrating comparisons of source term and X/Q-values, the control room doses for each accident are normalized to those of the MHA in Table 2. As can be seen from this Table, the MHA habitability dose bounds all other accidents and, therefore, MHA is the design basis accident for the TMI-1 Control Room Habitability analysis (Reference 11). Future plant changes or re-W.commonsTDR\\CR,Habitabiiity.aoc
ie Topical Report #121 Rev.0 Page 20 cf 29 analysis will be evaluated for potential impact on the calculated bounding CR habitability doses.
5.0
SUMMARY
AND CONCLUSIONS A revised methodology for calculating dose to the TMl-1 control room operators during an MHA has been developed. The revised methodology significantly enhances the control room dose assessmeqt by utilizing updated radiological methodologies and including thyroid dose? The analysis shows that the worst radiological consequences for the control room occur with an assumed failure of dampers AHD-37/39, and result in a dose of 4.0 Rem skin,0.3 Rem whole body and 29 Rem thyroid. These values are below the 10 CFR Appendix A, GDC 19 limits and the NRC Standard Review Plan 6.4 guideline of 30 Rem for the Control Room.
9 Further, an evaluation was performed to assure that the MHA was the most severe design basis accident for CR operator dose.
6.0 REFERENCES
1 NRC letter, J.F.Stolz to H.D. Hukill, "TMI-1 Control Room Habitability Review" dated August 14,1986 2 NRC letter, B.C. Buckley to J.W. Langenbach. "Three Mile Island Nuclear Station, Unit 1 Control Room Habitability Review" dated September 24',1997 3 TMl FTP-826.01," Control Room Habitability Envelope Pressure Differential Test" conducted on 1/31/98 to 2/01/98.
4 GPUN Calculation C 1101-826-E540-020 Rev 0," Atmospheric Transport X/Q's Using ARCON96 Code - Control Room Habitability", March 1998.
5 GPUN Calculation C-1101-826-E540-022, Rev 0," Estimate of X/Q values for TMI-1 Control Building Ventilation System Exhaust", March,1998.
6 GPUN Calculation C-1101-826-E540-025, Rev 0, "X/Q @ TMI Ventilation System Exhaust". March,1998.
7 STARDOSE User's Manual, PSAT C109.06, Rev 0, March 1998 W.
mmen\\ TOR'.CR H:m:::::::r/.:::::
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?,4 Topical Rcport #121 Rev.O Page 21 of 29 8 GPUN Memo No. E260-98-014, " Damper Repair Guidance for Control Room Habitability Submittal" March 24,1998 9 PLG-0433. "TMI-1 Control Room Habitability Study: Analysis of Radiological and Chemical Accidents for abnormal flow paths to the Control Room", August 28,1985.
10 GPUN Calculation C-1101-826-E000-026, Rev 0,"TMI-1 Control Room Habitability During the MHA", March,1998.
11 GPUN Calculation C-1101-202-E620-361, Rev 0," Design Basis Accident for Control Room Habitability", March,1998.
12 K.G. Murphy and K.M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19",13 th AEC Air Cleaning Conference, August 1974.
13 IMPELL Corporation Project # 0370-101," Failure Modes and Effects Analysis of the Three Mile Island Control Building Ventilation System" Rev 1,11/08/84.
l 4
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D Topical R2 port #121 Rev.O Page 22 of 29 Table 1 - Description of the Accident Release and Source-Receptor Configuration Accidents Radiation Release Description Release points Distance (meters)
(1)
(2)
MHA Source term based on R.G.1.4 with 0.1% vo!) day Containment &
91 containment leak rate and iodine removal through RB spray Aux. Bldg. (ES system. Two compartment model is used. ES leakage to both leakage)
BWST and Auxiliary Building are included.
MSLB Based on 0.35
- 10( 6) Ci/g of RCS dose equivalent iodine Intermeatate i12 with post accident iodine spike of 500. The average primary.
Building.
to-secondary leak rate of 6.45 gpm (hot RC)is assumed for 25.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />. No OTSG iodine partition factor is considered.
OTSG Tube failure RCS with 1% failed fuel leaks to OTSG for 34 min. at a rate Condenser offgass i12 of 435 gpm. Activity released into the secondary side is assumed to be released via the, condenser offgas. No direct release to atmosphere is assumed if the OTSG is to be isolated.
FHA (FHB) 56 fuel rods damaged and 100% of the gap activities released Auxiliary to Spent Fuel Pool at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the reactor shutdown.
All of the rods (208)in one fuel assembly are damaged at 72 Containment 91 hours0.00105 days <br />0.0253 hours <br />1.50463e-4 weeks <br />3.46255e-5 months <br /> after reactor shutdown. The isotopic breakdown of the release is in accordance with R.G.1.25.
Rod Ejection 17.5% of gap activity in the reactor core and entire RCS Containment &
91 activity (1% failed fuel) are released to RB. In addition,5 Condenser (5 gal.
gallons of RCS is leaked to the secondary side.
RCS)
WGDTR 1% of iodine activity associated with one reactor coolant Auxiliary 38 volume is released to the Auxiliary Building.
Building Rollup Door Loss of Load Release of secondary steam (205000 lbs.) for 55 min. based MSRV & ADV 112 on I gpm primary to secondary leak and 1% failed fuel.
Station Blackout Release of secondary steam (950000 lbs.) through MSRV &
MSRV & ADV i12 ADV and 540000 lbs through condenser off gas. Activity is (direct release) based on I gpm primary to secondary leak and 1% failed and fuel in RCS.
Condenser SFCDA 100% gap activity from l0 damaged assemblies is released Rail car at south-83 during the fuel transfer operation onto a rail car.120 days of east corner of decay time is allowed.
FHB Notes:
1.
All the accidents with no radiological hazard in Chapter 14 of the FSAR are not included 2.
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Topical Raport #121 Rev O Page 23 of 29 TABLE 2 - Summary of Control Room Habitability Dose Comparison Accidents Total Release Source-Receptor (x/Q)/(x/Q)w.
(DOSE)/(DOSE)..
(Cunes)
Distance (m)
(1)
(2)
(3)
(4)
MHA 8728 91 1,000 1.00E+00 MSLB 4741 112 0 915 4 97E-01 OTSG Tube Rupture 1.94 112 0.915 2.03E-04 FHA in FHB 2.76 83 1.046 3 31E-04 FHA in RB 155 91 1.000 1.78E-02 x
Rod Ejection Accident 707 91 1.000 8.10E-02 WGDTR 11 38 1.807 2.28E-03 Loss of Load O.0388 112 0 915 4 07E-06 Station Blackout 0.181 112 0 915 1.90E-05 SFCDA 4 33 83 1.046 5.19E-04 Notes:
(1) = total 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release of I-131 (2) = See Figure 5 for the source-to-receptor configuration (3) = ratio of X/Q value of j-th accident to that of MHA (4) = ratio of thyroid dose of J-th accident to that of MHA
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Rev.O Page 27 of 29 FIGURE 4 - STARDOSE MODELS MODEL FOR TE1 CONTROL ROOM DOSE ANALYSIS. BASE CASE (NQ FALURES) 10 CFM
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