ML19256E820
| ML19256E820 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/12/1979 |
| From: | Metropolitan Edison Co |
| To: | |
| Shared Package | |
| ML19256E818 | List: |
| References | |
| FOIA-81-48 NUDOCS 7911150302 | |
| Download: ML19256E820 (100) | |
Text
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THREE MILE ISLAND UNIT 2 REACTOR BUILDING PURGE PROGRAM SAFETY ANALYSIS AND ENVIRONMENTAL REPORT NOVDiBER 12, 1979 1334 255 s
7 gill 50 3Cz E' :
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THREE MILE ISLAND UNIT 2 REACTOR BUILDING PURGE PROGRAM SAFETY ANALYSC AND ENVIRONMENTAL REPORT TABLE OF CONTENTS PAGE 1.
Introduction 1
1.1 Need for Reactor Building Atmosphere Cleanup 1
1.2 Purpose of Report 1
1.3 Organization of Report 2
1.4 Conclusions 2
2.
Reactor Building Airborne Activity 5
2.1 Method 5
2.2 Source Term Evaluation 6
2.3 Reactor Building Source Term Results 6
2.4 Reactor Building Source Term Conclusions 6
3.
Purge Program Analysis 10 3.1 Purge Program Summary 10 3.2 System Design Basis 11 3.3 System Design 11 3.3.1 Component Description 12 3.3.2 Instruments, Controls, Alarms and Protective Devices 14 3.3.3 Monitoring 15 3.4 Design Evaluation 15 3.5 Tests and Inspections 15 3.6 Materials 16 3.7 Operation of Purge System 16 3.8 Accident Analysis 16 3.9 Failure Modes and Consequences 16 3.10 System Modifications for Purge 18 4
Safety Limits For Radioactive Gaseous Releases 23 4.1 Radiation Exposure in Perspective 23 4.1.1 Sources of Radiation in the Environment 23 4.1.2 The Person-Rem Concept 23 4.1.3 Effects of Radiation Exposure 23 4.2 Environmental Technical Specifications 24 4.3 10CFR20 - Standards for Protection Against Radiation 4.4 10CFR50 - Appendix I - Nuxerical Guides for Deaign Objectives 25 and Limiting Conditions for Operatien to Meet the Criteria -
25 ALARA 4.5 Limiting Conditions for Operation of Purge Program 26 5.
Environmental Effects of Purge Operation 28 5.1 Purge Compliance with Tech Specs 28 5.1.1 Method 28 5.1.2 Evaluation with Tech Specs 30 1334 256 i
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PAGE 5.1.3 Results of Purge Within Tech. Spee. Limits 35 5.1.4 Conclusion for Purge Release Within Tech.
38 Spec. Limits 5.2 Of f-Site Dose Determination for Purge 38 5.2.1 Method 38 5.2.2 Evaluation of Of f-Site Doses Due to Purge 43 5.2.3 Dose Effect Results 45 5.2.4 Dose Effect Conclusions 47 6.
Operational Effects of Purge Operation 60 6.1 Method 60 6.2 Filter Dose Evaluation 61 6.3 Results 62 6.4 Conclusions 62 7.
Environmental Effects of Purge Accident 64 7.1 Description of Accident 64 7.2 Accident Dispersion Model 64 7.3 Environmental Dose Consequences from Purge Accident 6S 8.
Alternatives to Reactor Building Purge Program 69 8.1 No Atmosphere Clean Up 69 8.2 Design Basis for Alternate Atmosphere Treatment and 69 Storage Systems 8.3 Charcoal Adsorption and Storage System 71 8.3.1
System Description
71 8.3.2 Design Alternates Considered 72 8.3.3 Cost and Schedule Est imate 72 8.3.4 System Evaluation 73 8.3.5 Charcoal Adsorption System Conclusion 73 8.4 Gas Cc.pression and Storage System 76 8.4.1
System Description
76 8.4.2 Design Alternate Considered 76 8.4.3 Cost and Schedule Estimate 77 8.4.4 System Evaluation 78 8.4.5 Gas Compression System conclusions 82 8.5 Cryogenic Processing and Storage System 82 8.5.1
System Description
82 8.5.2 Design. Alternate Considered 85 8.5.3 Cost and Schedule Estimate 85 8.5.4 System Evaluation 86 8.5.5 Cryogenic System Conclusions 87 8.6 Environmental Effects of Alternate 92 8.6.1 Normal Operation 92 8.6.1.1 Charcoal Adsorption and Storage System 93 8.6.1.2 Gas Compression and Storage System 93 8.6.1.3 Cryogenic Processing and Storage System 93 8.6.2 Accident Conditions 94 8.6.2.1 Charcoal Adsorption and Storage System 94 8.6.2.2 Cas Compression and Storage System 94 8.6.2.3 Cryogenic Processing and Storage System 95 1334 2c7 11
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LIST OF TABLES TABLE NO.
TITLE PAGE 1-1 Dose / Exposure Comparison for Reactor Building Atmosphere Clean-Up Alternatives 4
2-1 Gas / Atmosphere Sampling of TMI-2 Using HPR227 8
2-2 Best Estimate of RB Airborne Activity November 1, 1979 9
3-1 Hydrogen Purge to Atmosphere Design Performance and Equipment Data 19 5.1.2-1 Maximum Concentration of Kr-85 vs. Purge Rate 33 5.1.2-2 Time to Reach Each Limiting Concentration 34 5.1.3-1 Limiting Purge Rate 36 5.1.3-2 Limiting Purge Concentration 37 5.2-1 TMI Site Weather Instruments 48 5.2-2 Input Data for Dispersion Modelling 49 5.2-3 Approximate Terrain Elevations within 12 Miles of TMI Site 50 5.2-4 Assumptions for Purge Dose Calculations 51 5.2-5 Results of Purge Dose Calculations 52 5.2-6 Comparison of Dose for Purge sconarios 53 6.3-1 Hydrogen Purge Filter Dose Rate 63 7-1 Assumed Distance to Site Boundary in Each Direction 67 1334 258 iii
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LIST OF FIGURES FIGURE NO.
TITLE PAGE 3-1 Purge System Sketch 21 3-2 Flow Diagram Reactor Bld. - Ventilation and t
Purge 22 5.2-1 Population Di :ribution 0 to 10 Miles - 1980 54 5.2-2 Population Distribution 0 to 50 Miles - 1980 55 5.2-3 Generalized Flow Diagram for PURTST Program 56 5.2-4 Purge Dose Build up vs. Time 57 5.2-5 Selected Characteristics vs. Purge Ti=e Run No. 21 58 5.2-6 Selected Chara:: eristics vs. Purge '
e Run No. 26 59 8.3-1 Containment Atmosphere Charcoal Adsorption Conceptual System Design 74 8.3-2 Conceptual Layout - Charcoal Storage Arrangement 75 8.4-1 Containment Atmosphere Gas Compression Conceptual
System Design
79 8.4-2 Conceptual Layout High Activity Storage Arrangement 80 8.4-3 Conceptual :.ayout Low Activity Sterage Arrangement 81 8.5-1 Containment Atmosphere Cryogenic Cleanup System 88 8.5-2 Containment Atmosphere Cleanup System Cryogenic Treatmen_ Train 89 8.5-3 Cryogenic Treatment System Krypton Storage Secondary Contains Conceptual Arrangement 90 8.5-4 Cryogenic Treatment System Building and Equipment Layout 91 259 iv
.c THREE MILE ISLAND UNIT 2 REACTOR BUILDING PURGE PROGRAM SAFETY ANALYSIS AND ENVIRONMENTAL REPORT
1.0 INTRODUCTION
1.1 Need For Reactor Building Atmosphere Cleanup The unknown Three Mile Island Unit 2 core configuration poses a small but incalculable risk. Although much analysis has been com-pleted that tends to bound the limits of uncertainty with regard to the core configuration, this uncertainty caa hest be dealt with by timely entry into the reactor building and ultimate removal of the auclear fuel from the reactor pressure vessel.
In order to allow entry into the reactor building without signifi-cantly complicating the entry program and restricting the effective-ness of operations toward ultimate fuel removal, the reactor building atmosphere must first be cleansed of radioactive materials. Leaving the airborne materials in the atmosphere while other steps toward fuel removal proceed represents substantial risk of ultimate uncon-trolled release of these materials to the environment and unaccept-able increase in operations personnel exposure.
Airborne radioac,tivity within the reactor building has reduced con-siderably since the accident due to decay of the short-lived radio-active fission products such as Xenon and Iodine. The principal remaining airborne fission product is Krypton-85 which has a 10.76 year half-life. Due to this long half-life, additional delays in cleaning up the reactor building atmosphere will not materially reduce its radioactive concentration.
Several alternatives have been studied for removing the approximately 45,000 Curies of Kr-85 estimated to exist in the building. These alternates include charcoal adsorption and long term storage, gas compression and long term storage, cryogenic processing and long term storage, and atmospheric dis 7ersion of Krypton-85 by controlled purging of the reactor buila ng atmosphere. The optimum choice f rom an environmental impact standpoint when potential accidents are con-sidered is atmospheric dispersion through controlled purging of the reactor building atmospbere.
1.2 Purpose of this Report The purpose of this report is to document that atmospheric disper-sion through controlled purging of the reactor building atmosphere can be accomplished within all applicable safety limits and radia-tion protection standards and that per ge represents the optimum solution for reactor building atmosr.nere cleanup, considering the health and safety of the populatior. around the Three Mile Island Unit 2 plant. This report presents a description of the proposed program for controlled purging, the safety analysis for this purge program, the environmental impact of this proposea purge program, and the results of studies of the less desirable alternatives for reactor building atmosphere clean up.
260
e 1.3 Organization of Report This report is organized in the format of a combined safety analysis and environmental report. Following this introduction, the assess-ment of reactor building airborne activity is presented and then the analysis of the purge program. The purge program analysis includes a summary of the purge method, the design basis for the hydrogen control system and its modifications, the design evaluation, opera-ting description, and accident analysis for the system.
The Safety Limits for radioactive gaseous releases are discussed including some perspective on radiation exposure, and limiting conditions for the purge program.
The ef fects of purge operation on the environment and on operational exposure are presented, including the environmental effects of a postulated purge accident. Finally, the results of the studies on the alternatives to controlled purge are presented including the environmental ef fects of each alternate.
1.4 Conclusion The studies concerning disposal of the Krypton-85 f rom the contain-ment vessel result in the following conclusions:
1.
There are only four potentially feasible methods for disposing of the Krypton: purging to the atmosphere, charcoal adsorption and storage, storage as a compressed gas, and cryogenic separ-ation and storage. Of these four methods, three, charcoal adsorption, gas storage and cryogenic separation require a long schedule to implement, are of high complexity, but theoretically can provide a zero or near zero offsite dose.
The purge method can be implemented very quickly, is simple, but does yield a small finite offsite dose to the general population.
2.
Tha examination of radiation doses to the general population in' he event of accidents, for each alternative, shows just the reverse of the normal dose comparison, i.e., purge has an extremely small general population accident dose, whereas the other three have relatively large general population ac-cident doses. Of the alternatives studied, only purge to the atmosphere provides a permanent solution to the Krypton-85 problem. The oth_r three options require treatment and storage in systems which have the potential for accidental release of Krypton-85 during processing and especially during the long storage time required.
3.
The long schedule required for the storage options is con-sidered a significant safety disadvantage. There is no assurance that containment integrity can be maintained for the 2-3 years necessary to implement storage. As shown in Section 8.6.1, if the reactor building air cooling capability is lost, the reactor building pressure could rise to 1 to 2 psig. The uncontrolled leakage of Krypton 85, if the equivalent of a 1/2 inch diameter nole is present in the 2
1334 261
containment boundary, could result in an of f-site beta dose in the range of 15 to 80 mrem in a single day or 60 to 270 mrem if the leakage occurred over a 30 day period.
4 Kryptan-85 disposal is an essential prerequisite to perform-ing a is; 0..I work within the containment leading toward cleanur af the containment structure.
The delay in initia-ting : ch cleanup, which would be required by the storage options, can, in itself, be a significant safety hazard and cause large increases in radia-ion dose to the work force.
This increased dose would arise because of additional com-plexity in decontamination, but at ;his time cannot be cuantified.
5.
Purge of the Krypton-85 to the atmot phere can be performed under well-controlled conditions, and such purging can meet all technical specifications and Regulatory guidance. The est mated dose to the general population, as well as dose to the omsite staff, is extremely low or insignificant.
6.
Table 1-1 summarizes the radiation effects of each of the alternatives for reactor building atmosphere clean-up.
The expected dose / exposure shown in this table for each alternate uses expected meteorology based on historical data.
The system upset dose /expc sure analysis uses conser-vative, 5 probable extreme meteorology as specified in Regulatory Guides for accident analysis.
The coincidence of the postulated accident conditions and the extreme meteorology are highly unlikely as stated in Section 7.1.
It is recommended that cleanup of the containment atmosphere pro-ceed through purging as the safest, and mret effective permanent solution to the Krypton-85 problem.
3
TAlli.E l-1 t
Dose / Exposure Comparison For Reactor fluilding Atmospliere Clean-Up Alt ernat ives EXPECTED DOSE / EXPOSURE SYSTEM IIPSET/ACCII)ENT DOSE /EXPOSilRE OFF-SITE OFF-SITE ON-SITE OFF-SITE OFF-SITE SKIN DOSE Wil01.E WilOI.E SKIN IX)SE OFF-SITE Wil01.E (MREH)*
BODY DOSE
- 150DY IX)SE (MREM)**
Wilot.E IM)DY*
- IM)DY DOSE * **
BETA GAHMA PERSON-REM PERSON-REH llETA GAMMA DOSE (MREH)
_ PERSON-REM PURGE 5
0.1 0.75
<1 61 0.9 0.73 0.07 CilARCOAL 0
0 0
23.
104 1.4 1.24 0.13 CAS COMPRESSION 0
O O
58.
1730 24 20.7 2.1 s-CRYOGENIC
<1
<0.1 0
570.
4090 56 49.0 5.2
- CAI.CULATED IN ACCORDANCE WITil RG 1.109 USING THI-2 SITE IllSTORICAl. METEOR 01.OGICAI. DATA
- CALCUl.ATED IN ACCORDANCE WITil RG 1.109 and 5% PRollAlli.E POPill.ATION DOSE GIVEi1 TilAT Tile ACCIDENT OCCllRS i
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3 U
e 2.0 REACTOR BUILDING AIRBORNE ACTIVITY Three types of samples are being collected periodically from the reactor building atmosphere to determine the nature of airborne contaminants present. The samples are for noble gas. particulate, and radio-iodine activity.
2.1 Method In order to determine the activity content of the atmosphere in the reactor building, air samples utilizing the installed HPR 227 sample pumps and cabinet have been taken.
The sample procedure allows use of a normal sample path or an al:ernate sample path which connects penetration R-562B to the suction of the HPR 227 sample pump via valves AH-V147 and AH-V148.
Another sample path will be available in November through penetration R-626 (at elevation 358'), when the inner flange of that penetration is cut to allow camera and radia-tion monitor placement inside the reactor building.
To determine noble gas activity, a reactor building air sasple is collected in a 6 cc glass bulb and analyzed by gamma spectroscopy.
Isotopic identification is made on the basis of the discrete energy levels at which gamma rays are absorbed in a GeLi detector. The spectrum containing the various gamma peaks is then screened and compared against a library of known peaks vs. isotope to make final ident ificat ion.
The intensity of each peak at its discrete energy level is a function of the concentration of the respective radio nuclide. The process is commonly referred to as a " gamma scan."
To determine the particulate activity concentration in the atmos-phere, a sample of the reactor building air is pumped through a 100 millipore filter. Particulate activity is removed from the air by the filter and the filter is then analyzed using gamma spectroscopy as described above.
To determine the concentrations of the different types of iodine in the atmosphere, a sample of the reactor building air is pumped thsough a series of filters es shown below. Separation of the different forms of iodine is accomplished based on the relative affinity of each iodine species for a specific filter media.
Each filter is then analyzed using gamma spectroscopy as described above.
7 Flow U Millipore CdI2 ->. Iodaphenol -+ '
Ag M Carbon Carbon
! Filter
.i l
- Zeolite ;
J L.. -
4_.
Removes Removes Removes Removes Removes Particulate I2 HOI CH I all Iodine 3
Iodine 5
1334 264
2.2 Source Term Evaluation The sampling and analysis techniques desccibed in Section 2.1 pro-vide for the determination of noble gas activity, particulate ac-tivity and iodine in the reactor building atmosphere. The results of these samples to date are included in Table 2-1.
It should be noted that the sample results, especially for the samples taken prior to June 21, vary widely. Whereas the earlier samples were drawn under less controlled conditio.s, the current procedure requires extensive documentation to ensure accurate sample times are used and proper volumes are draen.
In addition, retained with each sample result is the documented condition of the sample lineup and reactor building ventilation system while the sample was drawn.
From Table 2-1, it can be seen that the dominant isotope inside the reactor building at this time is Kr-85 at e*0.78juci/ml. Particulate levels, primarily Cs-137, are on the order of 1 x 10-Of4Ci/ml.
The radio iodine levels inside containment are rapidly dropping due to decay. Latest results indicate Iodine to be below minimum detectable activity (MDA) levels of 10-9,0Ci/ml.
In order to determine a best estimate of the airborne radioactivity inventory in the reactor building, the results of all gas samples were reviewed and correlated.
The results of this review are given in Table 2-2.
When these activity levels for Kr-85, Cs-137, and Iodine 131 are evaluated against technical specification limits for allowable instantaneous and quarterly average allowable gaseous effluent release rates, the Kr-85 concentration is shown to control allowable release rates.
2.3 Reactor Building Source Term Results Results of reactor building air senples taken to date are shown in Table 2-1.
The best estimate of reactor building airborne radio-isotope activity projected to Nov. 1, 1979 is shown in Table 2-2.
This estimate includes consideration of all samples taken to date and projects reduced concentrations to Nov. 1, 1979 of tha shorter lived radio-isotopes.
2.4 Reactor Building Source Term Conclusions The airborne activity sampling process is suf ficiently defined and recent results are sufficiently consistent that the results given in Table 2-2 represent accurate values for airborne activity levels of the isotopes of concern namely noble gas Kr-85, particulate Cs-137 and Iodine-131.
The results are suf ficiently valid to serve as a basis for evaluating alternat a reactor building atmosphere clean-up options.
Also, from these activitf levels, it is clear that efforts should be taken to clean-up the reactor building atmosphere in order to 6
1334 265
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o reduce the total exposure during manned entry of the containment resulting from the Kr-85 present.
Finally, from the low levels of g trticulate (Cs-137) and Iodine, it is concluded that reactor building recirculation using the reactor building purification filtration system is not necessary to achieve further reduction in Iodine and particulate levels. Air discharged from the reactor building will be filtered through the in-line particulate and charcoal filters of the hydrogen control system during purge.
1334 266 7
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1134 267
TABLE 2-2 Best Estimate of R3 Airborne Activity November 1, 1979 Concentration Total Inventory Nuclide (l'Ci/ c e )
(Ci)
Kr-85 0.78 4.4 x 10+ '
Xe-131 m
<2 x 10-5
<1,14 Xe-133
<1 x 10-5
<5.7 x 10-1 1-131
<1 x 10-9
<5.7 x 10-5 Cs-134
<1 x 10-5
<5.7 x 10-1 Cs-135
<1 x 10-5
<5.7 x 10-1 Cs-136
<1 x 10-5
<5.7 x 10-1 Cs-137
<1 x 10-3
<3.7 x 10-1 1334 268 9
3.0 PURGE PROGRAM ANALYSIS 3.1 Purge Program Su= mary The reactor building purge program will purge the reactor building atmosphere using the hydrogen control subsystem of the reactor building ventilation and purge system. The purged atmosphere con-taining rauioactive gases will be released from the plant vent stack (160 ft. above grade level) at times when wind and other meteorolog-ical conditions are most favorable for atmospheric dispersion.
The hydrogen control subsys;em is designed for use as a back-up for the hydrogen recombiner. The reactor building atmosphere is drawn through a filter train by the hydrogen control exhaust fan before being discharged to the plant vent stack.
( A modification is being made to reroute flow f rom the inlet of the supplementary vent fil-ters to the plant vent stack to obtain an elevated release). The filter train consists of a prefilter, REPA filter, an activated carbon filter and another HEPA filter. The purge flow rate is con-trolled by a throttle valve, AH-V36.
(Valve AH-V36 is being modified to provide fine motion flow control). The air may be discharged at a rate of up to 1000 cfm.
( A modification is being made to increase the capacity f rom 150 cfm :o 1000 cfm.) The flow rate, temperature and radiation level are monitored during discharge.
The replacement air to the reactor building will be supplied through valves AH-V7 and AH-V33 such that the R3 atmosphere remains at slightly negative or atmopheric pressure throughout the purge process. Figure 3-1 gives a schematic of the purge system pumps, pipes and valve configuration.
The maximum purge discharge flow rate will be determined using the Technical Specif tmation instantaneous release rate for gross gaseous activity given by paragraph 4.2 and the latest assessment of Kr-85 level from the revised RB air sampling program. This release rate is expected to be in the range of 50 to 100 cfm initially. As the Kr-85 content drops within the reactor building, due to purging, the maximum allowable purge rate will increase until the 1000 cfm limit of system capability is reached. The actual purge rate during any time interval will be based on actual wind ccnditions such that for unf avorable meteorology, the maximum allowable release rate will be reduced to minimize dose accumulation at off-site locations. At the start of each purge period, wind data will be recorded and predicted incremental dose at the boundary vill be calculated and compared against an administrative limit of 0.1 mrem / hour beta skin dose.
If the dose rate is calculated to exceed this limit, the release rate will be reduced to stay within the limit. If the allowable release rate drops below 20 cfm, no release will be allowed during the period until meteorological conditions improve.
Acrumulated of f-site doses will be calculated thrcughout the purge process using actual meteorological and release data to assure that 10CFRSO Appendix I limits are not violated and that projections of' future purging will not cause the limit to be exceeded.
10 1334 269
Plant vent stack monitoring will provide continuous feedback and a complete record of actual stack releases to track total curie activity released and to compare actual releases with expected releases based on reactor building activity.
In addition, reactor building activity will continue to be monitored to determine the Kr-85 activity fo. establishing future purge re-lease rates in conformance with Technical Specification limits and to confirm that MFC limits or lower have been met at the end of the purge operation.
3.2 System Design Basis The purge program uses the installed hydrogen control (atmospheric purge) system with modifications to increase the f an capacity to 1000 cfm, to provide variable purge flow control with interlocks for rapid isolation on equipment failure or high radioactive levels at the fan discharge.
The atmospheric purge operation was originally designed to keep the reactor building hy.rogen concentration from reaching the lower flammability limit following a LOCA if the hydrogen recombiner system is not available. The atmospheric purge system is designed to 30 psig and 150* F and seismic Class I conditions. The system meets the requir'ements of ANSI B31.0 Code for Power Piping, Class 2, and ASME Boiler and Pressure Vessel Code,Section III, Class B.
3.3
System Design
In order to ef ficiently purge the reactor building of radioactive gases, the flow capacity of this system has been increased to 1000 cfm to match the filter train capacity. Variable system flow capability is added to control the atmospheric purge rate as a f unction of activity in the reactor building and meteorological conditions. Since the airborne activity within the reactor building will exponentialif decrease with purging, it is desirable to increase the maximum flow rate to 1000 cfm during the later stages of purge operation.
The reactor building atmosphere is drawn through a filter train by the hydrogen control exhaust fan before being discharged to the Station vent.
The filter train consists of a prefilter, HEPA filter, an activated carbon filter and another HEPA filter.
In the original design the hydrogen purge flow rate was controlled by throttle valve AH-V25.
The valve must be partially open for fan operation. The fan discharge valve, AH-V36 opens with fan start. For operation with the increased fan capacity, purge flow rate will be controlled by remote control of valve AH-V36 in place of AH-V25 for better control of flow rate over the full range of flow capacity. The air may be discharged at a rate of up to 1000 cfm.
Replacement air is supplied through AH-V7.
The flow rate, temperature, and radiation 1334 270 11
level will be monitored during discharge.
The atmospheric purge system is shown diagrammatically in Figure 3-2.
3.3.1 Component Description Design perfo-nance and equipment data are provided in Table 3-1.
Reactor Building Hydrogen Control Exhaust Unit The reactor building hydrogen control exhaust unit is located in the auxiliary building at an elevation of 328'.
The unit is comprised of a bank of filters housed in a steel cabinet and an exhaust fan connected to the housing.
The filter bank (Table 3-1) consists of the following filters listed as they occur in the flow path:
a.
Pre-Filter AH-F-36 b.
REPA Filter AH-F-33 c.
Activated Carbon Filter AH-F-34 d.
HEPA Filter AH-F-35 Access doors are located on top of the housing for easy maintenance.
There is a differential pressure switch connected across the filter bank which will initiate an alarm on high differential pressure.
Each filter is provided with a dif ferential pressure indicator.
Reactor Building Hydrogen Control Pre-Filter AH-F-36 The pre-filter is a replaceable bag filter designed for rough particle removal. It has a local differential pressure indicator.
Reactor Building Hydrogen Furce Absolute (HEPA) Filters AH-F-33 and AH-F-35 The HEPA filters (Table 3-1) are constructed of a dry fibrous high interception, sub-mi:ron glass fiber which has an efficiency of 99.97% for particles larger than.3 microns. The filters conform to ORNL-NSIC-65. The filters are counted in a steel frame and have alumimum separators. Each HEPA filter is fitted with a local differential pressure indicator.
Reactor Building Hydrogen Control Activated Carbon Filter AH-F-34 The activated carbon filters are designed to trap and remove gaseous contaminants (iodine) from the airstream.
The carbon filters (Table 3-1) are of activated charcoal impregnated type, and are of water repellant and fire resistant construction.
The adsorbent material (MSA 85851) is housed in a stainless steel flat bed type frame. The filters are tested in accordance with ORNL-NSIC-65.
2
- 334 27i
Each carbon filter is fitted with a differential pressure indica-tor.
A sprinkler system is built-in for each carbon filter tank.
for fire protection. Means for detecting radiation levels and leaks are provided through a flanged rubber sock port opening at the upstream and downstream f ace of each filter bank for insertion of radiation monitor probes.
Reactor Building Hvdrogen Control Exhaust Fan, AH-E-34 The Reactor Building Hydrogen Control Exhaust Fan will be replaced by a fan manufactured by Buf f alo Forge Company, Model No. 4RE, centri #ugal type, fabricated housing, direct driven, 1000 cfm capa-city at 48 inches of water static pressure at 3550 rpm.
This f an is located on the 328' level of the auxiliary building and driven by a Westinghouse, explosion proof induction motor with air cooled bearings rated at 15 horsepower at 3550 rpm.
If the Reactor Building Hydrogen Control Exhaust Fan ON/0FF switch on Panel No. 25 is in the ON positian, the motor can be powered from 2-11E3.
If the switch is in OFF the motor can be powered from 2-21EB.
There are two red lights to indicate which of the two sources are lined up to power the fan motor and its associated valves (AH-V-25, 36, and 52). Panel No. 25 has two PULL-TO-LOCK-STOP-NOR>iAL-START switches for each of the two power supplies. Additionally the motor has a local START /STOP pushbutton. Motor run indication is available on Panel 25 and locally. The fan will stop with a Fire Protection System signal or when its supply valve AH-V25 is fully closed. Fan start will automatically open its discharge valve.
Reactor Building Pressure Sensinz Line Penetration Isolation Valves AH-V5 and AH-V6 A solenoid operated I" stainless steel valve with a design pressure of 100 psig and a design temperature of 300*F is provided on both sides of reactor building penetration R-562 in the pressure sensing line. AH-V5 and V6 are located on the 305' Level of the reactor and auxiliary building. These valves close with an ES signal.
Con-trol is provided locally on Panel 25.
Indication is available locally and on Panels 13, 15 and 25.
Reactor Building Pressurization Valve AH-V7 An air cylinder operated,10" carbon steel butterfly valve with an ANSI Rating of 100 psig and a design temperature of 300*F is provi-ded in a branch connection of f the reactor building purge exhaust line between reactor building penetration R-552 and the outer isolation valve AH-V4B, on the 328' Level of the auxiliary building.
The valve is in full compliance with the " Draft ASME Code for Pumps and Valves for Nuclear Power", Section B, Nuclear Class II Valves.
The valve fails closed with a loss of instrument air.
The valve is normally locked closed with its outlet flow path blanked.
It is locally controlled. The valve is in the air compressor discharge path during containment leak rate testing.
1334 272
'3
Reactor 3 gilding Hydrogen Control Valve AH-V25 A motor operated 6",
carbon steel, butterfly valve with ANSI Rating of 150 psig and a design, temperature of 150*F is provided in the hydrogen control line upstream of the hydrogen control ex-haust fan.
The valve and fan receive their power from the same sources. The source is determined by an ON/0FF switch on Panel No.
25.
The valve must be partially open for the fan to start.
The valve is positioned locally and has local indication.
Reactor Buildinc Hvdrogen Control Discharge Valve AH-V36 A diaphragm operated, 6" carbon steel butterfly valve with an ANSI rating of 150 psig and a design temperature of 150*F is provided in the hydrogen control discharge line. The normally shut vent isola-tion valve will open when the hydrogen control exhaust fan is started.
The valve fails closed with a loss of instrument air.
AH-V36 is on the 328' Level of the auxiliary building.
Reactor Buildine Hvdronen Control Isolation Valve AH-V52 An air cylinder operated 10", carbon steel valve with ANSI rating of 100 psig and a design temperature of 300*F is provided in the hydrogen control line upstream of the hydrogen control valve, AH-V25.
The valve is in full compliance with the " Draft ASME Code for Pumps and Valves for Nuclear Power," Section B, Nuclear Class II Valves.
This containment isolation valve is padlocked shut and is only opened for hydrogen exhaust fan operation.
The power source is similar to that described for AH-V25.
The valve fails closed with loss of instrument air..AH-V52 is on the 328' Level of the auxiliary building.
3.3.2 Instrument, Controls, Alarms and Protective Devices All controls, indicators and annunciators described are located in the Control Room on Panel 25 unless stated otherwise.
All remotely controlled RB penetration isolation valves have position indicating lights on Containment Isolation Panel 15 in the Control Room. All ES operated valves have indicating lights located on Engineered Safety Features Panel 13 in the Control Room.
All instrumentation, controls, annunciators and computer inputs are included in Tables 11 and 12.
Reactor building air pressure indication is provided as part of the Building Spray (BS) System.
As discussed previously the power supply for the hydrogen control exhaust f an, AH-E-34, and its associated throttle valve, AH-V25, is determined by an ON/ OFT switch on Panel 25.
The throttle valve must be partially open (20%) for fan start. With fan start, the dis-charge damper opens. The flow path is instrumented with filter dif-ferential pressure alarms, radiation, flow and temperature recorders.
1334 273 14
3.3.3 Monitoring The reactor building radiation monitoring system will be used to obtain a sample of the building atmosphere for analysis of its isotopic composition. This system takes smnples from two points in the reactor building, which are located approximacely 10' 10" east and west of the north-south centerline of the re.ctor building dome.
The samples are transmitted through two lines inning from the dome down and outside to the reactor building air s smple gaseous monitor schematically shown as monitor RP-R-227, " Radiation Detec-tion and Sampling" on Figure 3-2.
The sampling lines are Seismic 1.
Redundant inlet and discharge lines are provided for the system to prevent a single active failure of any valve fe ;m impairing the function of the monitoring system.
In the nuclear sampling laboratory, the sample will be analyzed with a gas chromatograph to determine its hydrogen content. A gamma spectrum analyzer will be used to determine the isotopic composition of the sample.
During atmospheric purging, the purge exhaust flow is continuously monitored and recorded on Panel 25, so that the exact flow to the environment is known.
To replace the atmosphere exhausted from the building, a 10 in. pressurization valve (AH-V7), located outside the R.S. is provided to admit a controlled amount of outside air to the building.
3.4 Design Evaluation Given in Section 5 are evaluations of allowable flow rates as a fune-tion of time to stay within Tech. Spec. limits, and the expected off-site exposare as a result of full reactor building purging.
Also includeo in Sections 6 and 7 is an assessment of filter dose buildup and effect of accidental releases of radioactive gases during maximum flow. The exposure analysis shows conformance to limits in each case.
3.5 Tests and Inspection,s The atmospheric purge components are not continuously operated and therefore are accessible for out of service inspection.
The perfor-mance of the system components can be verified while the system is in operation. Pressure, temperature and flow instrumentation are provided as shown in Figure 3-2 to confirm performance of the system and its components. Radiation monitoring instrumentation is also provided in the system to check radioactive levels of the exhuast air.
In addition, means have been provided for pre purge DOP and freon leak tests of the filters.
The steel pipe duct work system was subjected to leak tests during manufacture, erection and af ter assembly in the field. Filters and filter housings were subjected to manufacturers performance and production tests as well
15 1334 274
The charcoal filter will be subjected to a Freon 11 leakage test at 1000 cfm, the maximum flow expected in the system.
The HEPA filters will be subjected to an Ef ficiency-Penetration Test (DOP). The filter will be tested according to MIL-STD-282, May 28,1956. Penetration will not exceed 0.1 percent of 0.3 micron diameter homogeneous particles of dioctyl phthalate (DOP).
3.6 Materials The ductwork is primarily mild carbon steel and has a 6 mil coat of Phenoline 368. The material for other components was given in Section 3.3.
3.7 coeration of Hvdrogen control Purge System Contro?.s for this system are located on HVAC Panel 25.
To start this system it is necessary first to open reactor building isolation valve AH-V3A (closed on ES signal) and AH-V52 (padlocked closed). A Blind flange on valve AH-V7 must be removed prior to this operation.
Open throttle valve AH-V25 to about 309' prior to starting the Hydrogen Control fan, AH-E-34.
Upon starting the f an the discharge valve, AH-V36, will open. Throttle AH-V36 as desired. When the reactor building pressure is slightly below atmospheric; open AH-V7 (normally locked closed) and then. open AH-V3B, to replenish the exhausted air.
The reactor building ar osphere is exhausted through isolation valve AH-V3A, a 10" branch i A containing valve AH-V52, a 6" line contain-ing throttle valve AH-V25, pre-filter, two absolute filters and an activated carbon filter. The f an then discharges through valve AH-V36 to the Station Vent.
The system is shutdown by stopping AH-E-34 and closing AH-V25, AH-V52 and AH-V3A and AH-V36.
3.8 Accident Analysis The worst case limiting accident is inadvertent and undetected initiation of the hydrogen control purge system at full capacity with ground level release for 30 minutes during worst case meteor-ological conditions. The off-site dose due to this accident is analyzed in Section 7 of this report. The results show maximum off-site exposure to an individual of 61 mrem beta skin dose or 0.73 mrem whole body dose, considerably less than 10CFR100 limits of 25000 mrem whole body dose.
3.9 Failure Modes and Consequences All failures and their consequences are evaluated with respect to increasing the probability of producing an uncontrolled radioactive release or a release at a faster rate than allowed by procedure.
- 1334 275
3.9.1 Loss of Instrument Air Loss of instrument air will affect only the air operated valves and dampers in the system. All air operated valves in this system fail shut upon loss of air pressure which stops all flow in the system and thereby prevents any release of radioactivity.
3.9.2 Loss of Power Fan AH-E-34 Loss of power to this fan wi? ' reduce flow rate through the system causing a reduction in the release rate.
Valves operated by air will fail shut on loss of power to the sole-noid operated air control valves.
Motor operated valve AH-V-25 will fail as is on loss of power.
Neither of these cenditions will cause any increase in the release rate.
Instrumentation Flow indicatian will be lost on loss of power, however, this indication does not control the process so no increase in release rate can occur.
If an increase in flow rate occurs due to some other cause, HPR 229 will alarm as a backup indication.
System operating procedures will require the operator to stop purging upon loss of flow indication.
Filter unit differential pressure alarm capability will be lost on loss of power.
Flow indication will provide a backup indication of filter blockage. Th is alarm does not cause any automatic action.
No increase in release rate will occur from loss of this instrument.
Reactor Building High Pressure or Loss of Power to AH-PS-5058 will cause AH-V-3A,B to shut thereby stopping flow in the system.
Radiation Monitor HPR 229 is being modified to cause fan shutdown upon loss of power to HPR 229.
Valve AH-V7 is also interlocked so that when the fan stops, this valve will close, isolating this potential path out of the reactor building.
3.9.3 System Leakage In order to ensure radiation will not be released from building ducts during operation, a leak test of ducting downstream of the containment isolation valves and the filter housing will be conducted prior to system operation at 18 inches of water positive pressure in accordance with ANSI N510, Section 6.3, 6.4, or 6.5 and shall indicate maximum leakage less than 6 cfm/1000 f t3 og system volume before acceptance.
}334 276 17
3.9.4 Fires The charcoal filter AH-F-34 is protected by a fire detector and an automatic deluge system which also secures fan AH-E-34 The filter housing drain is piped to floor drains which flow to a collection facility.
3.9.5 Duct Failure The steel pipe ducting is designed for 2 psig positive pressure.
The filter unit housing is designed for 11 inches H,0 negative pressure. The system is pr;tected from high pressure transients by a.5 psig containment pr,(ssure interlock f rom pressure switch AH-PS-5058.
Maximum internal dret pressure is limited to 1.5 psig, since the maximum dif ferential pressure will be caused by a
.5 psig reactor building pressure coupled with a minimum external prescurs of -1.0 psig caused by an external atmospheric disturbance.
This extreme pressure conditit,n is within the design rstings of the duct and filter unit and will not cause a duct failure which would result in an uncontrolled r;dioactive release.
3.9.6 Operator Errors Hisoperation of the ve.lves in the system could possibly increase the rate of radioactive gas release above the maximum allowable rate.
However, HPR 229 would alarm and automatically stop flow in the system if the allowable release rate were exceeded.
3.10 System Modification For Purge In order to use the hydrogen tontrol system to purge the reactor building safely at rates up to 1000 cfm, the following modifica-tions will be made:
1.
Replace the existing fan, AH-E-34, with a fan capable of at least 1000 cfm flow rate.
2.
Add manual jog control to valve AH-V36.
3.
Interlock AH-V7 to close on loss of power to the f an.
4.
Provide interlock to trip the fan on high activity as measured on HPR-229, or on failure of/ loss of power to HPR-229.
5.
Interlock AH-V3A&B to shut on high reactor building pressure.
6.
Uncap the stack.
7.
Provide gamma monitor probe in the hydrogen control filter housing to monitor filter activity buildup.
8.
Increase the measurement range of EPR-229.
i334 277
s Table 3-1 Hydrogen Purge to Atmosphere Design Performance and Eauiement Data a.
Hydronen Control Exhaust Fan Quantity 1
Type Centrifugal Exhauster with Direct Drive Flow, cfm 0 to 1000 Static Pressure, in W.G.
48 neg at 3550 rpm Fan (Motor) Speed, rpm 3550 Fan Motor Voltage /No. of Phases /El 460/3/60 15 Motor H.P.
b.
Hydrogen Purze Air Exhaust Filter Train One hydrogen purge air exhaust filter train in cabinet / housing containing the following filters listed in sequence with respect to air flow:
(1) Prefilter Quantity 1
Type Disposable bag filter Clean Pressure drop, in W.G.
0.8 Max. Capacity, cfm 1000 Face Velocity through Filter, fpm (max.)
500 Size of Filter, inches 24x24x36 Seismic Classification I
(2) Absolute Filter (REPA)
Quantity 1
Clean Pressure drop, in W.G.
1.0 Max. Capacity, cfm 1000 19 1334 278 w
e
Table 3-1 (Con't)
Size of Filter, inches 24x24xil-1/2 seismic Classification I
(3) Carbon Filter Quantity 3 per Bank Type Flat-bed radio Iodine Absorption activated carbon Max. Capacity, cfm 1000 Flow through cell, cfm 333 Clean Pressure drop, in W.G.
1.0 Size of Filters, inches 24x40x7-3/4 (4) 2nd Absolute Filter (REPA)
Quantity 1
Clean Pressure d.op, in W.G.
1.0 Capacity, cfm 1000 Size of Filters, inches 24x24xil-1/2 Seismic Classification I
1334 279 20
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4.0 SAFETY LIMITS FOR RADIOACTIVE GASEOUS RELEASES 4.1 Radiation Exposure in Perspective 4.1.1 Sources of Radiation in the Environment The average natural radiation exposure to persons living in the United States is estimated to be about 125 mrem per year. The source of this exposure is from cosmic r1ys and from naturally occurring radioactive elements in the aarth. Radiation is received directly from many minerals containing uranium and thorium isotopes in the ground or in the construction materials in homes. The mast significant radioisotope in food is potassium. An additional small amount of exposure is received through radioactive gases in the air. It is estimated that an additional exposure of 100 mrem per year may be received on the average from other than natural sources such as medical X-rays, luminous dials on watches, bomb detonations in the atmosphere, and television. 4.1.2 The Person Rem Concept It is appropriate to ccmpare new exposures to population groups relative to their exposure to natural background. One measure of the extent of population exposure i6 to add all the radiation exposures received by each individual in a population group. This resulting quantity is referred to as person-rem. The annual background populltion exposure within a 50 mile radius of the site is computed to be about 275,000 person-rem (125 mrem times 2.2 million people). By comparison, the person-rem for reactor building purging as discussed in this report is about 100,000 times less than the natural background amount. It should be noted that the whole body gamma doses listed for each event are most comparable to the background dose. The external body beta dose affects only the external parts of the body (skin or retina of the eyes) which are less sensitive to radiation than the whole body. 4.1.3 Effects of Radiation Exposure For many years standards committees have spent considerable effort to determine the effect of radiation on man. As a result, a set of guidelines have been developed to define maximum levels of radiation exposure which are acceptable for any individual to receive every year. These recommendations are embodied in the government regulation entitled 10CFR20 which limits whole body exposure to less than 500 mrem per year. Comparison of the site boundary doses from the events associated with reactor building purge considered in this report indicates that individual exposurer are well below both the natural background level of 125 mrem / year and the 10CFR20 limit of 500 mrem /yr. 23 1334 2n2
4.2 Environmental Technical Soeci.'ications Section 2.1.2 of the Environmental Technical Specifications contains specifications for the release of gaseous ef fluents. These are: 2.1.2.a. The instantaneous release rate of gross gaseous activity except for halogens and particulates with half iives longer than eight days shall not exceed: Oi 5 ,3 (MPC)i --<l.5 x 10 sec where Qi is the release rate in Ci/sec for isotope i, 3 and MPCi ( Ci/m ) is the maximum permissible concen-tration of isotope i as de fined in Appendix 3, Table II, column 1, 10CFR20, 2.1.2.b. The instantaneous release of I-131 and particulates with half-lives greater than eight days, released to the environs as part of airborne effluents, shall not exceed 0.3 gCi/sec. 2.1.2.c. The release rate of gross gaseous activity shall not exceed: 0 <2.4 x 10' 3 ( ), m i see when averaged over any calendar quarter. 2.1.2.d. The release rate of I-131 and particulates with half-lives greater than eight days, shall not exceed 0.024 gCi/sec., when averaged over any calendar quarter. The specifications above can be used to establish limiting reactor building release rates for instantaneous and quarterly average releases when the Kr-85 (paragraphs a. and c.) and the I-131 (paragraphs b. and d.) are known. Present estimates of airborne activity for Kr-85 of 0.78 uCi/ml and <l. x 10-9 uCi/ml for I-131 make the paragraphs a. and c. most limiting based on Kr-85 activity. Specification 2.1.2.c above assumes an average X/Q value of 4.2 x 10-5 sec/m3 Historical meteorological data for the TMI site, with elevated releases give ar. average X/Q value of 1.8 x 10-D sec/m3 for a margin factor of 23 built into the quarterly release rate limit. Since the limit is applied over any quarter, the limit for the quarter will not be any greater than one-fourth of the annual dose objective limit. Therefore a factor of roughly one hundred is achieved below tha 10CFR20 objective of 500 mrem off-site whole body dose when the quarterly average technical specifica-tion limits are imposed for radioactive gaseous effluent release from the plant vent stack. 1334 283
In other words, while the 100FR20 limit (to be discussed below) states as its objective a limic of 500 mrem annual whole body dose, the application in the Technical Specification achieves a factor of 100 reduction to 5 mrem annual whole body dose limit when using the quarterly average release rate limits. 4.3 10CFR20 - Standards for Protection Against Radiation Article 20.105, permissible levels of radiation in unrestricted areas, paragraph (a) states: "The commission will approve the proposed limits if the applicant demonstrates that the proposed limits are not likely to cause any individual to receive a dose to the whole body in any period of one calendar year in excess of 0.5 rem." Article 20.106, Radioactivity in ef fluents to unrestricted areas, paragraph (d) states: "For the purposes of this section the con-centration limits in Appendix B Table II of thic part shall apply at the boundary of the restricted area. The concentrntion of radio-active material discharged through a stack, pipe or similar conduit may be determined with respect to the pipe where the material leaves the conduit. If the conduit discharges within the restricted area, the concentration at the boundary may be determined by apply-ing appropriate factors for dilution, dispersion or decay between the point of discharge and the boundary." Table II, column 1 of Appendix B gives the following limits for the pertinent isotopes: Kr-85 3 x 10-7 uCi/ml Xe-131m 4 x 10-7 gCi/ml Xe-133 3 x 10-7 uCi/ml These isotopic limits are used with the limiting X/Q values in the Technical Specifications given above to determine allowable gaseous releases at the point of effluent release from the plant for instan-taneous and quarterly average limits. As pointed out earlier, use of these limits with the Technical Specification X/Q values to determine release rates will yield substantially lower actual values of the isotopic air concentrations in the unrestricted area because the expected X/Q values are at least an order of magnitude below the limiting X/Q values. For this reason, the expected whole body dose will be sub;tantially below the 0.5 rem yearly limit imposed by paragraph 20.105 (a). 4.4 10CFR50 Appendix 1 - Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion - ALARA. Paragraph B.1 states: "The calculated annual total quantity of all radioactive material above background to be released from each light-water-cooled nuclear power reactor to the atmosphere will not result in an estimated annual air do_a from gaseous effluents at any location near ground level which could be occupied by individuals '5 ^ 1334 294
in unrestricted areas in excess of 10 millirads for gamma radiation or 20 millirads for beta radiation." Paragraph B.2 states: "Not withstanding the guidance of para-graph B.1: (a) The commission may specify, as guidance on design objectives, a lower quantity 7f radioactive material above background to be released to the atmosphere if it appears that the use of the design objective in paragraph B.1 is likely to result in an estimated annual external dose f rom gaseous ef fluents to any individual in an unrestricted area in excess of 5 milli-rems to the total body; and (b) Design objectives based upon a higher quantity of radioactive material above background. to be released to the atmosphere than the quantity specified in paragraph B.1 will be deemed to meet the requirements for keeping levels of radioactive material in gaseous ef fluents as low as is reasonably achiev-able if the applicant provides reasonable assurat.ce tha t the proposed higher quantity will not result in an estimated annual external dose f rom gaseous ef fluents to any individual in unrestricted areas in excess of 5 millire=s to the total body or 15 millirems to the skin." As stated above in discussing 10CFR20 limits, the margin provided in bounding X/Q values in the Technical Specification to limit effluent release rates will bring the 0.5 rem total body dose limit stated in 10CFR20 down to the 10CFR50 Appendix I level when actual expected site meteorology is applied in calculating actual population doses. Nevertheless, the actual accumulate.4 doses during purge operation shall be calculated to demonstrate conformances to the 10CFR50 Appendix I limits. Of f-site dose calculations for the planned pu ge scenario using typical meteorological data are presented in Section 5. 4.5 Limiting Conditions for Operation of Purge Program In determining acceptability of the reactor building purge program, based on the above requirements, the limiting conditions are as follows: 1. Release rate for Kr-85, Cs-137, and I-131 are determined from Technical Specification paragraphs 2.1.2.a. and 2.1.2.b. to establish the most limiting release rate. For the current status of reactor building isocopic content, the Kr-85 content is most limiting. 2. For the total quantity of Kr-85 to be released, the quarterly average release rate limit stated in paragraph 2.1.2.c. can be verified not to be exceeded with the Kr-85 concentration below 0.988 ECi/ml. 1334 295 26
3. Based on actual meteorology of the site at the time of re-lease, administratively limit release rate to minimize actual dose to the unrestricted site boundary area so that total beta and gamma dose do not exceed limits stated in 10CFR50 Appendix I. The limiting value is 15 millirems skin dose due to beta activity in the Kr-85 release. 4. Continuous monitoring of meteorology will be provided to administratively control release as a function of actual characteristics and to calculate the dose build-up within allowable Appendix I limits. b0 27
5.0 iNVIRONMENTAL EFFECTS OF PURGE OPERATION Starting with the best estimates of reactor building airborne iso-topic activity, the Technical Specifications are used to establish limiting purge release rates from the plant vent stack. Using typical historical meteorology, the site boundary beta skin dose and gamma dose are calculated using the NRC prescribed methods defined in Regulatory Guide 1.109 " Calculation of Annual Doses to Man From Routine Release of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10CFR Part 50 Appendix I" and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water-Cooled Reactors." Since at the time of purge operation, the I-131 has decayed to in-consequential levels, the Kr-85 contribution to beta skin dose is the most limiting radioactive source. The results of this analysis yield a peak Jeta skin dose of 5.0 mrems when purging is limited during unf avorable meteorological conditions. This is compared to the 10CFR50 Appendix I limit of 15 mrems. For the case in which the only limit on purge operation is the instantaneous Technical Specification limit due to Kr-85 content in the released gas, and no limit is placed on purging during unf avorable conditions, the peak skin dose is 10.0 mrems. In both cases, the peak ga=ma dose is less than 0.10 mrems, compared to the 10CFR50 Appendix I limit of 10 mrem. 5.1 Purce Compliance with Technical Soecification 5.1.1 Method The computer code TIDRLS and manual calculations are used to ensure that the Technical Specifications for instantaneous and quarterly average release of iodine and gaseous activity are not exceeded. The appropriate Technical Specifications are: 5 3 f,1.5 x 10 m f see for instantaneous release of gross a. gaseous activity (applies for Kr-85) b. Instantaneous release rate of I-131 and particulates (applies for Cs-137) with half-life greater than eight days must be f,0.3aci/sec 0 3 f,2.4 x 10 m /see for average over a quarter of gross c. gp gaseous activity (applies for Kr-85) d. Average release over a quarter of I-131 and particulates (applies for Cs-137) with half-life greater than eight days must be j,0.024 aci/sec Where gi is the release rate in Ci/sec for isotope i and MPCi ( Ci/m ) is the maximum permissible concentration of isotope i as defined in Appendix B, Table II, Column 1,10CFR20. '8 1334 287 ~
Hand calculations are used to determine an initici purge rate that wc.1d not exceed the instantaneous Technical Specitt.r*lons discus-sed above. Ft. gases (Kr-85, Xe-133, and Xe-131m), the allowable purge rate is determined as follows: 01 = 1.5 x 105 m3 see / MPCi 3 Qi = (MPC ) (1.5 x 105 m /sec) i Then, purge rate, m3/sec = Q/ concentration, (where concentration 3 is in Ci/m ), For I-131, and Cs-137 the allowable purge rate is determined as follows: -6
- 0., uCi/see x 10
" 8" ## * "'(concentration) (1-charcoal ef ficiency) ' *3 ** (where concentration is in Ci/m3 as above). Af ter determining an initial purge rate, the decrease in contain-ment activity inventory resulting from the purge was calculated. No credit is taken for decrease in inventory due to decay. When containment activity decreases suf ficiently to allow approximately a doubling of the flow rate, a new purge rate is established. The same process is then repeated until a maximum purge rate of 1000 cfm was obtained. The decrease in activity in the RB is calculated as follows: Co T = _y in 7q-where T = time of purge m = purge rate in cfm M = volume of containment in ft3 Ci = concentration at starc of purge Co = concentration at end af purge Once the purge scenario is determined, it was input into the TIDRLS computer code for a detailed calcu.'atica of instantaneous and integrated releases. The integreted purge rate can be used to determine an average purge rate. Using average purge rate, the quarterly average release rate can be calculated as follows: Ci Quarterly release rate = -(-Average release in see) (purge time in seconds) (MPC) (Number of seconds in quarter) 29
5.1.2 Evaluation with Technical Specifications Initial Purge Rate Using the techniques described in Section 5.1.1, the initial purge rate is calculated that can be accomplished without exceeding the Technical Specification limits. For Kr-85, the allowable instantaneous purge rate is: Purge Rate = Q/ concentration Q = MPC (Kr-85) x 1.5 x 105 3/3 c 2 From 10CFR20 Appendix B, Table II column 1, Sec (Kr-85) = 3 x 10-7 Ci/m3 There fore, 3 x 10-7 (Ci/m ) x 1.5 x 10 3 5 (m /Sec) x 60 (Sec/ min) 3 Purge Rate = 0.78 (uci/ml) x 10-6 (Ci/pci) x 2.832 x 104 (ml/ft3) = 122 CFM The maximum allowable initial purge rate for release of Kr-85 is 122 CFM, based on an initial Kr-85 content in the reactor building atmosphere of 0.78juCi/ml. For Cs-137, the allowable instantaneous purge rate is: 0.3 (itCi/Sec) x 60 (Sec/ Min) Purge Rate = 1 x 10-5 94Ci/ml) x 2.832 x 104 (ml/ft3) (1_o,9) = 636 CFM The maximum allowable initial purge rate for release of Cs-137 is 636 CFM, based on an initial Cs-137 particulate content in the reactor building atmosphere of 1 x 10-5,0Ci/ml and a particulate filter efficiency of 90%. For I-131, the maximum allowable purge rate is determined the same way as for Cs-137 except that charcoal efficiency replaces parti-culate filter efficiency. This evaluation gives a maximum allow-able initial purge rate for I-131 of 636 x 104 CFM, based on an initi.1 I-131 concentration cf 1 x 10-9,o Ci/ml and a charcoal efficiency of 90%. From the above calculation, it can be seen that the limiting isotope for Tech. Spec. release limits is Kr-85. 1334 289 30
Purge Duration As the concentration of radioisotopes within the reactor building is reduced while purging, the allowable purge rate can be increased up to the maximum purge flow capacity of 1000 CFM. A typical scenario for purging might consist of stepwise purge levels of 100 CFM, 200 CFM, 500 CFM, 1000 CFM. The limiting concentrations for Kr-85 to stay within instantaneous Technical Specification limits are given in the Table 5.1.2-1. The tim to reach the limiting concentrations for each purge level can be calculated as described in the previous section assuming step wise increases in purge rate each time the Kr-85 content drops below the maximum concentration for the next increased purge rate level. The final time period at 1000 CFM purge rate is the time reguired to reduce the Kr-85 content from 0.095//ci/ml to the 1 x 10-) U Ci/ml MPC limit for restricted access to the reactor building per 10CFR20 Appendix B Table 1, Column 1. The time period at each purge rate and the initial and final Kr-85 levels are given in Table 5.1.2-2. The above analysis shows that the RB can be purged in about 31 days without exceeding the instantaneous release Technical Specifi-cations. The releases must also be compared to the quarterly average Technical Specifications. The quarterly average Technical Specification fcr I-131 and Cs-137 can be easily met using the above scenario. The quarterly average Technical Specification for gaseous activity can be met, so long as the initial gaseous concentration in the RB (Kr-85) is less than 0.988,0Ci/ml. This can be shown as follows: 3 O/MPC < 2.4 x 104 m /sec (from Tech Specs) Where MPC = 3.0 x 10-7) Mci /ml = 3.0 x 10-7 Ci/m3 for Kr-85, the limiting isotope Max allowable Q = (2.4 x 104 m3 sec) (3.0 x 10-7 Ci/m3 / = 7.2 x 10-3 Ci/sec max allowable' concentration (C in pCi/ml) = (0) (Number of seconde in quarter)
- But, RB volume lu ml 24 hr 60 min 60 sec)
(7.2 x 10_3 Ci/sec)(90 days x x x Max allowable C = (2.0 x 106 ft 3)(2.832 x 104 ml/ft3) C = 0.988 ml Maximum allowable concentration to be released within quarterly allowable Technical Specification limits can also be oetermined for Cs-137 and I-131. 31
For Cs-137, using Avg. Release Rate < 0.024u Ci/Sec, Max. allowable concentration released = 0.024 UCi/See x Total Seconds in Quarter RB Volume in ml Because a large fraction of particulate Cs-137 will be captured by the hEPA filters in the hydrogen control system, credit can be taken for filtered removal of Cs-137 prior to release. The HEPA filter at worst performance can be expected to be at least 90% efficient. Therefore 0.024 UCi/See x Total Seconds in Quarter Max. allowable concentration = RB Volume in ml x (1-Filter efficiency) , 0.024 (UCi/Sec) x 90 x 24 x 60 x 60 (Sec./Ouarter) 2.0 x 106 (ft3) x 2.832 x 100 (ml/ft3) x (1_,9) = 3.3 x 10"5 UCi/ml Therefore the maximum allowable Cs-137 concentration in the reactor building atmosphere for full release within one quarter within cur-quarterly average Technical Specification limits is 3.3 x 10-5 rent UCi/ml for a worst case filter efficiency of 907. Cs-137 removal. The maximum allowable I-131 concentration for full release within quarterly Technical Specification limits is determined in the same manner as for Cs-137 except that the HE?A filter ef ficiency is replaced by charcoal efficiency for Iodine removal. Using a 90% charcoal efficiency, the maximum allowable I-131 concentration within the reactor building atmosphere for full release within one quarter conforming to quarterly average Technical Specification limits is 3.3 x 10-5 ACi/ml. / 29l 32 k
Table 5.1.2-1 Maximum Concentration of Kr-85 vs. Purge Rate Purze Rate (CFM) Kr-35 Conc. (DCi/ml) 100. <0.95 200. 0.48 500. 0.19 1000. 0.095 1334 292 33
Table 5.1.2-2 Time to Reach Each Limiting Coccantration Initial Kr-85 Final Kr-85 Purge Purge Rate (CFM) Content (DCi/ml) Content (uCi/ml) Duration (days) 100 <0.95 0.48 9.4 200 0.48 0.19 6.5 500 0.19 0.095 1.9 1000 0.095 1 x 10-5 12,7 Total Duration 30.5 days 1334 293 34
5.1.3 Results of Purge Release Within Technical Specification Limits Table 5.1.3-1 gives the summary results for maximum allowable instantaneous purge rate within Technical Speoification limits for the three radio-isotopes most limiting on purge rate. Using the Kr-85 conce stration to establish purge flows as a func-tion of time, a full purge of the reactor building airborne radio isotope concentrations within MFC limits for restricted area access can be achieved in 31 days following the schedule shown in Table 5.1.2-2. The quarterly average release rate Technical Spscification limits can be maintained so long as the concentrations of radioisotopes within the reactor building atmosphere are less than the levels given in Table 5.1.3-2. The expected quarterly average release based on current Kr-85 levels will be 79% of the allowable Tech. Spec. limit for full purge of the reactor building atmosphere. I334 294 35
a Table 5.1.3-1 Estimated Limiting Purge Isotope Concentration (uCi/ml) te (CFM) I.a Kr-85 0.78 122 Cs-137 <1 x 10-5 636 1-131 <1 x 10-9 636 x 104 1334 295 36
e Table 5.1.3-2 Quarterly Avg. Liditing Purge Isotope Tech. Spec. Limit Concentration (CCi/ml) Kr-85 17.2 x 10+3 Ci/sec 0.988 Cs-137* 10.024pCi/Sec 3.3 x 10-5 1-131* 10.024uCi/Sec 3.3 x 10-5
- Assumes filter ef ficiencies of 90' for Cs-137 and I-131 removal.
1334 296 37
5.1.4 Conclusions for Purge Release Within Tech. Spec. Limits 1. Full purge of the reactor building atmosphere below instan-taneous Technical Specification radioisotope release rate limits can be accomplished in 31 days with quarterly average radioisotope release rates below 79% of the Technical Specification limits. 2. Kr-85 is the limiting isotope for determining purge rate. So long as the initial Kr-85 content is below 0.988u Ci/ml, the quarterly average limits can be met. 3. The purge rates can be allowed te vary from 100 to 1000 CFM with 9.4 days at 100 CFM, 6.5 days at 200 CFM, 1.9 day at 500 CFM and 12.7 days at the maximum purge rate of 1000 CFM. 5.2 Of f-Site Dose De termination for Purge 5.2.1 Method The PURTST and XDCALC computer programs were developed for use in a parametric study to assess the effects on offsite dose of dif-ferent purge start times and various purge procedures. These programs incorporate the environmental dose calculatian routines utilized in the dose assessment effort following the accident. The program computes not only beta, ga=ma, and thyroid doses but also checks for compliance with Technical Specification limits and MPC of fsite limits each hour during the purge. Both elevated and ground level release, points are evaluated. Two basic types of simulations were made. One type (XDCALC) uses a " predetermined" purge rate and the second (PURTST) uses a " variable" purge rate depending on meteorological conditions each hour. The basic objective of the study was to make a series of computer simulations to determine integrated doses of fsite while achieving the goal of reducing noble gas concentrations to below MPC in the reactor building during a purge time of one to two months, starting in October / November, 1979. Results are used to plan and evaluate the purge procedures to be utilized. 5.2.1.1 Meteorological Data Historical data from the onsite meteorological tower taken in 15 76, 1977 and 1978 were used for the same months of the year to simulate conditions in 1979. Hourly values of measured wind speed, wind direction and vertical temperature dif ference were available on computer files for access by the program. Specifications for instruments used to collect these data are given in Table 5.2-1. The meteorological program is designed and operated in accordance with NRC Regulatory Guide 1.23. r Vertical temperature difference was used to determine atmospheric stability (Pasquill-Gifford category) in accordance with procedures 38 1334 297
e also outlined in NRC Regulatory Guide 1.23. Wind speed measured at 100 ft. on the meteorological tower is adjusted to the 33 f t. or 160 f t level to be representative of the assumed ef fective heights used during the purge dose studies. The equation used for this adjustment is as follows: th U =U 1 RH 100 t h100/ where hRH (ft) and URH (mph) are the height and wind speed at the release height respectively, and n is a function of atmospheric stability as follows: Pasquill-Gif f o rd Value of Category n A, 3, C S 25 D 0.33 E,F,G 0.50 3.2.1.2 Initial Radioactivity in Reactor Building Simulations for variable purge rates requires knowledge of only the initial reactor building inventories since the PURTST program computes the remaining inventories af ter each purge hour. These initial inventcries are based on the RB air samples taken on June 26, 1979. For start dates beyond July 1, the initial inven-tories were reduced according to the half lives of the individual isotopes. The following table summarizes the assumed starting inventories in uCi/cc. Purge Start Date* Isotope July 1 August 1 Seotember 1 Kr-85 1.0 1.0 1.0 Xe-131M 0.085 0.012 0.0021 Xe-133 0.015 0.00017 negligible I-131 0.00012 0.00002 negligible
- Note that the Kr-85 level is conservative relative to the best estimate value of 0.78 quoted in Section 2.3.
Airborne iodine levels are low in the reactor building and are expected to remain low during the purge. In addition, the purge effluent will flow through charcoal absorbers prior to discharge. If most of the iodine is in an organic form, and more penetrates the charcoal absorbers compared with the elemental iodine, it would not be of particular importance in the environmental analysis because it is not taken up in the cow-milk pathway. Therefore, the controlling isotopes are the noble gases, (primarily Kr-85), and thyroid doses due to iodine releases are not calculated. 39 1334 298
5.2.1.3 Plant Characteristics Ef fluent from the reactor building purge will be directed into the plenum of the plant vent and out the plant vent stack. If the supplementary filter system plant release point were to be used (since this system is horizontal) no plume rise due to momentum of the existing gas is possible. Effluent will always enter the turbulent wakes surrounding plant buildings and undergo initial dilution in this region. Af ter leaving the influence of these buildings, concentrations in the plume at ground level will always decrease with distance. Plant parameters pertinent to the diffu-sion model are given in Table 5.2-2. Several runs were made to determine the beneficial ef fects of plume rise when the plant vent stack is used. For these runs, vent parameters are necessary and are given in Table 5.2-2. For elevated plumes, terrain must be subtracted, therefore, the assumed terrain at various distances downwind in each of the 16 direction sectors is provided in Table 5.2-3. 5.2.1.4 Standards 10CFR50, Appendix I - provides guidelines for of f site dose objectives for routine plant operation due to noble gases as follows: Noble Gas air gamma dose 10 mrad
- Noble Gas air beta dose 20 mrad
- Noble Gas whole body dose 5mren**
Noble Gas skin dose 15 mrem **
- applicable at site boundary
- applicable to real person 5.2.1.5 Atmospheric Dispersion Model For long-term ground level releases in the building wake, the sector average version of the Gaussian dispersion model is used.
.03 X/Q = gnd g
- where, z
2 2' +5 <VlTs 5 I = z z r z 3 X/Q = concentration at ground. level ( Ci/m ) + release rate Q ( Ci/sec) u = wind speed at 33 f t. level (m/sec) x = distance from plant (m) 3 s = Pasquill-Gifford dispersion coefficient (m ) z c = wake coef ficient (=0.5) H = building height (m) ~ 1334 299 'o
e o For the vent stack release runs, plume rise above tne building wake boundary under certain light wind conditions was accounted for. The model referred to as " mixed mode" from Regulatory Guide 1.111 was used for these cases. The elevated release model is defined as follows: !- S \\ + 2.03 X/Q =, exp ele ux c 2e2 where symbols are as before, and Hg = stack exit height above local terrain (m) Ah = plume rise due to momentum Jet (m) (determined using Briggs' model) X/Qgnd X/Qaixed " (I - E ) X/Qele + E t t mode where E is determined as follows t if E > 5.0, E =0 t u W if E < 1.0, Et= 1.0 u and Et= 2.58 - 1.58 (W /u) for 1 i W /u i 1.5 o o Et = 0.3 - 0.06 (W /u) for 1.5 i W /u i 5.0 o o Wo = Vertical exit velocity of plume 5.2.1.6 Dose Calculation Models in this analysis both whole body and skin dose analyses were made. The skin dose (Dskin) results from both beta and gamma radiation. Since beta particles are stopped by only a few centimeters of air, one must be submerged in the plume to receive a dose. Therefore, ground level concentrations of each isotope based on X/Q must be used. Gamma dose (Dgamma) can be received as a result of shine from plumes alof t as well as f rom submersion in the plume. A finite plume dose model is used to estimate gamma dose as described later. Beta dose to the skin is computed using the following relationship from NRC Regulatory Guide 1.109 Equation (11): Dskin = D amma + 0.114E Qi (X/Q) F t g i i 4 00 where 41
Q- = release rate ( Ci/sec) of isotope i XfQ = ground level dispersion factor (sec/m ) 3 Fi = dose factor Kr-85 = 1.34 x 10-3, Xe-131m = 4.76 x 10-4 3 Xe-133 = 3.06 x 10-4 (units of mrem-m /pci-yr) t = time (hour) 0.114 = constant for units correction Gamma dose (Dgamma) is computed using the finite plume model de-tailed as Equation (6) of Regulatory Guide 1.109. The integral "I" is solved using the method of Hamawi in accordance with the regula-tory guide and is not repeated here. For Kr-85, the only signifi-cant airborne isotope, the beta dose is about 100 times high.er than the gamma dose for an individual in the plume. Population doses are computed using the same dispersion and dose routines out to 50 miles. At each population segment shown in Figures 5.2-1 and 5.2-2, the dose is multiplied by the number of people and summed to determine person-rems. 5.2.1.7 Computer Programs The dose models were incorporated in a routine called PURTST which is used to assess the ef fect of varying a series of input para-meters. The general program flow diagram is shown in Figure 5.2-3. The routine starts by reading the meteorological parameters for the hour. Then, using the reactor building isotopic content for the hour, doses and concentrations at the site boundary are calculated for an assumed arbitrary flow rate of 100 CFM. Based on limits established at the beginning of the run for beta and ga=ma dose as well as MPC for each isotope, an allowable flow rate is computed. Tech Spec limits are also checked and if the release rate would exceed limits the flow is further reduced. Flexibility is provided to bypass checks of certain limits as part of the parameter study. For example, if Tech Spec limits are to be neglected, the Tech Spec limits are set to high values so they will not produce limiting flow rates. Similar provisions are made for other limiting con-dicions such as dose limit. This calculation is repeated every hour until the total period of record specified has been processed. At the end of each hour, the amount of each isotope released (based on the limiting parameter) is subtracted from the amount in the reactor building at the beginning of the hour. This provides the isotope concentration for establishing the release rate during the next hour. After each hour, summary tables of beta and gamma dose are incre-mented along with the total volume purged. Thus, at the end of the specified purge time the resulting doses are available along with total effluent release. No limits have been placed on doses within a direction sector to stop purge while winds are in the loaded sector. OI 42
o Several runs were made using the XDCALC program without regard to feedback from meteorological conditions. The XDCALC computer pre-gram has been used extensively in evaluating doses resulting from the accident. The XDCALC program uses the dispersion and dose models described above and a predetermined release source term that can be specified each hour. Doses are computed at several distances near the site and integrated each hour in the appropriate direction sector. 5.2.2 Evaluation of Of f-Site Doses Due to Purge Ceneral An evaluation has been made to assess the of f-site man-rem, beta-gamma integrated dose and instantaneous dose rate to the whole body, skin and thyroid as a result of various purge scenarios. The results vary as a function of release rate and meteorology. Alteration of the existing purge path to release via the station vent stack also affects the net environmental impact associated with RB purge. A series of runs were made using the PURTST and XDCALC programs and meteorological data for July, August, October, and November from 1976, 1977 and 1978. The result of these analyses are included in Table 5.2-5. Sensitivity studies were completed for varying purge rates, release locations and meteorological conditions. Cases Considered A summary of the cases evaluated is given in Table 5.2-4. The first case listed, Case No. 1, determines the expected boundary doses when a constant activity release rate is used to purge all ac*ivity over a one month interval (Kr-85 release rate is 2.15E4 pCi/see for one month). Estimated doses are calculated using typical July meteorology as taken from July 1976. The release point is taken to be the roof top supplementary filter vent. The second case listed, Case No. 2, uses the same meteorology and release location as Case No. 1 but the release rate is taken from the step wise purge rate scenario calculated in Section 5.1. Variations to Case No. 2 are treated in: Case No. 11, for vent stack elevated releases; Case No. 15, for Aug. 1977 reference meteorology; and Case No. 17, for August 1978 meteorology. The third case listed, Case No. 8, uses hi corical meteorology conditions to limit release rates to hourly limits for beta, gamma, and MPC off-site limits. This case uses July 1976 meteor-ological data and the ground level equivalent roof vent filter release point. Variations to Case No. 8 are treated in: Case No. 12, for vent stack elevated release; Case No. 13 for August 1976 meteorological data; Case No. 14, for August 1977 meteorological data; and Case No. 16 for August 1978 meteorological data. 43 1334 302
a Man-rem exposure to the surrounding population out to a radias of 50 miles from the site boundary was calculated for Case No. 1. Because the resulting man-rem was representative and quite small, this calculation was not repeated for the other cases. For Case Nos. 2,11,15, and 17 the ;tep wise purge scenario assumes 4 days at 50 CFM, 9.5 days at 100 CFM, 6.5 days at 200 CFM, 2 days at 500 CFM, 14.5 days at 1000 CFM, for a total purge duration of 36.5 days to reach MPC levels in the reactor building. Although 50 CFM was selected for the initial purge rate, since the I-131 has decayed substantially,100 CFM can be used for the initial purge rate to stay within Technical Specifi-cation limits as shown in Section 5.1. Fo r Ca se No s. 8, 12, 13, 14, and 16 the release is made in a series of steps in which purge rate each hour is varied in accordance with X/Q weather conditions to meet the following objectives: - Instantaneous Dose Rate (3) < 0.3 mr/hr (at site boundarv)* - Instantaneous Dose Rate (7) < 0.1 mr/hr (at site boundary) - Instantaneous activity release < 1.5 x 100,{3 - Peak hourly isotopic concentration at site boundary < 10 MFC - Maximum purge rate = 1000 cfm - Minimum purge rate = 20 cfm A number of iterations have been necessary to determine the appropriate selection of these limits to achieve f ull purge of the reactor building. The purge routine sets specific dose rate limits and uses real time meteorological input to meet these objectives by varying purge rate. Use of this scenario increases the length of purge, as compared to the technical specification release, but reduces totcl beta and gamma dose. It can be seen that the purge can be completed in 33-49 days using the dose objective release, depending on the release point and meteorology used. To confirm the effect of meteorology for months when the purge is most likely to be carried out, Cases 18 through 20 and Cases 21 through 23 and 26 were run with October and November meteorological data respectively. Cases 18 and 23 use the predetermined purge rate schedule as defined in Section 5.1, Cases 19 through 22 use meteorological feedback data to control purge rate within specified limits on hourly gamma and beta dose rates, MPC levels, purge flow capacity and Tech. Spec. limits, Case 26 uses only the Tech. Spec. limit on isotopic release rate for Kr-85 of 1.5 x 105 x 3p; for determining purge rate each hour. This is a slightly different scenario than the step wise purge increase rate case as in Case No. 2 since in Case 26, the purge rate will change slightly each hour as the KR-85 content remaining in the reactor building is reduced.
- For vent stack elevated releases, the beta dose rate limit is lowered to 0.1 mr/hr.
44 1334 30L3
0 The total integrated whole body dose in person-rem was calculated for Case Nos. I and 21. Because the levels were less than 1 person-rem it was not considered necessary to compleCe this calculation for each of the other cases. 5.2.3 Dose Effect Results Table 5.2-5 compares the actual calculated dose for each release scenerie discussed in -the-previous-cection. Frc= - this -::bic -it c an be seen that the total time to purge the reactor building to within 10CFR20, Appendix S, Table I limits for maximum permissible ce ncentration is in the range of 31 to 49 days depending on the purge constraints bsposed. In all cases, the 10CFR20 limit of 500 crem annual whole body dose is easily met, and the meteorology selected has only a slight effect on total dose when purging is controlled by meteorological feedback. Tabulated in Table 5.2-5 for each case are the average ef fective purge rates, the time to reach M10, peak skin doses at the site boundary due to beta and gamma radiation, the location of the peak doses, and the number of times specified ?imiting conditions are reached for the control-led purge cases. A review of the data presented in Table 5.2-5 indicates that a release made at a rate consistent with the current TMI-2 Technical Specifications will result in a small integrated dose to a person residing at the nearest residence (700 meters from the vent stack). The total whole body man-rem expenditure considaring populations out to a radius of 50 miles frem the site is less than 1 man-rem. By elevating the release point, an immediate reduction in the beta dose at the nearest residence can be achieved, however, the gamma dose is only slightly reduced. The beta dose reduction is primarily a result of increasing the height of the theoretical plume centerline containing the beta emitting Krypton above the nearesc residence. The mean free path of the gamma's however is much greater than that for beta's and gamma radiation dose is caused by a shine-mechanism which is relatively insensitive to the variation in distance between the source and the nearest residence. From Cases 1, 2, 10, 15, and 17, the skin dose due to beta radia-tion for ground level release varies between 24 to 50 mrem with no meteorological feedback to control purge rate. By introducing controlled purging using meteorological feedback to limit purging to specified hourly dose limits, the beta skin dose varies between 4.8 and 18 mrem for ground level releases. For elevated releases at the plant vent stack, the beta skin dose varies between 3.5 to 10.0 mrem with no meteorological feedback to control purge rate. Witn controlled purging (using dose limits and meteorological feed-back) and with elevated releases at the plant vent stack, the beta skin dose is in the 2.9 to 5.6 mrem range depending on the actual meteorology. I334 304 45
Table 5.2-6 gives a summary of the range of various purge scenarios calculated. In all cases, the skin dose due to gamma radiation is below 0.35 mrem at the site boundary and the whole body integrated dose throughout a 50 mile radius from the plant is approximately 1 person-rea for the worst case scenario. The results for elevated releases are all within the 10CFR50 Appendix I guidelines of 15 mrem annual skin dose for "as low as reasonably achievable." In particular, using elevated releases with controlled purging, skin dose can be held to within one third of the IGCFR50 Appendix I guideline limits. Figures 5.2-4, 5, and 6 present results of purge characteristics as a function of time for several of the purge scenarios analysed. The first figure (Fig. 5.2-4) compares the beta skin dose build-up at the limiting site boundary location as a function of tbne during purge for several cases. In all cases, most of the dose is generated during the early portions of the purge program when the reactor building contains the highest Kr-85 activity. After about 20 days, the purge is proceeding at maximum flow rate of 1000 CFM and the incremental dose is small from that point fo rwa rd. Comparing cases 19 and~20, the effect of tightening limits on the hourl-dose objective for beta skin dose can be seen. Although redacing the hourly beta limit from 0.1 mrem /hr to 0.05 mrem /hr lowers the initial dose accumulation, the total dose is not significantly ffected due to the fact that the initial savings in dose is compensated by increased incremental dose later in the purge cycle and an extension of the time at which the 1000 CFM flow rate limit is achieved. Figure 5.2-5 shows the additional characteristic features of the Case No. 21 purge as a function of time. This case is most typical of the recommended purge program for November purging, using meteorological feedback to control purge at specified hourly dose limits with an elevated vent stack release. The figure shows purge flow rate, integrated purged volume, remaining reactor building curie inventory, and accumulated beta skin dose. Again, this figure shows that the majority of the reactor building activity has been released by 20 days and that the remaining purge is completed at maximum flow rate to reach MPC levels. For comparison, Figure 5.2-6 shows the case of November meteorology, elevated vent stack releases but with purge rate controlled only by a instantaneous Tech. Spec. release rate limits for Kr-85. In this case the accumulated skin dose is twice as large as the hourly dose rate limited purge case, the maximu= purge rate is reached - much sooner and the reactor building Kr-85 curie activity drops more rapidiy. 1334 305 46
The total integrated whole body dose in person-rem was calculated for Case Nos. I and 21. Because the levels were less than 1 person-rem it was not considered necessary to complete this calculation for each of the other cases. 5.2.4 Dose Effect Conclusions Table 5.2-5 compares the actual calculated dose for each release scenario discussed in the previous section. Table 5.2.6 gives a comparison of the range of calculated doses for all evaluated meteorologies for use in evaluating alternative purge strategies. From these results the following conclusions are made: 1. All cases evaluated are substantially below the 10CFR20 objective of 500 mrer./yr whole body dose. Although the whole body dose is not shown, it can be taken as approxi- =ately 1.2% of beta skin dose giving results below 0.5 mrem whole body dose for all cases evaluated. 2. All cases evcluated with elevated vent stack release give skin dose belcw the 15 mrem /yr 10CFR50 Appendix I guide-line for "As Low As Reasonably Achievable." Results give beta skin dose in the range of 2.9 to 7.5 mrem. 3. Controlled purging with meteorological feedback reduces site boundary doses by 20 to 507. for elevated release. Results for controlled purge at elevated release give beta skin dose in the range of 2.9 to 5.6 mrem. The worst case dose is approximately 1/3 of 10CFR50 Appendix I guideline. 4 The total integrated person-rem whole body dose is less than 1 person-rem to the population within 50 miles of the reactor. Therefore, in perfo rming cost-benefit studies called for in 10CFR50 Appendix I paragraph II.D, the maximum benefit for other alternatives to controlled purging of the reactor building cannot be greater than 1 person-rem even if these alternatives achieve zero release. 5. The total time to purge the reactor building to maximum permissible concentration levels for Kr-85 is in the range of 30 to 36 days for elevated vent stack release. 6. Purge of the reactor building should be done through the station vent stack, using the meteorological feedback system described, in order to maintain of f-site doses as low as reasonably achievable. 1334 306 47
s TABLE 5.2-1 THI Site Westleer Instrumente Approximate Height Above Tower Base Sensed Recorded Rec o r de r e * * (It) Pa rame te rs Pa rame t e ra Analog Digitaf Instrument Description IDOA Kind speed & direction Wind speed & direction Estealine Angus Model Varian-Vil mint-Wind speed la Teledyne Model 50.l. Lill25 servo type two computer with auto-Aacmometers are three cup with a 1008 Wind speed & direction Wind speed & direction channel strip chart natic transmittal thneshold of 0.75 mph and a dia-recorder for each level. via phone line to tsnce constant of 5 ft. Accu racy Accuracy 0.11 of full disc storage. is 0.15 mph or II, whichever is scale. Eleven inch greater. Direction is measured chart width. using Teledyne 50.2. (Nic k-2 vanes are used witle tiireshold of 0.93 aph, a distance constant of 3.7 ft. and damping ratto of 0.4 at initial attack angle of 80'. Direction accuracy la 4 2*. 100 Vertical & horizontal Vertical & horisontal Same as above Same as above R.H. Young-Blvane Model [7Ud}C5 angle angle Threshold 0.4 mph, delay distance for 50% recovery is 3.2 ft. dampipg retto is 0.53. 150A Temperature T 150-31A ft Westronice Model HlI_D2. Same as sbove Dual Rosemont platinion 4 wire Twelve channel Selec-RTDs Model 104 HP (calibration 150s Temperature T 150-338 ft tronic potentiometric traceable to NBS) with 414 L dot printing recorder. linear triple bridges. Accu racy 334 Temperature-reference Ambient temperature 33 f t Accuracy better than (RSS for ayaten) le 0.11F. for comparison with 0.3% of full scale. other level. 338 Temperature-reference Ambient temperature 33 f t for comparison with other level. Cround level Rainfall Ra.nfall Same se above delfort Hodel 59I5 weighing rain gauge. Accuracy 1/21 of full scale. 140A Dew point temperature Dew point temperature $ame as above Same as above EG&C Hodel l10 SH-tliermoelectric ~~~~ dew point system. Accuracy +.5F. 4508 Dew point temperature Dew point temperature ~ U (,na 150A Ambient temperature Amhient temperature h 1505 Ambient temperature Ambient temperature ~^d TAmb 340 -T p 34o y O '%%j TAmh 33-Typ 33 f
- All parameters are continuously recorded.
TABLE 5.2-2 Input Data For Dispersior. Modelling METEOROLOGICAL DATA Parameter Characteristics Wind speed Measured at 100 ft, adjusted to 33 ft level for ground releases and adjusted exponentially to 160 ft for elevated releases Wind direction Measured at 100 f-Stability Based on At 150-33 ft and PG (A-G) dispersion categories in accordance with Regulatory Guide 1.23 ft(dispersion coefficient) Based on PG curves, limited to 1000m Site Specific Data Terrain height See Table 5.2-3 Population distribution See Figures 5.2-1 and 5.2-2 Plant Specific Data Vent Height 160 ft Vent exit diameter 3.0m Vent exit velocity 9.1 m/see Building Height for 170 ft computation of I,, the effective vertical disper-sion coefficient 1334 308 49
s TAltLE 5.2-3 Approximate Terrain Elevations (meters) Within 12 Miles of TM1 Site
- Distance (meters)
N NNE NE ENE E ESE SE SS!: S SSW SW WSW W WNW fM NNW 300 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 600 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 e 1,000 0 0 6 6 6 5 5 5 0 0 0 0 0 0 0 0 1,500 10 25 20 20 25 20 35 30 0 0 0 0 0 0 0 0 2,000 60 75 35 40 25 30 50 50 0 0 35 65 65 65 63 50 4,000 i 60 75 75 65 35 65 65 60 65 75 90 85 65 95 65 50 8 6,000 75 90 90 65 60 90 65 60 65 75 130 125 65 162 65 50 8,000 75 190 130 80 105 110 65 65 65 130 165 150 168 203 65 70 12,000 80 200 155 80 110 110 65 70 130 152 305 150 310 310 75 80 20,000 100 200 200 100 130 120 75 305 170 152 310 310 350 350 100 100
- Terrain height for distances beyond 20,000 was assumed to be unchanged f rom the last value since higher elevations would result in little change in dispersion at greater distances.
LN h C<
i e TABLE 5.2-4 Assumptions for Purge Dose Calculations Rele ase Meteorological Program Run No. Purge Scenario
- Point ***
Data Apolied** Used 1 Constant Ci/sec over 1 month to release GND July '76 XDCALC all containment activity. 2 Tech. Spec. Limit Purge Scenario at 100, GND July '76 XDCALC 200, 500, 1000 CFM per Section 6.1. 8 Release rate limit to hold hourly 5, Y, GND July '76 PURTST and MPC levels based on meteorology ( - limit = 0.3 mrem /hr). 10 Hourly release rate limit at 1.5E5 GND July '76 PURTST 3 instantaneous Tech. Spec. M /sec. limit. 11 Same as No. 2 VENT July '76 XDCALC 12 Same as No. 8 ( r limit = 0.1) VENT July '76 PURTST 13 Same as No. 8 ( A limit = 0.3) GND Aug. '76 PURTST 14 Same as No. 8 ( f limit = 0.3) GND Aug. '77 PURTST 15 Sane as No. 2 GND Aug. '77 XDCALC 16 Same as No. 8 ( 3 limit = 0.3) GND Aug. '78 PURTST 17 Same as No. 2. GND Aug. '78 XDCALC 18 Same as No. 2 VENT Oct. '78 XDCALC 19 Same as No. 8 (.4 limit = 0.1) - VENT Oct. '78 PURTST 20 Same as No. 8 ( - limit = 05, 50 CFM lower VENT Oct. '78 PURTST flow) 21 Same as No. 8 ( 3 limit = 0.1) VENT Nov. '78 PURTST 22 Same as No. 8 (.4 limit = 0.3) GND Nov. '78 PURTST 23 Same as No. 2 VENT Nov. '78 XDCALC 26 Hourly release rate limit at 1.5E5 VENT Nov. '78 PURTST 3 instantaneous Tech. Spec. M /sec. limit. All cases except No. I limit purge rate to 1000 CFM maximum.
- As santhly meteorological data changes from July to August, September or later, short lived isotope initial inventory is reduced based on decay half-life.
- GND = Ground level equivalent roof filter release point, VENT = Vent stack elevated releasa point.
- As santhly meteorological data changes from July to August, September or later, short lived isotope initial inventory is reduced based on decay half-life.
f33.f jjQ 51
TABl.E 5.2-5 Results of Purge Dose Calculations Avg Peak Pop-(I) Pop-(I) Wiiole Body Numbe r o f Ti me s Purge Time 9 Skin ulated Peak ulated Population Condition Limited Flow 4 Run Rate HPC is Dose Location [ ILse Location Dose MPC Tech Max fiin No. (CFM) reached (mrem) (m, dir) (mrem) (m, dir) (person rem) Gamma ik? t a 1imit Spec Flow Flow I 30d 50.0 700,E 0.35 700,E 1.02 2 36d 43.0 700,E 0.35 700,E 8 489 49d 13.0 700,E 0,12 700,E 0 392 0 211 746 84 10 643 31d 35.0 7CC,E 0.35 700,E 0 0 343 393 0 11 36d 6.4 1500,SE 0.i0 700,E 12 534 33d 5.6 1500,SE O.10 800,ESE 0 50 0 311 1072 0 13 491 41d 4.8 700,E 0.02 700,E 0 343 0 183 858 57 14 500 42d 12.5 700,E 0.08 700,E 0 362 0 188 648 18 E 15 36d 25.0 700,E 0.11 700,E 16 500 4 tid !0.0 700,E 0.05 700,E 0 394 0 162 528 12 17 36d 24.0 700,E 0.10 700,E 18 36 3.5 19 34 2.9 0.04 0 183 0 210 406 2 U 20 56 2.9 0.03 U h 21 36 5.0 0.06 0.'5 0 165 0 210 411 71 U 22 4b 18.0 0.09 23 36 7.5 26 32 10.0 0 0 0 316 416 0 (1) Location of highest dose to resident (meters, directica).
TABLE 5.2-6 Comparison of Dose for Purge Scenarios Peak Boundary 'a' hole Body Case Purge Skin Dose (wrem) Pop. Dose Nos Scena-io BETA GAMMA (person-rem) 12,19,21 Met. Feedback, Hourly Limit Beta = 0.1, Variable Meteorology, Vent Stack Release 2.9-5.6 0.04-0.10 0.75 11,18,23 Tech. Spec. Purge Schedule at 100, 200, 500, 1000 CFM; Variable Meteorolocy, Vent Stack Release 3.5-7.5 0.1 8,13,14 Met. Feedback, Hourly Limit Beta = 0.3, 4.8-18.0 0.02-0.12 16,22 Variable Meteorology, Ground Level Release 2,15,17 Tech. Spec. Purge Schedule at 100, 200, 24.0-43.0 0.10-0.35 1.0 500, 1000 CFM; Variable Meteorology, Ground Level Release LIMITS 10CFR20 Objective 500 mrem /yr. whole body
- 10CFR50 APP. I Guideline 5 mrem /yr. whole body
- FOR ALARA 15 mrem /yr. skin dose (BETA & GAMMA)
- Note that whole body is approximately 1.2?. of beta skin dose for Kr-85 therefore limit will not be approached so long as skin dose is met.
}2 53
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o Figure 3.2-3 ,a Generalized Flow Diagram for PURTST Progra= Read input These includes parameters Maximu= & -iniscm purge flow Beta and ga==a dose limits ^ w Factor allcwed abcVe MPC l Read met data for each isctope for the hour Assu=ed teen spec limit Starting uCi/cc value v Compute new isotopic in-ventory v Computo doses 1Ccmpute new for the hour ) flew to equal for 100CFM dose limit purge Compute frac-hemputenew tion of MPC " flow to equa' No for each iso-MPC limit j n e for 100 top Compute flow tc Yes Is time --) meet tech spec 6, Stop release rate limit eache Addtoaccumu-l Find most lated RC limiting ) at.osphere flow rate released
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o 6.0 OPERATIONAL EFFECTS OF PURGE OPERATION In order to complete the assessment of the radiological impact of purging the reactor building atmosphere using the hydrogen control system flow path, the filter dese rate has been evaluated to show acceptability for on-site exposure. The filter dose rate analysis for the purge process was completed assuming that the recirculation syster was not operated. To be conservative, airborne activity was assumed to include radio-isotopes present as of July 1, 1979. At the time of actual purge, the shorter lived I-131 will be many orders of magnitude lower. During a RB purge using the systems in the modes as described in this report, the most significant man-rem expenditure to personnel on site will be incurred during filter change out act ivit ie s. The following evaluation defines the change out sequence, estimates the required man-hour expenditures and utilizing the theoretical dose rate derived from the predicted curie buildup in the filter trains from the TIDRLS program, quantifies the total man-rem expenditures. 6.1 Method The computer code TIDRLS calculates time dependent radioactive transport into and out of a single node. The node used in this analysis was the TMI-2 containment building. It is a versatile code that may be used for ventilation studies, associating reactor primary coolant activity with an unidentified leakage rate for setting condit ions of operation, release of radioactivity, dose rate to personnel in control room, filter inventories of fission products and so on. Individual isotopes and in' ial concentrations are read into the code which then identifies the f amily to which it belongs and per-fo rms all calculations for both parent and daughter isotopes. Pro-visions are made to include recirculation and purge from the node. The node volume is assumed to have separate liquid and vapor regions and each isotope may be assigned a partition factor for transport between liquid and vapor regions. All gaseous daughters of isotopes in the liquid region are transported to the vapor region automatically. A gaseous daughter of a paren! isotope trapped in a filter is released from the filter. The radiological assessment calculation for filter change out used the TIDRLS code with recirculation through HEPA filters only (no iodine removal). The code was run for one hour intervals in the recirculation mode and filter fission product inventory at each time step was obtained. The individual isotopes and their initial concentrations were based on actual air sample data. A review of the fission product inventory at the time of the accident and their respective decay schemes determined that additional isotopes beyond 6 1334 319 O
e o those identified by gamma spectroscopic analyses performed to date do not reasonably exist. However, additional isotopes were assumed to be present in the filter change out dose rate calculations to develop a worst case calculation. The inventory of any isotope assumed to be present was at its minimum detectable concentration for this worst case. The dose rate calculation assumed the fission products trapped by the filters were evenly distributed across the face of the filter, and the filter was a disc source. The dose rate is given by the equation ^ D = Ko(E) E -- [ E l ( b ) - E1 (b sec 0)] where D = dose rate (r/hr) 2 see) Ko(E) = dose rate per unit energy flux (r/hr per Mev/cm E = photon energy (Mev) b = ootical length (cm) E (b) = exponential integral l 9 = angle between center of filter and point of measurement A source (photon /cm2 - sec) S = The calculation of dose rate was done for each gamma pres:ent in the filter. A filter ef ficiency of 90" was used in calculation of filter fis-sion product accumulation. 6.2 Filter Dose Evaluation 6.2.1 H2 Control Filter Change Out During the purge operation using the Hydrogen Control System, both HEPA and charcoal filters will be used. After being bagged nearby the filter housing, the filters will be transported to the equipment hatch. One man will be required to handle the REPA filter while two men are required to handle the charcoal filters because of their greater weight. 6.2.2 Filter Dose Rates Using the best estimate RB airborne activity dose estes on the the H2 control filters are calculated. The purge rate was based on the Tech. Spec, scenario given in Section 5.1. Assuming a purge without prior recirculation, the analysis showed that dose rate on the H2 control HEPA and iodine (charcoal) filters would be less than the design changeout point of 1 r/hr at 61 1334 .320
the end of a 35 day purge. This analysis assumed that iodine, gases and particulates in the RB would all be below MPC af ter this purge. Since filter changeout can be accomplished with dose rates up to I r/hr and since the filter dose rates expected are well below this value, the doses received by workers performing the filter removal should be acceptable. 6.3 Results Table 6.3-1 shows the dose rates that will occur on the hydrogen control system exhaust filter, if the reactor building is purged without prior recirculation. The maximum buildup on the REPA filter is 340 mr/hr at the end of the 840 hour purge period. The charcoal filter will have a maximum dose rate less than 1.2 R/hr after about two weeks of operation when July 1979 I-131 concentra-tion are assumed. The dose rate decreases from 1.2 R/hr after two weeks as a result of Iodine-131 decay. At the end ef the purge the dose rate would be less than 350 mr/hr. Using the I-131 levels expected to be present in November 1979, the charcoal filter doses will be several orders of magnitude lower. 6.4 Conclusions The dose rates calculated from best estimates of fission products in containment atmosphere during July 1979 for hydrogen HEPA and charcoal filters should not come close to the IR/hr design basis for filter changeout. Since the dose rate levels do not require filter changeouts during the operation of the purge system, the impact on man-rem expendi-tures will be minimal. 4 62
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o 7.0 ENVIRONMENTAL EFFECTS OF PURCE ACCIDENT The purge accident is discussed in Section 3. This section con-tains the analysis of the environmental effects of the postulated purge accident. The accident analysis is performed consistent with Regulatory Guide 1.145. " Atmospheric Dispersion Models for Potential Accident Consequences Assessments at Nuclear Power Plants." 7.1 Description of Accident The postulated purge accident that leads to worst case radiation dose to the environment is the extreme condition of uncorrected inadvert ent initiation of the modified hydrogen control system at 1000 cfm for 30 minutes before any controlled purging of the reactor building has been completed. The likelihood of this accident is extremely low because of the interlocks and procedures in place that allow purge system operation only when planned. Analysis of this accident using the conservative Regulatory Guide 1.145 meteorology makes the calculated environmental exposure even less likely. 7.1.1. Reactor Building Release During Accident For an assumed 30 minutes purge system flow at 1000 cfm with a building concentration of 1 mci /ml of Kr-85 prior to any controlled purge activity gives a total accident curie release of: 3 4 ft 2.83 x 10 ml 1 UCi 1 Ci Curies = 1000 x 30 min x x x min ft3 m1 106 Ci = 850 Ci released. 7.2 Accident Dispersion Model The accident dispersion parameter X/Q has been computed for TMI Unit 2 using the methodology outlined in the NRC's Regulatory Guide 1.145 which accounts for distance to the site boundary in each of 16 direction sectors and takes into account the reduction in dose due to wind meander under low wind speed conditions. Calibrated SF6 diffusion tests conducted at the site in 1971 (reported in Amendment 24 to the Unit #1 application) demonstrated the existence of this meander effect. Based on two years of site meteorological data, each of which had a combined recovery rate of more than 90%, the appropriate X/Q for use in short-term calculations (i.e., less than two hours) was 3 determined to be 6.8 x 10-4 sec/m. Following is a summary of methods used and input data. f334 j}} 64
Meteorological Data A two-year period of site meteorological data was used with wind speed and direction from the 100 ft level and AT taken between 150 f t and 33 ft. Speed was adjusted to be representative of the 33 ft level utilizing a power law relationsbip with the exponent being determined as a function of stability as follows: I 33 \\" u =u 1 33ft 100ft (H100 ) 0.25 for Pasquill Stability Classes A, 3 and C' where n = 0.33 for Pasquill Stability Class D 0.50 for Pasquill Stability Classes E, F and G and H= height (m) Data recovery (nercent) by parameter was as follows: First Year Second Year Parameter (7/1/76-6/30/77) (7/1/77-6/30/78) Wind speed 96.0 93.4 Wind direction 95.0 93.0 delta-T 95.4 93.5 Combined 93.5 91.0 Diffusion Class The Pasquill dif fusion class was determined using vertical tempera-ture difference (AT) and the categories given in NRC Regulatory Guide 1.23. Values of a and a were determined as a function of y z distance and stability class using the standard Pasquill-Gifford curves. The distance to the site boundary in each of the 16 direc-tion sectors wat variable with direction and was taken as the mini-mum distance to the site boundary in the central or either adjacent sector and are given in Table 7-1. Values of I were computed as follows: y I = Mc (x < 800m) = (M 1)4Y800m * 'Y where x is distance M = from 1.145 figure 3, f (wind speed) Equations Regulatory Guide 1.145 requires the use of three dif fusion equa-tions as follows: 1 (1) X/Q = yz+f } hh )24 u (ra a 65
a 1 (2) X/Q = E ( 3 'Y 'Z ) W 1 (3) X/Q = u'.y q where : 3 X/Q = relative concentration (sec/m ) u = wind speed at 33 ft (m/sec) 7" lateral plume speed coefficient (m) =
- z = vertical plume spread coefficient (m)
A = smallest vertical plume cross-sectional area of 2 containment (~2000m ) For each calculation, the following procedure is used to determine the appropriate X/Q. Determine the maximum of equation (1) and (2). Then determine the minimum of that equation and equation (3) if the wind speed is less than 6 m/sec and the stability is not ' unstable. This value is used for all calculations that follow. Calculations Values of site boundary X/Q were determined for each hour of the two year data bcse using the above equations. Cumulative probabil-icy distributions were then made for each direction and separately for the combined data independent of direction. An envelope was contructed around all 16 direction dependent curves and the 0.5% probable value (i.e., the value exceeded no more than 0.5% of the time) was determined to be 6.8 x 10-4 sec/m3 A second value required by the Regulatory Guide 1.145 procedure at the 5 level on the direction independent curve was determined to be 6.3 x 10-4 sec/m3 According to the Regulatory Guide, the first value must be used since it is higher. 1334 325 66
Table 7-1 Assumed Distance to Site Boundary in Each Direction Direction (from plant to site boundary) Distance (m) N 650 NNE 650 NE 630 ENE 610 E 610 ESE 610 SE 625 SSE 710 S 925 SSW 705 SW 705 WSW 705 W 1400 WNW 1400 NW 650 NNW 650 1334 326 67
7.3 Environmental Dose Consecuences From Purge Accident From Sections 7.1 and 7.2, the site boundary cloud concentration becomes: Curies, 850 Ci x 6.8 x 10-4 sec/m3 3 60 sec. M 30 min x g = 3.21 x 10-4 Ci/m3 From Regulatory Guide 1.24, the beta and gamma air dose for Kr-85 act ivity becomes: Beta Air Dose (mrads) = 0.23 x 0.67 x 3.21 x 10-4 x 1800 sec x 103 E = 98 mrads Gamma Air Dose (mrads) = 0.25 x.0052 x 3.21 x 10-4 x 1800 x 103 = 0.75 mrads Using Regulatory Guide 1.109 Table B-1 to convert from air dose to whole body dose gives: whole body dose due to accident 0.73 mrem. = This whole body dose is compared to 10CFR100 limits of 25000 mrem maximum allowable total radiation whole body dose to demonstrate that the accident consequences are well within 10CFR100 accident limits. 1334 327 68
8.0 ALTERNATIVE TO REACTOR BUILDING PURGE PROGRAM As discussed in Section 1, the unknown Three Mile Island Unit 2 core configuration poses a small but incalculable risk. It is possible that reactor bui". ding entry could be made and additional reactor building investigations completed (all necessary steps toward final disposition of the damaged reactor core) without cleaning-up the air, borne radioactivty in the reactor building. Because of the hazards involved and added precautions required when operating in an airborne contaminated environment, several methods were examined for clean-up of the airborne accivity prior to recommending the controlled purge option. This section discusses each of the evaluated alternatives. 8.1 No Atmospheric Clean-Up It is often tempting to conclude that no further action should be taken to reduce airborne radioactivity inside the reactor building. In view of the other acceptable alternatives available, the decision to take no action is not justified. The current level of activity is 80,000 times the maximum permissible concentration for restricted access per 10CFR20 Appendix B Table I, column 1, for Krypton-85. Since the activity is well defined, it may be possible to develop adequate shielding for a reactor building entrant to complete some assessment of reactor building conditicas with this high radiation environment. The risks to the entrant are quite high however and the opportunity to obtain useful information is very poor in this condition. More importantly, as activity continues in and around the reactor building, the likelihood of unplanned accidental releases of the contained gases under conditions of undesireable meteorology remains high. Although the reactor building is presently adequately sealed, the ability to maintain this sealed system indefinitely is questionable. 8.2 Design Basis for Alternate Atmosphere Treatment and Starage Systems The design bases for the alternate systems considered in this section are as follows: Current Noble Gas Activity Within Containment The noble gas activity within containment at this point in time consists entirety of the isotope Kr-85, with a half-life of 10.7 years. All other radioactive isotopes of xenon and krypton have decayed to negligible quantities. The concentration of Kr-85 within containment has been determined, based on sampling performed during the Summer of 1979, to be less than I mci /ml. bk )f0 69
Recuired Concentration of Kr-85 After Cleanuo The systems should reduce the Kr-85 concentration to the maximum permissible concentration (MPC) for occupational exposure, or 1 x 10-5 ;Ci/ml (see 10CFR20). While some work could proceed with j higher concentrations of Kr-85, a concentration of 1 MFC or less is considered essential for extensive recovery work inside of the containment building. Containment Volume and Reauired Process Volume The containment volume is two million cubic feet. At least 11.5 containment volume 9 must be processed using a " bleed and feed" type of operation to reduce the containment Kr-85 concentration to 1 MPC. This amounts to 23 million cubic feet of processed volume. It should be noted that the process volume is based on the assump-tion of perfect mixing of clean incoming gas with the gas ins ide of containment. A higher process volume could be required if this assumption is not realized. Desizn Bases Release of Kr-85 from the Site The design basis for the gas compression ar.a charcoal absorption systems is zero off-site release. Zero release inherently cannot be achieved by the cryogenic system, which can remove a part but not all of the krypton from the process ~ stream. The design basis off-site release for this system is 10-3 of the total containment inventory, or about 60 curies of Kr-85. This value is based on the removal efficiency specified in the existing system equipment specification. Seismic Design Category The seismic category for system components, supports, and buildings is Class 1. This category is the same as that for the TMI-2 con-tainment building and, accordingly, is considered required for components which would -catain the Kr-85 if it were transferred from containment. It should be noted that Regulatory Guide 1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Wcter Cooled Nuclear Power Plants," imposes less stringent seismic design requirements on current gaseous radioactive waste systems. This Regulatory Guide is not, however, considered appropriate for the situation at TMI-2. Design Code The design, fabrication, and installation of pressure boundary components is in accordance with the requirements of the ASME Code, Section III, Division 1, Class 3. Again, this code is more strin-gent than would be required of current gaseous radioactive waste 1334 ')29
systems by Regulatory Guide 1.143. It is considered appropriate, however, because it is consistent with the code for the existing containment vessel. 8.3 Charcoal Adsorotion and Storage System The first alternative considered for reducing the airborne activity in the reactor building is to draw of f the reactor building atmos-phere into a charcoal bed storage container so that the noble gases would remain adsorbed to the charcoal. This charcoal would then remain in storage indefinitely. In order to maintain the reactor building pressure within acceptable limits, the atmosphere is continuously replenished with outside air so the airborne concen-tration is reduced in exponential fashion. In order to achieve MPC levels within the reactor building the equivalent of 11.5 times the reactor building atmospheric volume must be processed. 8.3.1 System Description (See Figures 8.3-1 and 8.3-2) Gas withdrawn from containment is passed through HEPA and charcoal filters and then gas dryers which are needed to remove essentially all moisture. The gas then passes through tanks of activated charcoal in series which absorb the Kr-85. Once " break-through" occurs, the tanks are isolated and are used for storing the Kr-85 at ambient temperature end atmospheric pressure. Charcoal loses its ability to absorb krypton when it is exposed to significant humidity, i.e., in excess of about 3 percent. The total required charcoal weight is 34,000 tons (this represents approximately 40 percent of the total U. S. annual production). The charcoal volume is 2,000,000 cubic feet, which is equivalent to the TMI-2 containment volume. Storage tanks, rather than piping, are used to facilitate initial loading of the charcoal. A manhole would be required at the top of each tank for loading, and a second manhole would be required at the bottom of the tank for eventual disposal of the charcoal. Each tank would be provided with isolation valves, primarily for humidity control during filling operations. The valves would also be closed once containment cleanup operations were complete. The size of tank selected was based on fabricating the tanks in a shop and shipping them to the site. Twelve-foot diameter and 50-foot length represent about the upper Ibnit to shop fabricated tanks. Four hundred and fif ty such tanks would be required. Each would be an atmospheric tank, designed in accordance with Section III, Subsection ND, Class 3 components. The Code requires a minimum wall thickness for such tanks of 3/16 inch. The total tank metal weight would be 6,100,000 pounds. 1334 530
The building required to house the tanks (see Figure 8.3-2) would be 700 feet long, 150 feet wide, and 60 feet high. The charcoal provides significant self-shielding so that shielding is not con-sidered necessary for the tanks. 8.3.2 Design Alternates Considered Several alternates were considered for the charcoal adsorption system, and were rejected. In summary: a. Operation at Lower Temperature Sum =ary description of alternate -- Operate the charcoal at a lower temperature to increase its adsorption capa-bility and, accordingly, decrease the required amount of charcoal. A number of BWRs, for example, employ systems which operate at 0*F. At this temperature, the absorption capability is about 2.5 times greater than at 70*F. Basis for rejecting the alternate -- The required refrig-eration equipment increases system complexity. Malfunction of the equipment could cause an increase in charcoal temperature and therefore cause an uncontrolled release of Kr-85. The disadvantages are considered to cutweigh the advantage of decreased charcoal volume, particu?arly since the required amount of charcoal with a refrigerated system would still be very large (about 15 percent of the . total U. S. yearly production). b. Renenerate the Charcoal and Storr Gas Enriched in Krvoton Summary description of alternate -- Employ two parallel trains of charcoal, each with a few days holdup time for krypton. Process with one train until " break-through" occurs, while regenerating the alternate train. Store the regeneration product gas, which is enriched in krypton, using separate storage vessels. Potential regeneration techniques which are under development include (1) cycle the bed temperature, and (2) cycle the bed pressure. Basis for rejecting the alternate -- Laboratory scale tests show that such a system is potentially feasible, particularly temperature cycling systems which operate at cryogenic temperatures. However, such systems have not been employed for large scale applications, so that further engineering and development would be required. Accordingly, such systems are noc considered practical for near term use at IMI-2. 8.3.3 Cost and Schedule Estimate Cost Range (for component procurement, installation, building erection and materials, design and analysis, testing and checkout, charcoal, contingency) 1334 331 72
S120,000,000 to $160,000,000 (Note: More than S60.000,000 of this cost is the charcoal itself.) Schedule Range (for build;ng and equipment design, procurement, erection, inscallation, testing) 30 months to 40 months (Note: This schedule presumes charcoal would be available as required. A national commitment of U. S. production capacity would be required for this. ) 8.3.4 System Evaluation The charcoal adsorption system achieves full treatment and storage of the reacto'r beilding atmophere with zero radio-active release assuming no equipment failures or operator errors. This system is vulnerable to uncontrolled release during processing and long-term storage. In particular, charcoal lo'es its adsorption capability when exposed to moisture. Accordingly, gas dryer malfunction during processing, or contact of the char;oal by humid air during storage could result in inadvertent krypton release. Potential for fire also exists with charcoal, and could result in an un-controlled release of krypton. Use of a charcoal system does not resolve the problem of ultimate disposal of Kr-85. Long-term storage for more than one hundred years, and of f-site shipment are each considered less safe than controlled release of the Kr-85 by purging containment. Off-site shipment would be particularly impractical for this system because of the large volume of material. The extensive time required to build and install the charcoal adsorption system would in:rease the likelihood of inadvertent and uncontrolled leakage from the existing containment building, and thereby cause higher exposure to personnel. This extensive time delay to complete system installation would also delay TMI-2 cleanup operations. Finally, the cost of the charcoal adsorption system is high and no commensurate benefits are received. 8.3.5 Charcoal Adsorption System Conclusions In s'2mmary, when compared to controlled purging of the con-tainment building, the alternate charcoal absorption system is considered to be less safe -- it is less reliable, and clearly has the potential for uncontrolled releases of radio-activity with higher radiation exposures. 1334 332
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8.4 Gas Compression and Storage System The second alternative considered for reducing the airborne activ-icy in the reactor building is to draw off the reactor building atmosphere into a pressurized storage container so that this entire building atmosphere including the radioactive noble gases remains in pressurized storage indefinitely. The total volume to be stored is 23 million cubic feet. 8.4.1 System Description (See Figures 8.4-1, 8.4-2, and 8.4-3) Gas is withdrawn from containment using three compressors with a total capacity of 225 scfm. This permits containment cleanup in 71 days if the system ooerates with no malfunctions and if the total process volume does not exceed 23 million cubic feet. The gas passes first through HEPA and charcoal filters which are provided to remove any particulate radioactivity and minimize contamination of downstream components. Such filters would be required for each of the containment cleanup systems including the purge systcc. Accordingly, the cost and schedule associated with these filters were excluded from the evaluation. The storage container for the compressed gas is 36-inch 0.D. standard wall (0.375-inch thick) carbon steel piping. The design pressure for this piping is about 340 psig in accordance with the ASME Code, Section III, Subsection ND. At this pressure, a total pipe volume of 1,000,000 cubic feet is required for storage of the processed gas. The total required length of 36-inch piping is 150,000 feet. The pipe weight is 21,000,000 pounds. The piping is divided into two major sections to minimize shielding. The high activity piping section includes 20 percent of the piping ano contains 90 percent of the krypton-85. Six inches of concrete shielding are required. The high activity section is subdivided into five units to (1) ensure that the highest activity piping is at the center of the building (see Figure 8.4-2) and, accordingly, is shielded by outer piping; and (2) minimize the amount of uncontrolled Kr-85 release in the event of leakage. The Luilding which houses the high activity piping and the gas compressors is 260 feet long, 90 feet wide, and 30 feet high. A low activity pipe section contains 80 percent of the total piping and 10 percent of the krypton-85. No shielding is required for this piping. The building which houses the low activity piping is 220 feet long, 160 feet wide, and 60 feet high. 8.4.2 Design Alternates Considered Various alternates were considered for the gas compression system, and were tsjected. In summary: a. Storage in Higher Pressure Piping Summary description of alternate -- Store the containment atmosphere in high pressure piping, in order to reduce the }}}4 j}} 76
total storage volume. For example, thick-walled (1.0 inch) 36-inch piping would permit storage at 1,070 psig, and reduce the storage volume (or total pipe length) by a factor of three. Basis for rejecting the alternate -- The total pipe weight is not reduced by this alternate; the reduction in pipe length by three is balanced by the increase in wall thickness. Accordir. gly, pipe procurement costs would not be reduced. -- Standard wall piping is the most readily available. Accordingly, use of thick wall piping woald increase construction time. -- The likelihood of uncontrolled leakage, e.g., through system valves, is increased at higher pressure. b. Uje of a Single Large Storace Container Summary description of alternate -- Employ a large container instead of piping. For example, a vessel with 2 x 100 ft3 volume (equal to the TMI-2 containment volume) could contain the processed volume at a pressure of about 170 psig. The required wall thickness for such a vessel fabri-cated of carbon steel would exceed 8 inches. Basis for rejecting the alternato -- Such an alcernate would likely be significantly more costly and take longer to construct than a system which employs standard wall piping. c. Use of Standard High Pressure Gas Storage Bottles Summary description of alternate -- Employ standard 2,500 psig vessels which are used for storage and transport of commercial gas (e.g., A2 and H ). 3 Bacis for rejecting the alternate -- More than 80,000 such vessels would be required. The pipe and valve system for filling these bottles would be very complex and, accordingly, the likelihood of uncontrolled leakage would be significantly increased. 8.4.3 Cost and Schedule Estimate Cost Ranga (for component procurement, installacion, building erection and materials, design and analysis, testing and checkout, contingency) S50,000,000 to S75,000,000 Schedule Range (for building and equipment design, procurement, erection, installation, testing) 25 months to 35 months !334 336 77
s 8.4.4 System Evaluation The gas compression system achieves full treatment and storage of the reactor building atmosphere with zero radioactive release assuming no equipment failures or operator errors. Storage of krypten at high pressure for long periods of time in 28 miles of piping increases the likelihood of uncontrolled release compared to purging containment. While the dose rate resulting from such uncontrolled release is within the NRC limits in 10CFR100, it is relatively high. In particular, the dose at the site boundary due to release of gas from one of the five high activity storage secticas is about 2,500 mrads. Rele ase of all the gas would result in a corre;pondingly higher dose. It is considered that this system would not resolve the problem of ultimate disposal of the Kr-85. In particular: -- On-site storage for more than one hundred years would be required to reduce Kr-85 activity to the low levels which can be achieved by purging. The likelihood of an inadvertent uncontrolled release during this long period of time is significant. -- Off-site shipment of the compressed gas would incresse still further the likelihood of an inadvertent release, i.e., as a result of a shipping incident. (Off-site shipment would require additional equipment, not evaluated in this study, to reduce the volume of gas to a practical value). In addition, of f-site shipment would greatly increase the populated area which could be potentially af fected, and would create unnecessary problems. Venting the storage system over a long period of time, e.g., five to forty years, only varies the rate of personnel exposure compared to purging -- the total dose received is not changed. This would likely not resolve concern with release of the Kr-85.
- Further, equipment would be required to function reliably for a longer period of time than for purging and, accordingly, the likelihood of an uncontrolled release would be increased.
The extensive time required to build and install the gas compression system would increase the likelihood of in-advertent and uncontrolled leakage from the existing containment building, and thereby cause higher exposure to personnel. This extensive time delay to complete system installation would also delay TMI-2 cleanup operat ions. Finally, the cost of the gas compression system is high and no commensurate benefits are received. 1334 537 78
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8.4.5 Gas Compression System Conclusions When compared to controlled purging of the containment building, the alternate gas compression system is considered to be less safe -- it is less reliable and clearly has the potential for uncontrolled releases of radioactivity with higher radiation exposures. 8.5 Crvonenic Processing and Storace System The third alternative considered for reducing the airborne activity in the reactor building (primarily due to Krypton-85) is to draw off the reactor building atmosphere into a cryogenic processing system. This system would separate the noble gases from the re-maining gases cryogenically and the noble gases containing Krypton would be stared indefinitely in highly concentrated form. The total volume to be processed through the system is 23 million cubic feet. Gas removed from contair. ment passes through a cryogenic treatment system where most krypton is removed. The purified gas is dis-charged from the plant. (Note: The cryogenic unit effluent gas flow rate is greater than the input flow rate, because liquid nitrogen used for cooling vaporizes in the cryogenic units. The effluent must be discharged rather than recycled to containment in order to prevent containment pressure buildup.) Liquid krypton, xenon, argon, and methane are periodically withdrawn from the system, allowed to vaporize, and are stored at amb ient temperature in storage vessels. 8.5.1 System Description (See Figures 8.5-1, 8.5-2, 8.5-3, and 8.5-4) Gas withdrawn from containment is passed through HEPA and charcoal filters and then through the various remaining components which are shown in Figures 8.5-1 and 8.5-2. It is noted that all of the components shown in Figures 8.5-1 and 8.5-2, except for the cata-lytic reccLbiners and their associated preheaters and af tercoolers, are part of an existing system at a new BWR. The system has not been placed in operation. It is being scrapped and replaced by a new conventional type of charcoal system. We understand the major reasons for this decision by the utility which currently owns the systen are: Lifetime costs of the cryogenic system, including the cost of hydrogen and liquid nitrogen supply, maintenance, and operation, were considered to likely exceed the cost of a charcoal system. The cryogenic system was considered ill-suited for transient operations. For example, its ability to respond to sudden changes in input flow rate, such as could be caused by opening a vent valve to the main condenser, is questionable. 3dl 82
No significant operating experience is available on a cryo-genic system. It was considered likely that operation of this system would be significantly less reliable than a char-coal system because it contains many more valves, instruments, and other active components than a charcoal system. However, this is one of the cryogenic systems which could be made available in a reasonable period of time, and was theref)re chosen as a typical cryogenic system for this evaluation. The cryogenic system consists of three separate trains. The input flow rate is 75 scfm per train. After removal of oxygen by the recomb ners, the flow rate is 62 se fm. The ef fluent flow rate is 103 scfm, higher than the input, because some of the liquid nitrogen used for cooling the cryogenic units is vaporized in the units. The purified gas is discharged from the site via the reactor building roof vent. The cryogenic system can remove 99.9 percent of the krypton from the input gas in accordance with the original equipment specificat ion. The quantity of Kr-85 disenarged is approximately 60 curies. Shop tests were performed to establish the purification ef ficiency of the cryogenic units. These tests showed a removal efficiency by krypton greater than the value of 99.9 percent required by the equipment specification. However, the test was not performed under actual operating conditions. For example, pure nitrogen was em-ployed for the process gas rather than 'a gas containing moisture, carbon dioxide and argon in addition to nitrogen. Also, the bottom of the removal column (the "reboiler section") contained a mixture of nitrogen and krypton rather than a mixture of argon, methane, and krypton. Accordingly, it is considered that testing under actual eperating conditions would be required to prove out this system. Liquid krypton and xenon are removed from the cryogenic unit when they reach a concentration of 20 percent in the bottom of the removal column. The remaining 80 percent consists of argon (76 percent) and methane (4 percent). The liquified gases are vapor-ized and stored at amb ient temperature. The volume of stored gas would be about 800 standard cubic feet. This estimate is based on a concentration of Kr-85 within contain-ment of 1/4Ci/m1, which for a containment volume of 2,000,000 cubic feet, amounts to about 60,000 curies of Kr-85. This is about 60 percent of the total fuel inventory of Kr-85 expected after 90 days of operation. The total inventory of noble gas expected af ter 90 days of operation is 256 standard cubic feet. The estimate of 800 standard cubic feet total is, accordingly, the volume corresponding to about 60 percent of the total noble gas volume in a 20 percent rich mixture of noble gas with argon and methane. 1334 342 83
Many different components are employed for this system (see Figures 8.5-1 and 8.5-2). The function of each component is summarized as follows : Catalytic recombiners (Figure 8.5-1) -- Remove oxygen from the incoming air to prevent ozone buildup from irradiation of oxygen. Ozone in contact with light hydrocarbons, e.g., CH, 4 can detonate. (Note: The system supplier indicated this is a concern for operating BWR applications. He would have to per-form an evaluation to determine if catalytic recombiners are required for the TMI-2 application.) Hydrogen storage vessels (Figure 8.5-1) -- Provide hydrogen to the catalytic recombiners, 8,000,000 scf total required. Liquid nitrogen storage vessels (Figure 8.5-1) -- Provide liquid nitrogen for cooling the cryogenic units, 150,000 gallons total required. Krypton and xenon storage vessels (Figure 8.5-1) -- Store the Kr-85. Storage secondary container (Figure 8.5-1) -- Prevent Kr-85 release in the event of storage vessel, piping, or valve failure. Cryogenic unit feed compressors (Figure 8.5-1) -- Provide the required gas flow. Trace recombiners (Figure 8.5-2) -- Remove trace quantities of oxygen (up to 0.5 percent by volume) which may be present in the inlet gas. Prepurifiers (Figure 8.5-2) -- Remove water vapor and carbon dioxide from the gas stream to prevent plugging of the cryogenic columns. Cooldown heat exchanger (Figure 8.5-2) -- Reduce temperature of inlet gas (to -292*F) and increase temperature of outlet gas (to -40*F) and hydrogen. Removal column (Figure 8.5-2) -- Remove krypton from input gas (also Xe, A, and CH ) by counter flow of liquid nitrogen (at 4 -307'F) and the inlet gas. Condenser heat exchanger (Figure 8.5-2) -- Liquify gas output from the removal column. Phase separator (Figure 8.5-2) -- Remove excess hydrogen for recycle to the catalytic recombiners. Decay column (Figure 8.5-2) -- Provide three-hour decay time of the effl.mnt gas before it is released. f 334 id) 84
Cold box (Figurt 8.5-2) -- Contain all components which operate at cryogenic tet:peratures to prevent uncontrolled Kr-85 release in the event of equipment malfunctions. Ambient heater (Figure 8.5-2) -- Heat up effluent gas to the temperature required for prepurifier bed regeneration (+330*F for H O and CO2 removal). 2 Figure 8.5-3 shows a conceptual de.ign of the secondary storage con-tainer for t'... krypton storage vessels. This is, in effect, a small size containment vessel with two-foot thick reinforced con-crete walls, a stainless steel liner, typical piping penetrations with double isolation valves for the inlet and outlet headers, and a four-foot diameter equipment hatch for storage vessel installa-tion. It is designed for an internal pressure of 20 psig to with-stand the pressure resulting from f ailure of all the storage vessels. (This peak pressure would be 16 psig.) Figure 8.5-4 shows the building arrangement for the system equipment. It corresponds to the arrangement of the existing system, with .Tinor modifications to: incorporate the secondary storage container, Provide space for catalytic recombiner equipment, and Provide for above grade construction rather than below grade construction which was employed at the existing f acility. Two feet of con. crete shielding are required for the product storage vessels and the cold box components. 8.5.2 Design Alternates Considered Nc alternate designs were considered because, as described above, the evaluation is based on an existing design which is currently available. (Note that even though this is an existing system, there is no actual operating experience with this system, or with similar equipment at any commercial light water power reactor.) 8.5.3 Cost and Schedule Estimates Cost Range (for component procurement, installation, building erection and materials, design and analysis, testing and check-out, utilities, contingency) S10,000,000 to S15,000,000 (Note: This cost presumes that the cryogenic units will be available as surplus at a small fraction of their original cost or of the cost of new equipment.) Schedule Range (for building and equipment design, procurement, erection, installation, testing) 20 months to 30 months 1334 .544 85
8.5.4 System Evaluation This system is less costly and would require less time to in-stall than the gas compression or charcoal systems. It is considered, however, to be the least safe and most unreliable of any of the alternate systems evaluated for a number of reasons. -- The system produces highly concentrated Kr-85. Any leak-age or component failure could result in significantly greater amounts of uncontrolled radioactivity release than the other systems. Also, the radiation levels of the equipment and, accordingly, the exposure of plant personnel during maintenance and operation would be higher than for the other systems. -- The system is subject to plugging as a result of com-ponent malfunctions whica result in inadequate moisture or carbon dioxide removal. More than 100 automatic valves are used for prepurifier regeneration cycle control, and must function correctly. Electrical power supplies to the ambient heaters (for the final pre-purifier regeneration step) and cooling water supplies to upstream aftercoolers must also work. It is con-sidered likely that upsets will occur. In the event of plugging, the system must be thawed and purged to return to operation. The likelihood of uncontrolled Kr-85 release during such an off-standard operation is considered significant. -- Unless further analyses by the equipment supplier prove otherwise, catalytic recombiners would be required for system operation. Recombiners have been unreliable at many BWRs. For example, less than one week of continued operation of new off gas systems has been accomplished in the last several years as a result of difficulties with the recombiner system at several opera' ting BWRs. -- The system operates with excess hydrogen to ensure all oxygen is removed. Hydrogen leakage, e.g., from the hydrogen recycle circuit vslves, could result in hydrogen burning or detonation. -- Packed type valves, rather than diaphragm valves, are used throughout the system. The system operates at a pressure of about 85 psig. Accordingly, leakage will likely occur. An alternate would be to replace all valves, but this would require extensive refurbishment. This system was designed to remove greater than 99.9 percent of tb e noble gas activity from the input stream. Accordingly, some eff-site release would occur even if this system func-tioned properly, i.e., this is not a zero release system. 86 1334 345
Further, the specific removal ef ficiency which could be achieved has not been demonstrated. A shop test was performed, but not under actual operating conditions. For example, pure nitrogen was employed for the process gas, rather than a gas containing moisture, carbon dioxide and argon in addition to nitrogen. Also, the bottom of the removal column contained a mixture of nitrogen and krypton, rather than a mixture of argon, methane, and krypton. Accordingly, it is considered that testing under actual operating conditions would be required to prove out this system before considering using it at TMI-2. There is no significant operating experience with a cryogenic distillation system at any operating light water reactor. Accordingly, this is not a proven technology for reactor application. As with the other systems, use of a cryogenic system does not resolve the problem of ultimate disposal of Kr-85. Long-t e rm storage for over one hundred years and of f-site shipment are considered particularly undesirable for this system due to the highly concentrated form of the Kr-85. Venting of the t_ d5 over a long period of time, as with the other systems, . inly varies the rate of personnel exposure - not the total exposure - and increases the likelihood of an uncontrolled release. Even though the major process components for this system are available at an existing facility, the time required to achieve system operation would not be significantly less than for the gas compression system. In particular, time is required to design and erect the building which houses the system, provide required utilities hookups, provide inter-connecting piping for various system components, and to test the system. The extensive time required to build and install the cryogenic treatment system would increase the likelihood of inadvertent and uncontrolled leakage from the existing containment build-ing, and thereby cause higher exposure to personnel. This extensive time delay to complete system installation would also delay TMI-2 cleanup operations. Finally, the cost of the cryogenic treatment system is high and no commensurate benefits are received. 8.5.5 Cryogenic Processing System Conclusions When compared to controlled purging of the containment building, the alternate cryogenic treatment system is considered to be less safe -- it is less reliable, and clearly has the potential for uncontrolled releases of radioactivity with higher radiation exposures. 1334 546 87
A A A A ^ ^ [ HYDROGEN RECYCLE v FROM l CRYOGENIC TPAIN (SLE F IGURE 7) 4P HYDROGEN STORAGE aL 1.lgtil D N IT ROGt'N SloPAGE TO ] COOL 183G (FOR COOLING CRY Nit:lC l' NITS) PIACTOP BUI LDl";G WATER RVsr VL!;T o O (309 SCDMI 1' ~ 10" pCi/nl FR-t$ (62 St.n, (10 3 SCFM) 63 CCR!f s -MW N ~* h M-P. (M TOTAL BTLLASE - M -s CHARCOAL AND HEPA M FILTER g= n n [ "~ 2,, '. .w l N- ~' l (62 SCFM) (103 SCFM)
- 3:.
cio_. ww m tx1- *: M - +- FROM CC :!AI ::tI:;? -N-t (225 SCrm 1 LCi/cl ER-85 60,000 CURIES TOTAL CONTA!!.:!ENT DISCHARGE O O 8 N-i, ~ l (62 SCFM) l (103 SCFM) -m-e -e e m-: w _. l t'RLHEATERS -N6 (3) Q 1' (75 SCFM EACit) AFTEftCOOLI:RS CPYUGE: llc TREATML'NT T!iA1:3S JL CATALYTIC ( 3) g3y RECOMulNERS (SEE FIGURE 7) (3) r e (FOR OXYGEN c.TOl< ?J: C ^^ FLPkeVAL TO ' - - ' " - - ' -El'CO!!DARY CONTA!!:ER (SfI FICUkE 8) PREVINI' OZONE L _q_g _ g _,J B u l l.t aIJ P A!33 l'OTLNTIAL DETONATION) - M - H -( }-N-M -> U - - H -M -( }- H - H -. H .No -0 ____ g _ y _( ] _. g _. H _, u, 1111-2 reomCT COncrEssoR u C0lil AllillEN LAIMOSPliELCRYOLDilC_CLUWUP_SYS M (""""^'"*',""' - - N + 8-i } M - M --* i N FIGURE 8. 5-1 H - M (___)-H -- M-* FHVPTON A:iD Xi'!:o!! STUPAGE
s a NE: FURIFIED GAS PREFURIFIER RECENERATION CYCLE STEPS ARE: To REACTOR BUILDING ROOF VENT STEP 1: REIO\\E H O AND CO3 FROM CAS (103 SCFM) 2 STEP 2: RI:8.OVE PESIDUAL KR AND XE FROM PREPURIFIER ftI PLCYCLE TO CATAl.YTIC F ECO"31 ;ER-STEP,; RD OVE ff 0 AND CO FRQt iREFURIFIER 2 2 cong pgx LIT:!D CO:Ji>L ISL R ?41 T FOJE?I llLAT LXCllA::CER SLFPLY I "2 8 s 8 s AMDIENT Pl! ASE KRYPTON AND XE: DON FROM PREPURIFIER PURGE HEATER ElPAPATOR F'>V I*y f F E!"r/AL J,_ , + - y-STI:P 3 -->4 gS COOLING a ' VM ygg A (62 SCF:) ^w 1 W -,, + -y-- STLP 2 - - te n xit I. E;J 8 PREHEATER f ~~ M e nrAT ' txcuA:stR TRACE FCED AFTER s CO:1F RISSOR PICOttB Il4E R COOLER l pi;,,gg g g.,g d- ~ _ - 8 FOR i: STtT I COOLER SEPARATORS OXYGIN P RIM as 51 CT 14:3 RF.@ AL j I OR 11,0 AFD CO 204 *#P AIIU II DECAY COI.'- = l'FO1 Hbb r/A L H Y D.10CA RBO:3 (WITil Atil o:1AT IC 4 CC:.VE FS ION ynpi... :; 3 g,p Cit i ptf, cut 3gygon cyctg lin::1 OVAL Cui.UttN - SLE h0TI:) HYlipOCARfiON Clif.VI PSION UN IT (TO PREVf:JT bu t !.I,t:P GF Cif4) KR, X Fo A, CH 4 ] TO FPODUCT COMPhES50P TO FEED COMPRES50R INLt.T = (.x COMBt!STION A itil 2 AIR C010AlliMLNL AIMOSrilLPLCLEM'JP. SYSIBLCRY00LHILIREAltiLNL lPAIII U 10GL OL lHfLL) h 'CO F[6URE 8.5-2
o ~ KRYPTON AND XENON STORAGE CONTAINERS, 6 TOTAL INLET AND OUTLET HEADER PENETRATICNS (2 LOCATIONS) STAINLESS STEEL LINER -*= A [ REINFORCED CONCRETE 2' THICK . s
- .. w..
/ / 3
- j....
A s. g / .s. / .c EQUIP 24ENT HATCH ?- ( j 't 4' DIA?4ETER P b..< 10' 3 [ ~ ~ \\ ll. J '., ~ ' a 3 e s. l 6
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- ,, s.,.e io.
g = u-a .. at u VIEW A-A NOTES:
- SECONDARY CONTAINER TO BE SEISMIC CLASS 1.
- INTERNAL DESIGN PRESSURE = 20 PSIG TMI-2 CRYOGENIC TPEATMENT SYSTEM KRYPTON STORAGE SECONDARY CONTAINER
] } } 4 } f() CONCEPTUAL ARCANCEMENT FIGURE 8. 5-3
e f PRCDUCT COMPRESSC'R 2* CONCRETE SHIELDING FOR CO!D POXLS COLD BOK ( 3 PLACES) bA PPETPI' ATMENT EQUIFMENT SFID /., ..g. (I FIACES) _ "' /. _ _ 7"4 _a ~ g e$ g l I NOTES: ( lf .j ,/,' l. BUILDING AND CO'*PONENT SUPPOFTS AFTERCOOLF.R () P L ACt'S ) TO BE SEISMIC CLASS 1. ~~ ' (, b l#
- 2. LIQUID NITRMI.N AND HYDROCE's 9
STOPACF ELt*IP".f.!;T LOCATia IN f SLE A RATE CONV1:NTIONAL STRt;CTURE. L_ _/ CATAI.YTIC PITOMa!NER p ,e AND l'REliFATLR () I' LAC E S ) -_1 I d [l e COLD ttOX f / t__ ( i I e IF-a il 32-4 l l l \\ ( .- J r-- 3,, ] ] g--- B PI TFl:AT:tC?JT r k N _a 1 QUIP :LNT l l SKID r t \\ I /
- k A
= 11o' 73, N TRAN5MITTER RACK VALVE RACK (3 PLACES) (3 PI A t:S) PF' Pt'RIF I E R VA1.YE A*;D -HAIN INSTkUP' Etat RACK \\ IFIN, DID s SrCTION A-A (3 PLACES) ' PPEPLil4!FIER VA!VE AND PIPING SMID (Rv7ATI'D 90* CCW) CCNCRETE BLOCK SilIELDING 2' THICK II PI ACI'SI - FRYPTON STOR.%E St CONDARY CONTAINER ,W (sEE FIccRE el g PifN VIIH h 1H1-1 GYDOUllC Iff.AlnuiLSY51GI m J IUILDjilG NO U/J1PUUiLLAYOUI CD FibilLE 8. 5-4
~ 8.6 Environmental Ef fects of Alternates The radiological impact on the environment of each alternate considered is evaluated in terms of normal operation and accident conditions. Of f-site exposure is calculated for each condition. In addition, on-site dose is evaluated for each alternate consider-ing maintenance and surveillance during processing and storage. 8.6.1 Normal Operation Since each system is designed to collect and retain the noble gas, no substantial effect is expected due to normal operation. Because of the schedule delay associated with each, the alternatives to direct controlled reactor building purging carry with them the potential for accidental uncontrolled release of noble gases from the reactor building. A schedule delay for each alternative has been calculated to be in the range of 2 to 4 years. Likelihood of system leakage over this period is high enough that with 107. of the noble gases leaking during worst case meteorology, the off-site dose due to any of the alternatives would be 10 times as great as the 5.0 mrem boundary beta dose calculated for the controlled purge of the entire noble gas volume. An evaluation of possible pressure buildup in the TMI-2 containment has been made, since such a pressure buildup could be the driving force for leakage of noble gas from the containnent. This evalua-tion is based on the following assumptions: (a) The electrical air circulation fans within containment fail either due to failure of their motors or failures in their associated electrical circuitry. (Note: The containment temperature is currently being held at approximately 100*F with the fans running.) (b) The steam generator still continues to remove the bulk of the primary system's decay heat and the only heat trans-ferred from the primary system to the containment atmos-phere is the heat losses through the primary system insulation. (c) Solar heat is an additional source of heat to the containment buiding. Using the above described assumptions, the evaluation indicates that containment pressure can rise to between 1 and 2 psig. This positive containment pressure would be the driving force for emitting the noble gases out of any defects in the containment boundary. A study was performed to determine the impact of a 1 to 2 psig con-tainment pressure and various size containment leaks, to establish 92 1334 551
a . ~ the potential off-site doses that could result. As background information, the present containment design is based on an allowable leakage of 0.2 percent per day under a design pressure of 60 psi. This allowable design leakage is equivalent to having an 0.13-inch diameter hole in the containment. Taking this design basis hole size, the doses resulting from the leakage through it with a con-tainment pressure of 1 to 2 psig can be determined. Further, an assessment of the effect of increased leakage due to additional holes of 1/8-inch, 1/4-inch, and 1/2-inch diameter that could result from seal deterioration, cracks, corrosion, etc. has been made. The total off-site dose due to leakage caused by 1 to 2 psig pressure in containment is as follows: Off-Site Dose Off-Site Dase During a One-Day During a 30-Day ) ) Period, mrem Period, mrem Containment Condition 8 / / Design Basir, (DB) leak 1-5 0.01-0.04 4-19 0.04-0.16 (equivalent to 0.13-inch hole) DB + 1/8-inch hole 2-9 0.02-0.08 8-34 0.07-0.29 DB + 1/4-inch hole 5-23 0.04-0.19 20-84 J.17-0.71 DB + 1/2-inch hole 18-76 0.15-0.64 66-280 0.55-2.37 (1) The high dose numbers are based on using NRC meteorological parameters in Regulatory Guide 1.4, while the lower doses are based on using more realistic meteorology from the TMI-2 FSAR. As can be seen from this table, the doses from a leaking contain-ment, even if only a design basis leak exists, are greater chan the total 5 mrem off-site dose calculated for the entire controlled purging operation of the containment. 8.6.1.1 Charcoal Adsorption and Storage System The charcoal adsorption system is designed for full noble gas retention on charcoal beds and therefore no of f-site dose is calculated, assuming no operator error or equipment failures. The on-site whole body dose due to maintenance and surveillance during processing and storage is calculated to be 23 person-rem. 8.6.1.2 Gas Compression cnd Storage System The gas compression system is designed for full retention of the reactor building atmosphere and therefore no off-site dose is 93 1334 ,52
e calculated assuming no operator errors or equipment failure. The on-site whole body dose due to caintenance and surveillance during processing and storage is calculated to be 58 person-rem. 8.6.1.3 Cryogenic Processing and Storage System The design basis for the cryogenic. system is to achieve 99.9% noble gas removal. This results in 0.17. release of the Kr-85 to the environment. Based on the analysis of the of f-site dose for controlled purging of the reactor building, assuming ground level release and average meteorological conditons the site boundary beta skin dose is estimated to be 0.05 crem. The on-site whole body dose due to maintenance and surveillance during processing and storage is calculated to be 570 person-rem. 8.6.2 Accident Conditions Regulatory Guide 1.24 " Assumptions Used For Evaluating the Potential Consequences of a Pressurized Water Reactor Building Gas Storage Tank Failure" specifies that release of the entire contents of a single storage tank is to be postulated to occur over a two hour period for accident analysis. The atmospheric dispersion model used in this analysis is the same as that described in Section 7.2 in accordance with Regulatory Guide 1.145. The calculated accident X/Q value for each accident analyzed below is 6.8 x 10 ^ sec/m". ~ 8.6.2.1 Charcoal Adsorption and Storage System The noble gas is stored in 450 charcoal tanks with successively decreasing noble gas activity in each tank as the withdrawn activ-ity in the reactor building decreases due to the feed and bleed process. The highest activity tank contains 1430 curies. Using the same methoc as used in Section 7.3, the site boundary cloud concentration becomes: Curies = 1430 -4 x 6.8 x 10 sec/m3 = 3 M 2 hrs. x 60 x 60 = 1.35 x 10' curies /m The resulting site boundary doses are: Beta Air Dose = 150 mrads, Beta Skin Dose = 64 mrems Gamma Air Dose = 1.3 mrads, Gamma Skin Pose = 1.4 mrems Whole Body Dose = 1.2 mrems 8.6.2.2 Gas Compression and Storage System For the 3as compression and storage system, 42% of the noble gas activity is stored in the high activity storage volume. Using 94
e 3 the same method as used in Section 7.3, the site boundary cloud concentration becomes: Curies 23900 -4 3 x 6.8 x 10 sec/m = 3 M 2 x 60 x 60 = 2. 26 x 10' Ci/m The resulting site boundary doses are: Beta Air Dose
2510 mrads, Beta skin dose = 1730 mrems Gamma Air Dose = 21.2 mrads, Gamma skin dose
24 mrems Whole Body Dose = 20.7 mrems 8.o.2.3 Cryogenic Processing and Storage System For the cryogenic processing and storage system,the entire con-centrated noble gas volume is stored in a single compartment. Using the same method as used in Section 7.3, the site boundary cloud concentration becomes: Curies 56.600 ,.8 x 10 sec/m -4 3 = x 3 M 2 x 60 x 60 ~3 3 = 5.35 x 10 Ci/m The resulting site boundary doses are: Beta Air Dose
5930 mrads, Beta skin dose - 4090 mrems 50 mrads, Gamma skin dose
36 mrems Gamma Air Dose = Whole Body Dose = 49 mrems 1334 354 95}}