ML20038A946
| ML20038A946 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/20/1981 |
| From: | Boegli J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20038A943 | List: |
| References | |
| NUDOCS 8111240493 | |
| Download: ML20038A946 (20) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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TEXAS UTILITIES GENERATING COMPANY )
Docket Nos. 50-445
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50-446 (Comanche Peak Steam Electric
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Station, Unit 1 and 2)
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TESTIMONY OF J.S. 80EGLI CONCERNING CFUR CONTENTION NO. 9 Q1:
By whom.are you employed and describe the work you perform?
A1:
I am a Senior Nuclear Engineer, Effluent Treatment Systems Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
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Commission, Washington, D.C., 20555.
I am responsible for the review and evaluation of radioactive waste management and effluent control systems and for the calculation of effluent source terms for nuclear power reactors. A statement of my professional and educational qualifications is attached to this testimony.
3 Q2: What is the nature of the responsibilities you have had regarding the Comanche Peak Steam Electric Station?
A2:
I reviewed the por.tions of Applicant's Environmental Report and l
Final Safety Analysis Report pertaining to radioactive waste management and effluent control systems and also calculated the release of radioactive materials in effluents for the Comanche Peak f
4 8111240493 811120 PDR ADOCK 05000445-PDR
. Steam Electric Station.
I estimated the gaseous effluent source tenn provided in Table 5.6 of the Staff's " Final Environmental Statement related to the operation of Comanche Peak Steam Electric
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Station, Units 1 and 2" (NUREG-0775), September, 1981 (FES), based on my review of the gaseous radioactive waste management system for the Comanche Peak Station.E This source tenn was used by Radiological Assessment Branch for determining radiological impacts of Comanche Peak operation.
Q3: Would you descriN the scope of the subject matter addressed in your testimony?
A3:
I have been asked to address an issue raised by CFUR relating to Contention 9, in particular, whether, as CFUR alleges, the "As low As Is Reasonably Achievable" ("ALARA") criteria (defined in.,
10 C.F.R. Part 20) requires that planned " batch-releases" of radioactive effluents from Comanche Peak be made only under meteorological conditions which " minimize radiation exposures."
In addressing this issue, I will discuss batch releases from the gaseous radioactive waste management system at the Comanche Peak l
If The gaseous radioactive waste management system is described in Section 4.2.3 of the Staff's FES and evaluated in detail in Section 11.2.2 of the Staff's " Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2," (NUREG-0797), July 1981 (SER).
I prepared 9 4.2.3 of the FES and 911.2.2 of the SER (which are included as attachments to this affidavit) and their contents are true and correct to the best of my knowledge.
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. Station and the application of the Commission's regulations to releases of radioactive effluents from the Comanche Peak Station.
Q4: What is the purpose of the gaseous radioactive waste management system'at the Comanche Peak Station?
A4: The gaseous radioactive waste management system is designed to collect and process radioactive gases stripped fran the primary coolant and ve.1ted fran liquid process tanks. The gases are compressed, treated for hydrogen removal by recombiners and the remaining gas, mainly nitrogen, is stored in one of ten decay tanks for reuse or release, under controlled conditions, to the plant vent. The system is shared by Units 1 and 2.
As stated in the FES, 6 4.2.3, the Staff's detailed evaluation of the radwaste systems and the capability of these systems to meet the requirements of 10 C.F.R. Part 50, Appendix I are presented in Chapter 11 of the SER.
Q5: What are the controlled conditions for the release of gas from the decay tanks?
A5: The system of decay tanks is designed to collect and retain the radioactive gases on a batch basis, thereby permitting at least ninety (90) days for radioactive' gas decay.
During nonnal oper-ation, including anticipated operational occurrences, the decay tanks are filled with gas which contains variable amounts of fission product gases.
The treatment and decay processes remove radioiodine and particulate nuclides, and the remaining stored nitrogen contains some radioactive noble gases.
When the stored
4 volume inventory reaches a certain level, the gases from an isolated tank may be sampled and analyzed prior to a batch release-of gas to the atmosphere.
Such releases are continuously moni-tored and controlled, even under postulated accident conditions, in accordance with 10 C.F.R. Part 50, Appendix A, General Design Criteria (GDC) 60, " Control of releases of radioactive materials to the environment" and GDC 64, " monitoring radioactivity releases" and the technical specifications, which will assure that the
. release is tenninated prior to exceeding the limits of 10 C.F.R.
s 20.105, " Permissible Levels of Radiation in Unrestricted Areas" and 9 20.106, " Radioactivity in Effluents to Unrestricted Areas."
Results of the analysis of samples from the decay: tanks will be used to determine compliance with the requirements of 10 C.F.R. 9 50.36a and to confirm the adequacy of the equipment installed to meet the "as low as is reasonably achievable" (ALARA) criteria specified in 10 C.F.R. Part 50, Appendix 1.
Q6:
In the event of an accident, what are the restrictions on the gaseous releases of radioactivity?
A6:
In the event of an accident in which radioactive gases are con-tained and subsequently disposed' by a controlled release to the atmosphere, the releases would be restricted to the limits set forth in 10 C.F.R. QS 20.105 and 20.106.
Q7:
Do the ALARA criteria defined in 10 C.F.R. Part s 20.1(c) apply to planned batch releases from the decay tanks?
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. A7: Although the ALARA concept is defined in 10 C.F.R. 5 20.1(c), the design objectives and limiting conditions for operation to meet the ALARA criteria are provided in 10 C.F.R. Part 50, Appendix I.
As stated in Appendix I,Section I., design objectives and limiting conditions for operation conforming to the guidelines of Appendix I shall be deemed a conclusive showing of compliance with the ALARA requirements of 10 C.F.R. 5 50.34a and 50.36a. The design objec-tives of Appendix I require that the Applicants provide adequate treatment or gas holdup equipment to reduce radioactive materials in effluents to ALARA levels in accordance with 10 CFR Part 50.34a. My source term in FES Table 5.6 includes an estimate of the releases I
from the gaseous radioactive waste management system via the decay tanks. The NRC Staff has shown in the SER 511.2.2.5 that the gaseous waste processing systems for Comanche Peak are capable of maintaining releases of radioactive materials in gaseous effluents during normal operation (including anticipated operational occurences)
"as low as is reasonably achievable ("ALARA") within the requirements of 10 CFR Part 50, Appendix I and the Annex to Appendix I.
Q8:
Do the ALARA criteria of Appendix I require that planned batch l
releases of radioactivity from Co[nanche Peak Station be made during meteorological conditions which will " minimize radioactive exposures"?
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. A8: No.
The "ALARA" criteria of Appendix I do not include planning batch releases fran the decay tanks during meteorological conditions which " minimize radioactive exposures", since the meteorological conditions do not reduce the levels of radioactive materials in gaseous effluents to "ALARA" levels, as required by 10 CFR Part 50, Appendix I.
Attachments:
FES, 5 4.2.3 FES, Table 5.6 SER, 9 11.2.2 e
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J. S. Boegli Professional Qualifications Effluent Treatment Systems Branch Office of Nuclear Reactor Regulation My name is J. S. Boegli.
I am a Senior Nuclear Engineer in the l
Effluent Treatment Systems Branch in the Office of Nuclear Reactor Regulation.
I attended Case Institute of Technology and received-a Bachelor of Science Degree in Chemical Engineering from Indiana Tech-nical College in 1951.
In 1952, I received a Master of Science Degree in Chemical Engineering from Kansas State College.
From 1955 to 1956-l I completed advanced courses in chemical and nuclear engineering at the University of Michigan and applied Health Physics training at the Oak Ridge National Laboratory.
From 1953 to 1973 I was employed by the Nat.ional Aeronautics and Space Administration and held positions as research engineer in heat and mass transfer, design engineer in nuclear reactor coolant, utili-ties, ventilation and radwaste systems, process systems supervisor, and technical consultant at the NASA Plum Brook Reactor in Ohio.
In July 1973, I joined the Nuclear Regulatory Commission (formerly AEC) as a Senior Nuclear Engineer in the Effluent Treatment Systems Branch, Division of Technical Review.
In this position I am responsi-ble for the review and evaluation of radwaste treatment systems and fnr the calculation of releases of radioactivity from nuclear power reactors.
My duties involve developing analytical models and per-forming calculations on the effectiveness of proposed radwaste systems, studying technological improvements and developing criteria governing radwaste processing, monitoring, shielding and handling.
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Final Environmental Statement related to the operation of Comanche Peak Steam Electric Station, Units 1 and 2 Docket Nos. 50445 and 50446 Texas Utilities Generating Company U.S IVuclear Regulatory
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4.2.3 Radioactive-Waste-Management Systems 10 CFR 550.34a (Section 50.3'4a of Title 10 of the Code of Federal Regulations) requires an applicant for a permit to construct a nuclear power reactor to include a description of the preliminary design of equipment to be installed for keeping levels of radioactive materials in effluents to unrestricted areas "as low as is reasonably achievable." The phrase "as low as is reasonably achievable" takes into account the state of technology and the economics of improvement in relation to benefits to the public health and safety and other 4-3 S
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societal and socioeconomic considerations and in relation to the utilization of nuclear energy in the public interest. Appendix I to 10 CFR Part 50 pro-vides numerical guidance on design objectives for light-water-cooled nuclear pcwer reactors to meet the requirement that radioactive materials in effluents released to unrestricted areas be kept as low as is reasonably achievable.
To comply with the requirements of 10 CFR Part 50.34a, the applicant has elected to meet the requirements of the Annex to Appendix I to 10 CFR Part 50, dated September 4, 1975, in lieu of performing a cost-benefit analysis as required by Section II.D of Appendix I.
The applicant has provided final designs of radwaste systems and effluent-control measures for keeping levels of radioactive materials in effluents to unrestricted areas as low as is reasonably achievable within the requirements of Appendix I and the Annex to Appendix I.
In addition, the applicant has provided an estimate of the quantity of each principal radionuclide expected to'he released annually to unrestricted areas in liquid and gaseous effluents produced during normal operation,
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including anticipated operational occurrences.
The staff's detailed evaluation of the radwaste systems and the capability of these systems to meet the requirements of Appendix I are presented in Chapter 11*
- of the Safety Evaluation Report.
Also, the quantities of radioactive material calculated by the staff to be released from the facility are presented there, 4-4
and in Section 5.8 of this environmental statement, along with the calculated doses to individuals and to the population that will result from these, effluent quantities.
Technical Specifications in the operating license will require that the appli-cant:
(1) establish release rates for radioactive material in liquid and gaseous effluer,ts and, (2) provide for the routine monitoring and measureme..!
of all principal release points to assure that the facility operates in con-formance with the requirements of Appendix I to 10 CFR Part 50.
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NUREG-0797 Safety Evaluation Report re atec to t7e operation of Comanc7e Pea < Steam E ectric Station, Units 1 and 2 Docket Nos. 50-445 and 50-446 Texas Utilities Generating Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 1
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Table 11.2 Calculated re1 cases of radioactive materials in gaseous effluents from Comanche Peak Units 1 and 2 (Ci/ year / reactor)
Waste Gas Main Condenser Processing Building Ventilation Vacuum Pump Nuclides System Reactor ~AuslTiary TUfbine Exhaust Total Kr-83m a
a a
a a
a Kr-85m a
2 2
a 1
5 Kr-85 254 6
a a
a 260 Kr-87 a
a 1
a a
1 Kr-88 a
2 4
a 3
9 Kr-89 a
a a
a a
a Xe-131m 1
10 a
a a
11 Xe-133m a
19 2
a 1
22 Xe-133 a
1900 110 a
68 2100 Xe-135m a
a a
a a
a Xe-135 a
10 6
a 4
20 Xe-137 a
a a
a a
a Xe-138 a
a a
a a
a TOTAL NOBLE GASES 2400 Mn-54 4.5(-3)b 2(-6) 1.8(-4) e c
4.7(-3)
Fe-59 1.5(-3) 6.7(-7) 6(-5) c c
1.6(-3)
Co-58 1.5(-2) 6.7(-6) 6(-4) c c
1.6(-2)
Co-60 7(-3) 3.1(-6) 2.7(-4) e c
7.3(-3)
Sr-89 3.3(-4) 1.5(-7) 1.3(-5) c c
3.4(-4)
Sr-90 6(-5) 2.7(-8) 2.4(-6) c c
6.2(-5)
Cs-134 4.5(-3) 2(-6) 1.8(-4) c c
4.7(-3)
Cs-137 7.5(-3) 3.4(-6) 3(-4) c c
7.8(-3)
TOTAL PARTICULATES 4.3(-2)
C-14 7
1 a.
a a
8
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A-41 a
25 a'
a a
25 H-3 a
a 1100 a
a 1100 1-131 a
2(-4) 4.4(-3) 1.6(-4) 2.8(-3) 7.6(-3)
I-133 a
2.4(-4) 6.3(-3) 2.3(-4) 3.9(-3) 1.1(-2)
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a - less than 1.0 Ci/yr noble' gases and carbon 14, and less than 10 4 Ci/yr for iodine.
b - exponential notation: 4.5(-3) = 4.5 x 10 s, c - less than 1% of total for nuclide.
11-4
Table 11.3 Principal parancters and conditions used in calculating releases from radioactiec material in tip id and gaseous effloents frcm Comanche F ue. Units 1 and 2 Reactar pcwer icvel (cegtwatts thermal) 3565 Plant capacity factor 0.80 Failed fuel 8
0.121 Primary steam Mass of coolant (1b) 5.4'x 10s Letdc.n rate (gpm) 75 Shim bleed rate (gpd) 1.9 x 108 Leakage to secondary system (ib/ day) 100 Leakage to containment building b
teakage to auxiliary building (Ib/ day) 160 Frequency of degassing for cold shutdewns (/yr) 2 Secondary system Steam flew rate (1b/hr) 1.5 x 107 Mass of liquid / steam generator (1b) 8.8 x 104 Mass of steam /stea, generator (1b) 8.5 x 108 Secondary rociant mass (1b) 2.47 x.10s Rate of steam leakage to turbine area (Ib/hr) 1.7 x 108 Number of steam generators 4
Containment building volume (ft )
2.9 x 10' 8
Annual frequency of containment purges (shutdown) 4 Annual frequency of containment purges (at power) 20 Iodine partition factors (gas / liquid)
Leakage to auxiliary building 0.0075 Leakage to turbine area
- 1. 0 Maincondenser/airejector(volatilespecies) 0.15 Liquid rad.aste subsystem decontamination factors Subsystem stream Iodines Cs, Rb Other nuclides Boron recovery system 104 2 x 104 105 Equipment drains 104 2 x 104 105 Liquid waste system (LWS drain) 104 105 105 Liquid waste system (chemical waste) 104 105 105 Principal equipment decontamination factors Equipment Iodines Cs, Rb Other nuclides I
Boron recovery system evaporator 10.
108 108 Liquid waste system evaporator 103 104 104 Anions' Cs, Rb Cations BR5 feed demineralizer 10 2
10 BR5 evaporator polishing demineralizer 10 10 10 LWS evaporator polishing demineralizer 10 10 10 Iodines Particulates Auxiliary and radwaste area a'nd ventilation (HEPA/ charcoal) 10 100 Reactor building purge air and refueling canal exhaust 10 100 Reactor building-internal cleanup (HEPA/ charcoal) 10 100 a - This value is constant and correspovls to 0.12% of the operating Wwer fission product source term as given in NULG-0017.
b - 1% per day of the primary coolant noble gas inventory and 0.001% per day of the primary coolant iodine inventory.
11-5
11.2.2 Gaseous Waste Processing System The gaseous radioactive waste processing sy,te'n and the plant ventilation system are designed to collect, store, process, monitor, recycle, and/or discharge potentially radioactive gaseous wastes which are generated during no operation of the plant.
necessary to reduce releases of radioactive gases and particulates to the The principal sources of gaseous waste are the effluents from environment.
the gaseous waste processing system, condenser vacuum pu 11-10
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Table 11.5 Design parameters of principal components considered in the evaluatjon of liquid and gaseous radioactive waste treatment systems Component Number Capacity, each Liquid Systems:
Chemical and volume control system Volume control tank 1/ reactor 400 fts Mixed bed demineralizers 2/ reactor 120 gpm Cation demineralizer 1/ reactor 120 gpm Thermal regeneration demineralizer 5/ reactor 250 gpm Boron recycle system (BRS)
BRS feed demineralizers 2 shared 120 gpm BRS holdup tanks 2 shared 56,000 gal Reactor coolant drain tank 1/ reactor 350 gal BR$ evaporator package 1 shared 15 gpm BRS evaporator condensste demineralizer 1 shared 120 gpm Liquid waste processing system (LWPS)
LWPS waste holdup tank 1 shared 10,000 gal LWPS waste evaporator condensate tank I shared 5,000 gal LWPS floor drain tanks (I or II) 2 shared 10,000 gal LWPS floor drain tani (III)
I shared 30,000 gal LWPS monitor tanks 2 shared 5,000 gal Chemical drain tank 1 shared 600 gal 1 shared 10,000 gal Laundry and hot shower drain tank LWP5 evaporator packages (floor drain & vaste) 2 shared 15 gpm LWPS demineralizers 2 shared 35 gpm s
Laundry reverse osmosis unit I shared 10 gpm Gaseous Systems:
Gaseous waste processing system (GWPS)
GWPS compressors (design pressure, 150 psig) 2 shared 40 scfm GWPS recombiners (design pressure, 150 psig) 2 sharad 50 scfm GWPS decay tanks (design pressure, 150 psig) 10 shared 600ft8 Design Classification and Seismic Design Criteria per Regulatory Guide 1.143.
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The gaseous waste processing systea (G'.'PS) for Ceuanche Peak is shared between Units 1 and 2.
Iba GWPS collects and stores the hydrogenated fission product gases stripped from the prir.iary coolant letdown, the volume control tanks, and the reactor drain tanks, by compressing the gas and removing the hydrogen in catalytic recombiners prior to storing the gas in decay tanks.
The stored gases are reused or san,11ed and analyzed before they are released to the environment.
A schematic diagram of the gaseous waste processing system is oiven in Figure 11.2.
Ventilation exhaust air from the containment and from the auxiliary building is processed through itEPA filters and charcoal adsorbers before it is released to the environment.
Turbine building exhausts are released without treatment.
Condenser vacuum pump exhaust is processed through 11 EPA filters and charcoal adsorbers before it is released to the environment.
11.2.2.1 Gaseous Waste Processing System The GWPS is designed to collect and process gases stripped from the primary coolant and from the hydrogenated gases vented from the volume control tanks, the boron recycle system holdup tanks, and the reactor drain tanks.
The gases are compressed by redundant 40-scfm compressors, treated for hydrogen removal by redundant recombiners and the remaining gas, mainly nitrogen, is stored in decay tanks for reuse.
There are 10 storage tanks included in the GWPS; each has a design pressure of 150 psig and a volume of 600 ft!.
In its evaluation, the staff determined that the 10 tanks provided have the capacity to store the radioative waste gases approximately 90 days for decay.
The staff believes that the system capacity and design are adequate for meeting the demands of the station during normal operation, including anticipated operational occurrences.
11.2.2.2 Containment Ventilation System Radioactive gases are released inside the containment when primary system components are opened or when primary system leakage occurs.
During normal operation, the gaseous activity is sealed within the containment but will be released during containment purges.
Based on information submitted by the applicant, the staff assumed there would be 24 purges /yr through HEPA filters and charcoal adsorbers released to the environment.
Four 24-hr purges /yr are assumed to occur after shutdown for the purpose of reducing /yr are assum radioactivity concentrations prior to operator access. Twenty 2-hr purges occur while the reactor is operating at full power to control the containment pressure, temperature,. humidity, and airborne radioactivity level above a predetermined value.
In its evaluation, the staff assumed a particulate decontamination factor.of 100 for HEPA filters and an iodine decontamination factor of 10 for charcoal adsorbers 11.2.2.3 Ventilation Releases From Other Buildings Radioactive materials are introduced into the plant atmosphere as a result of leakage from equipment transporting or handling radioactive materials.
The staff estimated that 160 lb/ day of primary coolant will leak to the auxili'ary building with an iodine partition factor of 0.0075.
Small quantities of radionuclides are released to the turbine building atmosphere based on an estimated 1700 lb/hr of steam leakage.
The plant ventilation systems are 11-12 O
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designed to induce air ficws from potentially less radioactive contaminated areas to areas having a greater potential for radioactive contamination.
The staf f calculations assumed that of fluents from the auxiliary building sa feguards building, containment purge exhaust, and main condenser vacuum pump ex,haust are released through itEPA filters and charcoal adsorbers before they are released through the plant vent.
The turbine building offluents are released directly to the environment without treatment.
11.2.2.4 Main Condenser Vacuum Pumo Offgas from the main condenser vacuum pumps contains radioactive gases as a result of primary-to-secondary leakage.
In its evaluation, the staff assumed a primary-to-secondary leak rate of 100 lb/ day. Noble gases and iodine are contained in the steam generator leakage and are released to the environment through the main condenser vacuum pumps in accordance with the partition factors listed in Table 11.3.
The vacuum pump exhaust is released to the environment through HEPA filters and charcoal adsorbers.
11.2.2.5 "onformance With Federal Regulations and Branch Technical Positions The proposed seismic design and quality group classification and capacities of the principal equipment in the GWP5 are listed in Table 11.5.
The staff finds the applicant's design criteria for the gaseous waste processing system and structure housing this system is acceptable in accordance with Regulatory Guide 1.143.
The GWP5 provides for monitoring hydrogen and oxygen upstream and dcwnstream of the waste gas processing system catalytic recombiners.
If the hydrogen or oxygen content exceeds a predetermined level, an alarm will sound in the reactor control room, alerting the operator to the condition.
If remedial action does not correct the hydrogen or oxygen concentration, automatic controls will discontinue the oxygen feed and/or hydrogenated waste gas feed to the recombiner.
The hydrogen and oxygen monitoring syste.n meets the acceptance criteria for dual gas analyzers for systems with recombiners, as given in SRP Section 11.3(Revision 11 s
Thenormalventilationexhaustsystemshavebiendesignedandwillbemaintained and tested in accordance with the guidelines presented in Regulatory Guide 1.140.
The staff calculated that ce proposed gaseous radwaste treatment and plant ventilation systems are capable of reducing the release of radioactive materials in gaseous effluents to apprbximately 2400 Ci/yr/ reactor for noble gases, 0.076 Ci/yr/ reactor for iodine-131,1100 Ci/yr/ reactor for tritium, and 8 Ci/yr/ reactor for carbon-14.
Using the calculated releases of radioactive materials in gaseous effluents l
from Units 1 and 2, given in Table 11.2, the staff calculated that the annual ganma and beta air doses at or beyond the site boundary are less than 10 mrad /
site and 20 mrad / site, respectively.
As shown in Table 11.2, the suaif calcu-lated the release of iodine-131 to be less than 1 Ci/yr/ reactor.
Using the calculated releases for iodine-131 given in Table 11.2, the staff calculated the dose or dose commitment to any organ of an individual in an unrestricted area to be less than 15 mrem /yr/ site.
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r Based on its evaluation, the staff concludes that the gaseous waste processing systeas for Co:.:anche Peak Steam Elr:ctric Station,-Units 1 and 2, are capable of maintaining releases of radioactive materials in gaseous effluents during nereal operation (including anticipated operational occurrences), so that the calculated doses are less than the numerical design objectives of Sections II.B and C of Appendix I of 10 CFR Part 50.
The staff evaluation also shows that the applicant's design of the gaseous waste treatment systems for Units 1 a'nd 2 conforms to the numerical des!gn objectives of RM 50-2, as specified in the option provided by the Conaission's September 4,1975 acendment to Appendix I of 10 CFR Part 50.
Therefore, the systems meet the requirements of 10 CFR Part 50.
The staff concludes that the' gaseous radwaste processing systems are capable of reducing radioactive materials in effluents to ALARA levels in accordance with 10 CFR Part 50.34a and are, therefore, acceptable.
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