ML20136F394
| ML20136F394 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Vogtle |
| Issue date: | 10/11/1984 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082840446 | List:
|
| References | |
| FOIA-84-663 NUDOCS 8410240413 | |
| Download: ML20136F394 (100) | |
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Docket File 50-424 Docket File 50-425 (w/o enc 1)
Docket Nos. 50-424/425 METB Docket Files METB Reading File MEMORANDUM FOR: Thomas M. Novak, Assistant Director ADRP Reading File for Licensing, DL FROM:
Daniel R. Muller, Assistant Director for Radiation Protection, DSI
SUBJECT:
METB DRAFT SER INPUT FOR V0GTLE, UNIT NOS. 1 AND 2 PLANT NAME: Vogtle Electric Generating Plant, Unit Nos. I and 2 LICENSING STAGE: OL DOCKETNUMBER(S):
50-424/425 RESPONSIBLE BRANCH: LB#4; M. Miller, LPM REQUESTED COMPLETION DATE: October 1, 1984 As requested in the August 24, 1984 memorandum from E. Adensam, enclosed is the Meteorology and Effluent Treatment Branch (METB) draft input to the Vogt,le SER.
If there are any questions concerning this submittal, please contact J. Fairobent (x29427), Meteorology Section, or C. Nichols (x27634),
Effluent Treatment Systems Section, METB.
Original signed WS Daniel R. Muller, Assistant Director for Radiation Protection Division of Systems Integration
Enclosure:
As stated cc:
R. Bernero W. Gammill E. Adensam M. Miller C. Willis I. Spickigr.s
@:Eid566ent C. Nichols govko@ #
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METEOROLOGY AND EFFLUENT TREATMENT BRANCH INPUT TO DRAFT SAFETY EVALUATION REPORT FOR l
V0GTLE ELECTRIC GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-424/425 l
2.3 Meteoroloqy l
Evaluation of regional and local climatological infonnation, including extremes of climate and severe weather occurrences which may affect the design and siting of a nuclear plant, is required to assure that the plant i
can be designed and operated within the requirements of Comission regulations.
Information concerning atmospheric diffusion characteristics i
of a nuclear power plant site is required for a determination that radioactive effluents from postulated accidental releases, as well as
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routine operational releases, are within Comission guidelines.
Sections 2.3.1 through 2.3.5 have been prepared in accordance with the review procedures described in the Standard Review Plan (NUREG-0800),
l utilizing infonnation presented in FSAR Section 2.3, responses to requests for additional information, and generally available reference materials as described in the appropriate sections of the Standard Review Plan (SRP).
f 2.3.1 Regional Climatology The plant is located in eastern Georgia along the Savannah River, about 20 l
miles southeast of Augusta.
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Maritime tropical air masses dominate the region in summer and alternate
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l with continental air masses in winter. The mean annual temperature in the i
area is about 17.4*C (63*F), ranging frcm about 7.8'C (46*F) in December and January to about 26.7'C (80*F) in July. Annual precipitation in the area is l
about 1090 m (43 inches).
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The Vogtle plant is located near a principal track of cyclonic. storms that originate along the Gulf Coast and move northeastward along the East Coast, f
resulting in a variety of severe weather phenomena. About 77 thunderstonns I
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j can be expected on about 56 days each year, being most frequent in June, July and August. Considering the frequency of thunderstorms in the region, 4
j the applicant has estimated about 10 lightning strikes per year in the
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square kilometer area containing the Vogtle plant. Hail often accompanies
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severe thunderstorms.
In the period 1955-1967, 6 occurrer.ces of hail with l
diameters 19 m (3/4 inch) or greater were reported in the one-degree
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latitude-longitude square containing the site.
i dj Tornadoes also occur in the area. About 30 tornadoes have occurred within j.
the one-degree latitude-longitude square containing the site in the period L
l 1954-1983, resulting in an annual tornado occurrence frequency of 1.1.
The j
~ applicant has conservatively computed a recurrence interval for a tornado at the plant site to be about 500 years. The staff has performed an 4
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independent assessment of tornado occurrences in the Vogtle region and h
computed a recurrence interval for a tornado at the plant site to be about i
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-4800 years. Waterspouts are not considered likely on the Savannah River in the vicinity of the Vogtle plant.
The design basis tornado characteristics selected by the applicant conform to the recommendations of Regulatory Guide 1.76, " Design Basis Tornado for
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Nuclear Power Plants," for this region of the country. These r
l characteristics are: rotational speed - 290 mph; translational speed -
70 mph; and a total pressure drop of 3 psi occurring at a rate of 2 psi /sec.
s Hurricanes or remnants of hurricanes pass through the region occasionally.
1 1871 1982, 40 tropical cyclones (tropical depressions, During the period trcpical storms and hurricanes) passed within 100 nautical miles of the site.
High wind speed occurrences in the area are associated with severe thunderstorms, extratropical cyclones, tropical storms, and hurricanes. The I
highest " fastest mile" wind speed reported at Augusta was 62 mph in June I
l 1965. The applicant has identified the " fastest mile" wind speed at a height of 30 feet with a return period of 100 years of 105 mph. However, for design of seismic Category-I structures, the applicant has used a 5
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" design wind velocity" (operating basis wind speed) of 110 mph at 30 feet above the ground, including gust factors and vertical velocity profiles developed in accordance with the criteria of American National Standards Institute A58.1, " Building Code Requirements for Minimum Design Loads in BuildingsandOtherStructures"(1972).
The applicant has identified the basins of the nuclear service cooling water towers (circular mechanical draft) as the ultimate heat sink for the Vogtle plant. The applicant examined meteorological data from August, GA for the period 1947-1981 to determine the meteorological design conditions for the i
ultimate heat sink. The conditions to ma'ximize water temperature and water i
usage were selected as the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> combination of dry bulb and wet bulb temperatures in the period of record resulting in the highest water usage
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and in the maximum temperature in the cooling tower basins followed by the 24-hour period resulting in the maximum average temperature in the basins and highest one day water usage. These conditions were repeated to i
synthesize a 30-day period. The synthesized meteorological conditions i
selected by the applicant for the design of the ultimate heat sink appear
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appropriately conservative.
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Heavy snowfall is not common in the region, but roof loads may accumulate due to a wintertime precipitation mixture of snow, ice and rain. Average annual snowfall at Augusta is only about 25 mm (1 inch). The applicant has reported the maximum snowfall in a 24-hour period in the area to be 350 mm (13.7 inches) in February 1973. The applicant has estimated the weight of the 100 year return period snowpack at ground level to be 8 psi.
Ice storms, which can plug drains and scuppers as well as disrupt offsite power, occur in the area. The applicant has indicated that freezing rain occurs on about 2 days each year. The applicant has also indicated that freezing rain has been reported to last as long as 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />. Severe and extreme environmental loads for consideration in the design of the roofs of l
safety-related structures at Vogtle are most likely due to the accumulation of water on the roofs. The adequacy of roof loadings due to water accumulation is discussed in Section 2.4.3 of the SER.
l The applicant has considered the following meteorological conditions in the i
design of the HVAC systems for all safety-related buildings: 98 F dry l
bulb /80*F wet bulb temperatures for summer air conditioning; 93*F dry l
bulb /78*F wet bulb temperatures for summer ventilation; 17'F dry bulb i
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6 temperature for winter heating; and average wind speeds of 7.5 mph and 15 mph for summer and winter, respectively. The bases for the selection of the temperatures for air conditioning and heating were the 1% probability of occurrences (summer) and 99% (winter) probability of occurrence values from the distributions presented by ASHRAE. The bases for the summer ventilation conditions was the 5: ASHRAE values for temperature. Wind speeds are
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supposed to be characteristic seasonal averages. The applicant has indicated in the response to RAI 451.5 that safety-related auxiliary systems and components (including the diesel generator air intake, service water valves, main steam isolation valves, and impulse lines) are enclosed in Seismic Category I structures maintained within acceptable environmental conditions by safety-related HVAC systems. Extreme temperatures of 41.7 C
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(107*F) and -16.1*C (3*F) have been reported at Augusta. Temperatures in excess of 32.2*C (90*F) are expected at Augusta about 60 days each year.
The applicant has analyzed onsite data for the periods December 1972-December 1973 and April 1977-December 1983 and determined that "the maximum consecutive duration of the dry bulb and wet bulb temperatures exceeding the design values is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, respectively," and that "the maximum consecutive duration of the dry bulb temperature lower than the design value is 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />." Conditions.with return periods of 100 years are normally considered for design of safety-related auxiliary systems and components.
The 100 year return period extreme temperatures in the Vogtle area are e
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r approximately 43.9'C (111*F) and -20*C (-4*F).
Although temperature excursions beyond the design values are infrequent, further justification of the adequacy of the extreme temperatures considered by the applicant for the design of safety-related auxiliary systems and components is required.
Also, the applicant has not directly addressed the diesel generator air intake.
l Large-scale episodes of atmospheric stagnation occur frequently in the region. About 90 atmospheric stagnation cases totaling about 360 days were reported in the area in the period 1936-1970. Ten of these cases lasted 7 days or more.
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As discussed above, the staff has reviewed available information relative to the regional meteorological conditions of importance to the safety design and siting of this plant in accordance with the criteria contained in Section 2.3.1 of the Standard Reveiw Plan. Based on this' review, the staff concludes that, with the exception of design basis temperatures for auxiliary systems and components, the applicant has identified and considered appropriate regional meteorological conditions in the design and siting of this plant, and, therefore, meets requirements of 10 CFR Part 100.10 and 10 CFR Part 50, the Appendix A, General Design Criterion 2 (GDC). The design basis tornado characteristics selected by the applicant Ih*
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conform to the position set forth in Regulatory Guide 1.76, and, therefore, t
meet the requirement of 10 CFR Part 50, Appendix A, GDC 4 to determine an accept'able design basis tornado for missile generation.
2.3.2 Local Meteorology Climatologicai data from Augusta, Georgia and available onsite data have been used to assess local meteorological characteristics of the plant site.
Precipitation is well-distributed througout the year, ranging frcm about 55 mm (2.2 inches) in October to ib'out 130 mm (5.1 inches) in July. Maximum and minimum monthly amounts of precipitation observed at Augusta have been 355 mm (14.0 inches) in July 1906 and a trace (less'than 0.01 inches) in October 1953. The maximum amount of precipitation in a 24-hour period at Augusta was 250 mm (9.8 inches) in October 1929. Snowfall is not common at Augusta, although snow has occurred in each month frcm November through March. The maximum monthly snowfall at Augusta was 355 mm (14 inches) in February 1973, and the maximum amount of snowfall in a 24-hour period was 350 nm (13.7 inches) also in February 1973. The annual average total precipitation measured at the site for the composite period December 1972-November 1973 and April 197.7-March 1979 is about 675 mm (26.6 inches) compared to the annual average at Augusta of 1035 mm (40.7 inches) for the same period of record. Annual average precipitation recorded at Augusta for l
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9 the same composite 3-year period as the onsite data showed reasonable agreement with the long-term climatological averages. Although spatial variability in precipitation occurrences is expected and may contribute to these differences, the most likely source of the differences in annual average amounts is the applicant's measurement of precipitation and data reduction techniques.
Wind data taken from the 10 m level of the ensit,e meteorological tower for a T
3-year composite period of record (April 4,1977-April 4,1979 and April 1, 1980-March 31, 1981) indicate that winds. are well-distributed, with' wind direction frequencies varying from about 4% to about 8%.
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The average wind speed at the 10 m level is about 4 m/sec. Calm conditions (defined as wind speeds less than the starting threshold of the anemometer) occur infrequently, at less than 0.5% of the time.
9 Slightly stable (Pasquill type "E") conditions predominate at the Vogtle site, occurring about 34% of the time as defined by the vertical temperature gradient between the 45.7 and 10 m levels. Moderately stable (Pasquill type "F") and extremely stable (Pasqill type "G") conditions occur about 16% and 9% of the time, respectively, for the same stability indicator. Moderately stable and extremely stable conditions were observed with relatively the ee
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As discussed above, the staff has reviewed available information relative to local meteorological conditions of importance to the safe design and siting 4
of this plant in accordance with the criteria. contained in SRP Section 2.3.2.
Although the staff is concerned about apparently anomalous precipitation measurements at the Vogtle site, the staff concludes that the applicant has identified and considered appropriate local meteorological conditions in the design and siting of this plant and, therefore, meets the requirements of 10 CFR Part 100.10 and 10 CFR Part 50, Appendix A, GDC 2.
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2.3.3 Onsite Meteorological Measurements Program i
Meteorological measurements at the Vogtle site were initiated in April 1972.
The meteorological tower used to provide data to support both the l
Construction Permit and Operating License applications is located about 1500 m (5000 ft) south-southwest of the Unit 1 containment building. Wind speed and wind direction are measured at the 10 m (33 ft) and 45.7 m (150 ft) levels, and vertical temperature gradient is measured between the 10 m and 45.7 m levels. Ambient dry bulb and dew point temperatures are measured at the 10 level, and precipitation and solar radiation are measured near the i
ground. The applicant has performed an analysis of the overall measurement 1
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11 system accuracies for each parameter, and concluded that the system accuracies for analog recording are not within the specifications presented in Regulatory Guide 1.23. System accuracies for digital recording appear to
. comply with the specifications presented in Regulatory Guide 1.23. The meteorological data provided with the Operating License application have been checked for reasonableness. The preliminary results indicate that the data collected by the meteorological measurements program are reasonable compared to other data coll,ected in the area. However, the check is not yet complete.
Three years (April 4,1977-April 4,1s. ) and April 1,1980-March 31,1981) of meteoroTogical data were provided with the Operating License application.
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Meteorological data from all the collection periods (including data for the period December 4,1972-December 4,1973) have been compared, and no significant differences have been identified. The three most recent years of onsite data have been combi:1ed into joint frequency distribution of wind speed and wind direction by atmospheric stability for use in the atmospheric dispersion assessments presented in Sections 2.3.4 and 2.3.5.
Wind speed and wind direction data for these assessments were based on measurements at the 10 m level, and atmospheric stablity was defined by the measurement of vertical temperature gradient between the 10.a and 43.7 m levels.
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12 Analog strip charts have been used to record meteorological data provided with the Operating License application. Calibration of the system since 1977 has been performed twice per year. Joint data recovery of wind speed and wind direction at the 10 m level by atmospheric stability (defined by the vertical temperature gradient between the 10 m and 45.7 m levels) was 92% for the 3-year composite period described above. Because the periods of missing data were sufficiently random during the 3-years of record, the composite data set is expected to reasonably reflect expected diurnal, seasonal, and annual airflow and stability patterns at the Vogtle site. The 3-year period of record is also expected to reasonably represent occurrences.
of extreme atmospheric conditions of importance for assessments of local transport and diffusion characteristics. The frequencies of occurrence of moderately stable and extremely stable conditions at Vogtle agrees reasonably well with other sites in the southeastern United States.
Dose consequence assessments based on available onsite meteorological data are expected to be reasonably conservative. Extreme meteorological conditions for design of safety-related structures, systems, and components (discussed in Section 2.3.1) were based on long-term (30 years or more) climatological data from nearby National Weather Service stations, and not directly on the 3 years of onsite data. However, the representativeness of long-term offsite data was determined by comparisons of concurrent offsite data with available onsite data.
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j For the postoperational meteorological measurements program, the applicant has installed a new meteorological tower located in the vicinity of the old tower. The tower will be instrumented at the 10 m and 60 m levels.
Although the applicant has not specified the parameters to.be measured, most likely wind speed and direction will.be measured at the 10 m and 60 m levels l
and vertical temperature difference will be measured between the 60 m and 10 m levels. The new tower was installed in January 1984, and the applicant l
has indicated that one full year of data from this tower will be available in February 1985. Data from both the old and new towers will be correlated.
To address meteorological requirements for emergency preparedness planning outlined in 10 CFR Part 50.47 and Appendix E to 10 CFR Part 50, the
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i applicant will be required to upgrade the operational meteorological measurements program to meet the criteria in NUREG-0654, Appendix 2,
" Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants." The upgrades f
must be in accordance with the schedule of NUREG-0737, III.A.2, "Clarificaticn of TMI Action Plan Requirements," or its supplements. The
. incorporation of current meteorological data into a real-time atmospheric dispersion model for dose assessments will also be considered as part of the upgraded capability.
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14 The staff has reviewed the onsite meteorological measurements system in accordance with the criteria contained in SRP Section 2.3.3.
The applicant
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has indicated that instrumentation and data reduction proceduret for analog l
recording do not conform to the recommendations of Regulatory Guide 1.23,
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j "Onsite Meteorological Programs." The staff is continuing to check the reasonableness of the data collected to date, and the staff will ensure that the new meteorological measurements program conforms to the specifications of Regulatory Guide 1.23.
The current meteorological measurements program appears to have provided data to represent onsite meteorological conditions as required in 10 CFR Part 100,10; however, the staff is continuing its evaluation of the adequacy of these data. Nevertheless, the staff concludes i
that the site data provide a reasonable basis for making preliminary
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conservative estimates of atmospheric dispersion conditions for estimating
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consequences of design basis accident and routine releases from the plant because the resulting wind speed, wind direction, and atmospheric stability distributions appear reasonable for the location of the Vogtle site.
Additional analyses will be performed to confirm this conclusion, n
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2.3.4 Short-Term (Accident) Diffusion Estimates 1
[i To audit the applicant's estimates, the staff has performed an independent L
assessment of short-term (less than 30 days) accidental releases from buildings and vents using the direction-dependent atmospheric dispersion
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i 15 model described in Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants,"
with consideration of increased lateral dispersion during stable conditions accompanied by low wind speeds. Three years (as described in Section 2.3.3) of onsite data were used for this evaluation. Wind speed and wind direction data were based on measurements at the 10 m level and atmospheric stability was defined by the vertical temperature gradient measured between the 10 m and 45.7 m levels. A ground-level release with a building wake factor, cA, of 1184 m2 was assumed. The relative concentration (X/Q) value for the 0-2 hour time period was determined to be 1.8 x 10-4 sec/m3 for the 5% overall site limit at an exclusion boundary distance of 1097 m.
Virtually identical X/Q values were calculated at the exclusion area boundary in the s
east-northeast, south, south-southwest, and southwest sectors. The 5%
overall site limit X/Q at the outer boundary of the low population zone (LPZ) was also slightly higher than the X/Q values in individual sectors.
The X/Q values for appropriate time periods at the LPZ distance of 3218 m are:
Time Period X/0 (sec/m3) 0-8 hours 3.1 x 10-5 8-24 hours 2.2 x 10-5 1-4 days 1.0 x 10-5 4-30 days 3.4 x 10-6 em'*g-4
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Based on the above evaluation performed in accordance with the criteria contained in SRP Section 2.3.4, the staff concludes that the applicant has considered appropriate atmospheric dispersion estimates for assessments of the consequences of radioactive releases in accordance with the requirements of 10 CFR Part 100.11. The atmospheric dispersion estimates provided in this section have been used by the staff in an independent assessment of.the a
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consequences of radioactive releases for design basis accidents.
2.3.5 Long-Term (Routine) Diffusion Estimates To audit the applicant's estimates, the staff has performed an independent calculation of annual average relative concentration (X/Q) and relative deposition (D/Q) values using the straight-line Gaussian atmospheric dispersion model described in Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors." The results of this I
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using the correction factors contained in NUREG/CR-2919.
Releases from the plant vents (atop the containment building) were
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considered as a mixture of elevated and ground level, except for the o
transportdirections(affectedsectors)ofeast-northeastandeast,where j
the natural draft cooling towers could significantly affect atmospheric dispersion. For the transport directions of east-northeast and east, i
j-releases from plant vents were considered as ground level. Releases from
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the turbine building (including the air ejector exhausts) were considered as ground level, with mixing in the turbulent wake of the major plant structures. Releases from the radwaste building were also considered as 4
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ground level, with mixing in the turbulent wake of that building. The same 3-year period of record described in Section 2.3.4 was used for this r
evaluation.
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Based on the above evaluation performed in accordance with the criteria i
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contained in SRP Section 2.3.5, the staff concludes that site-specific atmospheric dispersion conditions have been considered in demonstrating compliance with the numerical guides for doses contained in 10 CFR Part 50, Appendix I.
The atmospheric dispersion estimates developed by the staff are j
included in the assessment of the radiological impact to man resulting from I'
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18 routine releases to the atmosphere contained in the staff's environmental statement.
6.5 Engineered Safety Feature Atmosphere Cleanup Systems 6.5.1 System Description and Evaluation FSAR Section 6.5 contains information pertaining to engi'neered safety feature (ESF) atmosphere cleanup systems, their design bases, and applicable '
acceptance criteria.
The staff has reviewed tne applicant's design, design criteria, and design bases for the ESF atmosphere cleanup systems for Vogtle, Unit Nos. I an,d 2.
The acceptance criteria used as the, basis for 'ts evaluation are in 6
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Section II of SRP Section 6.5.1 (NUREG-0800). These acceptance criteria include the applicable GDC, ANSI N509-1980, ANSI N510-1980, Regulatory Guide 1.52, and other documents identified in Section II of the SRP.
Conformance to the acceptance criteria provides the bases for concluding that the ESF atmosphere cleanup systems meet the requirements of 10 CFR Part 50.
i The ESF atmosphere cleanup system at Vogtle, Unit Nos. I and 2, consists of process equipment and instrumentation necessary to control the release of radioactive iodine and particulate material following a design-basis
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19 accident (DBA). At Vogtle, Unit Nos. I and 2, the following four filtration systems have been designed for this purpose:
i (1) control room heating, ventilation and air conditioning system described in FSAR Section 6.4, Subsections 6.5.1 and 9.4.1; (2) fuel handling building post-accident exhaust system described in FSAR i
Subsection 9.4.2; (3) piping penetration filter exhaust system described in FSAR Subsection 9.4.3; and (4) electrical penetration filter exhaust system described in FSAR Subsection 9.4.5.
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Each system is designed to function automatically upon receipt of an ESF.
actuation system signal. EaSh of these systems was reviewed in accordance with the SRP. The results of these reviews are discussed below.
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(1) Control Room Heating, Ventilation and Air Conditioning System j
The control room heating, ventilation and air conditioning system contains two 100% capacity essential air filtration systems, with each system designed to filter up to 25,000 fta/ min of air. Each filtration a
l system includes, in order, a demister, an electric heater, a 0
high-efficiency particulate air (HEPA) fil'ter, a 4-in.-deep charcoal i
a adsorber, another HEPA filter, and a fan. The purpose of the control room heating, ventilation and air conditioning system is to limit the amount of radioactivity introduced into the control room following an accident by pressurizing the control room and by filtering the air I'
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1' 20 entering the control room, and to filter radioactivity already in the control room so that doses to control room operators will be within the design criterion of GDC 19. A safety injection signal or the detection
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of high radiaticn levels in the control room outside air intake causes
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the initiation of the control room isolation signal. This signal causes activation of the essential air filtration units followed by closure of the isolation dampers between the normal and essential systems, which automatically trips the nomal air handling units. One of the essential air filtration trains then may be manually transferred
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to the emergency standby mode. Air within the control room is recirculated continuously through the essentjal air filtration unit.
1 The outside gir required for pressurization is mixed with the return -
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air upstream of the filtration unit. The system design provides no potential bypass pathways around the essential air filtration units while they are operating in this mode. The staff has credited the system with 99% remcval efficiency for all forms of radioiodine, pending resolution of the open items discussed later regarding this subject.
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(2) Fuel Handling Building Post-Accident Exhaust System I
The fuel handling building post-accident exhaust system consists of two 100% capacity filtration systems with each designed to filter up to 5,000 ft3/ min of air.
Each filtration system includes a demister, an i
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l electric heater, a HEPA filter, a 4-in.-deep charcoal adsorber, another HEPA filter, and a fan. The fuel handling building post-accident f
exhaust system is designed to maintain a minimum negative pressure 1
within the fuel handling building and to filter exhaust air following a j
fuel handling accident, and thereby to minimize the release of airborne l
radioactivity to the outside atmosphere. This ensures that offsite i
i radiation exposures are within the guidelines of 10 CFR Part 100 and exposures to operating personnel in the control room are within the design criterion of GDC 19. The system design provides no potential i
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bypass pathways around the air filtration units while they are operating in this mode. The staff has credited the system with 99%
rqpoval' efficiency for all forms of radiciodiile, pending r6 solution of the open items discussed later regarding this system.
(3) Piping Penetration Filter Exhaust System The piping penetration filter exhaust system consists of two 100%
capacity filtration systems with each designed to filter up to i
16,000 ft3/ min of air. Each filtration system includes a demister, an electric heater, a HEPA filter, a 4-in.-deep charcoal adsorber, another HEPA filter, and a fan. The system is designed to maintain a minimum negative pressure on the piping penetration area of the auxiliary building and to filter exhaust air following containment and penetration area leakage under accident conditions. This ensures that f,
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r 22 the offsite radiation exposures resulting from the postulated post-LOCA leakage in recirculation piping are within the guidelines of 10 CFR Par' 100 and exposures to operating personnel in the control room are c
within the design criterion of GDC 19.
It all ensures that the ECCS and containment spray pump rooms can be purged to allow access for equipment repair and maintenance. The system design provides no potential bypass pathways around the air filtration units while they
..are operating in this mode. The staff has credited the system with 99%
removal efficiency for all forms of radiotodines, pending resolution of the open items discussed later regarding 'this system.
(4)' Electrical Penetration Filter Exhaust System The electrical penetration filter exhaust system consists o'f two 100%
2 a
capacity filtration systems with, each designed to filter up to 6,000 ft3/ min of air. Each filtration system includes a demister, an electric he.ater, a HEPA filter, a 4-in.-deep charcoal adsorber, another HEPA filter, and a fan.< The system is designed to maintain a minimum i
negative pressure on the electrical penetration area of the control building and to minimize release of airborne radioactivity following postulated post-LOCA containment leakage by filtering recirculated and exhaust air. This ensures that the offsite radiation exposures resulting from these accidents are within the guidelines of 10 CFR Part 100 and exposures to operating personnel in the control room
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23 resulting from these accidents are within the design criterion of The systeur design p' ovides no potential bypass pathways around GDC 19.
r the air filtration units while they are operating in this mode. The staff has credited the system with 99% removal efficiency for all forms of radioiodine, pending resolution of the open items discussed later regarding this system.
The ESF filtration systems were reviewed according to SRP Section 6.5.1 (NUREG-0800) and Regulatory Guide 1.52, Revision 1.
The applicant has.provided a comparison of the design of the Vogtle, Unit
~
I Nos. 1 and 2, ESF filtration systems with the acceptance criteria of the SRP
~
in FSAR Subsection 6.5.1.7.
The staff has determined that the applicant has propused a significant exception to the SRP acceptance criterion concerning conformance to the guidelines of Regulatory Guide 1.52 and the recomendations of ANSI N509 in that the proposed minimum instrumentation, readout, recording, and alarm provisions for the ESF atmosphere cleanup systems are not in conformance with Table 6.5.1-1 of the SRP, as follows:
(1) no local indication is provided of unit inlet or outlet flow; (2) no local high alarm signal is provided of the pressure drop across the prefilter (demister in the Vogtle design);
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.(3) no local status indication is provided for the electric heater; (4) no local indication, high alann, and low alarm signals are provided and 1..
no high alarm, low alarm, and trip-alarm signals are provided in the control room for a tamperature sensor located betweet.the heater and 3
thefirstHEPAfiltery (5) no local higri alann signal.is provided and no recorded indication is provided in the control room of the pressure drop across the first HEPA filter; (6) no local two-stage high alarm signal is provided for a temperature sensor located between the adsorber and the second HEPA filter; (7) no local high alarm signal is provided for the pressure drop across the
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second HEPA. filter; and
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(8) no local hand switch and status indication.is provided for the deluge valves.
The applicant has stated that the ESF filtration systems are designed to i operate only during post-accident conditions and do not operate under normal conditions, except during testing. Because the ESF filtration unit equipment rooms are potentially high radiation areas under post-accident i
conditions, the Vogtle design relies on control room instrumentation for monitoring of the filtration units. A high humidity alann in the control N
. room provide's' direct indication of high humidity rather than the indirect indication a low temperature alarm would provide. Based on our review, we l
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25 conclude that the local instrumentation and the icw temperature alam in the f
control room identified above as not provided in the Vcgtle design are not 1
needed to assure that the ESF atmosphere cleanup systems will perform their design safety functions. We further conclude that the appl.icant has not r
J provided justification for the other exceptions taken to the minimum instrumentation listed in Table 6.5.1-1 of the SRP, namely:
(1) no high alam and trip-alarm signals are provided in the control room i
for a temperature sensor located between the heater and the first HEPA filter; and j
(2) _ no recorded indication is provided in the control room for the pressure
,' drop-across the first HEPA filter.
1 This, th'erefore, is a'n open item.
4 The applicant has proposed a further exception to the SRP acceptance criterion concerning conformance to the guidelines of Regulatory Guide 1.52 1
J in that no cooling mechanism has been provided for the ESF filtration units
{
charcoal adsorber sections which has been demonstrated to satisfy the i
single-failure criterion. The applicant has stated that an analysis was j
performed to conservatively model the heating of the charcoal due to a 1
ll postulated loss-of-coolant accident; and that the results of this analysis showed that a cooling mechanism is not needed for the ESF filter systems. A water spray system is provided to allow flooding of the charcoal bed to I
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prevent bed ignition. Based on our review, we conclude that the applicant has not provided adequate justification for not providing a cooling mechanism which satisfies the single-failure criteria. This, therefore, is an open item.
The staff concludes that the design of the ESF atmosphere cleanup. systems, including the equipment and instrumentation to control the release of radioactive materials in gaseous effluents following a postulated DBA, is
- acceptable except as noted. This conclusion is based on the applicant having met the requirements of GDC 19, 41 and 61 by providing ESF atmosphere
_ cleanup systems on the control room habitability, containment, and associated systems. The applicant has met the requirements of GDC 41, 43
~ ~
and 64 by providing for the inspection and testing of the ESF atmosphere cleanup systems and monitoring for radioactive materials in effluents from these systems.
In meeting these regulations, the applicant has demonstrated that the design of the ESF atmosphere cleanup systems meets the guidelines of Regulatory Guide 1.52 and the ANSI N509 and N510 industry standards, as referenced in the SRP. The staff has reviewed the applicant's system descriptions and design criteria for the ESF atmosphere cleanup systems. On the basis of its evaluation, with respect to the SRP criteria, the staff finds the proposed ESF atmosphere cleanup systems acceptable, except as noted.
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The filter efficiencies given in Table 2 of Regulatory Guide 1.52 are appropriate for use in accident analyses, pending resolution of the open items discussed above.
10.4'.2 Main Condenser Evacuation System i.
' FSAR Section 10.4.2 contains infonnation pertaining to the main condenser evacuation (airremoval) system,thesystemdesignbases,andtheapplicable~
acceptance criteria. The staff has reviewed the applicant's design, design criteria, and design bases for the main condenser' evacuation system (MCES)
[
for Vogtle, Unit Nos. I and 2, in accordance with Section II of SRP Section 10.4.2 (NUREG-0800). The SRP acceptance criteria include GDC 60 and~
-64 and Heaf Exchanger Institute Standard, " Standards for Steam Surface
~
- Condensers." Guidelines for implementation of the requirements of the 4
acceptance criteria are provided in the regulatory guides referenced in Section II of the SRP. Conformance to the acceptance criteria of the SRP provides the bases for concluding that the MCES meets the requirements of
~10 CFR 50.
The MCES is designed to establish and maintain main condenser vacuum by removing noncondensible gases from the main condenser. Two two-stage steam i
jet air ejectors and two mechanical vacuum pumps are provided to remove noncondensables and hold vacuum during nonnal operation. The steam jet air E
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ejectors are provided with water-cooled inter-and after-condensers. The I
mechanical vacuum pumps are used to draw initial condenser vacuum during plant startups and may also be used during normal operation. Air and m
noncondensible gases removed from the main condenser shells by the steam jet a
air ejectors and mechanical vacuum pumps are continuously monitored for radioactivity prior to discharge through the turbine building vent. The exhaust gases are routed through the condenser vacuum exhaust filter system prior to discharge whenever a high level of radiation is detected by the radiation monitoring system.
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The applicant's provisions for quality assurance for the design, R
~
" construction, and operational phases of the MCES should be reviewed to j
determine conformance with Regulatory Guides 1.33 and 1.123, as provided in 1
the 3RP. The applicant has proposed an exception to the criteria in that the MCES is not safety related and Regulatory Guides 1.33 and 1.123 establish requirements only for safety-related structures, systems, and
[
components. However, the applicant has provided no further information regarding quality assurance for the design, construction, and operational phases for the main condenser evacuation system. Therefore, this is an open f
item. Equipment quality group classifications were reviewed to detemine conformance with Regulatory Guide 1.26, as provided in the SRP. No exceptions were noted. The MCES cn acity was reviewed to determine t
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conformance with Heat Exchanger Institute Standard, " Standards for Steam Surface Condensers," as provided in SRP Section 10.4. No exceptions were t
noted.
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The MCES includes equipment and instruments to establish and maintain condenser vacuum and to prevent an uncontrolled release of gaseous i
radioactive material to the environment. The scope of the staff's review l
I included the' system's capability to transfer radioactive gases to the l
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. ventilation exhaust systems and the design provisions incorporated to monitor and control releases of radioactive materials in effluents. The
[
staff has reviewed the applicant's system descriptions, piping and 4
instrumentation diagrams, and design criteria for the MCES components in
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accordance with the SRP.
It concludes that the MCES design is acceptable except as discussed above.
10.4.3 Turbine Gland Sealing System l
l FSAR Section 10.4.3 contains information pertaining to the turbine gland
/
sealing system, the design bases, and applicable acceptance criteria.
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The staff has reviewed the applicant's design, design criteria, and design bases for the turbine gland sealing system for Vogtle, Unit Nos. I and 2, in i
accordance with Section II of SRP Section 10.4.3 (NUREG-0800). The acceptance criteria include GDC 60 and 64. Guidelines for implementation of l
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30 the requirements of the acceptance criteria are provided in the regulatory guides identified in Section II of the SRP. Conformance to the acceptance criteria provides the bases for concluding that the turbine gland sealing
-system meets the requirements of 10 CFR Part 50.
The turbine gland sealing system provides sealing steam to the main turbine generator shaft to prevent the leakage of air into the turbine casings and the potential escape of radioactive steam into the turbine building. The turbine gland sealing system uses three steam sources: main, auxiliary.
and/or extraction steam. The steam supply is passed through the' turbine gland seals and condensed in the steam packing exhauster condenser. The condensate is returned to the main condenser and noncondensible gases are
- ~
continuously monitored for radioactivity prior to discharge through the turbine building vent. The exhaust gases are routed through the steam packing exhauster filtratio;; unit prior to discharge whenever a high level of radiation is detected by the radiation monitoring system.
The applicant's provisions for quality assurance for the design, construction and operational phases of the turbine gland sealing system should be reviewed to determine conformance with Regulatory Guides 1.33 and 1.123, as provided in the SRP. The applicant has proposed an exception to the criteria in that the turbine gland' sealing system is not safety-relate'd Nn m --.
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31 and Regulatory Guides 1.33 and 1.123 establish requirements only for safety-related structures, systems and components. However, the applicant has provided no further information regarding quality assurance for the design, construction, and operational phases for the turbine gland sealing system. Therefore, this is an open item.
The staff concludes that the turbine gland sealing system design is acceptable except as discussed above.
11.0 RADI0 ACTIVE WASTE MANAGEMENT The radioactive waste management systems for Vogtle, Unit Nos.1 and 2, are designed to pr. ovide for the controlled handling and treatment of liquid, gaseous and solid wastes. The liquid radioactive waste management system processes wastes from equipment and floor drains, sample wastes, decontamination and laboratory wastes, and chemical regeneration wastes.
i j
The gaseous radioactive waste management system provides (1) waste gas decay tanks to allow decay of short-lived noble gases, and (2) treatment of l
ventilation exhausts through high-efficiency particulate air (HEPA) filters and carbon adsorbers, as necessary, to reduce releases of radioactive materials to as low as is reasonably achievable (ALARA) levels in accordance j
with 10 CFR Parts 20 and 50.34a. The solid radioactive waste management system provides volume reduction by drying and incineration and' f
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32 solidification by using cement and polymer binders. The radioactive waste l
management review area also includes the process and effluent radiological monitoring and sampling system provided for the detection and measurement of radioactive materials in plant process and effluent streams.
1 The staff has reviewed the applicant's design, design criteria and design bases for the radioactive waste management systems for Vogtle, Unit Nos.1 and 2.
The acceptance criteria used as the basis for staff evaluation are in Sections 11.1, 11.2, 11.3, 11.4, and 11.5 (NUREG-0800). These acceptance j
criteria include the applicable GDC (Appendix A to 10 CFR 50),
10 CFR 20.106, Appendix I to 10 CFR 50, and American National Standards 1
Institute (ANSI) Standard N13.1, " Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities." Guidelines for implementation of the requirements of the acceptance criteria are provided in the ANSI standards,
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regulatory guides, and other documents identified in SRP Section II.
l Conformance to the acceptance criteria provides the bases for concluding 1
]
that the radioactive waste management systems meet the requirements of 10 CFR Parts 20 and 50.
11.1 Source Terms ll The applicant provided the expected annual radioactive releases from Vogtle,
]
Unit Nos. I and 2, in FSAR Tables 11.2.3-1 and 11.3.3-2.
The staff has
- f. -
33 performed an independent calculation of the primary and secondary coolant concentrations and of the release rates of radioactive materials using the information supplied in the FSAR, the GALE computer program, and the j
methodology presented in NUREG-0017. Table 11.1 presents the principal parameters that were used in this independent calculation of source terms.
These source terms were used to calculate individual doses in Sections 11.2 and 11.3 for Vogtle, Unit Nos. 1 and 2, in accordance with the mathematical models and guidance contained in Regulatory Guide 1.109, " Calculation of Annual Average Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR Part 50, Appendix I."
11.2 Liquid Waste Management System 11.2.1 System Description and Review The liquid radioactive waste (radwaste) management system consists of process equipment and instrumentation necessary to collect, process, monitor, and recycle or dispose of radioactive liquid wastes from the operation of Vogtle, Unit Nos. I and 2.
The liquid radwaste system is designed to collect and process wastes according to the source, activity, and composition of the fluids.
All liquid waste processing system waste will be processed on a batch basis to permit optimum control and disposal of radwaste. Before these wastes are 4
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. :. 1 34 released, samples will be analyzed to determine the types and amounts of radioactivity present. On the basis of the results of the analyses, the waste will be recycled for eventual reuse in the plant, retained for further processing, or released under controlled conditions. The released waste is combined with the effluent from the cooling tower blowdown sumps and the combined flow is then discharged to the river through the discharge pipe. A radiation monitor in the discharge line will automatically terminate liquid "
waste discharges if radiation measurements exceed a predetermined level.
P The liquid waste management system consists of the boron recycle system (FSAR Subsection 9.3.4.2), steam generator blowdown processing system (FSAR Subsectio'n 10.4.8), turbine building floor drain system (FSAR Subsection 9.3.3), and the liquid waste processing system (FSAR
~
Section 11.2). The liquid waste processing system consists of (1) the reactor coolant drain tank subsystem which collects nonaerated, reactor grade effluent from sources inside the containment for recycling; (2) drain channel A which collects aerated, reactor grade effluent that normally can be recycled; and (3) drain channel B which processes all effluent that is normally to be discharged to the environment and is not suitable for recycling.
In addition, the liquid waste processing system provides capability for handling and storage of spent ion exchange resins.
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The boron recycle system, which is shared between the two generating units, processes reactor coolant effluent that can be readily reused as makeup.
The system processes the affluent by mixed bed demineralizers, an evaporator
- with gas stripping and a polishing anion domineralizer. The system collects and processes reactor coolant system effluent, most of w'hich is from the letdown and process drains. The staff estimated that the boron recovery subsystem waste input will be approximately 2000 gpd per generating unit, of' which 1700 gpd will come from the letdown line in the chemical and volume control system, and 300 gpd will come from the reactor coolant drain tank.
The staff assumed that 25% of the processed water will be released to the river. The remainder will be recycled for reuse within the plant. The design capacity of the system (based on the design flow of the recycle 1
evaporator package) is 15 gpm (21,600 gpd). The difference between the expected flow and the design flow provides adequate reserve for processing surge flows.
The steam generator blowdown processing system (1 per generating unit) accepts water from each steam generator blowdown line, processes the water as may be required by mixed bed demineralizers, and delivers the processed water to the condensate system or to the waste water retention basin. The staff estimated that the steam generator blowdown processing system waste input will be approximately 54 000 lb/hr (108 gpm) and assumed that 10% of
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the treated process stream will be discharged to the river via the waste water retention basins, cooling tower blowdown sump, and discharge pipe.
The remainder will be recycled for reuse within the plant. The design flow l
capacity of the steam generator blowdown processing system.is 360 gpm, based on a maximum flow of 90 gpm from each of the four steam generators. The di.fference between the expected flow and design flow provides adequate reserve for processing surge flows.
The turbine building floor drain system collects in sumps the normally nonradioactive turbine building floor drains, equipment drains, sampling
[
wastes, and other miscellaneous drains. The collected fluid is usually sent i'
to the oil separator prior to discharge to the waste water retention basins.
If the fluid becomes radioactive, it is directed to the turbine building drain tank, from which the waste water is pumped to an oil separator, an activated charcoal filter, and demineralizers to remove oil and radioactive j
materials prior to discharge to the waste water retention basins.
l The reactor coolant drain tank subsystem of the liquid waste processing system collects valve leakoffs from the No. 3 seal leakoffs, reactor coolant pumps, reactor vessel flange leakoff and other deaerated tritiated water sources inside the containment. This liquid is normally processed by the baron recy'cle system for reuse; otherwise, it is sent to the liquid waste processing system drain channel A for processing. The staff estimated that
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i 37 the reactor coolant drain subsystem waste input flow will be approximately
.300 gpd per generating unit and assumed that 25". of the treated process stream will be released to the river through the discharge pipe. The remainder will be recycled for reuse within the plant. A separate reactor coolant drain tank subsystem is provided for each generating unit. The design capacities of the reactor coolant drain subsystem pumps and heat exchanger are each approximately 100 gpm. The reactor coolant drain tank usable volume is 350 gal. The difference between the expected flow and the design capacities provides adequate reserve for processing surge flows.
Drain channel A of tne liquid waste processing system collects, through lines connected to the waste holdup tank, liquids from accumulator drains (via reactor coolant drain tank pump suction); sample room sink drains (excess. primary sample volume only); ion exchanger, filter, pump, and other equipment drains; and condensate from the volume reduction system.
If the quality of the water in the containment sump or auxiliary building sump is acceptable for recycling, it may be directed to the waste holdup tank.
Otherwise, it will be directed to the floor drain tank for disposal. The wastes will be processed through the waste evaporator and, if necessary.
through the waste evaporator condensate demineralizer.
If further processir.; is required, the condensate can be returned to the waste holdup
' tank for additional evaporation. The staff estimated that the drain channel
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38 A waste input flow will be approximately 775 gpd per generating unit and assumed that 25% of the treated process stream will be released to the river through the discharge pipe. The. remainder will be recycled for reuse within the plant. A separate drain channel A subsystem is provided for each of the two generating units.
Drain channel B of the liquid waste proc,essing system collects and processes' wastes from floor drains, equipment drains containing nonreactor grade water, laundry and hot shcwer drains, and other nonreactor grade sources.
Water may enter the floor drain tank from leaks inside the containment through the containment sump, from leaks in the auxiliary building through auxiliary building sumps and floor drains, and from chemical laboratory
- ~
drains.
If necessary, the floor drain tank liquid can be processed by th waste evaporator or by the waste monitor tank demineralizer prior to release.
The staff estimated that the floor drain tank waste input flow will be approximately 2050 gpd per generating unit and assumed that 100% of the treated process stream will be released to the river through the discharge pipe. A separate floor drain tank and associated equipmei.c are provided for each generating unit.
The design capacity of the drain channel A subsystem and the floor drain tank processing, combined (based on the design flow of the waste evaporator) is 15 gpm (21,600 gpd). The difference between the expected flow and the design flow provides adequate reserve for processing surge flows.
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_The laundry and hot shower tank is provided to collect and process waste effluents from the plant laundry and personnel decontamination showers and hand sinks.
If necessary, the water can be directed through the Unit 1 or Unit 2 waste monitor tank domineralizer for cleanup prior to dis':harge to the river through the discharge pipe.
In its evaluation of the liquid radioactive waste management system, the
[
staff considered (1) the capability of the system to maintain releases below I
the limits in 10 CFR 20 during periods of fission product leakage (at design levels) from the fuel, (2) the capability of the system to meet the ALARA criterion in accordance with 10 CFR 50, Appendix I, Sections II.A and II.0,
-(3) the system design objectives for equipment necessary to control releases
~ ~
l-of radioactive effluents to the environment in accordance with 10 CFR 50.34a, (4) the system design to ensure adequate safety under normal i
and postulated accident conditions in accordance with GDC 61, and (5) the design features that are incorporated to control and monitor the releases of radioactive materials in'accordance with GDC 60 and 64.
The estimated releases of radioactive materials in liquid effluents were
'j calculated using the PWR-GALE Code described in NUREG-0017. The PWR-GALE Code is a computerized, mathematical model for calculating the routire releases of radioactive material in effluents from pressurized-water 4.
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reactors (PWRs). The basic code has been in use since 1976 for all PWR licensing reviews. The calculations in the code are based on (1) data generated from operating reactors. (2) field and laboratory tests, e
_ 3) standardized coolant activities derived from American Nuclear Society l
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(ANS) 18.1 Working Group recommendations, (4) release and transport mechanisms that result in the appearance of radioactive material in liquid streams, and (5) the Vogtle Plant, Unit Nos.1 and 2, radwaste system design
- l features used to reduce the quantities of radioactive materials ultimately released to the environment. The principal parameters used in these f
calculations are given in Table 11.1 of this SER.
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11.2.2 Evaluation Findings The liquid radwaste system includes the equipment necessary to control the releases of radioactive materials in liquid effluents in accordance with GDC 60 and 64. Capacities of principal components considered in the liquid waste processing system evaluation are listed in Table 11.2. The staff
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concludes that the design of the liquid waste management system is
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acceptable and meets the requirements of 10 CFR 20.106, 10 CFR 50.34a, Appendix I of 10 CFR 50, and GDC 60, 61 and 64, as referenced in the SRP.
i This conclusion is based on the following:
9
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(1) The applicant has met the requirements of 10 CFR 20.106. The staff has e
considered the potential consequences resulting from reactor operation
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41 and has detennined that the concentrations of radioactive materials in i
liquid effluents in unrestricted areas will be a small fraction of the limits in 10 CFR 20, Appendix B, Table II, Column 2.
(2) The applicant has met the requirements of Section II.A of Appendix I of 10 CFR 50 with respect to dose-limiting objectives by proposing a liquid radwaste treatment system that is capable of maintaining releases of radioactive materials in liquid effluents so that the calculated individual doses in an unrestricted area from all pathways of exposure are less than 3 mrems to the total body and 10 mrems to any organ.
In its evaluation, the staff considered releases of radioactive materials in liquid effluents for normal operation, including anticipated operational occurrences, based on expected radwaste inputs,
- ~
over the life of the plant for Vogtle, Unit Nos.1 and 2, in accordance with SRp Section 11.1.
The applicant has met the requirements of the Commission's September 4, 1975 Annex to Appendix I to 10 CFR 50 with respect to meeting the "as low as is reasonably achievable" criterion, and, therefore, need not perform a cost-benefit analysis as otherwise would be required by Section II.D of Appendix I to 10 CFR 50.
(3) The staff has reviewed the applicant's quality assurance provisions for the liquid radwaste systems, the quality group classifications used for O
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42 system components, and the seismic design applied to structures housing these systems. The design of the systems and structures housing these systems meets the intent of the criteria given in Regulatory Guide 1.143. The staff has reviewed the provisions incorporated in the applicant's design to control the release of radioactive materials in liquids resulting frcm inadvertent tank overflows and concludes that the measures proposed by the applicant are consistent with the criteria given in Regulatory Guide 1.143.
(4) The applicant has met the requirements of GDC 60, 61 and 64 with F
respect to controlling and monitoring the releases of radioactive material to the environment. The staff has considered the capabilities i
of the proposed liquid radwaste treatment system to meet the demands of
~
the plant resulting from anticipated operational occurrences and has concluded that the system's capacity and design flexibility are adequate to meet the anticipated needs of the plant.
f 11.3 Gaseous Waste Management System 11.3.1 System Description and Review The gaseous waste processing and plant ventilation systems are designed to collect, store, process, monitor, and discharge potentially radioactive gaseous wastes that are generated during normal operation of the plant. The systems consist of equipment and instrumentation necessary to reduce i
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t releases of radioactive gases and particulates to the environment. The principal sources of gaseous waste are the effluents from the gaseous waste L
processing system and ventilation exhausts from the containment, auxiliary, fuel handling, radwaste solidification, radwaste transfer, and turbine buildings.
The gaseous waste processing system is designed to collect, process and store gaseous wastes generated by normal plant operations including anticipated operational occurrences. The system consists mainly of two closed loops, each associated with one of the generating units and comprised I
of a waste gas compressor, a catalytic hydrogen recombiner, and seven waste gas decay tanks. Waste gas is pumped by the waste gas compressor to the hydrogen recombiner where oxygen is added to oxidize the hydrogen to water After removal of the water vapor, the gas stream is circulated to a vapor.
waste gas decay tank and then back to the waste gas compressor suction. A waste gas decay tank is valved into the recirculation loop for 1 to 2 days after which it is isolated and another tank valved into service. Hydrogen is continuously removed in the recombiner and therefore does not build up in the system. The largest contributors to the nonradioactive gas accumulation are helium generated from baron in the reactor and impurities in the l
hydrogen and oxygen supplies. With continued plant operation, the gas
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l pressure in the system will gradually increase as the nonremovable gases e
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i-44 accumulate in the system. At pressures over 20 psig, the valves are aligned so that the gases flow from the compressor to the decay tanks and then to the recombiner and back to the compressor suction. This arrangement is suitable for pressure up to 100 psig. Although the system is designed to
~
accommodate continuous operation without atmospheric releases, the system design permits controlled discharge of gas from the waste gas decay tanks to the plant vent.
l 1
j' Ventilation air from the containment building is filtered prior to i
exhausting through the plant vent..f,The exhaust air flow rate is 15,000 ft3/ min for normal purge operations during refueling and 5,000 ft3/ min for a minipurge during power access periods.
3 I-The auxiliary building nonnal ventilation system is designed to maintain the i
building at a negative pressure to prevent release of radioactivity to the atmosphere. The air is filtered prior to exhausting through the plant vent.
9 The fuel handling building ventilation system normal subsystems are designed to maintain the building at a negative pressure and to filter the air prior to exhausting through the Unit 1 plant vent.
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45 The radwaste building ventilation system filters potentially contaminated tir from the radwaste solidification building prior to exhausting through the radwaste solidification building stac.k and ducts potentially contaminated air from the radwaste transfer building to the. auxiliary building nonnal exhaust filtration system prior to exhaust through the plant vent.
In the turbine building, gases from the condenser vacuum exhaust and from the steam jet air ejectors are routed through the condenser vacuum exhaust filter system when a high level of radiation is detected by the radiation monitoring system. Also in the turbine building, gases released from the steam packing exhauster are routed through the steam packing exhauster
~
filter system when a high level of radiation is detected by the radiation monitoring system. Otherwise, no filtration is provided prior to exhausting gases from the turbine building ventilation system through the turbine building roof exhaust fans.
In its evaluation of the gaseous radwaste management system, the staff considered the following SRP criteria: (1) the capability of the system to meet the processing demands of the station during anticipated operational occurrences,(2)thequalitygroupandseismicdesignclassificationapplied to the equipment and components and structures housing the system, (3) the
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46 design features that are incorporated to control and monitor the releases of radioactive materials in accordance with GOC 60 and 64, (4) the potential for gaseous releases resulting.from hydrogen explosion in the gaseous radwaste system, and (5) the capability of the system design to meet the ALARA criterion in accordance with 10 CFR 50 Appendix I, Sections II.B.
II.C, and II.D.
The estimated releases of radioactive materials in gaseous effluents were calculated using the PWR-GALE Code described in NUREG-0017. The PWR-GALE Code is a computerized mathematical model for calculating the routine releases of radioactive material in effluents from PWRs. The basic code has been in use since 1976 for all PWR licensing reviews. The calculations in the code are based on (1) data generated from operating reactors, (2) field and laboratory tests, (3) standardized coolant activities derived frcm ANS 18.1 Working Group recommendations, (4) release and transport mechanisms that result in the appearance of radioactive material in gaseous streams, and(5)theVogtlePlant,UnitNos.Iand2,radwastesystemdesignfeatures used to reduce the quantities of radioactive materials ultimately released to the envircnment. The principal parameters used in these calculations are given in Table 11.1 of this SER.
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s The staff has reviewed the applicant's quality assurance provisiors for the gaseous radwaste system, the quality group classifications used for system components, the seismic design criteria applied to the design of the system and structures housing the radwaste system. The design ofothe system and 1
i structures housing this system meets the intent of the criteria given in i
I Regulatory Guide 1.143 and referenced in the SRP.
t The staff has reviewed the provisions incorporated in the applicant's design l
to control releases resulting from hydrogen explosions in the gaseous t.
radwaste system and concludes that the measures proposed by the applicant i
are adequate to prevent the occurrence of an explosion.
1 The staff has reviewed the provisions incorporated in the applicant's design to control and monitor radioactive materials in the normal ventilation exhaust systems during normal plant operation, including anticipated operational occurrences, and concludes that the system design is adequate to control and monitor airborne radioactivity.
A 11.3.2 Evaluation Findings
}
The staff concludes that the design of the gaseous waste management system i
j is acceptable and meets the requirement of 10 CFR 20.106; 10 CFR 50.34a; I
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GDC 3, 60, 61, and 64; and 10 CFR 50, Appendix I, as referenced in the SRP.
.j i
This conclusion is based on the following findings:
(1) The applicant has met.the requirements of GDC 60 and 64 with respect to controlling releases of radioactive material to the environment by ensuring that the design of the gaseous waste management system includes the equipment and instruments necessary to detect and control the release of radioactive materials in gaseous affluents. Capacities of principal components considered in the gaseous waste processing system evaluation are listed in Table 11.2.
(2) The applicant has met the requirements of Appendix I of 10 CFR 50 by meeting the ALARA criterion as follows:
(a) Regarding Sections II.S and II.C of Appendix I, the staff has
~ ~
considered releases of radioactive material (noble gases,
[
radiciodines, and particulates) in gaseous effluents for normal operation, including anticipated operational occurrences, based on expected radwaste inputs over the life of the plant. The staff has determined that the proposed gaseous waste management system is capable of limiting releases of radioactive materials in gaseous effluents so that the calculated individual doses from releases of radiciodine and radioactive material in particulatt form in an unrestricted area from all pathways of exposure are t
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49 l
less than 5 mrems to the total body, 15 mrems to the skin, and 15 arems to any organ.
(b) The applicant has met the requirements of the Commission's
~
September 4.-1975 Annex to Appendix I to 10 CFR 50 with respect to i.
meeting the "as low as is reasonably achievable" criterion, and, therefore, need not perform a cost-benefit analysis as otherwise would be required by Section II.D of Appendix I to 10 CFR 50.
e i
(3) The applicant has raet the requirements of 10 CFR 20. The staff has considerad the potential consequences resulting from reactor operation t-and determined that the concentrations of radioactive materials in
~
gaseous effluents in unrestricted areas will be a small fracticn of the limits specified in 10 CFR 20, Appendix 8. Table II, Column 1.
~
(4) The staff has considered the capabilities of the proposed gaseous waste I
management system to meet the demands of the plant resulting from anticipated operational occurrences and has concluded that the system's capacity and design flexibility are adequate to meet these demands.
(5) The staff has reviewed the applicant's quality assurance provisions for the gaseous waste management system, the quality group classifications used for system components, and the seismic design applied to the design of the system and of strue.tures housing the radwaste system.
The design of the system and of structures housing the syrtem meets the criteria given in Regulatory Guide 1.143.
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(6) The staff has teviewed the provisions incorporsted in the applicant's design to control releases resulting from hydrogen explcsions in the gaseous waste management system and concludes that the measures 1
proposed by the applicant are adequate to prevent the occurrence of an explosion in accordance with GDC 3.
I 11.4 Solid Waste Management System L
11.4.1 System Description and Review The solid waste management system consists of equipment and instrumentation i
necessary for the solidification or packaging of radioactive waste resulting from operation of the reactor water letdown purification system, the condensai;e demineralizer system, the liquid and gaseous radwaste systems,
~
and the miscellaneous debris resulting frcm nonnal operation and maintenance f
of the plant.
i i
I The solid waste management system consists of the following subsystems:
(1) resin transfer system; (2) bauflushable filter syttem; (3) backflushable filter crud transfer syster, 1(
(4) liquid / slurry waste solidification system; f
(5) dry waste system; 4
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- 51 (6) volume reduction system; (7) filtar handling system; (8) dry product transfer system; and (9) dry product solidification system.
The resin transfer system provides for the remote transfer of spent regenerative resin from the radwaste transfer building to the radwaste solidification building. The backflushable filter system provides for the removal and delivery of radioactive crud from certain process streams to the backflushable filter crud transfer system, which provides for the remote transfer of the radioactive crud from the radwaste transfer building to the radwaste solidification building.
The liquid / slurry waste solidification system provides for the solidification of spent resins and backflushable filter crud and, as a backup to the volume reduction system, provides for the solidification of chemical drain wastes and evaporator concentrates. A process control program will be used to ensure complete solidification in a cement binder.
The dry waste system provides for the collection of dry wastes, shredding of combustible waste, compaction of compressible noncombustible dry wastes, handling of activated components and equipment, and packaging and storage of compressed dry wastes and handled activated components and equipment.
52 The volume reduction system consists of two subsystems; the fluid bed dryer processes evaporator concentrates or chemical drain tank wastes, and the fluid bed dryer waste processor incinerates combustible dry wastes, contaminated oil and low activity spent resin from the condensate polishing domineralizer system and the steam generatcc blowdown system. The system is c
described in detail in the topical report, " Radioactive Waste Volume Reduction System, Topical Report No. AECC-?.-NP," Aerojet Energy Conversion Company, December 1981. The system produces a dry solid waste product, a liquid condensate process stream which is returned to the liquid radwaste system for processing, and an effluent gas process stream which is filtered t
prior to discharge through the radwaste solidification building vent. The filter handling system provides for the semi-remote removal of spent l
' radioactive cartridge filters and filter housings and their placement in j-shielded drums for transport to the.radwaste solidification system. The dry product transfer system provides holdup capacity for the dry product from the volume reduction system and provides for the transfer of the dry product to the dry product solidification system, which solidifies the particulate and ash from the volume reduction system. A process control program will be' used to ensure complete solidification of the dry product in a polymer
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binder.
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53 Solidified wastes, spent filter cartridges, and solid compactible wastes will be packaged in 55-gallon drums and stored onsite until they are shipped in shielded casks for offsite disposal.
The review of the solid waste management system, which was conducted in accordance with the SRP, included line diagrams of the system, piping and instrumentation diagrams (P& ids), and descriptive information on the solid waste management system and those auxiliary supporting systems that are essential to its operation. The applicant's proposed design criteria and design bases for the solid waste management system and the applicant's anaysis of these criteria and bases were reviewed and compared with those of the SRP. The staff also reviewed (1) the capability of the proposed system to process the types and volumes of wastes expected during normal operation and anticipated operational occurrences in accordance with GDC 60, (2) the provisions for the processing and packaging of wastes relative to the requirements of 10 CFR 20, 61, and 71 and applicable Department of Transportation (DOT) regulations, (3) the applicant's quality group classification and seismic design relative to Regulatory Guide 1.143, and (4) provisions for onsite storage before shipment. The basis for acceptance in the staff's review was conformance of the applicant's designs, design criteria, and design bases for the solid radwaste management system to the
u 54 regulations, guides, staff technical positions, and industry standards referenced in the SRP.
11.4.2 Evaluation Findings
'1 The annual quantities of solid wastes without volume reduction are estimated to be approximately 16,000 ft3 of solidified wet wastes,'containing approximately 10,000 Ci of activity, and approximately 7500 ft3 of dry waste. With volume reduction, the volume of solidified wet wastes will be reduced to approximately 4,000 ft3 and the volume of dry waste will be reduced due to the incineration of combustible dry wastes. The onsite i
storage capacity exceeds the expected quantity of drummed waste for 1 year of plant operation. Because the staff's guidance specifies storage space
~
for 1 month's capacity of waste, the staff finds the storage volume adequate for meeting the demands of the plant.
l
'Ili In the FSAR, the applicant stated that the process control system for the liquid / slurry waste solidification system and the dry product solidification system will be available prior to fuel load. This is a confirmatory item.
d The staff has not completed its review of the topical report concerning the volume reduction system, as referenced in the FSAR. Therefore, the I
following conclusions are pending the resolution of this open item.
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i 55 The staff concludes that the design of the solid waste management system is acceptable and meets the requirements of 10 CFR 20.106; 10 CFR 50.34a; GDC 60, 63, and 64; and 10 CFR 71, as referenced in SRP Section 11.4. This conclusion is based on the applicant demonstrating.that the solid waste management system includes the equipment and instrumentation used for the processing, packaging, and storing of radwastes before shipment offsite for
~
burial.
1 i
i The basis for acceptance in the staff's review has been conformance of the i
applicant's designs, design criteria, and design bases for the solid i
radwaste system to the regulations and guides referenced above and in SRP Section 11.4, as well as to staff technical positions and industry
~ ~
standards. On the basis of the foregoing evaluation and the condition that the applicant provide an acceptable process control program, which includes a compliance program to meet 10 CFR 20.311 and 10 CFR 61 for waste classification and waste minimum and stability requirements. The staff concludes that the proposed solid radwaste system is acceptable.
g i
11.5 Process and Effluent Radiological Monitoring and Sampling Systems 11.5.1 System Description and Review i
i The process and effluent radiological monitoring systems are designed to provide information concerning radioactivity levels in systems throughout
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56 the plant, indicate radioactive leakage between systems, monitor equipcent performance, and monitor and control radioactivity levels in plant discharges to'the environment.
I 1
Table 11.3 provides the proposed locations of continuous monitors. Monitors
)
on certain effluent release lines will automatically tenninate discharges if
. radiation levels exceed.a predetermined value. ' Systems that are not amenable to continuous monitoring or for which detailed isotopic analyses are required will be periodically sampled, and the samples will be analyzed in the plant laboratory. The potantial airborne radioactive releases to the f
environment from Vogtle, Unit Nos. I and 2, are from the following nonnal release points:
~ ~
'(1) Unit 1 plant vent; (2) Unit 2 plant vent; (3) Unit 1 turbine building exhaust; (4) Unit 2 turbine building exhaust;'and (5) radwaste solidification building vent.
The plant vent includes discharges from the containment purge system, gaseous radwaste system, fuel handling building HVAC system, and auxiliary building HVAC system. Condenser air ejector and steam packing exhaust is released through the turbine building exhaust.
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c 57 The plant vent effluent radiogas, air particulate, and iodine monitors are designed to provide representative data on the gaseous activity, particulate activity, and gaseous iodine activity released to the plant environs. This l
data is collected and displayed on a CRT and hardcopy printer on demand.
Alams will be annunciated on high radiation signals.
The containment vent effluent radiogas, air particulate, and iodine monitors measure radiogas, air particulate, and gaseous iodine activity in the containment purge vent. The monitors initiate automatic closure of the i
containment pua e supply and exhaust valves ~ for high radiation levels, c
The waste gas processing system effluent radiogas monitor isolates the waste i
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gas processing system on a high radiation alam signal. The fuel handling building effluent radiogas monitor initiates, on a high radiation signal, the switching of the fuel handling building ventilation system from the normal operating mode to the accident mode.
l i
Gases released from the condenser vacuum exhaust and from the steam jet air ejectors in the turbine building are routed through the filtration system when a high level of radiation is detected by the condenser air ejector and steam packing exhauster radiogas monitor.
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58 The waste solidification building effluent radiogas, air particulate, and iodine monitors are designed to provide representative data on radioactive i
releases to the environs. There is no provision, however, to isolate the volume reduction system on a high radiation signal. The staff has not completed its analysis of the need for such an automatic control function to maintain releases below the limits of 10 CFR Part 20. Therefore, this is an open item.
The potential radioactive liquid effluent normal release points are as follows:
(1) liquid waste discharge line into the discharge pipe, which discharges into the river; (2) steam generator blowdown liquid process discharge line into the water
~
retention basins, which discharge into the main condenser cooling tower blowdcwn sump, which discharges through the discharge pipe into the river; (3) turbine building drain liquid line into the waste water retention basins; (4) nuclear service water system cooling tower blowdown discharge line into the main condenser cooling tower blowdown sump; and (5) control building sump discharge line.
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t 59 The waste liquid effluent monitor, on a high alam signal, initiates automatic valve closure o,n the liquid waste discharge line. The steam generator blowdown liquid process monitor, on a high alam signal, I
automatically closes the steam generator blowdown processing system isolation valves and discharge lines. The turbine building drain liquid i
effluent monitor, on a high alarm signal, stops the flow from the turbine building drain system to the waste water retention basin. The nuclear service water process monitor provides indication of leakage from equipment
. processing radioactive liquid into the nuclear service cooling water. The l
control building sump effluent monitor, on a high alarm signal, initiates
[
automatic isolation of the discharge line.
~
11.5.2 Evaluation Findings The following evaluation findings are pending the resolution of the open item previously noted. The staff concludes that the process and effluent
{
radiological monitoring instrumentation and sampling system for the liquid f
and solid radwastes are acceptable and meet the relevant requirements of f
10 CFR 20.106 and GDC 60, 63, and 64. The process and effluent radiological monitoring and sampling systems for the liquid and solid radwaste include the instrumentation for monitoring and sampling radioactivity in I
contaminated liquid and solid waste process and effluent streams. The I
staff's review included (1) the provisions proposed to sample and monitor i
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- - - ;.-n 60 all liquid effluents in accordance with GDC 64; (2) the provisions proposed to provide automatic termination of liquid effluent releases and ensure control over discharges in accordance with GDC 60; (3) the provisions proposed for sampling and monitoring plant waste process streams for process control in accordance with GDC 63; (4) the provisions for conducting sampling and analytical programs in accordance with the guidelines in
' Regulatory Guides 1.21 and 4.15; and (5) the provisions for sampling and monitoring process and effluent streams during postulated accidents in accordance with the guidelines in Regulatory Guide 1.97, Revision 2.
The review included P& ids and process flow ~ diagrams for the liquid, gaseous, and
- solid radwaste systems and ventilation systems, and the location of monitoring points relative to effluent release points shown on the site plot
~
diagrams..
On the basis of its review, the staff has determined that the applicant's designs, design criteria, and design bases for the process and effluent radiological monitoring instrumentation and sampling systems for the liquid and solid radwastes meet the guidelines of SRP Appendix 11.5-A and industry l
l standards and concludes that the systems are acceptable.
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y-61 Item II.F.1 Attachment 1, Noble Gas Effluent Monitor and Attachment 2, Sampling and Analysis of Plant Effluents
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Potential gaseous accident release pathways are the folicwing:
(1) The plant vent, which includes discharges frem the containment purge system, the auxiliary building HVAC system (which includes discharges from the gaseous radwaste system and the containment piping penetration area filter and exhaust system), the fuel handling building HVAC, and the containment electrical penetration area filter and exhaust system
~
in the control building.
(2) The condenser air ejector and steam packing exhauster system.
(3) The steam generator safety relief valves' and atmospheric dump valves.
(4) The auxiliary feedwater steam turbine exhaust vent.
(5) The steam generator blowdcwn line break overpressurization relief damper.
I Effluent radiogas, iodine, and particulate monitors are provided at the plant vent and at the condenser air ejector and steam packing exhauster. A strap-on monitor is also provided at the main steam line. The applicant has proposed to use the main stcamline monitors to estimate the releases from the actuation of the steam generator safety relief valves, atmospheric dump A
valves, and the auxiliary feedwater steam turbine exhaust vent. The applicant also has proposed to use the steam generator liquid monitor and I
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62 the appropriate steam generator blowdown line flow instrumentation to l
h estimate the' activity released from the steam generator blowdown line break L
overpressurization relief damper. The SRP provides for the instrumented monitoring or sampling and analysis of identified gaseous effluent paths in the event'of postulated accident releases. This is an open item. The applicant has stated that design information pertaining to accident range
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' noble gas effluent monitors and continuous sampling of gaseous effluents for post accident releases of radioactive iodines and particulates has not been,
finalized and will be available after May 1983. This is a confirmatory i
item.
i
~ Item III.D.I.1 Integrity of Systems Outside Containment Likely to Contain Radioactive Material The applicant is committed to a program to reduce leakage from systems outside containment which could contain highly radioactive fluids during a serious transient or accident to "as low as practical" levels; and has stated that the specified information on the proposed program will be submitted no later than 4 months prior to fuel loading. This is a i
confinnatory item.
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Table 11.1 Principal parameters and conditions used in calculating releases of radiaoctive material in liquid and gaseous effluents from Vogtle Generating Plant, Unit Nos.1 and 2 L
h Parameter Value/ Unit n
Reactor power level (MWt) 3,565 Plant capacity factor 0.80 Failed fuel (%)
0.12*
Primary system 5
Mass of coolant ( b) 5.1 x 10 Letdown rate (gpm 75 3
Shim bleed rate ( pd) 1.7 x 10 Leakage to secondary system (1b/ day) 100 g
F Leakagetocontainmentbuilding(lb/ day)
Leakage to auxiliary building (lb/ day 160 Frequency of degassing for cold shutdowns (times /yr) 2 Letdown cation demineralizer flow (gpm) 7.5 Secondary system 7
Steam flow rate (lb/hr) 1.51 x 105 Mass of liquid / steam generator (lb) 1.11xIg Massofsteam/steamgenerator(1b) 6.4 x 10 6 Secondary coolant mass (lb) 2.0 x 103 Rate of steam leakage to turbine area (lb/hr) 1.7 x 10 6 Containmentbuildingvolume(ft3) 2.75 x 10 Frequency of containment purges (times /yr) 4 Containment low volume purge rate (fts/ min) 5000 Iodine partition factors (gas / liquid)
Leakage to auxiliary building 0.0075 i
Leakage to turbine area 1.0 Main condenser / air ejector (volatile species) 0.15 LIQUID RADWASTE SYSTEM DECONTAMINATION FACTORS I
Baron recycle Liquid waste Floor drains Material system processing (clean waste)
(dirty waste) f 5
Iodine 1 x 10 1x10) 1x10g Cesium 2x10j 1 x 10 1 x 10 4
4 q
Other 1 x 10 1 x 10 1 x 10 U
U
- This value is constant and corresponds to 0.12% of the operating power
{
product source term as given in NUREG-0017 (April 1976).
p
- 1%/ day of the primary coolant noble gas inventory and 0.001%/ day of the
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primary coolant iodine inventory.
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INDIVIDUAL EQUIPMENT DECONTAMINATION FACTORS (1) Evaporator All nuclides System except iodine Iodine 4
Radwaste evaporator 10 103 Boron evaporator 103 102 (2) Demineralizers,
10 Boron recycle evaporator polishing 102 1
1 demineralizer LIQUID WASTE INPUTS Decay Flow rate Fraction Fraction Collection time Stream (god) of PCA discharged time (days)
(days)
Shim bleed rate 1,700 1.0 0.25 4.2 2.1 Equipment drains 300 1.0 0.25 4.2 2.1 Clean wastes 713 0.127 0.25 5.6 0.19 Dirty wastes 2,050 0.019 1.0 1.95 0.19 GASEOUS WASTE INPUTS ***
Holdup time for xenon (days) 90 Holdup time for krypton (days) 90 Fill time of decay tanks (days) 0
- There is no continuous stripping of full letdown flow.
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Table 11.2 Design parameters-of principal components considered in the evaluation of liquid, gaseous and solid radioactive waste treatment systems for Vogtle Generating Plant, tinit Nos. 1 and 2 p._ _ Component Number Capacity, each LIQUID SYSTEMS
- Wasteholduptank(perunit) 2 10,000 gal Wasteevaporatorcondensatetank(perunit) 1 5,000 gal Flow drain tank (per unit) 1 10,000 gal Waste monitor tank (per unit) 2 5,000 gal Chemical drain tank (shared) 1 600 gal Laundry and hot shower tank (shared) 1 10,000 gal Waste evaporator concentrates holdup tank 1
2,000 gal l
(shared) l Wasteevaporator(perunit) 1 15 gpm l
Waste evaporator condensate demineralizer 1
35 gpm (perunit)
Waste monitor tank demineralizer (per unit) 1 35 gpm
~ Baron recycle holdup tank (shared) 2 112,000 gal Baron recycle evaporator (shared)(shared) 1 15 gpm Boron recycle feed demineralizer 2
120 gpm Boren recycle evaporator polishing 1
120 gpm
- demineralizer(shared)
GASEOUS SYSTEMS
- 1 l-Waste gas compressor (2 units) 4 40 scfm Waste gas decay tank (7 per unit plus 16 600 ft3, 2 shared) 100 psig Catalytic hydrogen recombiner (2 units) 3 50 scfm, 3 scfm Hp, SOLID SYSTEMS
- Polymer storage tank 1
6,000 gal Cement storage tank 1
2,400 gal Steam generator blowdown spent resin 1
5,980 gal transfer tank Waste processing system liquid spent resin 1
5,980 gal transfer tank Backflushable filter crud transfer tank 1
2,185 gal Product storage hopper 1
80 ft2 Volume reduction system spent secondary resin 1
18.4 gal /h
[
- -evaporator concentrates chemical 1
35 gal /h 5-drain wastes dry-combustible wastes 1
95 lb/hr j
i
= Quality group and seismic design in accordance with Regulatory Guide 1.143.
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i Table 11.3 Process and effluent radiation monitoring systems for Vogtle Electric Generating Plant, Unit Nos. I and 2 Detectable range, Stream Monitored Detector Type uCi/cc
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LIQUID Waste liquid affluent Gamma scintillation 10~ to 10~1 Steam generator liquid Gamma scintillation 4 x 10~ to4x10-f Nuclear service water Gamma scintillation 4 x 10~ to 4 x 10~2 Steam generator blowdown Gama scintillation.
4 x-10~ to 4 x 10-Component cooling water Gama scintillation 4 x 10~ to 4 x 10 2 Turbine building drain Gama scintillation 4 x 10~ to 4:x 10~2 l
Control building sump Gama scintillation 4 x 10~ to 4 x 10~
effluent 3
e GASEOUS Plant vent effluent (low range)
Particulate Beta scintillation 10-11 to 10-6 1
l Iodine Gama scintillation 10-11 to 10-6 Radiogas Beta scintillation 5 x 10~7 to 5 x 10-2
' ontainment vent effluent C
Particulate Beta scintillation 10-11 to 10-6 Iodine Gama scintillation 10-11 to 10-6 Radiogas Beta scintillation 5 x 10~7 to 5 x 10-2 l
~Radwaste solidification building effluent 10~11 to 10-6 13 Particulate Beta scintillation 10-8.to 10-6 10-Iodine Gama scintillation to 10-3 Radiogas Beta scintillation 4
Waste gas processing system G-M tube 10-1 to 10 effluent Fuel handling building Thin-walled G-M tube 10-6 to 10~1 effluent Plant vent (high range)
Particulate Passive particulate filter NA Iodine Passive iodine filter NA 4
Radiogas Beta scintillation 10~6 to 10 for Xe-133 4
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Table 11.3 (Continued)
P Detectable range, Stream Monitored Detector Type uCi/cc
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GASEOUS (Continued)
Main steam line Strap-on gamma 10-1 to 103
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Condenser air ejector and steam i
packing exhauster Particulate Passive pa'rticulate filter NA Iodine Passive iodine filter 5x10p N
5 to 10 Radiogas Beta scintillation for Xe-133 i
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