ML20092J934
| ML20092J934 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/23/1984 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| References | |
| OL-A-038, OL-A-38, NUDOCS 8406270315 | |
| Download: ML20092J934 (173) | |
Text
-
F t y pct m8tR ,3 () L
- . mo. a ;Tn. Ft c.. j sfg g A99t_ E x . 38 LGS FSAR 09QETED e I.D.2 PLANT SAFETY PARAHETER DISPLAY CONSOtt C Position iy gy ,7 Each applicant and licensee shall install a safety parameter .
display system (SPDS) that will display to operating ~ personnel a
~
l minimum set of parameters which define the safety status of.the ,
plant. This can be attained through continuous indication of ,
direct and derived variables as necessary to assess plant safety status.
Response
The Limerick Design includes an Emergency Response Facility Data System (ERFDS). T3.is system is based on the General Electric Emergency Response Information System. The SPDS at Limerick will be a part of this ERFDS.
The hardware design is essentially complete with installation scheduled to begin in April 1983. The software effort is proceeding on schedule to support system chec,kout in August 1983.
It is expected that the system will be operational by fuel load.
The parameters included in the SPDS display are based on the entry conditions for the Limerick symptom-based emergency I, procedures. As changes and improvements are made to the reactor '
pressure vessel control and containment control procedures, the system can be modified to reflect these changes. The SPDS parameters are a subset of the parameters available in the ERFDS data base, which is based on the Regulatory Guide 1.97 Rev. 2 BWR ;
parameter list. The system has been designed in accordance with the guidance provided in NUREG-0696.
A safety analysis describing the basis on which the selected :
parameters are sufficient to assess the safety status of each identified function for a wide range of events will be available by August 1983.
Prior to fuel load, the SPDS will be reviewed in conjunction with the other NUREG-0737 supplement initiatives. As one of the final steps in integrating all of the control room modifications (i.e.,
Regulatory Guide 1.97, SPDS, Emergency Operating Procedures), an emergency procedure walk through is scheduled as part of the control room review.
- I.G.1 TRAINING DURING LOW-POWER TESTING Position We requ. ire applicants for a new operating license to define and commit to a special low-power testing program approved by NRC to be conducted at power levels no greater than 5 percent for the Rev. 20, 05/83 1.13-18b 8406270315 840423 PDR ADOCK 05000352 0 PDR L>
y J , _
b Y ,
e v
5.S. IWlHAR REGULATORY COMMISSIGI - -
- g7 <
, - e AppEcant 7 $taff Intervenor identified --Received " Rejected Date* ' W-23-ffi - '
. Reporter: -
i .
9 t
L i L :
+- 1 l
i l :
l LGS FSAR There are no outstanding mineral rights within the exclusion )
area.
g
! 2.1.2.2 Control of Activities Unrelated to Plant Operation l I
t Activities unrelated to plant operation that occur within the !
i . exclusion area, aside from transit through the area, are those l associated with the Limerick Atomic Information Center, located i l
approximately 1500 feet from the plant along Longview Road.
About 37,000 people visited the information center in 1979. This !
is expected to increase to 40-45,000 in the initial year of plan * !
operation. The number of visitors to the center seldom exceeds l 100 at any one time. Evacuation of these people is discussed in ,
t Section 13.3.
l l 2.1.2.3 Arrancements for Traffic Control on Public Passaceways {
r r i
Arrangements have been made with the Pennsylvania State Police to l control public access to the exclusion area in the event of an ;
l emergency, j Arrangements have been made with Conrail to control rail traffic <
t thecugh the exclusion area in the event of an emergency.
Letters of' agreement between PECo and the State Police and :
Conrail are provided in Appendix'A of the LGi Emergency Plan, l FSAR Section 13.3.
i l
2.1.2.4 Abandonment or Relocation of Roads I
Prior to station construction, Longview Road traversed the site !
in a southerly direction from the juncture of Sanatoga Road and !
Possum Hollow Road to the railroad right-of-way on the eastern !
! bank of the Schuylkill River. This portion of Longview Road, ;
approximately 6000 feet, was abandoned and relocated to the l eastern edge of.the Limerick site on a portion of roadway !
formerly known as Lozark Road. New sections of the roadway were !
constructed to realign Longview Road and Lozark Road between Keen ;
i Road and the existing paving on Longview Road South of Brook !
l Evans Creek. Both Longview and Lozark Roads are township roads. ;
l l 2.1-3 Rev. 1, 09/81
'9 i i
2.3'.2.2.2.2 Emergency Spray Pond. f L
The ultimate: heat sink at the Limerick Generating Station is a !
spray pond. .During routine operations this pond will not be !
heated, and water temperatures will fluctuate in response to ambient meteorological conditions in the same manner as any i natural pond of the same size. This should produce no adverse !
impact to the local meteorology. !
2.3.2.3 Topooraphy l
The topography of the LGS site is described in Section 2.1.1. j The' topography of the region surrounding the site, out to a l distance of 50 miles, is summarized in Table 2.3.2-86 which lists the offsite' terrain elevation-(in feet above mean sea level) i versus distance from a point midway between the Limerick vents. !'
The~value listed is the maximum elevation on or outside the site '
boundary which occurs within each of the sixteen 22-1/2 degree sectors at the' distance listed.
-These terrain elevations were obtained from USGS maps. {
2.3.3 ONSITE METEOROLOGICAL MEASUREMENTS PROGRAM The onsite meteorological measurements program at the Limerick .
. site began on December 10, 1970 with preliminary wind ,
measurements taken from a six-bladed Aerovane located 30 feet ;
above. grade on a temporary pole. Windspeed and direction data ,
were continuously collected at the temporary pole until December ;
28, 1971 when it was removed from service. Prior to the sensor i removal, the onsite meteorological measurements program was !
expanded on December 10, 1971 with the installation of Weather !
Station No. I near the temporary pole location. The main tower l (Tower No. 1) extending about 281 feet above grade (250 feet MSL) !
was erected on high ground, northwest of the reactor locations. ;
Windspeed, wind direction,-and temperature.from three elevations i are continuously recorded. Instrument elevations are listed in Table 2.3.3-1. Additional onsite measurements of horizontal and !
vertical wind direction fluctuations, relative humidity, !'
~
barometric pressure, and precipitation complete the observation at Weather Station No. 1. [
In order to evaluate the effects of.the shallow Schuylkill River
-Valley, the onsite meteorological measurements program was again ,
- expanded on December 28, 1971 with the installation of a second !
weather station. Weather Station No. 2 is located across the Schuylkill River from the main tower and is onsite in an open field having a base elevation close to that of the valley floor.
Tower No. 2 at this location, extends 314 feet above grade '
-(121 feet MSL). Windspeed, wind direction, and temperature from three. elevations are continuously recorded. Tower 2 was :
t I
2.3-25 L
.-e. , . _ - . , _ . _ . -.
t I
established-to provide supplementary site' data on the temperature profile in the valley during the preoperational period. This ;
tower was instrumented at mean sea level elevations coincident with.those of Tower 1.in order to. compare meteorological
-conditions over the valley with those over the adjacent low ,
hills. The locations and relationships between the various wind and temperature instruments are shown in Figures 2.3.3-1 and 2.3.3-2.
t The. overlapping arrangement of the facilities, which allows a ;
comparison of wind and temperature measurements from each towe.
at two corresponding levels, produces a complete picture of wind
' flow and lapse rates from the valley bottom to a point about 270 feet above the higher terrain. ;
To-determine the typical flow over the river and adjacent low :
terrain, a satellite to Weather Station No. I was established and
' data collection began on November 20, 1974. The sensors are located 32 feet above grade 106 feet (MSL) and are capable of i continuously measuring windspeed and wind direction. [
In 1983,.the complete system is being upgraded to comply with the ,
criteria.of Regulatory Guide 1.23, Rev. 1, and NUREG-0654. Data L
.from each of the three meteorological locations will be !
transmitted to the control room where it will be recorded on stripcharts and logged by a data-logger. The data will also be i transmitted to the Technical Support Center as input to the ;
' radiological and meteorological monitoring system (RMMS). ,
2.3.3.1 Precoerational Meteorolocical Measurement System (1970-1983)
This meteorological system was used to obtain measurements as described in Table 2.3.3-1. .
2.3.3.1.1 Measurements and Instrumentation i 2.3.3.1.1.1 Siting i As shown in Figures 2.3.3-1 and 2.3.3-2, the main meterological I weather Tower 1 located at Weathir Station No. 1 is a 280-foot ,
tower situated approximately 3000 feet NW of the Limerick
. structure vents. Weather Tower 1 is also located approximately.
2000 feet NNW of the center of the Unit 1 cooling tower ~ location and approximately 2400 feet NW of the center of the Unit 2 cooling tower location. Grade elevation at Weather Station No. 1 l is'250 feet MSL.
The wind instruments on Tower 1 are mounted on retractable booms extending upwind 10 feet 0 inches west of the tower. Each face ,
of the triangular tower is 3 feet 6 inches wide. The temperature '
sensors are located in aspirators and are 2 feet 0 inches from Rev. 21, 06/83 2.3-26 i
r
sc l
I i
-the tower. Weather Station No. I has a base surface made of I yardstone. The relative humidity sensor is located in a standard ;
IU.S. Weather-Bureau-type shelter 5 feet above grade and the !
surface beneath the instrument shelter is wood.
~
l Meteorological Weather Tower 2 located.at Weather Station No. 2 I is a'310-foot tower situated approximately 2100 feet west of the !
- L'..nerick structure vents. -
Weather Tower 2 is also located !
approximately*1950 feet WSW of the center of the Unit 1 cooling :
tower-location, and approximately 2600 feet WSW of the center of ;
the Unit 2 cooling tower-location. ;
The wind instruments on Tower 2 are mounted on retractable booms
. extending upwind 10 feet 0 inches WNW of the tower. Each face of l Ethe triangular tower is 3 feet 6 inches wide. The temperature sensors are located in aspirators and are 2 feet 0 inches from !
r- the tower. Weather Tower 2 has a base surface made of yardstone.
f 2.3~.3.1.1.2 -Inctrumentation and Performance Specifications l l
- The instrumentation systems installed on the Limerick site were designed to meet the NRC requirements at the time of installation ,.
and they' generally meet those of Regulatory Guide 1.23. Any ;
. deviations from Regulatory Guide 1.23 are described in the~ -i
. -following subsections. !
~ The manufacturers' specifications and accuracies for the sensors !
and associated equipment are given in Table 2.3.3-2. Deviation i from paragraph C4 of Regulatory Guide 1.23 regarding the system ;
accuracies is discussed and justified in the following sections t on each type of measurement.
'2.3.3.1.1.3 Windspeed l
' The Bendix Aerovane Wind Transmitter, Model 120, measures ;
windspeed by means of a six-bladed rotor coupled to the armature ;
of a: tachometer magneto located in the nose of the instrument. :
The output voltage is directly proportional to the impeller rotation speed and, therefore, is a function of windspeed. This [
< Aerovane system is used on Towers 1 and 2 at Limerick. f As shown on Table 2.3.3-2, some of the instruments do not meet the required starting speeds. This presents no problem because ;
- real calm conditions with absolutely no, air motion are extremely ?
rare at most sites. Measured calms can be far more frequent, depending on the threshold speed of the instrument used.
At Limerick, the number of calm hours recorded on the six-bladed !
Aerovane is shown in Table 2.3.3-3. All levels of both Towers 1 and 2 are instrumented with these six-bladed sensors. The ;
175-foot instrument at Tower 1 is at the elevation representative '
of vent releases. With only 1.7% calm hours, a more sensitive !
i 2.3-27 Rev. 21, 06/83 t
f LGS'FSAR instrument could not produce any significant improvement. The 30 foot level of Tower 2 does have aWith highthis percentage oflight calmwind hours due ta) .its_ valley location. in mind, instruments meeting the recommendations of Guide 1.23 were installed in the valley on the satellite tower. As shown in
-Table 2.3.3-3, the light wind sensor also produces a large (11.5%) number of calm hours. Experience with both types of instruments indicates that the' continued durability and accuracy of the six-bladed Aerovane far outweighs the advantage of the l
slightly lower threshold _ speed offered _by the light wind instruments. Regulatory Guide 1.23 also specified 90% data
[
j i
recovery, which is considered equally important.
The satellite tower.uses a Bendix-Friez 3-cup anemometer, P/N 2416914, to determine windspeed. The 3-cup anemometer has cone-shaped cups formed of 0.010 inch thickness aluminum. The
' cup wheel is attached to a stainless steelThe shaft which rotates, output voltage is t
via coupling, the tachometer generator.
l directly proportional to the speed of rotation and, therefore, is a function'of-windspeed.
^
f l' 2.3.3.'1.1.'4' Wind Direction The Bendix Aerovane Wind Transmitter, Model 120, measures wind direction by coupling a streamlined vane to a type 1HG synchro.
1
-This synchro electrically transmits the position of the vane and,
! therefore, the wind direction to the recorder, r
l' The satellite tower uses a'Bendix-Friez Wind Vane, P/N 2416970, to determine wind direction. This wind vane is very light and' Changes in azimuth I
sensitive having a low moment of inertia. The signal angle are transmitted, via coupling, to a synchro.
output from this synchro is directly proportional to the angular position of the vane and, therefore, wind direction is j-transmitted to a synchro in the recorder.
l 2.3.3.1.1.5 Temperature i
The ambient temperature measuring system uses Leeds and Northrup 100 ohm copper thermohm sensors (resistance thermometers). to These j
thermohms are accurate to 20.20F across the range of -100 1100F, the detectors use four leadwires, two of which are L
connected to a constant current source and the other two leadwires are connected, via electronic amplifiers to an analog recorder. - . Contained in the constant current loop is the copper causing measuring coil, whose resistance varies with temperature, l
! the voltage drop across the coil to change proportionally. This voltage drop is then sensed by the measuring loop of a null balance potentiometer having a scale calibrated in degrees
( fahrenheit.
l I ' Rev. 21, 06/83 2.3-28 I
LGS FSAR 2.3.3.1.1.6 Temperature Difference l
[
'The temperature difference sensors at the Limerick site are !
identical to the ambient temperature sensors, except for the
~
-selection of matched sets. These_ sets have an accuracy of 20.loF
- across a;-120 to +120F temperature difference range. The reference thermohm (el 26 feet) is connected (opposite in .
polarity) to both. upper elevation thermohms. The two voltage [
~ drops (one from each set) are algebraically added, and the ;
resulting output is equivalent to the temperature difference !
-reading.
l lBoth the ambient temperature and delta temperature sensors are I located in a Teledyne/Geotech aspirated thermal radiation shield, j Model 327. This is to_ ensure the aasurement of ambient t
- . temperature and temperature grad. eats substantially independent !
of solar, atmospheric, and terrestrial thermal radiation.
2.3V3.1.1.7 Relative Humidity l
'The Bendix Hygrothermograph, Model 594, is used at and around the ;
Limerick ~ site to determine both relative humidity and ambient air ;
temperature. I r
The relative humidity portion of the instrument consists of a :
hair-type humidity-responsive. element, a lever system, and a ;
cylindrical chart. The accuracy of the humidity unit is 5% l which includes the temperature effects to which the instrument +
.may be subjected. ;
The temperature-responsive unit consists of a Bourdon tube, a I lever system and a cylindrical chart (same cylinder used for ;
humidity). The accuracy of the hygrothermograph temperature unit -
is 'loF.
Regulatory Guide'I'.23 suggests that at sites where there may be [
oan increase in atmospheric moisture content (i.e., cooling i towers) dew point or humidity should be measured on the tower.
The results of published field studies (Refs 2.3.3-1 through 2.3.3-4) prove conclusively that the only changes in atmospheric "
-moisture characteristics'which may be experienced from cooling tower operation would occur at the plume elevation,.not at the ;
j -ground level. The results of approximately 400 flight. test '
- -observations indicate that the cooling tower plumes would rise clear of the ground and have no effect on the low level moisture characteristics. For dewpoint or humidity measurements to have ,
any relevance to cooling tower effects, they must be obtained at i elevations ranging from approximately 1000 feet to 5000 feet above ground, which is not possible on a continuous basis. ,
Since there is little or no potential for fogging or icing L conditions resulting from the Limerick cooling towers, there is i
2.3-29 Rev. 21, 06/83
~
no need for a dewpoint measurement at the 10 meter level on the i tower.
2.3.3.1.2 Calibration and Maintenance Procedures 2.3.3.1.2.1 C'alibration All sensors and related equipment are calibrated according to '
written procedures designed to ensure adherence to NRC Regulatory
~
Guide 1.23 guidelines for accuracy. Calibrations occur at least every six months, with component checks and adjustments performed when required.
All meters and other equipment used in calibrations are, in turn, calibrated at scheduled intervals. All instruments used in calibrations are traceable to the National Bureau of Standards. ;
I g
2.3.3.1.2.2 ' Maintenance Inspection and maintenance of all equipment is accomplished in accordance with procedures in the instrument manufacturer's manuals. This. inspection occurs at.least once a week by qualified technicians capable of performing the maintenance, if required. The results of the inspections and maintenance performed are kept in a log at the site. The information ;
contained in.this log is a~1so transmitted to the environmental engineering section and meteorological consultant. i i
In-the event that the required maintenance could effect the instruments calibration, another calibration is performed prior to returning the instrument to service.
~
2.3.3.1.2.3 Data Output and Recording systems ,
All meteorological outputs, at this time., are recorded by analog systems. The-charts from these systems are sent on a weekly basis to the applicant's meteorological consultant, Meteorological Evaluation Services, Amityville, New York, for i inspection to detect discrepancies or evidence of malfunction and
. data analysis. ;
-The analog recording systems for the weather towers are enclosed I in a structure with thermostatically controlled temperature.
2.3.3.1.3 Data Analysis Procedures 2.3.3.1.3.1 Data Quality Control All data are subject to a quality check by Meteorological
. Evaluation Services. These analog charts are inspected for the ;
following items:
Rev. 21,. 06/83 2.3-30 ;
1
r
- a. Verification of log sheets versus actual charts received
- b. Time continuity Ec. Instrument malfunction ,
i
- d. Inking problems ;
4
- e. Directional switching problems i
- f. Negative speeds
- g. Missing data l
.An evaluation of system performance is made monthly. The [
percentage of data recovery for Limerick Weather Station instrumentation is shown from 1972.through 1976 in Section 2.3.2.
2.3.3.1.3.2 Data Reduction l LAll readings that are taken from the strip charts represent ,
. ~ hourly averages (except where noted).
Data are reduced into the different categories as follows: }
Wind i a._ Windspeed: hourly average speed. Negative speeds are ,
recorded as read,
- b. Wind direction: hourly average direction
- c. Span: The' span is read-from the same portion used to obtain-the average direction. Span is defined as the
-width of the' direction trace excluding any-abnormal spikes. Maximum span read is 3600 !
- d. Gustiness: Th'e gustiness is read from the same portion of the chart used to obtain the average direction.
Gustiness and its characteristics are described in i Ref 2.3.3-5. !
Temocrature and Humidity [
- a. Hygrothermographs: All relative. humidity and temperature readings taken from a hygrothermograph are 3 instantaneous readings on the hour. l i
.b. Ambient' temperature: Recorded on a strip chart; hourly i average temperature manually recorded.
l l
2.3-31 Rev. 21, 06/83 i
_ _ _ . _ _ _ _ ~ . _ _ _ . _ - - _ _ _ _ _ _ . . _ _ . _ _ _ _ _ . _ _ - _ _ _ _
. 7 F , .
e LGS FSAR-
- c. Delta temperature: Recorded on a strip-chart; hourly average temperature manually recorded.
.f 2.'3.3.1.3.3 Analyses ;
~The-hourly-data obtcined (as described) have been compiled into :
the series of summary tables described in.Section 2.3.2. These !'
data are used as inputs to the computation of the X/O estimates descr'ibed in Sections 2.3.4 and 2.3.5.. l 2.3.3'.2' Operational Meteorolocical Measurement Svstem (1983) >
The meteorological measurement system is being upgraded to comply t with Regulatory Guide'1.23, proposed Rev. 1, and NUREG-0654 f criteria. Table 2.3.3-6 is a list of the meteorological measurements made by the system. !
.2.3.3.2.1 ' Measurements and Instrumentation >
i 2.3.3.2.1.1 Siting As shown in Figures 2.3.3-1 and 2.3.3-2, the main meteorological weather Tower 1 located at Weather Station 1 is a 280-foot tower j situated approximately.3000 feet NW of the Limerick structure vents. Weather Tower 1 is also located approximately 2000 feet
- NNW of the center of the Unit 1 cooling tower location and approximately 2400 feet NW of the center of the Unit 2 cooling >
tower location. Grade elevation at Weather Station 1 is 250 feet i
MSL.
- The wind instruments on Tower 1 are mounted on retractable booms extending upwind 10. feet.0 inches west of the tower. Each face i of the triangular tower is 3 feet 6 inches wide. The temperature sensors are located in aspirators and are 2 feet 0 inches from I
.the tower. Weather Station 1 has a base surface made of yardstone. A. dew point sensor is located on the temperature :
aspirator at the 26-ft elevation.
Meteorological Weather Tower 2 located at Weather Station 2 is a 310-foot tower situated approximately 2100 feet west of the Limerick structure vents. Weather Tower 2 is also located approximately 1950 feet WSW of the center of the Unit 1 cooling tcwer location, and approximately 2600 feet WSW of the center of the Unit-_2. cooling tower location. !
The wind instruments on Tower 2 are mounted on retractable booms
- extending upwind 10 feet 0 inches WNW of the tower. Each face of the triangular tower is 3 feet 6 inches wide. The temperature ;
sensors are located in aspirators and are 2 feet 0 inches from the tower. The dew point sensor is located on the temperature j aspirator at the 26-ft elevation. Weather Tower 2 has a base ,
surface made of yardstone.
Rev. 21, 06/83 2.3-32 l l
t
9 LGS FSAR 2.3.3.2.1.2 Instrumentation and Performance Specifications l The instrumentation systems installed on the Limerick site were designed to meet the requirements of Regulatory Guide 1.23, proposed Rev. 1.
The manufacturers' specifications and accuracies for the sensors and associated equipment are given in Table 2.3.3-7.
2.3.3.2.1.3 Windspeed l The Bendix Aerovane Wind Transmitter, Model 120, measures windspeed by means of a six-bladed rotor coupled to the armature of a tachometer magneto located in the nose of the instrument.
The output voltage is directly proportional to the impeller rotation speed and, therefore, is a function of windspeed. This Aerovane system is used on Tower 1 at the 270-ft elevation.
This instrument is not required by Regulatory Guide 1.23, proposed Rev. 1, and does not meet the required starting speed.
It is included in the system only to provide a comparison with other wind instrumentation.
All other wind speed measurements are provided by Climatronics three-cup anemometers, Model F460. A 30-hole photochopper with an LED photochopper device provides a frequency output directly proportional to wind speed.
2.3.3.2.1.4 Wind Direction l The Bendix Aerovane Wind Transmitter, Model 120, measures wind direction by coupling a streamlined vane to a type 1HG synchro.
This synchro electrically transmits the position of the vane and, therefore, the wind direction to the recorder. This is used on Tower 1 at the 270-ft elevation.
All other wind direction measurements are provided by Climatronics Wind Direction Sensor, Model F460. The sensor consists of a counterbalanced, light weight vane and a precision, low torque, highly reliable potentiometer to yield a voltage output proportional to the wind direction.
2.3.3.2.1.5 Sigma theta (Standard Deviation) l Standard deviation of wind direction is computed using a CMOS microprocessor (1812) with a sampling rate of 1 per second. Each wind sample is converted from polar to rectangular coordinates with northern and eastern components. From this, the standard deviation (<,) is computed over an interval of 15 minutes. 48" 2.3-32a Rev. 21, 06/83
p- - .
LGS FSAR 2.3.3.2.-1.6 Temperature The ambient temperature measuring system uses a standard 100 Ohm, 4-wire platinum RTD. Climatronics platinum temperature translator, P/N 100950, provides a voltage output of 0 to 5 volts corresponding to an ambient temperature of -30 to 1200F. Each
~
~ translator is calibrated to its comparison sensor which enables precise measurement accuracy. The nonlinear sensor response is compensated for by the translator electronics to provide an excellent fit.to the Callendar-Van Dusen Equation. These sensors are housed in a Climatronics TS-10 PWA motor-aspirated temperature / dew point shield.
~ 2'.3.3.2.1.7 ' Temperature Difference The temperature sensors used in determining temperature difference are identical to those used for ambient temperature.
Climatronics platinum temperature difference translator, P/N 100950-1, provides a voltage output of 0 to 5 volts corresponding to a temperature difference of -10 to 200F. These sensors are housed-in the TS-10P motor-aspirated temperature shield.
2.3.3.2.1.8 Dew Point Dew point temperature is measured using Climatronics Lithium Chloride dew point sensor, P/N 101197. Dew point translator, P/N 100089, provides a voltage output of 0 to 5 volts corresponding to -30 to 1200F. The sensor is housed in the TS-10 PWA motor-aspirated temperature / dew point shield at the same elevation as the ambient temperature sensor.
2.3.3.2.1.9 Precipitation Precipitation is measured using a gauge of the " tipping bucket" type. Each tip of the bucket-equals 0.01 inches of water. The gauge has an electrical heater for operation with snow or freezing rain. The gauge is manufactured-by Climatronics Corp.,
P/N 100097-1. The translator for this unit, P/N 100747, consists of an electronic counter and an A-D converter and converts the output counts'to a 0-5 volts analog signal. The counter will recycle to 0 on the 100th count.
2.3.3.2.1.10 Data Communication and Display Data from Tower 1 and from Tower 2 will be recorded on strip-charts at the base of the respective tower and in the control
' room. Data from the satellite tower will be recorded on strip-charts in the control room only. Data from all three locations will also be logged by a data-logger in the control room and input to the radiation and meteorological monitoring system
-(RMMS) in the Technical Support Center. Towers 1 and 2 interface to the control room by means of independent communication lines.
Rev. 21, 06/83 2.3-32b
I LGS FSAR Data from this system will be presented to the EOF and control !
room on CRT displays. The meteorological data is also used by the Class A model for accident dose assessment.
2.3.3.2.2 ' Calibration and Maintenance Procedures l i
(
2.3.3.2.2.-1 Calibration l
All sensors and related equipment are calibrated according to l written procedures designed to ensure adherence to Regulatory Guide _1.23,' proposed Rev. 1, guidelines for accuracy.
Calibrations occur at least every.six months, with component i checks and adjustments performed when required. !
i' All meters and other equipment used in calibrations are, in turn, .
calibrated at scheduled intervals. All instruments used in. '
calibrations are traceable to the National Bureau of Standards.
2.3.3.2.2.2 Maintenance. l Inspection and maintenance of all equipment is accomplished in I accordance with procedures in the instrument manufacturer's l manuals.- This inspection occurs at least once a week by ,
qualified technicians capable of performing the maintenance, if (
required. The results of the inspections and maintenance performed are kept in a log at the site. The information
~ contained in this-log is also transmitted to the environmental i
. engineering section and meteorological consultant.
t
, In'the event that the' required maintenance could effect the ;
instruments calibration, another calibration-is performed prior to returning the instrument to service, j i
2.3.3.2.2.3 . Data Output and Recording Systems l All meteorological outputs, at this time, are recorded by analog i systems. The charts.from these systems are sent on a weexly !
. basis to the applicant's meteorological consultant, :
Meteorological Evaluation Services, Amityville, New York, for )
' inspection to detect discrepancies or evidence of malfunction and l data. analysis. $
'The analog recording systems-for the weather towers are enclosed ,
in a' structure with thermostatically controlled-temperature, j 2.3.3.2.3 Data Analysis Procedures l [
l 2.3.3.2.3.1 Data Quality Control l l All data are subject.to a quality check by Meteorological !
Evaluation Services. These analog charts are inspected for the !
'following items: i i
2.3-32c Rev. 21, 06/83 I i
I
[
- ,y Q'
- 7 ~;--Q ~
- ~ ,
'" j y
.h q' W ^ (4
& '. 7 . ,
i1, .
-u -
e eN . <
7 ,'
w, q Wt C w
q i m
& +
7p y_w(L ,
5 QNm 4Gs1FSAR
-N g. ( '., .
C o,.
.3 x 3% u . ; : x
- a. .Verificction of-log sheets - versus actual charts received n.- , -
' ,t_
- w. sig s \ i a' , e 'n
'b. ATidie'corciauihye , a 3'q s 3 ,
V x, ,~. , , ,
Sy .'c. Instrument :r alfunction
'l'T
...t ,t s
E.'d. JInkf.n -N' 4 ; J . % g,p;roblems ( (, 5 \( ~ c y ,
s 7
Dire'C.,.io ah GwitChing problC C ' ,
- ff
- Negative speedm..
. C - LR r
[ g ,.Q
- g. xa Missing x5 data @~'
s ,
s
- An' eval. Aye: ion o'f. system',pe;fory,*nce is made monthly to ensure s - >
that data i r) recovery /is satisfactory,
~ ..' '
2.-3.375.3.2 Data Reduction '
l All.sreadings that are taken from the strip charts represen* '
hourly averageo (except where npted).
C N, ;s;' 'N ~
s . !
Da.,ta are.reduceddinto^the diffqrent categerits as follows:
c4 Wind . ., A L m .- is 4 .
s a '. ' WE6d::peQ ' hourly average speed. Negative speeds are recorded as read '
-p ,
- ,N% r
- b. Wind direction.N hourly average direction 3 .
t vs x c i
- c. SpanF.Jhd span is 5'ead from the same , portion used to obta'in' the average direction. Span is' defined as the
_ width 3cf the direction trace excluding any., abnormal spikes. Maximum span read is 360 degrees. .
s \ !
5'. -Gustiness ITower 1 sAe cvane only): The gustiness is '
read from the same porticn of the chart used to obtain theJaverage directich 'fors the aerovane. Gustiness and ,
s
'u,._
'its'characte(tstics ,'are described,in'Ref ,' s 2.3.3-5. '
.s ' '
i Temeerature-and Dew Point ,
'a. Dew Point: Recorded 'cn a strip' chart; hourly average i dew point temperatu W manually recorded. '
1 , s
' b. ),hh.ient temperaturstrsRecorded on r strip chart; hourly average temperature -
manually cocorded. l c.'
'DeltaLtemperature: Jecceded'en a strip chart; hourly g average tempe nture manually recorded.
- p, ,r . , q _.
l N .(. -
.7 -a
[
s L .Rev. 21, 06/83 N ?
u ,_2.3-32d I 1
, ~ -
l' 4
- s ,
- *. s
. A L
,, , -w.,,-,,-n-,,.,--r- , ,..,,w--,-n,.-,,-------,-,.-..-~.<,,-a,\-,...~..-~-e-. -<---,e---,
i
+ / -- -
i e
- d. Precipitation: Recorded on a strip chart; each discrete step represents 0.01 inches of' liquid water. The number -
of steps are added to obtain the total precipitation for
.the hour. . ;
2.3.3.2.3.3 Analyses l The hourly data obtained (as described) have been compiled into l the series of. summary tables described in Section 2.3.2. These i data are used as inputs to the computation of the X/O estimates described in Sections 2.3.4 and 2.3.5. !
2.3.3.3 Offsite Meteorolocical Monitorino Locations l l l
TheLimerickmeteorological.dabafromthepreoperationalperiod !
have been compared with offsite data from the Philadelphia and :
Allentown, Pa.. National Weather Service (NWS) Stations and with !
' the data from the Philadelphia Electric Company Peach Bottom .
Atomic Po6er Station. Whenever possible Limerick parameters were !
compared.with concurrent data from the regional stations to !
assess their-similarity, as well as with the longer term records i
' from the regional stations to assess the climatological
- --- representativeness of the. time period during which the Limerick site data were obtained. I i
The following are brief descriptions of the offsite measurement i locations: ;
2.3.3.3.1 Philadelphia l l i
'The Philadelphia NWS station is presently located at the Philadelphia International Airport, approximately 31 miles ,
southeast of the Limerick site. The airport is located on the southern edge of the city, bordered on its southeast side by the l Delaware River. The area is relatively flat, with no appreciable , i terrain roughness to influence the data.
7 The Philadelphia NWS meteorological sensors have been moved several times during the period of record used in the longterm comparisons. In 1960,-the NWS established standard elevations -
for all meteorological sensors, and the instrument locations have :
remained unchanged since that time. A complete history of the !
sensor locations at the Philadelphia NWS station is shown in ,
Table 2.3.3-4. !
2.3.3.3.2 Allentown l [
The Allentown NWS station is located approximately 31 miles north f of the Limerick site at the Allentown-Bethlehem-Easton Airport.
The station is 5 miles northeast of the city of Allentown in the Lehigh River Valley.
2.3-32e Rev. 21, 06/83 ;
.%, , , , , - - m, , % ._ _.,--..n-----w_..-,--...---..,.---,-,---...--w..- .....,,.-w--..-.%,...---.m.m-
. - ,_m- ,.-, . - . .---,m_-m._
c- , , ,
e- '
- f .,
s s ,
. u .
. .a ,
p '-
+ ,
a, ,
-f f . s 3he river' valley is surrounded 'by rolling' terrain and numerous 1/smaill , streams,, but .there are .also some farger terrain features in
//5the. area._. Blue Mountain..is~afridge located 12' miles north of
' Q A11entown, ranging' between 1000 to 1800 feet high'. South
. ? Mountain,-ranging between 500 and 1000 feet high, is located on
'the southern. edge <of Allentown. 'However, neitherfof these two
- 3f:.: 2 q 3 ,.
v f .
1-g i
. s.
.5
' e9. p !?>i
'.$ f f s j $' ,4-t il.
I
- j .s l' e
- q. .p f l.\. - ,.
> , 9; ,
pi i;p -
4>( 3
,, ., .w
.;c J y
+
f : '{' ef [ t .,
,/ t-
+ ,-
y ; 8 e
if s
. .- j ,
~
k.
s
'i
.4 ', /
- ,, / -,
'I j '. ' r -y',' .\ . ,i ./
s ,,.. n ,
l > i v '
- I
- . ) s / 5 g ,
,i '
3
/ -- -
L.- [,N , I i, . / ,1, ,.
- 4
.,M / 'e> '
t i
. \, > ' -,
8 l
-s( ., <
p .c q ./
1/
- s. ; .)i \ )l , .'
. ' 6 at ,, <',.i
^-
l jy . ?.
sk g '.; . ,
iL
'}
\
.).
! J~ ,s -3q/ . ,
-s
.r p .a p p ,':
n ., ,-
. .,i +
,',4 , j 4' t, > ,
s-
.,%)f / 't6
^ f , '
p . ,, . f f ./
t s e+ , /
4 3,
.4 1 .
!- .. : ,h, = .g -
st
'l
- c < ,
s2 ,
< v
./f y '
- g. ~ .,
/
-> j ; -
[
< +
t
, ',. j ,i 'r, . .. . - * ~
)
r -
'. S 4 7
< i-
. ' ' .> Revt . l 2,1, 0 6',f. 8 3 t _ 2.3-32f ,
l
'.- , q ,; ,
a-s r.
'3 f ^! , , ;
p i .
.p ,
L ,, , .
4 ,
c.:
I i
- mountains is close enough to the Allentown NWS station to have l any' direct effect on the local meteorology. The Allentown NWS t meteorological sensors have been moved between various elevations
- and locations during the' period of record used in the long-term !
comparisons, but were moved to the standard NWS elevations in 1965, and have remained-unchanged since that time. The complete i history of the sensor locations and elevations is shown in Table 2.3.3-5.
i 2.3.3.3.3 Peach Bottom Atomic Power Station (PBAPS) l Weather Station No. 2 at the PBAPS is located approximately
~48 miles southeast of the Limerick site. The Peach Bottom plant
- is' located in the Susquehanna River Valley, but Weather Station l No. 2 is a 320 ft. tower situated on a hill overlooking the ,
valley. The 320 ft. wind sensor at Weather Station No. 2 is at an elevation comparable to the upper level Limerick wind sensors, I and therefore provides a useful check of the representativeness j of the-meteorology. ;
- 2 .' 3 . 4 SHORT-TERM (ACCIDENT) DIFFUSION ESTIMATES 2.3.4.1 Obiective
- Estimates of atmospheric diffusion (X/Q) are made at the ,
exclusion area boundary and the outer boundary of the low !
population zone (LPI). These estimates are made for periods of i 2, 8, and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, and for 3 and 26 days following a postulated accident. -The section-dependent model.in Draft Regulatory ]
Guide 1.XXX (Ref 2.3.4-1) is.used.
2.3.4.2 Calculational Progedure J
j The calculation procedure used to determine X/O for the l i
-appropriate time periods-following a postulated accident is described in Draft Regulatory Guide.1.XXX. The diffusion model l presented in-this guide is used to determine X/O values for the X/O values for longer l
first two hours following the accident. l l time periods are determined by logarithmic interpolation between ;
i.
-1 the two-hour accident value and the annual X/O at each receptor point. The: annual-X/O values are calculated using the model
- described in Regulatory Guide 1.111 (Ref. 2.3.4-2). l g
The Limerick emission is classified.as a low-level release This >
4 according-to the criteria of Draft Regulatory Guide 1.XXX.
requires that the source be treated as ground level. This i F assumption is also made in the annual X/O calculations. ;
f i
f i
2.3-33 Rev. 21, 05/83 we-e =* wm-*,r- -m.*--y+---=---+we'w 'p -- '-gNurt**iw eg--Tw---w-4y y-1 m 1*'Mw - '
l
[
LGS FSAR together with 1971 actual use and consumptive use Table 2.4-4, details of individual entitlements (references are listed in Similar information is presented in Table 2.4-5 Section.2.4.15). l
. for industrial users. Groundwater users are discussed in Section 2.4.13.2. ,
2.4.2 FLOODS j i
2'.4.'2.1 Flood History Flood history is discussed in detail in Section 2.4.3.5.2. ,
. Table 2.4-6 shows peak recorded flows at several stations in the :
Schuylkill Basin. Historically, the greatest recorded flood at Pottstown was due to Hurricane Agnes in 1972; however, according to Philadelphia records, there may have been greater floods before the beginning of the Pottstown record. None of the ;
I historic floods on the'Schuylkill River (see Secti.on 2.4.3.5.2) was caused by ice jams or landslides. Surges, seiches, and tsunamis are not relevant to the Limerick site. The consequences t of the failures of upstream dams are discussed in Section 2.'4.4.
2.4;2.2 Flood Desian Considerations The design basis flood level with respect to the Schuylkill_ River j
.is conservatively estimated at el 207 feet. This stage is !
derived from an SPF, combined with the wave crests from three !
simultaneous dam breaks and the 1% wave runup generated by a l 40-mph wind. .Without the wave, the maximum level is estimated at !
el 201 feet. The three dams are Ontelaunee, Blue Marsh, and ;
Maiden Creek-(in the early planning stage). The derivation of i this flood is discussed in Section 2.4.4. i The lowest grade-level entrance to any safety-related structure is at el 217 feet, which is 10 feet above the design basis flood [
Therefore, Schuylkill River floods cannot affect level (DBFL).
any of the safety-related facilities.
The Schuylkill niver PMF is conservatively estimated at 500,000 ,
cfs, based on Appendix B of NRC Regulatory' Guide 1.59, l Revision 2. When combined with a simultaneous dam-break flood wave due-to a PMF-induced failure of Ontelaunee Dam, the highest :
stage obtained at Limerick was el 181 feet (without wind-wave).
This-is well below the stage obtained from the multiple dam-break, as given above. The Schuylkill River PMF is discussed in detail in Section 2.4.3. ;
The water surface (el 201 feet) resulting from the dam-break ,
analysis is transient. It is estimated that the time required :
for the dam-break flood wave at the river cross-section at ;
l Limerick to rise above and subside back to the SPF elevation (152~ feet) is approximately 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The length of time for ;
i 2.4-4 i
f
.O -
6 LGS FSAR which-the flood wave will be above el 177 feet and 195 feet would i be-about 4.5 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively. -
The shortest distance from the el 201 feet (dam-break flood elevation without wave runup) contour to the nearest safety-related struct ure inland (diesel oil storage tanks) is about i 126 feet. The foundation of-these structures is at el 194 feet. l For floodwaters to reach these structures, percolation through !
the embankment would have to occur. An analysis of the !
percolation was made using the following equation, assuming that !
Darcy's law is valid and that flow is one-dimensional: [
e i
ab g .ax2 3A .11,@l (2.4-1) !
,as.
ax ax k at i where: ;
b(x,t) = saturated thickness b = average saturated thickness k = hydraulic conductivity n = , porosity ,.
z(x) = elevation of base of porous medium -
l Subject to the following-conditions .
b(0,t) =
Ho.
b(x,0) = 0 l where:
Ho = the available head at x = 0 It.can be shown that the solution to the above equation can be written in the form:
- x - -
x -
b = h e * .erfc 2a G -acG
~ . .
+ erfe 2a G +ac/T (2.4-2)
!. 0 2.4-5 l-b
t ,
-where: '
-a = .c= g a = slope'of'the base of the porous medium
_The base'of the porous' medium-(base of embankment material) was Lassumed to rise linearly from el 170 feet at the toe of the embankment'tc'el 194 feet at the tanks. The results of the
! analysis indicate that, if a~ conservative permeability of 5x10*:ft/yr is assumed and if1the flood crest remains at el.202 feetJ(Section 2.4.3 shows maximum flood elevation of 201 feet) for a duration of two hours, groundwater would not'be above x el 194 feet anywhere within 80 feet of the tanks. Therefore the
' dam-break flood wave would not affect hydrostatic pressures on
-the foundations of. safety-related structures.
2.4.2.3 Effects of Local' Intense-Precipitation The latest updated e' stimate-(1976) of point (10 square miles) PMP
.was obtained from the National Oceanic and Atmospheric Administration (Ref 2.4-2). In the updating,'which was subsequent to Hurricane Agnes, relatively small increases were made compared to pre-Agnes PMP estimates from Hydrometeorological Report No.J33 (Ref 2.4-3). The. updated values are given in Table
- 2. 4 -7. -
i.
' As explained later, the' impact of-the onsite FMP was analyzed by
-dividing the site area into smaller subareas. For this purpose, the 72-hour PMP was divided into 6-hour increments using the
' distribution given in-Ref 2.4-2. The 6-hour PMP was divided into
- l. 1-hour increments using the U.S. Army _ Corps of Engineers l'
distribution for standard project storms (Ref 2.4-4). These distributions are shown in Table 2.4-7. The 1-hour PMP was further subdivided into 54 10- and 15-minute increments using the
. distribution given in Ref 2.4-29.
~
To estimate the peak runoff from each subacea, the rational formula was conservatively used:
0 = CIA (2.4-3) where:
0 = ' peak flow rate in cfs
. 2.4-6
(
3.- !
O LGS FSAR C= runoff coefficient I= rain fall-intensity in inches / hour corresponding to a duration equal to the time of concentration, and
_A=. area in acres. j
- The runoff coefficient, C, was conservatively assumed to be 1. I Kirpich's formula (Ref 2.4-34) was used to estimate the times of concentration for different subareas.
To obtain conservative results, the following assumptions were {
used in the analysis - l
- a. The site drainage system was assumed to be completely l blocked. :
i
- b. The peak flows from the respective subareas were assu. d :
to reach the respective collection points at the same ,
time, with no lag. It was also assumed that storage due to.ponding did not attenuate the outflow rate from any of these subareas.
- c. To estimate the rate of flow over embankments or
- roadways, a weir coefficient of 3.0 was assumed. This ,
assumption was based on the results of tests conducted '
by the USGS (Ref 2.4-5)._ These tests indicated that the !
discharge coefficient is not.affected by downstream :
submergence of up to.80%. To account for storage in the i spray pond, the impact of FMP on the contributing- !
. drainage area was analyzed using standard flood-routing !
techniques (Section 2.4.2.3.2). l
' 'As shown in Figures 2.4-4 and 2.4-5, the plant' site is divided
-into three main functional subaceas: l
- a. The turbine-reactor complex area, at el 217 feet .
- b. The cooling-tower. area, at el 257 to 265 feet
- c. The spray pond area' ;
The cooling-tower area is the highest of the three functional !
areas and is-located to the north of the turbine-reactor complex i area.. The spray pond is located north of the cooling-tower area l but is separated ~from it by approximately 300 feet. The .
' intervening land rises'(Figure 2.4-5) to a hilltop at i el 285 feet, about 800 feet east of the north-south plant :
centerline (E4000). Runoff from this hilltop would be directed [
partly to the cooling tower area; partly to the spray pond area; partly to Possum Hollow Run; partly to the turbine-reactor complex area; and partly to Sanatoga Creek. .
Sanatoga Creek is ;
f 2.4-7
--*e-- -d .tw.,,4y_g,g%,-.-owg-,7,-+,-e. yay..,,,,,ew,--, ,w,,,y,-,,q-y,--wewr-ygy,,ws_,,.--c-ww--5,-,w,,,p.ww,.w,_v,.,wm.,_ey.w.,.-,,,-e,_,- .,.,,,,m,,.,-,y.ww_wy,m, , , .
not shown-on Figures 2.4-4 or 2.4-5. It is located north of the spray pond area and drains to the Schuylkill River upstream of the plant site.- Figure 2.4-11 shows the confluence of Sanatoga Creek with the Schuylkill River-near Sanatoga.
Runoff from the three main functional areas drains toward several
' low-points, which in turn drain away from the site. Numerous local drains and small. surface ditches have been provided in the site drainage system to facilitate the drainage of normal storm runoff.- However, as noted above, for the investigation of onsite
.PMP, none of these drainage. facilities is assumed to function, except for the open channel portion of a ditch draining the cooling-tower area. Otherwise, all flow is assumed to be surface flow, over land or over roadway. All drain pipes and culverts are ascumed plugged.
The impact of a PMF in the Possum Hollow Run, which passes along the eastern and southern sides of the site area, is examined in Section 2.4.2.3.5. The impact of a PMF in the Sanatoga Creek is discussed in the following paragraph.
Sanatoga Creek drains an area of less than 10 square miles. With respect to the plant site, the nearest point on the creek is approximately 1400 feet upstream of its confluence with the Schuylkill River. At this location, the thalweg of the creek is approximately at el 127 feet. As shown in Figure 2.4-4, the spray pond is located mostly within.the Sanatoga Creek Basin.
The cooling towers'are. located on a ridge that rises in an east-northeast direction and separates them-from the spray pond area. The same ridge forms the drainage boundary between Sanatoga Creek and Possum Hollow Run and isolates the turbine-reactor area from Sanatoga Creek. The lowest elevation in the i vicinity of the spray pond is el 240 feet, and the crest of the spray pond spillway is at el 251 feet. The SPF elevation of the :
i Schuylkill River near its confluence with the Sanatoga Creek is !
! estimated to be 155 feet. It is inconceivable that the water f L surface elevation in Sanatoga Creek,.with backwater that is due :
to a concurrent SPF in the Schuylkill River, would rise higher ,
than el 240 feet, which is 113 feet above the creek thalweg and j 85 feet above the SPF elevation of Schuylkill River. For reasons !
l given in Section 2.4.3.5, it is considered unlikely that a PMF on !
-the Schuylkill River would coincide with a PMF on the Sanatoga ;
Creek. Therefore, it was concivded that a PMF in Sanatoga Creek f would not endanger safety-related structures, and further >
detailed analysis was not considered necessary. l
. The following sections describe flow-routing assumptions, l drainage areas, and the water-surface elevations resulting from a :
'PMP on the site area. Table 2.4-7a summarizes the drainage :
characteristics, probable maximum rainfall intensity for the critical duration and the resulting peak discharge for each of +
the subareas. ;
i Rev. 6, 06/82 2.4-8 f
I
2.4.2.3.1 Drainage from Cooling-Tower Area
-As shown in Figure 2.4-5, the cooling-tower area and nearby ;
natural topography are divided into three drainage areas, designated as DA-2, DA-3, and DA-4. ,
Area Desicnation -Area (Acres) Peak Dischtrae (cfs) i DA-2 25.0 620 l DA-3 4.7 151 l l t
DA-4 13.0 357 j 2.4.2.3.1.1 Drainage Area DA-2 i The runoff from DA-2 collects in part at the road junction at el 257.7 feet, located at the northern central part of the cooling-tower area. This point is designated as CP-1 (collection point-1) on Detail 3, Figure 2.4-6, Sheet 2. From CP-1,.the ,
runnoff drains westward along the roadway and the adjacent ditch !
and picks up runoff from the remainder of DA-2. Finally, the !
entire runoff from DA-2 passes through a low. point in the roadway '
(CP-2) and enters a ravine that drains to the Schuylkill River (Detail 4, Figure 2.4-6, Sheet 2). To the south of CP-2, the :
-access roadway, which partly encircles the Unit 1 cooling tower, i rises to a'high point that forms the boundary between DA-2 and -l DA-3.
l The most important aspect of drainage from DA-2 is the height to '
which water is ponded against the roadway fill that spans the distance between the two cooling towers on their south flank.
This fill has a nominal top of el 264 feet. As long as runoff [
from DA-2 follows the route described and does not overtop this roadway fill, it need not be added to the flows entering the ,
, turbine-reactor complex area. i
-To estimate the maximum water depth against this embankment, it is assumed that all the runoff from DA-2 collects at CP-1 and :
passes through critical depth over a. conservatively assumed i effective flow width of 30 feet. Critical depth must occur at :
this point, since flow into the ditch downstream is ;
supercritical. This is shown in Section E-E' on Detail 3 in Figure 2.4-6, Sheet 2, which shows the assumed effective width ,
and the discharge rating curve for CP-1. The water surface i elevations at CP-1 and along the roadway embankment are'shown on :
Section A-A' in Figure 2.4-6, Sheet 3. The effective width of ;
30 feet was selected arbitrarily to approximately correspond to ;
the width of the main stream entering the ditch. However, flow l would spread to a larger width of'92 feet, as shown in Detail 3 and Section E-E' of Figure 2.4-6, Sheet 2. Consequently, the {
maximum water surface elevation is less than the estimated value j 2.4-9 Rev. 6, 06/82 I I
I
, _ - . , , , - . , , .. , . - _ _ __, . _ . _ . - _ _ . , _ . . - _ _ , _ , . . _ _ . - ~ . _ - . - . - _ . _ _ - - . _ . . - - - . _ . . .
=of cl 261.3 feet, shich is based on an effective flow width of 30 feet. 'As shown on Secti9n A-A', Figure 2.4-6, Sheet 3 and on Section C-C', Figure 2.4-6, Sheet 2, downstream of CP-1, the flow enters a ditch (supercritical flow with a normal depth of 4.1-feet) and then backs up over the roadway at CP-2 to a depth sufficient to pass the peak flow of 620 cfs. To compute the
- resulting water surface elevation at CP-2, a rating curve was
. prepared for'Section D-D' and is shown in Figuie 2.4-6, Sheet 2.
Por Section D-D', an effective horizontal flow width of 60 feet is assumed at el 243.3 feet. As can be seen from figure 2.4-6,
>- Sheet 2 (Section-D-D'), this is a conservative assumption. Using ;
i
.the D-D rating curve, the water surface elevation at CP-2, :
corresponding to a discharge of'620 cfs, is estimated as !
el 245.6 feet. This is well below CP-1 (el 257.7 feet) and the f high point-(el 250 feet) on the access road separating drainage !
l : basins DA-2 and DA-3. Thus, it is concluded that, through CP-2, DA'-2 drains to the Schuylkill River without overflowing into- ;
- DA-3. ;
2.4.2.3.1.2 Drainage Area DA-3
'As shown in Figure 2.4-5, DA-3' comprises the southwest part of the cooling-tower area. It drains generally into the western
- half of the turbine-reactor complex area, both down the roadway and.down the-south face of the slopes of the access roadway embankment. . DA-3 includes natural ground and the roadway to the 220-kV switchyard, but not the switchyard itself, which is sloped generally south to drain toward the Schuylkill River. A peak flow of 151 cfs is generated from DA-3.- The disposition of this discharge is discussed in Section 2.4.2.3.3 with the l- turbine-reactor complex area.
2.4.2.3'.1.3' Drainage Area DA-4 Like DA-3, this area.also drains directly to the turbine-reactor complex area on the eastern side of the plant centerline.
l However, it includes a relatively larger proportion of natural catchment, most of which is on the east of the cooling-tower excavation arna, The handling of the peak discharge from this area,.357 cfs, is discussed in Section 2.4.2.3.3.
2.4.2.3.2 ' Drainage from Spray Pond Area The spray pond drainage area, shown as DA-1 in Figure 2.4-5, includes the spray pond.itself and the cut-slope areas draining IV toward it, as well as two small pieces of natural topography that drain toward the cut slopes (one on the northwest and the other
- en the southeast perimeter). The total drainage area is
[
- 17.2 acres. Drainage is to the pond itself, which has'a normal operating level of El. 251 feet and a 60-foot wideThis spillway set i
at El. 252 feet (shown as CP-5 on Figure 2.4-5). spillway h
Rev. 19, C4/83 2.4-10
i
~
drains to a natural channel, across a road embankment, and then l northward.to Sanatoga Creek. ,
The spray pond spillway is designed to pass a routed PMF (48-hour !
storm) preceded by an SPF assumed equal to one-half the PMF !
ordinates. No infiltration is assumed. The. pond is assumed to be at normal water surface at the beginning of the storm. A ;
30-minute routing period is used, with a flood routing computer '
program adapted from the U.'S. Army Corps of Engineers program for hydrograph combining and routing. Precipitation increments for the most critical six hours of"the storm are arranged as <
followsr !
Precipitation Precipitation ;
Period (Percent Period (Percent !
(30 min) of 6 hr) (30 min) of 6 hr) j 1 5 7 27 2
l 2 5 8 8 l
3 6 9 7 ;
i 4 . 7 10 6 l 5 7 11 6 ;
i
.6 11 12 5 l t
The remainder.of the storm is distributed in accordance with l Ref 2.4-4. i
- The pond is filled to a maximum of El. 252.5 feet at the ,
I beginning of the 28th hour.- The spillway discharges the surcharge above El. 252 feet before the beginning of the second storm. The secondRflood peaks at a maximum mean inflow cf ,
251 cfs, with a corresponding maximum outflow of 194 cfs and a ,
maximum water surface of El. 253 feet. l ;
[
Further details for the pond are given in Section 2.4.8. l 2.4.2.3.3 Drainage from Power Plant Complex Area 3 The finished floor elevation of relevant safety-related [
structures is el 217 feet. The coordinate line (E4000) divides i the plant area approximately in half, with the divide at i el 217 feet. Drainage is generally away from the structures. ;
The roadways form broad-crested weir-type berr.s that vary from !
el 216.6 feet to el 215 feet. In the western half of the area, l the encircling railroad, which is generally at el 217 feet, channels all runoff to a low section on the west side at i el 215 feet. The west half is designated as DA-5 and the east !
2.4-11 Rev. 19, 04/83
half as DA-6, as shown on Figure 2.4-4. The drainage. areas and l
- peak flows;from these areas were estimated to be as follows: j
- Area Designation Area (acres) Peak Flow (cfs) i DA-5 8.3 206 DA-6 8.5 211
~
2.4.2.3.3.'1- Drainage Area DA-5
.The service railroad track forms the boundary for DA-5, except l' for~the northern edge, which is the southern edge of DA-3. All along its length, the top of the rail is set at el 217 feet, !
except for 400 feet on the west side of DA-5, which is at el.215 feet,.and 400 feet of vertical ~ curve on each side of the i depressed reach. The depressed reach of the track forms an embankment weir discharging all storm runoff from DA-5 westward !
- over the plant bench, and from there to the Schuylkill River (see !
Figure 2.4-4).
t While the railroad is the final outlet from the area, a much .
I narrower and more critical control exists within the area bounded by.the railroad. . This is CP-3, a depressed section of the j service road, shown in Detail 1 of-Figure 2.4-6, Sheet 1. The
. yard and road elevations encompassed by the railroad track vary ;
within'the bounds of 217 feet and 215 feet. The roadways l
-entering the structure have crowns set at el 217 feet or lower e and are sloped to drain away from the structures. All roof ,
drains were assumed plugged. The roads in the southwest part of DA-5 are set ~at el 216.6 feet. Drainage from the area contained :
between the roads and the structure ponds to el 216.6 feet before t
. flowing over the roadway and draining either towards the railroad !
track or CP-3. The area containing the refueling water storage -
tank and the condensate storage tank has a dike and wall ;
surrounding it with a top elevation 223 feet. The dike and wall j cause ponding of the entire storm (under the assumption of clogged drains), to a water depth of 40 inches for a 72-hour i storm. - If the dike fails, the entire runoff drains through CP-3. !
- For computation of the flow at CP-3, it is assumed that the ;
entire runoff from DA-5 drains through CP-3. The attenuation !
provided by the'above storage is not considered. !
In addition to waters originating in DA-5, there is a significant amount of runoff coming from DA-3 (the southeastern cooling tower l drainage area) into DA-5. Some of these waters pass directly to i the depressed track section without going through CP-3. It is !
also possible that waters from DA-3 coming down the cooling tower- l access road may have sufficient momentum to flow into DA-6. To !
?
be conservative, both CP-3 and CP-4 were checked for their ability to handle 100S. of the DA-3 flow, in addition to the respective local flows (CP-4 is discussed under drainage area ;
i 5 4
2.4-12 i
DA-6, below). . These assumptions result in a total discharge of 357 cfs at CP-3, 206 cfs from DA-5, and 151 cfs from DA-3. '
i The location of CP-3 is shown in Detail 1 of Figure 2.4-6, :
.Sheetc1. It can-be-seen that the depressed section at el 215 feet lies on a horizontal 900 curve in the service road.
The base width at el 215 is 78.4 feet. To be conservative, overflow through the triangular portions of the trapezoidal
- section shown in Figure 2.4-6, Sheet 1 (Section F-F') is neglected. The resulting rating curve for CP-3 is shown in
' Figure 2.4-6, Sheet 1. From this curve, the water surface ;
elevation at CP-3, corresponding to a peak flow of 357 cfs, is ;
estimated as el.216.3 feet. !
l It is.therefore concluded that even if the runoffs from both DA-3 :
and DA-5 pass.through CP-3, the water surface elevation upstream j of CP-3 does not. rise above el 216.3 feet, except in the ponded ;
areas enclosed by the roadway where water can pond to el 216.6 :
l: feet.and.in the diked areas where ponding to el 220.3 feet can l L occur. ;
2.4.2.3.3.'2 Drainage Area'DA-6 In the south and east parts.of DA-6, some ponding occurs between .
the turbine-reactor complex area and the roadway at el 216.6 feet.. However, when the water surface rises above L, el 216.6 feet, it drains to the south and east along the entire i length of the roadway. The lowest point within DA-6 is at CP-4, j l located at the northeastern corner of DA-6. Although some runoff <
i .. .from-the south and east sides of DA-6 runs directly off, it is !
! -conservatively assumed that all drainage from DA-6 (211 cfs) discharges through CP-4. In addition, it is assumed that the t y . peak flow of 357 cfs from DA-4,.together with the peak flow of
-151 cfs-from DA-3, also. passes through CP-4. This results in a ,
combined peak flow of 719 cfs passing through CP-4.
'The horizontal geometry of CP-4 is shown in Detail 2, Figure 2.4-6, Sheet 1. An effective weir length of 120 feet is .
assumed at el 215 feet. Section G-G',~ Figure 2.4-6, Sheet I and '
the associated rating curve for CP-4 show that a water surface
. elevation of 216.6 feet is sufficient.to discharge the estimated peak-flow of 719 cfs. Thus, it is concluded that even if the a entire flow from DA-3, DA-4, and DA-6 passes through CP-4, the resulting water surface elevation (el 216.6 feet) is below the .
Iowest grade-level entrance.to any safety-related structure, i
-which is at el 217 feet.
2.4-13 1
. 1
-2.4.2.3.~4 Roof Loads-on Safety-Related Structures That Are Due 3
. to PMP Onsite ;
In! the' previous analysis of surface drainage, it is assumed that i
~
' all-roof drainage overflows'to_the ground and then-to the various (
control points.' If all roof drains and scuppers are blocked,
- i water could pond on the roofs of'some safety-related structures [
to a depth controlled.by the height _of the roof-parapets. The a
. highest parapet on any safety-related structure is 4 feet '
3 inches. 'The. maximum 24-hour PMP is 34.4 inches. The latter
~ produces an equivalent uniform roof load value~of 180 lb/fta. - :
, However, assuming that some accumulation and overflowing occurs, t
- the maximum water depth could equal'the height of the parapet ;
plus a!.small amount of head providing flow over the parapet. The i maximum hourly rate of 10.2 in/hr is equivalent to 4
0.9 fts/fts-hr. A depth of 6 inches over the parapet provides a t
- capacity of 3900 fts/hr/ft , for a head of 0.5 feet on the crest.
.Thus, a;1-foot length of parapet, with a head of 0.5 feet, can i discharge the runoff that is due to the maximum hourly PMP i covering a roof area of 4300 square feet. If all drains are i
~
assumed to be plugged, the maximum accumulated water depth over j the roof is 4.75 feet, equivalent to a roof load of 296 lb/fta, The design roof loads of all the safety-related structures exceed j this value (Section 3.2). ,
2.4.2.3.5 PMF in Possum Hollow Run l i
Possum Hollow Run has a drainage area of 1.3 square miles. It I
rises approximately 2.5 miles northeast of the Limerick site and flows southwesterly, entering the Schuylkill River through a ;
I gorge along the south side of the station. ,
To assess the flood hazard to the Limerick site posed by Possum Hollow PMF, the PMF is assumed to occur in Possum Hollow Run at the same time that an SPF is occurring in the Schuylkill River.
The Schuylkill River SPF is assumed to be 50% of the PMF, or 250,000 cfs, which results in a Schuylkill River stage of ,
el 152 feet, i 1
'It is unlikely that a PMF en the Schuylkill River would be l coincident with the PMF on the Possum Hollow Run. A PMF on the .
Possum Hollow Run (drainage area = 1.3 square miles) is caused by [
a local intense thunderstorm, while a PMF'on the Schuylkill River i (drainage area = 1170 square miles) is due to a basin-wide PMP :
f storm system whose center lies well upstream of Limerick.- The two storms would have different characteristics, and the joint i probability that the'y produce peak runoffs at Limerick at the ;
same time-is very low. .
Using-the slope area method, a rating curve was developed for the l Possum Hollow Run. For this purpose, a typical cross-section is ;
taken at the point where the el 152 feet contour crosses the L !
i 2.4-14 i
i
- . I l
LGS FSAR stream. The bedslope of the Possum Hollow Run is 0.02. Based on !
field inspections and comparison with photographs of streams with !
- known n values-(Refs 2.4-6 and 2.4-33), a Manning's n value of l 0.05 is assumed. :
A PMF hydrograph was developed for the Possum Hollow Run using !
the procedure outlined in Ref 2.4-6. The 6-hour PHP
. (Table 2.4-7) is divided into half-hour increments, following the
- distribution used for the spray pond (Section 2.4.2.3.2). The resulting hydrograph peak is 3840 cfs, and the base is i i 15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. j e
Using the rating curve developed for the Possum Hollow Run, the ;
water surface elevation corresponding to a PMF discharge of ;
3840 cfs is estimated to be 159 feet. The corresponding normal l depth, velocity, and Froude number are 7.0 feet, 8.6 ft/sec, and (
0.76, respectively. !
I
- Jur stated earlier, the SPF elevation in the Schuylkill River is l The bed elevation of the Possum Hollow Run at its
' I 152 feet MSL.
- ~ confluence with the Schuylkill River is 105 feet MSL. Therefore,
~
at the time of the SPF, the flood water of the Schuylkill River l enters the Possum Hollow Run up to a point where its thalweg is ;
i at el 152 feet. This point is about 2400 feet upstream of the i confluence of the Possum Hollow Run with the Schuylkill-River. !
The PMF on the Possum Hollow Run is superimposed on the SPF'in !
the Schuylkill River. It is conservatively assumed that in this ,
backwater reach of 2400 feet the cross-sectional area of the i Possum ~ Hollow Run below el 152 feet is ineffective, and the i entire PMF flow passes through the remaining cross-sectional i
- area. This results in a water surface elevation of 159 feet at a j section 2400 feet upstream of the confluence. This flood level :
is about 57 feet below the plant bench. Upstream of this !
section, the Possum Hollow Run is separated from the plant by 1 high ground that is more than 60' feet above the bed of the Possum l Hollow Run. It is not credible that the PMF of the Possum Hollow Run_(3840 cfs) causes a water depth of 60 feet or more.
Therefore, it is concluded that the PMF in the Possum Hollow Run ;
coincident wih the SPF in~the Schuylkill River would not flood any safety-related structure at Limerick.
l 2.4.2.3.6 Summary of Results - Local Intense Precipitation Table 2.4-8~ summarizes the results of local intense precipitation ;
investigations at the associated collection points (CP) and l drainage areas (DA). These investigations show .that water levels j are below the access elevations of the safety-related structures, j The Possum Hollow PMF is treated apart from the onsite local !
intense precipitation. A maximum PMF discharge of 3840'cfs was !
calculated. An SPF occurring simultaneously in the Schuylkill
[
b 2.4-15 l s
+-v+--- -- . . , - - - ,.*- , y. ,-,.,,.--#,. ,,,,,-,n.,,- .,,,,,~m,w,,,,,,em ,,,,-.,s,.m_,.%,,, -.-..,%,.
I LGS FSAR River; produces a backwater effect, but conservative estimates !
-indicate that the water surface elevation would be well below the plant grade el 216.5 feet.
2.4.3 PROBABLE MAXIMUM FLOOD (PMF) ON STREAMS AND RIVERS The PSAR included a PMF developed by the U.S. Army Corps of i Ingineers for Pottstown, adjusted to the site by the ratio of l drainage areas. This PMF was estimated at 356,000 cfs, with a i stage at the plant site of 158 feet. After the PSAR was 7 prepared, Hurricane Agnes occurred (1972). In 1975 the PSAR was l
- updated to reflect the effect of Hurricane Agnes. At that time !
tnere was no reason to change the PMF. But later the National i Weather Service revised the values of probable maximum ;
precipitation for the basin. Another significant change occurred }
Construction began on Blue Marsh with respect to upstream dams.
Dam, and Maiden Creek Dam became an authorized project, whereas ,
at.the time of the PSAR, construction on Blue Marsh had not .
started, and Maiden Creek was not expected to be authorized. l These changes combined to require a new flood analysis. j Since the Limerick PSAR was prepared, NRC Regulatory Guide 1.59 L Revisions 1 and 2 were issued, giving the option of using either i detailed flood routing studies-(Appendix A) or enveloping maps i for determining peak PMF flows (Appendix B). The latter method
~
is simpler, but more conservative, giving a value of 500,000 cfs '
as compared with the PSAR estimate of 356,000 cfs (both before i adjustment for a dam bresk). The Appendix B method was selected for this report because it is conservative and indicates a ;
-" dry-site" condition during the PMF, with a large margin of ,
safety.
PMFs were computed for the Schuylkill River at Limerick and for Possum Hollow Run. The computations show that the maximum i resulting stage in the Schuylkill River is el 181 feet. The maximum stage that is due to a PMF in Possum Hollow Run was not !
computed, but was found to be less than el 186 feet at a point nearly due east of the turbine-reactor area complex. The PMF for [
the Schuylkill River is covered in this section. The PMF for Possum Hollow Run is discussed in Section 2.4.2.3.5. j 2.4.3.1 Prebable Maximum Precipitation (PMP) i
'This section is not applicable because the PMF is estimated in accordance with Appendix B, NRC Regulatory Guide 1.59. !
2.4.3.2 Precipitation Losses This section is not applicable because the PMF is estimated in !
accordance with Appendix B, NRC Regulatory Guide 1.59. l i
2.4-16 ;
t L f
I i
c
~2.4.3.30 Runoff and Stream Course Models This section is not applicable because the PMF is estimated in accordence with Appendix B, NRC Regulatory Guide 1.59.
2.4.3.4 Probable Maximum Flood Flow
-- :From Appendix B of-NRC Regulatory Guide 1.59, the PMF for the Schuylkill River at the Limerick site, corresponding to a drainage area of 1170 square miles, is 500,000.cfs.
[
The design flood for Ontelaunee Dam is 41,000 cfs. The PMF~ i inflow to Maiden Creek Dam (which may be built at the headwater i of Lake Ontelaunee)1has been estimated by the U.S. Army Corps of Engineers at 118,000 cfs, with-a spillway peak outflow at l 92,000-cfs. t Blue Marsh Dam was designed by the U.S. Army. Corps of Engineers '
to pass a PMF without failure. Under current authorization plans, the U.S. Army Corps of Engineers plans to design the Maiden Creek Dam spillway for the PMF. Both of these projects attenuate their PMF inflow peaks considerably. For Blue Marsh, [
the. attenuation is from~ 128,600 to 74,800 cfs, and-for Maiden ,
Creek, itfis from 118,000'to 92,000 cfs. .In the analysis of l flooding that is due to a dam break, it is conservatively assumed ,
I
.that no attenuation occurred during the PMF passages through either Blue Marsh or Maiden Creek Dam. .
The flood stage'at-the-Limerick site for a 500,000 cfs PMF' peak ,
is el 174 feet, not accounting for a hypothesized failure of !
10ntelaunee Dam-(Section 2.4.3.5). .To assess stages produced by the flood wave caused by the PMF-induced failure of Ontelaunee l i Dam,'the results of a study of dam failure permutations are used. !
<' The method of analysis used in this study is the'same as described in Section 2.4.4.2 (Refs.2.4-23 and 2.4-24). One of -
the permutations'in this study postulates the failure of Blue Marsh Dam superimposed on the SPF in the Schuylkill River. This !
results in a water surface elevation of 177 feet at Limerick. !
The flood wave caused by the failure of Ontelaunee Dam is [
superimposed on the above condition, resulting in a water surface -
Thus, the rise in water 1 elevation of 184 feet at Limerick. :
surface elevation that is due to the failure of Ontelaunee Dam is 7 feet. f t
'The assumption is made that a failure of Ontelaunee Dam also !
produces an incremental increase in stage of 7 feet at Limerick. l Thus,_the water' surface elevation for the PMF plus a failure of i Ontelaunee, is estimated as el 174 + 7 = 181 feet. This ;
elevation (el 181 feet) is,19.7 feet lower than the water surface ,
elevation (el 201 feet) caused by the hypothesized seismically ,
induced failure of three major dams upstream of Limerick (Section i
2.4.4.2). Therefore, the latter elevation governs plant safety, j t
2.4-17 l f
J i
I
+
4 - - and a more refined analysis for the water surface elevation -
caused by1the PMF combined witn the failure of Ontelaunee Dam is
- not warranted.
!2.4.3.5 ' Water Level Determinations ,
r A-discharge rating curve for the Schuylkill River near the I project site.(Figure 2.4-7) was-developed using observed flood j
~
levels, computer backwater studies, and slope-area methods. The procedure that was used-is discussed in the following sections.
2.4.3.-5.1 Data Availability l
- Stream flow data are available in USGS publications (Refs 2.4-7 ;
through 2.4-11). Within the basin and upstream of the plant !
- site, 22 gaging stations have been operated, with 11 presently sactive. The stations are listed in Table 2.4-9, and their locations are indicated on Figure 2.4-8. Additional data are !
~
contained in References 2.4-12, 2.4-13, and 2.4-14. Data on '
- river and' flood profiles are available from the Commonwealth of Pennsylvania, U.S. Army Corps of Engineers, and U.S. Geological l
- Survey'(USGS), in addition to a special high-water mark survey commissioned by Philadelphia Electric Company (PECo) in 1972 *
(Figure 2.4-9).
2.4.3.5.2 Historical Floods l
. Flood-producing storms in this area are normally associated with ,
tropical disturbances. Although flooding from snowmelt occurs annually, snowmelt runoff ucually has not been associated with *
,: major historic floods. Peak stages and discharges published by the USGS and the Corps of Engineers for the major historic floods are given in Table 2.4-6 for several stations on the Schuylkill .
River. At Pottstown, the 1902 flood, with a peak discharge of 53,900 cfs,'was the highest known until. June, 1972. However, the '
Reading and Philadelphia data indicate that the 1902 flood was very likely exceeded in 1850 and 1869 and may have been exceeded in 1757 and 1839. ;
In June, 1972, Hurricane Agnes proiuced the flood of record on ;
many Pennsylvania streams. The flow at Pottstown has been evalaated as 95,900 cfs by the USGS (Ref 2.4-15). Figure 2.4-10 shows the flood-frequency curve for the Schuylkill River at .
Pottstown. This curve is based on composite regional flood :
discharge relationships given in Ref 2.4-16. It is not expected l that the 1972 flood alters these regional relationships. ;
2.4.3.5.3 Water Level-Determinations up to 356,000 cfs :
Table 2.4-10 gives the values from which the rating curve, shown
-in Figure 2.4-7, was drawn. For flows above 20,000 cfs,-bridge clogging is assumed (Section 2.4.3.5.3.5).
2.4-18 i
,m.. .._._m . . . - - , . , , . . . ~ _ , _ _ , . _ . . . . . . _ _ , _ . _ . _ . . _ , _ , . , _ . - - - - , - , , _ _ - _ _ , , _ , , . . ,,_,,m.,__.. _.,_.m.__._.
e'.
LGS FSAR After completion of the studies that resulted in the rating curve shown.in Figure 2.4-7, Hurricane Agnes produced the flood of
-record at Pottstown, 95,900 cfs. Some seven hours before the
> -flood's peak, an oil lagoon at-Pottstown was overtopped by the
~ flood waters, producing a slick along the river that left oil
- ' (marks for a considerable distance downstream (Ref 2.4-17).
Figure-2.4-9 shows the results of a special survey commissioned by-PECo in July, 1972 to determine high-water marks between Sanatoga'(1.4 miles upstream from the plant) and Cromby (8.6 miles downstream). All readings were taken along the east side of the Schuylkill River and reflect the top of the oil marks where they were visible.
If an upper envelope of that portion of the profile near the i- plant site is taken as indicative of the actual high water, it appears'that the 1972 flood rose to about el 131 feet. In addition, there appears to have been almost no clogging of the bridges during this flood.
The rating curve-(Figure 2.4-7) indicates'a water level at about el 134.3 feet for a flow of 95,900 cfs. . This is 3.3 feet higher than the elevation of 131' feet given by the flood level survey shown in Figure 2.4-9. The assumption of bridge clogging (Section 2.4.3.5.3.5)- accounts for about 2 feet of the difference. The rating curve is apparently_ conservative when
- used-for estimating water level for a given discharge.
2.4.3.5.3.'1 Method's of Computation The geometry of the channel near the site is taken from a survey made in 1969. Normal uniform flows are assumed for the low
' flows, and an-approximate roughness is obtained using the average water-surface slope shown in Ref 2,4-18. These low-stage 7
computations were checked by field observations in December, 1969.
The-flood levels for flows from 20,000 to 356,000 cfs.are
- obtained using an adaptation of the Corps of Engineers Standard l ' Step Backwater-Program (Ref 2.4-19). Flood levels for flows
!. above'356,000~cfs are discussed in Section 2.4.3.5.4. The.
l _ program employs a-computing method similar to Method 1 in Ref 2.4-20. Whenever data are insufficient, conservative estimates are made. The computations consider the 14.1 miles of river between Pottstown-(5.5 miles upstream from the plant site) and'the village of Cromby (8.6 miles below the plant site).
2.4.3.5.3.2 Topographic Data
! - Figure 2.4-11 shows the locations of cross-sections used in the backwater studies. Data are obtained'from four sources:
2.4-19
- a. ' Commonwealth of Pennsylvania cross-sections taken in 1967.for the vicinity of Pottstown (river stations 3680+00 through- 3810+00)
.b.: Corps of Engineers cross-sections taken in 1969 in the same reach to supplement the commonwealth sections 1: . USGS topographic sheets revised in 1968, Pottstown and Phoenixville 7.5-minute quadrangles (10-foot contours) for the sections downstream of Station 3810+00 d.- Specific surveys made for these studies in 1969 and 1970 by PECo of all bridges between Pottstown and Vincent Dam and.of the river bottom close to the plant site ,
The only' modifications made to existing topographic data consist !
of assuming that impounding basins along the river are full of l dredged material. For the sections taken from the topographic maps, approximate bottom elevations are developed using information contained in the 1950 report to the commonwealth by the Schuylkill River project engineers (Ref 2.4-18). The !
surveyed cross-sections indicate that the main channel could be represented as having a horizontal bottom two feet above the thalweg elevation. This approximation is extended to the main
- ' channel portions of the unsurveyed sections, if they are assumed to be rectangular in section with their base elevations two feet
- above the thalwegs shown in the 1950 report.
' Lengths of reach for both the main and overbank sections are :
scaled from the USGS topographic maps. (
I USGS aerial photographs taken in mid-1968 are used to determine floodplain character for roughness evaluation. Floodplain use culture undoubtedly changes from time to time and place to place, but it is assumed that the net effect of such changes is to keep ;
the floodplain roughness approximately constant. ;
2.4.3.5.3.3 Selection of Observed Flood Profile i
Roughness coefficients for a natural stream are best determined by a trial-and-error process of matching observed high-water l profiles with those obtained from backwater computations using }
various roughness coefficients. ;
Flood profile information is not available for the 53,900-cfs :
flood of February, 1902. l Several flood marks are available from the USGS and the U.S. Army l Corps of Engineers (Ref 2.4-21) for the 50,800 cfs flood of ,
May, 1942 but were collected before the river restoration project was completed in 1950. Changes to the river, the installation of !
the impounding basins and dredging, have been sufficient to 2.4-20 t i
i
i significantly alter the hydraulic characteristics of the river !
and floodplain. :
e The largest flood between 1950 and the time that the rating curve ;
was computed occurred in August, 1955, with a peak of 42,300 cfs. l
'The flood level at the USGS gage on the Hanover Street Bridge in ,
e Pottstown was obtained from Ref 2.4-10, and'other flood marks !
from Ref 2.4-21. These data are as follows. ;
OBSERVED MAXIMUM WATER LEVELS, 1955 ;
Flood Level ;
Location (ft) f
., t Pottstown Hanover St. Bridge Station 135.84
. S. Pottstown Madison (Keim) St. Bridge 133.55 Linfield Highway Bridge 120.03 Spring City Highway Bridge 105.97 I
t i
See Section.2.4.3.5.3 for a discussion of the-1972 flood.
2.4.3.5.3.4 Derivation of Manning's n l In 1969, the Philadelphia District Office, Ccrps of Engineers, i performed backwater studies in connection with a proposed channel !
improvement project near Pottstown. The following values of l Manning's n'were apparently used by the Corps in their 1969' ;
study: j Natural main. channel (before improvement) 0.042
- l. Built-up areas 0.062 ,
. Fields (cultivated and fallow) 0.062 Wooded areas 0.078 Aerial photographs support the values 0.062 and 0.078 as '
reasonable for overbank roughness, considering the types of culture along the river. Because of lack of sufficient definition of observed profiles, it was decided to adopt those :
- i. values for overbank flows and to use main channel roughness- ,
- coefficients that result in a surface profile matching the
! observed'1955 flood levels. !
Conversation with a witness to the 1955 flood at Pottstown [
revealed that only small flotsam was involved and that most ;
i 2.4-21 s
m __ _ . _ _ , _ ,- _. _ _. _ _... _ _ _ .__ ____..__ __ _ _ _ _ _ .__ _.~.__ _ __
~
3 :
i
pecbably no' clogging occurred at any of the bridges. Thus, the !
bridge openings were considered clear in the calculations l
'arformed to evaluate n. Table 2.4-11 compares the computed and .
vaserved values of water levels and gives the n values developed !
.for each' reach. [
The increase in computed roughness going downstream could be
-caused by actual-discharge increases along the 11-1/2-mile reach, j or it could reflect the attenuation of the flood wave as it moved downstream. These, however, are probably of minor importance, !
and the increase in n is probably due to an increasing lack of :
definition of the physical description of the channel.
-It should.be no'ted that.a weighted mean of the calculated values !
tagrees well with that used by the Corps -- 0.041 versus 0.042.
In terms of flood elevations at the site, it is conservative to '
useelarge roughness factors downstream of the project site. j Additional conservatism is introduced by assuming the n value to !
be constant for all flow stages. !
2.4.3'.5.3.5 Bridge Clogging The following assumptions are made regarding clogging at the i various bridges openings in handrails and trusses are assumed i to be fully clogged; bridge openings flanked by woods are assumed !
50% clogged; and main channel openings are taken as 20% clogged. ,!
In the'1955 flood, as noted above, the bridges did not become !
clogged with debris. However, in a flood such as the PMF, large I debris would be common, and the bridges could be subject to i clogging. Because the bridges are over a mile downstream and .
well. submerged during' extreme floods, the net effect of clogging i is approximately a 2-foot increase in water level at the site. j 2.4.3.5.3.6 Selection of Starting Water Surfaces }
To obtain proper convergence, it is necessary to extend the .
computations to Cromby, 8.6 miles downstream of the project site. l At this downstream location, the slope-area method is used to obtain an approximate water level. A typical cross-section and ,
the average slope observed during the 1955 flood are used for ,
this. It is assumed that the true water surface at this location falls between the two elevations bounding a 25% band around the [
com'puted conveyance (ara /s/n). Table 2.4-12 shows the j convergences obtained in the various runs made in this study. i
)
1 4.3.5.3.7 Flood Discharges Studied i
i The lower portion of the curve in Figure 2.4-7 was developed from !
several computed water levels. The results in Table 2.4-10 have i f already been noted, but some discussion of the flood quantities !
is in ceder. '
[
2.4-22 :
i I
. _ . . ~ _ . . ~ . _ _ . , . . _ , , , .--..~...____,_,.,_.,______-.._._,,._._._____m 4
- a. The average annual flood discharge, 21,000 cfs is based on 42 years of data at the USGS gage at Pottstown.
- b. The average annual flood at Pottstown, 28,000=cfs is computed from regional data presented by the USGS in Water Supply Paper (WSP) 1672.
- c. .The second highest flow of record, 53,900 cfs is at
? Pottstown,
- d. The 100-year flood, 99,000 cfs is computed from the regional data in WSP 1672.
- e. An arbitrary discharge used to obtain 200,000 cfs is better. definition.of the rating curve.
f._ When the PSAR was prepared, 356,000 cfs was the original modified Corps of Engineers estimate of the PMF. In this report, the PMF is estimated at 500,000 cfs, using the more conservative estimating procedure given in NRC Regulatory Guide 1.59, Appendix B. The calculation of levels for the 356,000 cfs flow with the backwater program provided an additional point for graphic extension of the rating curve.
Figure 2.4-12 shows the computed water surface' profiles for these floods for the reach between Sanatoga Highway Bridge and Linfield Railroad Bridge. These two points are approximately 4500 feet upstream and 7500 feet downstream from the plant site, respectively.
2.4;3.5.4 Water Level Determination for Flows Above 356,000 cfs The discharge rating curve shown in Figure 2.4-7 is extended to above the 356,000 cfs flow (el 158 feet) using the approximate buticonservative method described in this section.
The variation of friction' slope with discharge in the backwater studiesLis plotted in Figure'2.4-13. The actual ~ conveyance provided at the depth obtained in the backwater studies is estimated from the formula:
K = 1.486 (ara /s)n-1 (2.4-4) where:
K = conveyance , A = flow area R = hydraulic mean depth , n-1 = coefficient of roughness The friction slope is computed from the formular o
2.4-23
LGS FSAR Sf = (Q/K): (2.4-5) where
..Sf = th'e friction slope
-Q = the discharge K = the conveyance of the river at the Limerick site.
This-is done for each of the six discharges used to construct the-rating curve at and below 356,000 cfs (Table 2.4-10)
In Figure 2.4-13, the points obtained in the backwater study are indicated'with circles. As discharge increases from 21,000 to 28,000 cfs, there.is essentially no change in friction slope.
However, as discharge then changes to 53,000-cfs, there is a sharp decrease in friction slope; this continues to 99,000 cfs, which seems to be at or near a minimum friction slope; as discharge increases beyond 99,000 cfs, the friction slope increases. The minimum slope obtained here, however, is not the same as the minimum slope of uniform flow, critical slope. All the flows considered are vell into the suberitical range, being M1_ type backsater curves. Differentiating equation 2.4-5 with ,
respect to Q, the following equation is obtained:
AS 266K "Y g gg G' (2.4-6) g " T 5--
l This equation shows that if the ratio O/AQ is larger (smaller) than K/AK in a certain range of depth, there is a corresponding i decrease (increase) in slope and an increase in discharge. Thus, I the shape of slope-discharge relation (Figure 2.4-13) is a consequence of the backwater starting elevations and the ;'
consequent variation of conveyance with depth at Limerick.
There appears to be a trend toward an asymptote of Sf = 0.00038 as 0-increases beyond 356,000 cfs. However, without additional backwater computations to confirm this trend, its extrapolation would be questionable. To be conservative with respect to depth, a decreasing slope versus discharge relation was chosen instead of the apptrent asymptotic value. The relationship chosen is the '
best straight-line fit-by-eye to the six points in Figure 2.4-13.
At extreme flood stages above el 156 feet at the Limerick site, the left bank 'ftcing downstream) curves away from the river alignment northeasterly, toward Possum Hollow. When the stage of '
the river rises above this level, this side of the cross-section is less effective in carrying discharge, because of local separation of flow in the bend. A conservative extension of the rating curve is made by neglecting a portion of the left side of :
r 2.4-24 i
l l s
. --v- - - - - - - - - - - , , - - - - - , , ,,----,--w<r , . - - , n.----~,-,,--r-----,,m -- -
w-LGS FSAR the stream'(looking downstream) in computing cross-sectional area. A plot of'the Limerick cross-section looking upstream is given in Figure 2.4-14 and shows the area used in computing conveyance above el 158 feet.
The slope-discharge relation of Figure 2.4-13 is used to determine the required conveyance and associated stages, as given in. Table 2.4-13.
The conservative nature of this extension is shown by comparing the conveyances required at 356,000 cfs. The backwater data for
~356,000 cfs yield a friction slope of 0.000366, with a required conveyance, K, of 356,000 (0.000366)-*/a = 18.6 x 10* cfs. The straight-line relation in Figure 2.4-13 indicates a slope of 0.00028, which gives a conveyance requirement of K = 356,000 (0.00028)-1/a = 21.3 x 10* cfs. This is 14% higher than the conveyance actually required, based on the backwater study, and indicates a correspondingly higher stage at Limerick. While the stage for 356,000 cfs is based on the backwater study (el.158 i' feet), this comparison demonstrates the conservative nature of the straight-line relation shown in Figure 2.4-13 when used in extending Figure 2.4-7 above el 158 feet.
The required stage given in Table 2.4-13 is determined from a stage conveyance curve, Stage = f(K), in which the conveyance K is determined'from equation 2.4-4.
In this case, a weighted n value was computed for a water surface E of el 135 feet based on 0.042 for the channel and 0.07 for the
.overbank area. The weighted value obtained is n = O.063, and this value is used for all depths. For computing the area and hydraulic radius above el 158 feet, the cross-section portion noted above on the left bank is omitted..
The final stage relation is shown in Figure 2.4-7.
The conservative nature of the extension above el 158 feet i L (356,000.cfs) is clearly-shown by the change in slope of the L curve at that point. If the straight-line stage-discharge l relation that prevails below el-158 feet is extrapolated, it l-gives a lower stage for the same discharge. The extended stage rating curve should give conservative estimates of stage at Limerick for all discharges covered. This rating curve is not L applicable to dam-break waves. The stage and discharge during
! the passage of such waves are discussed in Section 2.4.4. The
- ~ calculations for the above rating curve are based on the i assumption that the river channel carries steady, gradually
- varied flow with.the origin of the backwater curve at Cromby.
l l
L
! 2.4-25 l
l
--. . -- . . , , _ . , , - . - . , _ . _ - ~ - _ _ , - . . - - _ - _ --
4 t
2.4.3.6 Coincident Wind-Wave Activity i
The water surface obtained with the PMF without wind is less than that obtained with the SPF and a. multiple dam break, as discussed in Section 2.4.4. Since that case is more critical, the discussion of wind-wave activity is omitted here; see Section '
2.4.4'.3 for a. discussion of wind-wave activity under the more adverse dam-break case.
2.4.4. POTENTIAL DAM FAILURES, SEISMICALLY INDUCED The reservoirs in the Schuylkill Basin upstream of Limerick can l be classified as.being either minor or major with respect to a l seismically-induced failure.
Table 2.4-2 is a list of minor dams that are either too small or l too remote to cause significant flooding at Limerick in the event of'their seismic failure. The table indicates that these are all less than 100 feet high and less than 4000 acre-feet in volume. ;
Except for three Schuylkill River dams, their drainage areas are :
~
all less than 16 square miles. Their locations are shown in ,
Figure 2.4-3.
Table 2.4-3 lists three major dams whose seismic failure could, j under certain circumstances, generate significant waves in the 4 Limerick reach of.the Schuylkill River. These structures are discussed further, and their various failure. permutations are
- considered here in detail.
Ontelaunee Dam has a storage volume of 11,900 acre-feet and a height of 52 feet. ;t is principally of concern because it is
' located immediately below the site of another and larger dam that is authorized for future construction by the U.S. Army Corps of l Engineers - Maiden Creek Dam. Maiden Creek Dam will be situated 5 miles upstream of Ontelaunee Dam and will have a volume of l 114,000 acre-feet, of which 38,000 acre-feet will be dedicated to flood control storage; the maximum height will be 110 feet. ;
Blue Marsh Dam, completed in 1979, is about 35 miles upstream of Limerick on Tulpehocken Creek. This U.S. Army Corps of ,
Engineers' project has a total storage of 50,000 acre-feet, of which a minimum of 22,900 acre-feet is ded.;ated for flood ,
storage; the maximum height is 96 feet. l
' Figure 2,4-14 presents a location map and a schematic profile :
relating Ontelaunee, Maiden Creek, and Blue Marsh dams in elevation to Limerick Plant. All three dams have, or will have, uncontrolled open spillways.
Felix Dam is a recreation dam on the Schuylkill River in the city I of Reading, which is downstream of the Maiden Creek confluence. i f
2.4-26 i
- , - - -- e.,,,_._.--,.-._.-.,..-_..-,m - - , _ _ - - _ - . - , _ - -
t l
l
~
l LGS FSAR s -The pool has a volume of 1470 acre-feet, and the dam is a 24-foot i high rock-filled timber crib overflow structure, with a spillway i elevation 237.5 feet MSL. This dam will be drowned out (or will
. be prirviously destroyed) by the SPF, and the flood. wave resulting l
, from its failure therefore need not be added to the flood waves !
from the failure of the Ontelaunee and Maiden Creek dams !
upstream.
Appendix'A of Regulatory Guide 1.59 suggests that an acceptable !
combination of runoff floods and seismic events would be the safe shutdown earthquake (SSE) with a 25-year flood and the operating !
basis earthquake (OBE) with a standard project flood. The OBE !
has a design acceleration of 0.075 g horizontal and 0.050 g [e vertical. The SSE has a design acceleration of 0.15 g horizontal and 0.10 g vertical. !
- 1 On-Blue Marsh Dam,-the'U.S. Army Corps of Engineers provides for l
' horizontal design accelerations of 0.05 g for concrete structures ;
.and 0.10 g for the embankment; no provision is made for a vertical acceleration component. Plans for completing final l design and for construction of Maiden Creek Dam, also a Corps of -
Engineers project, has been indefinitely deferred; the Corps !
presumably would apply its Blue Marsh seismic design values to- !
- Maiden Creek Dam.
There is no ln'dication that Ontelaunee Dam, an earthfill and masonry spillway structure complete in 1934 and now owned by the i City of Reading, has any specific design provision for seismic loadings.
l The U.S. Army Corps of Engineers design accelerations do not On clearly fall into either the OBE or the SSE design category. ;
the one hand, the embankment horizontal design acceleration !
-exceeds the OBE horizontal design value, but on the other hand ,
the concrete structure horizontal' design value is less than the .
OBE. In addition, no provision is included for vertical design !
acceleration. While the U.S. Army Corps of Engineers das designs
- i embody a considerable resistance to seismic failure, the amount !
of damage accompanying either an OBE or SSE is difficult to' !
assess without a detailed seismic' analysis of those structures.
Therefore, for the purpose of this report, the two Corps of l Engineers dams'and the Ontelaunee Dam are all considered as i non-seismic Category I structures. Their total instantaneous j failure-is considered simultaneously with the Limerick SPF, which ,
b thus corresponds to an OBE design failure condition. This !
~
simplification negates the need for study of the SSE combined j with a 25-year flood. since that case is simply the same total ;
' seismic failure, but with a lesser coexisting flood. j i
~
t 2.4-27 Rev. 6, 06/82 (
i I
, , p. , , .y .
.s w u
T) ,. -
,g s
u
,r a 1-s N q ,
- N i
,7;. w. ,s A
- x. - ,
~7 q
, - M ~ ] ~. w ggg yggg.,-
y
\v ,
"% , 4 D'. 1 2.4.4.1 Dam Fal, lure @ rmutatit.ns
.. i
~ s N
. '7 l If.amajorselhmiededureoccubredat e future Maiden Creek Dam,+ dec.ino-type f all'ure would be (1likely at Ontelaunee because i of the limited spiEway capacity.\ Moreover, these two dams will i be a"pproximately 5 miles apar't,s and Blue Marsh Dam on Tulpehocken !
thus a seismic event Creek severeisenough only 6.7vmiles ;to ciustfrom Onkelaunes;s tii any or of these.could cause failure ;
severe damage to'the other' two. Therefore, a' multiple failure '
analysis is considsred in uhlch all three structures are considered'to. fail in such-a way thatgLhe'ir' peaks arrive at
' ~
Limerick' simultaneously. i. i ,
i The failure,e sseismic or otherwise, of 'anr o;f the six Schuylkill River navigation dams downstream of(Limerick would not affect s
safety-related.yster supplies, sirWe safet~p% elated water [
suppites de not yely on the Schuylkill Rives.g , '
l s .N 1. ;
ics 2 specifies that'the ,
NRC appropriate Regulatory SPF GuideN1.59, at the dam shoul Revis,dibe coincident with the dam fallure in the OBE case and thatithe flood-control pool should be l filled. This specification is egaalje$ or exceeded'in the analysis at all reservoirs, as indicated *10 T4ble 2.4-3. The SPF peak flow fer.415: trick of 250,000 cfs is estimated at 50% of the PMF, using theicGriservative criteria i Guide 1.59,yevision2fordetermin{n,92AppenilixBofRegulatory ' +
g;the PMF.
q ., s .
Landslides, induced by seismicipc %on could blo'ck the river, causing dam-break-type waves 7 downstream when.swater pressure :
butids up-and brSches the slice material. However, topographic consideratic'hsNaldnbappear to preclude any hazard to the plant i L
'40from miles land:tli(eainduced
.upstrNm from waves.S the 5th, theAny '
Allegheny Mountains lie over i
. landslides across the river ,
y1 back up water that could later breach Limerick,ktwoul the s\i'de;d
\
+.odever, b be greatly att,enuated.in Below the Allegheny that area, m,
[
Mountains', in the Great Valley above' Reading and the Triassic ,
Lowland' below' Reading, the riveris meander \ floodplain is !
l typically 3000 to 5000 feet wide. ' Blockage,oT'this wide channel by a landslide is not conceivaMe. The topography on both sides v of the floodplain,is generally gentle, but with some steeper :
hills occurring, particularly on the south side of the !
floodplain, between the site and Reading; however, a' study of the !
USGS 1:24,000 maps indicates that, except at Reading, none of these appear to have the volume required to block the river.
i At Reading, there are several steep hills on both sides of the I l
river. Of' theses the closest to the river and the steepest is Neversink Mountain, on thejnorth side (Figure 2.4-14). The river l
is 200 feet wide and has no floodplain at tbjs point. If Neversink Mountain were to collapse and block the river, with the slide subsequently breached b/ the river, water could back up to i
+ (
s \'
~ !
('24-28 l Rev. 6, 06/82 ' .
p 3 .,,
)
s
- - --- - - - - , ~ _ - _ _ - , . . , - A.__ ,_.,______.___.______,_...{,,_,,,____._,_____,_.,
l i
.el 255 feet before spillage occurs over Poplar Neck, a long ridge ,
on the south bank. The river bottom here is at about el 170 i l -feet; therefore, the maximum depth of the water backup would be 85 feet, after which it would spill over Poplar Neck. The {
l resulting wave could not produce a water surface elevation at !
- Limerick that would endanger the plant.- The joint occurrence of l a wave at Limerick from a Neversink Mountain slide and a failure l i 'of one or'more upstream dams is an event much less likely than is j j required by Regulatory Guide 1.59. The peak flow rate that is due to simultaneous failure of the three dams, Maiden Creek, Ontelaunee, and Blue Marsh, is conservatively estimated at
762,000 cfs. When superimposed on the SPF flow of 250,000 cfs, the comoined event results in a total flow rate of 1,012,000 cfs. ;
l i 2.4.4.2 Unsteady Flow Analysis of Potential Dam Failures j i
l' The dam break problem has been discussed in the literature for- !
various idealized cases. The most generalized and convenient l approach developed to date is by Sakkas, Ref 2.4-23, as most [
j recently described by Sakkas and Strelkoff, Ref 2.4-24. With :
L -this method, cognizance is given to channel friction and volume l of the flood wave. However, it was derived by dry bed conditions l and requires consideration of a uniform prismatic channel j downstream.
i In Ref 2.4-23, dimensionless graphs are given that are determined ,
by numerical integration of the equations for unsteady flow, j using the method of characteristics. A wide variety of -
combinations of dam height, channel slope, and cross-sections is ,
analyzed so that the various graphs permit an analysis of many i practical dam failure situations.' Channel geometry is assumed as !
prismatic, and procedures are given in Ref 2.4-23 for deriving [
the constants characterizing the channel cross-section, yielding ,
i a relation between the cross-section width 5 and the depth Y. l This relation is given by: :
m I = UY (2.4-7) i L !
where:
I = the width at depth Y !
l E and a = constants derived from the cross-section properties. ;
'Ref 2.4-23 gives dimensionless curves of the time of arrival of the !
wave front, the flood depth, and its time of arrival as a function of the distance downstream, empirical cross-section constants C and a, and the initial Froude number F at the dam j i
i 2.4-29 Rev. 6, 06/82 !
I I
y , ,
1 y( ' ' '
Miu uI '
c - a 7.f"") r -(? ?.T4 '
l
'where: I
~
v.- - (ef) l i
l
= The hydraulic radius at the instant of failure !
N.
Yo = the initial depth at the dam -
g- = acceleration of gravity l
-5 =' channel slope f n =-Manning's' friction coefficient for channel downstream Symbols ~with bars.over the. top are dimensicaed variables, and !
symbols without the bar are dimensionless, following Sakka's <
notation..
r
.For the assumptions specified, the results of the Sakkas !
~ procedure-are considered reasonably accurate,. However, adapting ;
the general model to a particular situation'and then superposing multiple dam failures requires careful consideration and ;
. interpretation-to obtain conservative ~ estimates of incremental.
~
^
and total flood depth. -A set of simplified surrogate regimes is ,
assumed whose cumulative effect on flood depth certainly is :
greater than the depth in the actual hydraulic regime that would prevail.in such an event. This conservatism is justified only
-because of the Limerick plant's location high above the Schuylkill-River.-
To incorporate the prismatic channel assumption, it is necessary '
to adopt a single representative channel cross-section between
.the failed dam and the Limerick site. The river section at a dam is typically narrow, wnereas the-river floodplain below tends to -
be wider. - In.the Schuylkill River and its tributaries, the basic !
. valley is U-shaped.or V-shaped;-i.e., there are no relatively ;
. wide'floodplains:in the reaches between the dams and Limerick, ;
and the main stream channel does not vary greatly in basic ;
-geometry, particularly under the initial condition of the SPF.
Under_these conditions an approximation of a constant prismatic ;
section is considered realistic.
Rev.~6, 06/82 2.4-30 ,
i i
/
= - ---. --4-w,..,.4%.r -. ,m, _ , _ . . , , , , , _ . _ . , , , , _ , . , _ _ , . , , _, ,, _, _ _ _ , ,
i LGS FSAR A single representative cross-section is defined by taking the geometric mean of two, paired characteristic breadth and depth dimensions at three cross-sections: one at the hypothetical dam described later; one at the Limerick Plant; and one at a section of the Schuylkill River located about 1 mile upstream of the town of Birdsboro, or about 15 miles upstream of Limerick. These three cross-sections are presented in Figures 2.4-7a, 2.4-7b and 2.4-7c respectively. Figure 2.4-7d is the defined representative ;
cross-section. The geometric mean section is then used for estimating the effect of a particular dam break. -
The results given by the Sakkas procedure apply to a flood wave progressing down an initially dry bed. To approximate this i condition in the analysis, the channel slope and the geometry of the two downstream sections are set before a mean section is !
determined. They allow for the existence of the SPF and any r
P e
I i
L 2.4-30a Rev. 6, 06/82
I This page is blank Rev. 6, 06/82 2.4-30b
(* 1
. t stage-increments from previous dam breaks. In effect, the peak ,
of the SPF is " frozen" at 250,000 cfs at Limerick for the time !
that the peak of a dam-breck flood whve is enroute to Limerick. l Conservatism is embedded in this. approach, since an earthquake of ;
multiple dam break failure magnitude is more likely to be a r single massive event dispensing flood waves from each dam at the l same time,-rather than timed so that their peaks would be l additive at Limerick, simultaneously with the crest of the SPF. l Since Maiden Creek and Ontelaunee dams are located on the same stream within 5 milesLof each other, the SPF peak at Ontelaunee is estimated from that at Maiden Creek, using the ratio of drainage areas. This gives.the SPF peak at Ontelaunee as i 58,400 efs. If'it is conservatively assumed that there is no l attentuation that is due to the storage in Ontelaunee reservoir !
above the spillway crest (el 294 feet), the SPF peak of f 58,400~cfs discharges at a reservoir elevation of 304.2 feet. It is estimated that, at this elevation, the volume of water impounded behind Ontelaunee Dam would be 29,000 acre-feet.
l Based on the-information available from the U.S. Corps of Engineers, the volumes cf water stored in Maiden Creek and Blue Marsh reservoirs under SPF conditions are estimated as 130,000 !
.and 76,700. acre-feet, respectively. Thus, the total storage of {
the three dams combined together is 235,700 acre-feet.
To obtain a conservative water surface elevation that is due to :
simultaneous. failure of the three dams,.it is assumed that the !
total volume of water-(235,700 acre-feet).would be stored [
upstream of a hypothetical dam located near Reading, !
I approximately 30 miles upstream of Limerick and 5 miles i downstream of Blue Marsh Dam. As the hypothetical dam is downstream of the three'real dams, the channel storage assumed ,
- l. available for attenuation of a flood wave released from the '
l hypothetical dam would be smaller than actual. This results in a l
higher, and therefore conservative, estimate of the water surface ~
elevation at Limerick.
l The area-capacity characteristics.of the hypothe'tical dam are assumed to be the same as at Blue Marsh Dam, which is the nearest ,
. upstream dam. Actually,-the channel section at the location of I
the hypothetical dam would be wider than that at Blue Marsh, and so the above assumption provides conservative estimates of the l water' surface elevation. corresponding to a given storage. By this assumption,.the water surface elevation in the hypothetical l reservoir for the combined storage of 235,700 acre-feet is l estimated to be 333.5 feet. The river bed elevation at this ,
location is 213 feet. Thus-the depth of impoundment of~the hypothetical dam at the instant of the postulated failure was 120.5 feet.
- . i l
2.4-31 l
L .
. .., .-..m., . ..,,_ ,,_. - . . . _,,_,,,,,m _ . , , . , . . . _ , . . . . _ , . , - , , . . - . . . . . - , . . _ .
^ ~
.t i
-Using a slope-area mathod and: assuming uniform flow conditions, the water surface elevations corresponding to.the SPF at '
Birdsboro and Limerick are estimated to be 188 and 152 feet, respectively. To' simulate pre-existing SPF conditions in the river, cross-sections at Birdsboro and Limerick are assumed to be
. represented.by the portions of the flow sections above the SPF elevations. For the cross-section just downstream of the-With the hypothetical dam, dry channel conditions are assumed.
Sakkas-method, it is conservative to ignore any water in the l
' channel immediately downstream of a failing dam because a deeper ;
than actual reservoir storage results. As stated earlier, to use l l
Sakkas' dimensionless_ charts, a representative prismatic section
-for_the routing reach from the hypothetical dam to Limerick is obtained by taking the-geometric mean of the characteristic '
-breadth and depth of the_ river sections at the hypothetical dam (above dry bed),.at Birdsboro.(above SPF elevation), and at-Limerick (above SPF elevation).
ETo be conservative, the channel slope between the-hypothetical I dam and Limerick is computed using the dry-bed elevation at the hypothetical dam site and the SPF elevation at Limerick. .The actual channel slope during SPF conditions is steeper than that used. ,
5 To- obtain a representative value of the coefficient of roughness, n, for the river reach between the hypothetical dam and Limerick, weighted average values are computed for.the sections at !
n Birdsboro and Limerick using the equation (Ref 2.4-33):
ARh
- n=- /' Ai1 R Sh (2.4-8)
O' ( "i j/
where:
r 'A- = total-flow area R = hydraulic mean depth for the' entire section ;
A,R i
= flow area and hydraulic mean depth for the
- channel-section A,, R, = flow area and hydraulic mean depth for ;
overbank flow !
ni, n n- = coefficients of roughness for the channel and !
overbank sections i
I 2.4-32
i i
- For the section at the hypothetical dam site, where dry bed conditions are assumed, a value of n = 0.062 is used for the entire section. This value is recommended for floodplains with light-to-medium brush and trees (Ref 2.4-33) and conservatively :
represents the conditions at the hypothetical dam section. For .
the sections at Birdsboro and Limerick, where the dam-break flood l wave is superimposed on the SPF, a value of 0.062 is is used for l
- the overbanks, and a conservative value of 0.03 is used for the l l interface between the water surface at SPF and the dam-break flood wave. The arithmetric average of the n values for the i three sections was assumed to be applicable for the entire reach. ;
. i
.Using Sakkas' curves (Ref 2.4-23), the maximum flood depth that
^
is due to the dam break is determined at the downstream section L - at' Limerick. The calculated depth thus determined is a depth in [
the derived geometric mean cross-section. To relate the computed '
[ depth to.the actual section, conveyance relations are used.
L Using the derived properties of the geometric'section (C and M), ,
- the area is given by. ;
A= C Y f t.
(M+ t
! (2.4-9) i f
From this, the conveyance.(Equation 2.4-4) required in both the !
geometric section and the natural section can be calculated.
Using the stage-conveyance characteristics of the natural river l- - section above the SPF elevation at' Limerick, the water surface i elevation that gives incremental conveyance (above SPF elevation) equal td that corresponding to the calculated depth in the i
~
geometric mean cross-section is. computed. This results.in a
-water surface elevation of 201 feet at Limerick.
! 2.4.4.'3 . Water Level at Plant Site ,
( .:
l - 2.4.4.3.1 Maximum Computed Water Surface and Wave Effects l l This section describes the effect of wind-waves on the maximum l - water surface (el 201-feet) estimated at Limerick that is due to '
tne dam-break condition described in the previous section.
I n
TheJdam-break waves are transients'and do not contain enough i volume to cover the entire Schuylkill Valley above Limerick to '
f~ the computed. maximum depth at Limerick. However, for an L approximate wind-wave analysis, the following conditions are ;
!~ assumed; the water surface at Limerick is el 201 feet, and wind ;
velocity is 40 mph. >
! . Fetch is based on~a map study of a level pool upstream at el 201 feet. - At-this water surface elevation, the ridges and the two :
O 2.4-33 ,
+
w = -- - - -...-.+3.,e....m- ,e-e.---,--,w.-3-= o+. -w s e r en-w.w--,we,.% ,--r--,-w-r,,--e---wwwvw-wn--eem--.sm-as.-,-siww--=*w-we-t--w,--wwww----ew-www-+-
e -
"I i
i
bends between Pottstown.and the site are submerged. This permits a roughly rectangular fetch of'31,000 feet average length and a ,
width.of 4,500 feet, on a bearing of about N 72c W. The ;
centerline of the rectangle intersects a ridge about a mile west of the village of Stowe. It is conservative to assume a rectangular fetch area, which according to Ref,2.4-25 gives a
-fetch effectiveness ratio of 0.18 and an effective fetch of 2.-65 miles. Using Ref 2.4-25, these parameters yield a :
significant wave Hsv (33% frequency) cf 3.8 feet, with a period _ i of.3.9 seconds and a length of 78 feet. The maximum wave (1% ;
frequency) is estimated as 1.67 Hs = 1.67 x 3.8 = 6.4 feet. !
At;the Limerick plant: site bench, the maximum water surface ;
without wind. action,.el 201 feet, intersects a sloped-fill surface to the windward-(west) side of the plant (Figure 2.4-4). j 1The fill surface has an average slope of 1 vertical to 2 -
horizontal. The toe of the slope varies between el 170 feet and .
.200; feet. The top of the fill varies from el 213 feet at the !
northwest end (adjacent to the switchyard bench), to el 216 feet ,
on the. south side of the plant, where the top of the fill curves :
toward the east. From the top of this fill, the ground rises gently on a.long flat slope-to the railroad track at el 215 feet ,
'to 217 feet, which encircles the west side of the turbine-reactor !
-area complex. The railroad track's low part (el 215 feet) is 90 !
to 350 feet from the top of the slope. The fill material i '
broken rock and fines, with_ sizes up to 24-30 inches. For ,
computing wave runup, this material can be considered as graded '
riprap. To compute-runup, curves given in Figures 7-19 of Ref 2.4-25 are used, which are for a 1:2 slope. However, the wave from the maximum-fetch direction is not perpendicular to the T 1:2. slope surface, but intersects it at an' angle of about 640 ;
from normal. This results in an effective slope normal to the !
wave of about 1:4.5, so that actual runup is less than computed. ;
. Runup (R) from the significant-(33%) wave and the maximum wave
'(1%) is calculated from Ref 2.4-25 as follows: i Ho= Height- Max el l Wave (ft) Hn/cT2 R/Hn R (ft) Runup (ft)
- Significant '3.8 0.0078 1.3 4. 9_ 206 i
Maximum 6.4- 0.013 1.0 6.4 207 l
'Thus,-the. highest water surface elevation at Limerick that is due !
to the most-severe dam break permutation coincident with wave :
activity induced by a 40 mph wind would be 207 feet. This is !
.10 feet below the plant grade (el 217 feet) and 8 feet below the i
)
railroad track (el 215 feet).-
2.4-34 i
4;
, . . . . - . - - - - - _ - -- -- , .,_ . . . ~ . , , _ _ . . . . - _ . - . . - - . _ _ . _ , ,
t
?
i I
LGS FSAR f 2.4.4.3.2 Recapitulation of Conservative Steps in Dam-Break i Analysis !
The computed water surface elevations (with and without waves) resulting-from the dam-break analysis are the result of a
-compounding of conservative assumptions. :
r 1The series of conservative steps used to compute the water surface are listed below. ;
- a. The.SPF is estimated at 50% of PM7, as determined from ;
Appendix B, NRC Regulatory Guide 1.59, Revision 2. The PMF, as estimated by this procedure, has a peak of f 500,000 cfs (unadjusted for dam breaks), whereas the ;
peak PMF computed by conventional methods in the PSAR was.356,000 cfs, approximately 30% less. The SPF peak
- at 50% of the Appendix B PMF, or 250,000 cfs, is thus ,
also conservative. Conservatism is further confirmed by '
l
'a recently published U.S. Army Corps of Engineers estimate (Ref 2.4-26).of 128,000 cfs for the SPF at Pottstown; adjusted to the site, this would be -
131,000 cfs, 52% of the SPF given by Appendix B, NRC !
l Regulatory Guide 1.59, Revision 2. j
- b. .The effective cross-sectional area used at Limerick i
, omitted part of tie left overbank area (looking [
i- downstream) to co.iservatively allow for local flow i separation produced by the river bend.
- c. The earth fill dams should not fail instantaneously as is assumed.
i
- d. A dry channel is assumed' downstream of the hypothetical.
- dam, resulting by a larger effective reservoir depth by ;
using Ref 2.4-23.
L e. The concept of the hypothetical dam implies' seismic
- L . failures of the three real dams generated at different
! times, corresponding to three different travel times to i
- Limerick. .This is an improbable series of selectively destructive tremors. -It is more likely that a single catastrophic tremor would be involved in a multiple dam i failure, with the resulting flood waves arriving at "
Limerick at different times.
- f. The actual channel slope in the river reach from the hypothetical dam to Limerick is steeper than the slope ,
used in the analysis. This results in a higher water !
l surface elevation at Limerick, computed by Sakkas'
- ' procedure. ,
?
- 2.4-35 l
L . . . . - - . . _ - . . - . - . - - - . . - - - - _--. - - .-.-... .
LGS FSAR Since'all of.these assumptions are-conservative, it is concluded
.that the maximum stages computed with and without waves are well above the stages.that a precise analysis would indicatt. It is finally concluded that the most severe seismic dam-break
.p'ermutation of the three dams, Blue Marsh, Ontelaunee, and Maiden Creek, would.not endanger safety-related structures. The
' simplified-analysis is justifiabla because the plant area is high above the Schuylkill river.
~
~2.4.5. PROBABLE: MAXIMUM SURGE AND SEICHE FLOODING l This section is not applicable to Limerick Generating Station.
2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING This section is not applicable to Limerick Generating Station. l 1
2.4.7 ICE-EFFECTS l This section is not applicable to Limerick Generating Station.
- Spray pond icing is' discussed in Section 9.2.6.
2.4.8 COOLING WATER CANALS AND RESERVOIRS In this section, only the hydrologic engineering aspects of the spray pond components of_the ultimate heat sink are covered.
There_are no canals in the cooling water system. The spray pond ,
serves as the ultimate heat sink for the residual heat removal !
service water (RHRSW) system and the emergency service cater
'(ESW) system after a possible accident. The pumps of the two systems take water from the spray pond, and circulate it through coolers and heat exchangers. The warm water is returned to the :
spray pond through a network of spray nozzles. A complete ,
description of the spray pond system is given in Section 9.2.6. {
2.4.8.1 General Descriotion of the Soray Pond The spray pond is located about 500 feet north of the cooling
~
towers. The bottom of the pond is at el 241 feet and is composed -
of a.600-by-400-foot rectangular midsection, with a semicircle (radius =200 feet) on each side. The spray pond system is common to both Units 1 and 2. The system consists of a spray pond, uncontrolled emergency. spillway, pump structure complex, and associated piping and valves.
As shown-in Figure 2.4-5, the spray pond is constructed by excavation only. The slopes of the excavation are 1:1 in rock and 4:1 in soils. Around most of the spray pond, random 2.4-36
- s. l l
I
\
compacted fill about 3 feet deep overlies the original natural l ground, with'a bench at the soil-rock interface. The water !
surface area of the pond at El. 251 feet is approximately 9.9 l acres. An additional 7.3 acres of the surrounding area, !
including roads, cut surfaces, and natural te.cain, drains :
towards the pond. Runoff onto the cut face is normally ,
intercepted by a drainage ditch along the outside edge c_ a :
peripheral service road at El. 255 feet and is directed to l culverts that discharge to the pond. Along the north edge of the ]
pond,.the finished roadway is constructed to E1. 252 feet for a ;
distance of 60 feet, wi*.h 9% slopes upward at either end to ;
El. 255 feet; this low portion of the roadway is designed to function as the crest of an uncontrolled emergency spillway.
- Spill is directed to a draw that drains northward into Sanatoga l Creek. ,
j 2.4.8.2 Hydrolocic Desion Bases j i
Derivation of the PMF that forms the basis for the hydrologic l design of the spray pond and emergency spillway is discussed in l Section 2.4.2.3.2. The emergency spillway is sized so that the
- design flood.w'ill be evacuated safely. The elevation of the
! roadway around the pond and the slope protection of the pond are !
dictated by. anticipated wave action.
2.4.8.2.1 Design Basis Flood Level-(DBFL)
To arrive at a conservative elevation that -i:s due to severe l l
floods, with coincident wind-wave activity in the spray pond, the ;
i-following cases were analyzed for waves, freeboard, and slope l E protection, using the procedures given in Ref 2.4-25.
i Still Pond :
Max Water Maximum L i Surface El. Wind l l l Case . Flood (ft) (mph) !
1 PMF 253.0 40 l i 2 1/2 PMF-(SPF) 252.5 90 l l
.Results are as follows: I l: Maximum Wave I Significant~(33-1/3%) Maximum (1%) Runup l Case Wave Heicht (ft) ~ Wave Heicht (ft) Ft El. l t
i l: ~1 0.8 1.3 1.1 254.1 l L 2 2.0 3.3 2.4 254.9 l
, t l
2.4-37 Rev. 19, 04/83 l
LGS FSAR Based on the above, the DBFL is set at el 254.9 feet. The i roadway surface has been set at el 255 feet, or-0.1 foot higher ;
than the DBFL. r Using criteria in Ref 2.4-25, the mirtimum riprap stone requirements _are: a minimum weight of 7 pounds, a 50 percentile weight 1of 30 pounds, and a maximum weight of 108 pounds (assuming i a stone density of 165 lb/ft3). .The gradation and design of the riprap for ti'e spray pond soil slopes are discussed in Section '
2.5.5. The riprap is capable of resisting the wave action and '
therefore protects the soil. slopes. No protection is necessary l
.for rock-cut slopes. .
[
2.4.8.2.2 Safe Shutdown and Operating Basis Earthquakes [
According to Regulatory. Guide 1.59, the higher of the following :
two alternative combinations of events is considered to be an i adequate design. basis for floods that are due to seismic failure ;
.of dams: ;
i
- a. Alternative 1 year flood coincident with SSE and
.2-year extreme windspeed from the critical direction and ;
length of effective fetch !
2-year extreme windspeed from the critical direction and length of cffective fetch ,
i When routed through the spray pond, the SPF yields a maximum I water surface elevation of 252.5 feet. The 25-yearTheflood yields i a maximum water surface elevation of 251.8 feet. starting !
water surface elevation in.the spray pond for both these cases is r assumed to be the normal _ operating water surface elevation .
(El. 251 feet). In the latter case, the entire volume of the 25-t year 24-hour precipitation, without any losses over the contributing drainage area of 17.2 acres, is superimposed on the normal pond elevation of 251 feet. Because the flood is contained below the spillway crest (El. 252 feet), flood routing i computations are not required'for this case, j To estimate the height of waves due to earthquakes, the following !
equation, developed for determining the height of waves generated i by a piston-type wave generator (Ref 2.4-31) is used: l (2rd) i L
H = 2s M +(*2 L
sinh sinh * (# N ML cash (2.4-10) l Rev. 19, 04/83 2.4-38 L
H = wave height vertical distance between the wave !
crest and trough (feet) :
s = design displacement (amplitude) caused by the ;
earthquake (feet) d = initial depth of water (feet) ;
L' -= wave length; a function of the period of the DBE
. (feet) j
- From t'he response spectra of the-safe-shutdown earthquake (SSE) and operating basis earthquake (OBE), the design displacement (amplitude) and pe'riod are estimated to be 6.5 inches and !
2.0 seconds for the SSE and 3.0 inches and 2.0 seconds for the ,
OBE, respectively. Conservatively, assuming that a negative wave '
is generated at the opposite boundary of the spray pond and that
. the amplitudes of the positive and negative waves _are in phase and additive,.the maximum possible wave height would be 2H. This results in a maximum water surface elevation H above the still water. The computed values of H and the resulting water surface i elevations are tabulated below:
Earthquake Wave Max Water ,
~
Concurrent (ft) 2-Yr Wind Surface ,
Earthauake flood H 2H Effect-(ft) Elev (ft) :
OBE SPF 1.0 2.0 0.25 253.8 l SSE. 25-year 2.1 4.2 0.25 254.2 l Because the spillway crest is at El. 252 feet, there would be sor.e splash over the spillway for a short duration. However, ,
water loss _due to such splashes would be negligible compared to I the total capacity- of the spray: pond and would not impair the !
safety-related water supply functions of the pond.
For th'e design'of riprap protection, the wind-generated _ waves ,
described in Section 2.4.8.2.1 are more critical than the short duration earthquake-induced waves. >
For the spray-head pipe supports and the pump structure, f hydrodynamic forces due to the OBE and SSE are computed. The forces on the pipe-supports are computed using-the virtual mass formula, F = CmVsa (Ref-2.4-30), in which F is the hydrodynamic force on a body of submerged volume V that is due to an acceleration a, in a fluid of mass density p. Based on
- information available in Ref'2.4-30, a value of 1.5 is used for
~!
l 4 2.4-39 Rev. 19, 04/83 t l
i LGS FSAR the coefficient Cm. The resulting forces on the spray-head pipe i supports are 431 and 890 pounds for the CBE and SSE, '
respectively; the forces are assumed to be applied at mid-depth, El. 247.3 feet and 247.5 feet, respectively. i 1
- For the pump structure, hydrodynamic forces are computed both in -
~the north-south-(N-5) and east-west (E-W) directions. .For.the
- N-S direction, the TVA method (Ref 2.4-31) is used. For the E-W direction, where only part of the structure is exposed to water, tho method given by Sarpkaya (Ref 2.4-32) is used. The resulting l forces and their points of application are given bdlow: j Point of Max Force Application !
Earthquake Direction (kips) (ft) l SSE N-S 144.0 246.2 E-W 70.0 247.5 OBE N-S 67.0 2a6.0 E-W 34.0 247.5 [
These hydrodynamic forces on the pipe supports and the pump ,
structure.do not include the hydrostatic forces that are due to i normal water-depth, flood surcharge, or earth pressures, nor do they include dynamic forces that are due to waves.
-2.4.8.3 Low Level-Cutlet Facilities Evacuation of the normal storage of the spray pond, if needed, is t accomplished by using the ESW system and/or RHRSW system pumps to pump water to the cooling tower basins via the cooling-tower-spray pond intertie line down to the minimum operation level of ,
the pumps, and by other means below that level. )
2.4.9 CHANNEL DIVERSIOF This=section-is not applicable to Limerick Generating Station.
1 2.4.10 FLOODING PROTECTION REQUIREMENTS
-As discussed in Section 2.4.2.2, the safety-related structures <
i and facilities are secure from flooding. Hence, flooding protection requirements are not necessary.
t
~
2.4.11 ' LOW WATER CONSIDERATIONS [
Extreme low flow in strea=s does not affect the ability of any [
safety-related facilities to perform adequately, including the ;
r I Rev. 19, 0s/33 2.4-40
[
[
- , _- - - . . - . _ . - . - , , _ , , . - , - _ . - _ ~ , , , . . , . . . . , , , . . - - - - , _ . - . . . _ _ , . - - - - - _ - - _ . . _ - - -
l l
LGS FSAR I 3.5.1.3 Turbine Missiles i
'An analysis was performed to evaluate the probability of damage from postulated turbine missiles to safety-related components at
-Limerick. The probability of unacceptable damage to safety- -
related components due to turbine missiles has been calculated as 4.11x10-10 per unit per year for the two turbine trains. In the '
following, the data used in this analysis, along with salient features of the analysis, are described. Details concerning i j turbine characteristics, overspeed protection, and valve testing are discussed in Section 10.2. ;
+ >
3.5.1.3.1 Turbine Placement and Orientation The safety-related structures that could be struck by a i postulated turbine missile are the reactor enclosures, diesel- I generator enclosure, the control structure, and the spray pond !
pump. structure.
The locations of-these structures with respect to the turbines are shown in Figure 3.5-1. The figure also shows the 2250 !
missile ejection zone with respect to the low-pressure (LP) ;
' turbine wheels for each turbine unit and the areas within these l zones. Specific equipment and its location within these' zones '
are referenced in Figure 3.5-1. .
3.5.1.3.2 Missile Identification and Characteristics The turbine-generators at Limerick are manufactured by GE. Each ,
unit consists of a tandem-compound six-flow, non-reheat, 1800 rpm j
-turbine, directly connected to a synchronous generator. The '
turbine has 38-inch-last-stage buckets. e l i GE has performed-studies (Refs 3.5-2 through-3.5-5) to determine
- the characteristics of the missiles that can be expected as a .
result of a turbine burst. The methodology is discussed in their memorandum reports on hypothetical turbine missiles. Each of the ,
seven stages in a typical LP turbine were analyzed. Significant .
similarities were found in the dimensions, shapes, weights, and ,
, initial energies of' missiles from adjacent stages. These
-similarities justify grouping the stages to simplify the probability calculations. Three stage groups are considered.
Turbine stages 1 through 3 are included in Stage Group I; stages .
4 through 6 are included in Stage Group II; and stage 7 is i included in Stage Group III. It is postulated that four types of missiles can be ejected by wheels in each stage group. Fragment [
Group a is a 1200 portion of a turbine wheel nub; fragment Group b is a 600 portion of a turbine wheel hub; and fragment Groups e and d are miscellaneous missiles. These missile types are illustrated in Figure 3.5-2. The characteristics of the missiles postulated for each stage group are given in Table
'3.5-1.
3.5-5 -
. -3.5.1.3.3- Probability Analysis The probability of turbine missile damage is expressed as:
P. = P P,P3 (3.5-1) where:
P. = probability of turbine missile damage, per year-P =. probability of a turbine failure resulting in the ,
ejection of a missile, per year !
P, = probability that a missile will strike a barrier '
that protects a critical plant component, given-that a missile has been e]ected from the turbine P, = probability that a missile will spall a barrier, thus damaging a critical plant component, given ;
that a missile.has been ejected from the turbine and.has struck-the barrier P , P,, and P 3 are evaluated using a methodology that considers '
turbine characteristics, turbine failure mechanisms,. plant layout, and barrier types.
The analysis considered 18 missile ejection points representing It was assumed that Stage Group I and the two turbine trains.
Stage Group II missiles are ejected from the centers of the six '
hoods, so an ejection point was placed at each of these locations. Each of these ejection points includes six Stage Group I wheels and six Stage Group II wheels. Stage Group III ,
missiles can originate at each of the two end wheels in each hood. The remaining 12 ejection points were located accordingly. &
The procedure for calculating the total P, and P,xP 3 for each target is discussed below. The probabilities for each target i~
~
were obtained from: .
II II II 18 III III 6~
+ W Pg ) +t Wi (3.5-2)
Pr =E (W Pi i=7 Pi i=1 where the superscripts refer to tne stage groups and the subscript i refers to a particular ejection point.
In P,xP3 calculationsPfisdefinedas:
l t
3.5-6
~
= Max ((P,xP3 ) i ,(P,xP3 ) i ,(P,xP3 )i ,(P,xP 3) } (3.5-3) ,
I
= (P,xP 3)( ,
(3.5-3)
-where the superscripts a, b, c, and d refer to missile type. .
InP,calculationsPfisdefinedas:
I Ia Ib Ic Id PL-= Max [(P,) i ,(P,) i 3(P,) i , ( P,){ ] if I
(P,xP 3 )('= 0 I Ia I Ia Pg.= (P,){ if (P,xP3); = (P,xP3 )i I Ib I Ib Pi = .(P,)i if (P,xP3)i_= (P,xP 3 )( ,
I Ic I Ic t
Pi = (P,)-i. if (P,xP3)i = (P,xP3 )i I Id- I Id Pi = (P,)i if (P,xP3); = (P,xP3 )i E El
, P g and P are similarly defined. '
Wf,Wfl ,
andW) are weighting factors' associated with each ejection point. They are the probabilities that a particular '
wheel included in ejection point i fails, given that the turbine has failed. GE states in Refs 3.5-2 and 3.5-4 that the wheels fail with equal probability. Since there is a total of 42 wheels -
-per turbine, it-follows that:
I II Wi = Wi = '6/42 and III Wi = 1/42 3.5-7
m LGS FSAR The contributions from each turbine are computed in a similar fashion.
If E.=P r 7 +E,7 then 3 II II II 12 III III 2
7
=I (Wi Pi + Wi . Pt ) + E Wi Pi (3.5-4) i=1 i=7 6 I I II II '18 III III 2,=I 7 . (Wi Pi =Wi Pi ) + E Wi Pi (3.5-4) i=4 i=13 where P 73 and PT2 are the contributions to the total probability, P, from the Unt:~1 and Unit 2 turbine trains, respectively.
T I II III ~
Note that if Pi =Pi = 1 (i=1,2,3) and P i = 1(i=7,- ,12) then 3 6- 6 12 1 12 1 P
T1
= I (42 + 42) +r (42) = 3 x 42 + 6 x 42 = 1 i=1 i=7 The P. for the target is obtained by multiplying its P,xP3 by P 3.
The P,, P,xP 3 ,-and P. for each of the units on the site are computed by summing the probabilities ice the targets in the unit. The average values of P3 for the individual targets and units are calculated by dividing P,xP3 by P,.
The analysis assumes that each missile damages at most one target
-- and ignores ricochets. The first assumption avoids double-counting the effects of the missile under consideration. The second assumption is justified by the geometry of the targets that were conservatively defined to be entire structures. The calculation of P, and P,xP3 'is discussed below in detail, as is the value of P: used in the analysis.
Missile Generation-Probability P
- In general, two specific overspeed conditions are postulated for which the missile generation probability values (P ) and the i missile characteristics are evaluated:
3.5-8
_ _ _ _. _ _ _ m .
i l
i
- LGS FSAR l O a. Design overspeed (low speed burst)
- this is 120% of l rated speed of the turbine and is based on the precept !
that, should the turbine speed-governing system be
. incapacitated so that the turbine is tripped by the i overspeed trip mechanism, the attained speed does.not !
exceed 120% of' rated speed. Disc failure could occur at ;
this' speed as a result of undetected material
. deficiencies leading to brittle fracture. i
- b. Destructive overspeed (high speed burst): this is 180%
of. rated speed and is the lowest calculated speed at I which any LP rotor disc bursts, based on the average !
tangential stress being equal to the maximum ultimate i tensile strength of the disc material, assuming there [
are no flaws or cracks in the disc. !
t i
At either overspeed. condition, it is postulated that the rupture of one-disc does sufficient damage to the unit that further !
overspeeding and additional missile generation do not occur. j GE has. established that the probability of missile generation at !
the design overspeed conditions is statistically insignificant. i The probability of disc failure leading to the ejection of a !
missile at the. destructive overspeed is calculated by GE as (
1.5x10-7 in 30 years of operation of the turbine. This }
corresponds to a yearly probability of occurrence of destructive !
l overspeed. turbine missiles.of 5x10-'. For further details of l
,' this analysis for' estimating the missile generation i probabilities, reference is made to GE's memo reports ,
(Ref 3.5-4).
- Calculation of Strike ProbabilityL(P,)
t Figure 2.5-3.il'ustrates.a Cartesian coordinate system used te !
L .specify the direction of missile ejection from the turbine. Tne
! _x_ axis. corresponds to the turbine shaft, and the y axis is normal !
p to the snaft in the horizontal plane.
The direction of missile ejection-is specified by two angles: e !
subtends the y axis and the projection of the ejection vector on !
the y-z plane, and e is.the angle from the y-z plane to the t l
ejection vector. Two additional angles are derived from e and e: ;
L s' is the vertical ejection angle measured from the x-y plane to i
! the ejection vector, and e' is the horizontal angle subtended by l i- the y-axis and the projection of the ejection vector on the x-y t plane. The angles are related by the following formulas: !
I
'e' = . sin-1 (sine cose) (3.5-5)
- i 3.5-9 Rev. 7, 06/82 !
i
, . . . , , . . . . . _ _ , . _ . . , , . . .- ...__,.__,_m._ _ _ _ . . . . . . _ _,...
t o' = tan-1 tane (3.5-6)
Cose ;
L . i If the effect of air-resistance is discounted, the missile follows a parabolic trajectory that lies within the vertical
. plane defined ~f the formula: x/y = tan +'.
An equation 'or the trajectory may be derived from elemen:ary
- physics: ,
cra ;
- z. ~= r tans' - 2(Vcose')2 (3.5-7) i In the.above equation, r is the horizontal distance from the i point of missile ~ ejection; V is the ejection velocity; e' is the vertical ejection angle' defined in Formula (Eq. 3.5-5); and g is the gravitational constant.
From Equations _3.5-5, 3.5-6, and 3.5-7, it can be seen that a missile-trajectory is determined by the two angular variables e and *,.which determine o', and by the ejection velocity V. The
' principle of the calc lation of the strike probability P, is to determine,.out of the range of possible values, the range of ,,
-+, and V, which define missile trajectories intersecting the target. Functions must be defined, P(*), P(e), and P(V), which determine the probability of missile ejection over the range of each of the three variables. P,-is then the product of the probability distributions integrated over the range corresponding to missile strike trajectories:
e2 +2(*) V2(e,$)
P,= P(c'P(*)P(VidedsdV '3.5-E) et * ( e )- Vele,*)
The-ejection probability distribution for the angle e is normally assumed to-be uniform over the 3600 arc about the turbine axis:
de P(e) de = 2r (3.5-9) t Th'e probability distribution for the angle , is assumed to be uniform.within some range + to e specified by the turbine
-vender:
de ;
- 5,5, P(+)d, = . -, ; e 5*5* (3.5-10)
P(*)d, = 0;,,
3.5-10
The ejection _ probability distribution P(V) for the specified ;
- range of possible ejection velocities is uniform with V ,
c dV I (3.5-11)
P(V)dV = Vmax-Vmin; Vmin 5 V 5 Vmax !
~P(V)dV = 0; V<Vmin or V>Vmax l f ,
Since Equation 3.5-8 is not readily integrable over e and e, a numerical integcation is necessary to complete the evaluation of ,
P ,.- ;
L -
l An appropriate range, ei ~ to er ,.is selected for which a target !
strike is possible. The range is divided into I discrete ;
ejection angles, ei, each representing an angular increment of '
L . - width A * . :
L - .
For each_value'of ei,Lthe angular interval + (ei) to e,(ei) ,
L- - corresponding to target strike trajectories must be determined. i The. projection of.*: (vi) and +,(vi) on the x-y plane defines the angles.vi '
and e,', which subtend the target at the missile !
+
origin as illustrated in Figure 3.5-4. *' and +,' may be
- calculated from simple geometry; + (ei) and.+,(ei) are obtained -
by inverting Equation 3.5-6.
- (di) = tan-1 (tan * ' cosei) (3.5-12) j f' i
+,(oi) = tan-1 (tan e,' cosei) i If +, (ei)~<
- min or if + (ei) > emax, no strikes are possible.
- - If + (ei) < + min, it is set equal to emin, l- If 4, (ei >
- mar, it is set equalfto $=ax.
5 The angular range'in + is divided into J discrete ejection angles ~
c: - eij, each-representing an angular increment of width ae . The l numerical integration thus treats J values.of
- for each value of e, or a1 total of I x J missile ejection directions, each representing a solid angle of area AeA+i.
L l For each direction of missile ejection.specified ei and vij, ;
Equation 3.5-7 for the parabolic trajectory-is inverted to solve i for the range of. ejection velocities that correspond to ;
. trajectories intersecting the target.
.The vertical plane'containing the trajectory intersects the ;
target at a minimum distance ri and a maximum distance r, from L [
. the turbine. Equation 3.5-7 is inverted to solve for V,, the maximum ejection velocity corresponding to a target strike, by inserting r, for e and inserting 2 , the target roof elevation l relative to the turbine, for 2. If V, 2 Vmax, V, is set equal to [
?
i 3.5-11 t
.a,--, ,,.---,-,,,,-_m,, ,,,,,,,,--,-,.,.,,g,.c,~my, m,,.m n ,mgen ,
,y.,.-my-,., .-,pn,_,,,, -e-gg,-,-m--
t
'Vmax'. .The-minimum velocity corresponding to target strike, Va, can be determined.by inserting r for r in Equation 3.5-7 and inserting:the ground elevation with respect to the turbine for 2.
If Ve s Vmin, Vo.is set equal to Vmin. If Vo 2 Vmax, no strikes are possible. Thus a missile ejected at velocity V, intersects 4
the far edge of the target roof. A missile ejected at velocity
.V o intersects the lower edge of the target wall nearest the
.. turbine. Missiles ejected at velocities greater than V, overshoot the target; those ejected at velocities less than Vo fall-short.
To distinguish between strikes on the target' roof and strikes on the wall, an additional ejection velocity is. determined. V3 corresponds to a trajectory intersecting the upper edge of the target wall. It is determined by inserting r for r and Ir for 2 in Equation'(3.5-7).- If V 2 Vmax, V is set equal to Vmax. .The
. ejection velocity range Vo to Vs then corresponds to wall strikes, and the range Vi to V, corresponds to roof strikes. The trajectories' corresponding to.these three ejection velocities are
-illustrated in Figure 3.5-5.
lFor the ejection angles e n and eij, an increment in the strike probability, Prij, is calculated for.the formula:
g+ 1C asi x V, (+i, +11) - Vr (+1, +1il P, = -2r , -, V -
V (3.5-13) -l ij max min max min The increments are summed to obtain t'he strike probability:
1 3 P, = r -r P, (3.5-14) i=1 j=1 ij
-Calculation of the Damage Probability (P3)
A missile striking a concrete wall with sufficient impact to cause spalling, which is the ejection of concrete fragments from
-the inner wall face, may constitute a. hazard, even if the wall is not perforated. Thus, a missile impact that causes spalling from the walls of a target structure is normally considered the threshold of target damage.
This analysis used the formulation presented below to predict the minimum velocity required to initiate spalling on a barrier. The
' data from low velocity missile impact test (Eq. 3.5-12, 3.5-13) have enabled development of empirical relationships defining the concrete element thickness for threshold of spalling by low velocity solid steel missiles.
Rev.-7, 06/82 3.5-12 4
-~ --.+..-e-.. ,,,...,m. . _ , , _
~
From solid steel missiles: l !
o 0.4 0.5 .
Tss'= 15.5 W Vs (3.5-15) 0.2
/YTc D ;
where_ }
- Tss =' thickness'for threshold of spalling for solid steel !
missiles (in.) ;
T = thickness of the barrier (in.) l l W- = missile weight (1b). l D = missile' diameter (in.) l .
f'c = concrete strength (psi). l :
E Vs =- missile striking velocity (fps) l Equation (3.5-15) defines the threshold of spalling for low velocity missiles. -The Corps of Engineer's equation.(Ref. 3.5-
- 14) is selected 1to define the threshold of spalling for high velocity nondeformable missiles. This equation is rewritten in the following form:.
1.5 Tes = 304 W vs C 2.80D (3.5-16) ,
1.785 1000 vM lD :
i where Tes = thickness for threshold of concrete spalling l t
For high velocity missiles, this equation is considered:more reliable.than other available empirical relationships because it is based on the most extensive accumulation of experimental data ;
from tests involving high velocity nondeformable missiles with a large variation of missile size and weight. Also, a comparison of Equations 3.5-15 and 3.5-16 reveals a convergence in predicted thicknesses at the intermediate velocities. ,
.The thickness versus velocity curves defined by these equations for-a particular missile may or may not intersect, depending on ,
~
missile weight and diameter. However, the difference in predicted' thickness is small at a velocity Vt where the two curves become parallel. For solid steel missiles, the value of Vt would be: t i
3.5-13 Rev. 7, 06/82 i i
pv - - - - - -
Vt = 425 0.6 W
where Vt = transition velocity (fps)
. The velocity Vt defines the transition. velocity between Equations 3.5-15 and 3.5-16. For striking velocity less.than Vt, threshold of spalling is defined by Equation 3.5-15.
At' striking velocities ~ greater than Vt, the value of Ts would be between Tes'and Tss avd converge toward Tes as Vs approaches 2Vt.
. The'value of T in this velocity range would therefore be closely
. represented by:
~
Ts = Tcs - (Test - Tsst) (2 - Vs); 1 < Vs < 2 Vt Vt (3.5-18) where.
= th'ickness for threshold of spalling for the range Vt to Ts 2Vt (in.)
Vt = transition velocity from Equation 3.5-17 Test = Tes at velocity Vt from Equation 3.5-16' Tsst = Tss at velocity Vt from Equation 3.5-15 For striking velocities greater than 2Vt, the threshold of spalling is determined from Equation 3.5-16.
To apply.the equations. developed above, it is necessary to use an equivalent diameter. This diameter, D, is calculated on the
' bases of the. projected rim area of the turbine. missile from:
/4A D= r (3.5-19) where A is the projected area of the missile.
The' projected rim area of a missile depends on the missile orientation about its axis.
For. fragment group A and fragment group B let:
.A1 = R3 T2 + R2 (T1 - T2) (3.5-20a)
A2 = (R3 - R2) T2 + (R2 - R1) T1 (3.5-20b)
.Rev. 7, 06/82
- 3.5-14
I LGS FSAR Where R1, R2, R3, T1 and T2 are illustrated by Figure 3.5-2.
Values of these parameters for each fragment group are given in :
Table 3.5-1.
The projected area of fragment group A missiles is then given by: l A1 - (Al-A2) ces (w/6 + e) 0 5 e 5 r/6 l A(e) = [A1 (1 + sin (9-r/6)) s/6 < e 5 r/3 l l
/3Al sin e w/3 <e 5 r/2 = ts (3.5-20c) l
< where e describes the orientation of the missile. e=0 corresponds to the orientation with minimum projected rim area ;
and es corresponds to the orientation with maximum projected i area.
Similarly, for fragment group B missile: l A1 -
(A1 - A2) cos (w/6 + e) 0 5 e 5 r/6 l A(e)- = / Al sin e + A2 cos (w/6 + e) r/6 < 0 $ r/3 l 1
Al sin e w/3 < e 5 r/2 = es l (3.5-20d) l For fragment group C and D missile: l A(e) = A3 (cos 0+ sin e) 0 $ e 5 r/4 = es l (3.5-20e1 l 2
where A3 = XT2 and X and T2 are illustrated by Figure 3.5-2.
Values of X and T2 are given in Table 3.5-1 for each stage group.
Equations 3.5-15, 3.5-16 and 3.5-18 are used to generate ranges of damage initiation velocities Vwo to Vwl and Vr0 to Vrt, corresponding to minimum (e = 0) and maximum (e = es) projected areas of the missile, for the walls and roof of the target, respectively. :
For normal impact velocities between Vw0 and Vwl, there will be some' projected areas for which wall damage is not possible. The minimum normal impact velocity is calculated for two atditional o values of G. The results of these calculations are used to generate a third order polynomial which gives the maximum value .
of e for which wall damage is possible for normal impact velocity V.
l 3.5-15 Rev. 7, 06/82 ,
ew (V) e AwV3 BwVa + CwV - Dw (Vwo 5 V $ Vw1) (3.5-21)
The probability that wall damage will occa is staply.
Gw (V)/es (3.5-22) where it is assumed that the distribution of missile orientation is uniform. A similar expression is developed for.the roof.
er (V) = ArV3 - BrVa + CrV - Dr (Vro 5 V 5 Vrl) (3.5-23)
'For strikes on the target wall, the. calculation cf the normal [
component of : impact velocity is simplified, since, for parabolic trajectories, the horizontal velocity component is constant,
- i. Vcose'. .With aw defined'as the angle between the wall normal and -
the vertical plane containing the trajectory, the normal impact velocity component is given by:
V = Vcose' cosa (3.5-24) c nw w The range of the normal components of the missile velocities which result'in wall strikes is:
V.o 1 =. Vocose' cosa $V $ V cose' cose i
=V i )
n w nw w n (3.5-25)
For each value of et and vij, an increment in the damage probability, P ijw, 3 can now be calculated.
Let **ij = sin-1(sinei cossij) and C = cose' cose
'If V $ V o, P3 =-0 ni w. ijw e.(V )
$V {=V ni w n dVn If V-no 5V wo
< V ni wi
, P3 ijw j V e G (V max
-V min 1 C 4 w s ij (3.5-26)
Rev. 8, 07/82 3.5-16
i LGS FSAR If V. -$ V. <V <V , P3 =
no -. wo wi ni ijw I y v -v :
wi i wi e (V ) C ;
w n dVni C + 11 V e. (v -v i tj (v -v 3 ;
wo s max min max min !
I (3.5-27) ;
If V <V <V $V , P3 = '
.wo no ni wi ijw V e (V ) ;
ni w n dVn C V e (V -V ) ij 3 no s max min !
(3.5-28)
If V < V. <V . <- V , P3 =
wo no ni wi ijw i V V -V ni i wi ,
e (V )
- C !
w n dVn C + ii .
V e (V -y ) ij (V -Y )
wo s max min max min (3.5-29) e If V 2V ,P 3
= V V no wi ijw i- o (V -V ) (3.5-30) i
. max . min
{
To simplify the .alculations for the roof, it was assumed that l the roof is at the same elevation as the turbine. This is '
conservative for s'. abs that are above the turbine. Under this assumption,-the range of vertical velocities with which the missile. strikes the roof is:
- V 3 = V i sine' 5 V 5 V 3 sine' = V (3.5-31) n nr ns The computation of the increments P3 ijr due to roof strikes is similar to the calculationof the increments Psijw. To obtain 3.5-16a Rev. 7, 06/82
z y ,
i,- .
s
{
.~,,%,* t
,4+
I y {< -
P 3 ijr, Vno is replaced Vni, Vni, Gw or er, Vwo ylth Vro , Vw with
~
Vr and Cij with sine'ij in the ex7ressions for Psijw. The probability that the target is,di^. aged, given h turbine failure, is: 1 I J ,
P3xP 3 = I I at a+1 -(P3' + P3 ) (3.5-32) 1-1 j=1 2n * -t ijw ijr max min An average P3 for the target is calculated by setting:
P? = P,xP, (3.5-33)
P, Probability of Turbine Missile Damage (P.)
The results of this probability analysis _are shown in Tables 3.5-2 and 3.5-3. Table 3.5-2 presents the values of P,, P 3, P,xP 3 , and P, computed for Unit 1 targets due to each of the two Target areas are shown on_ Figure 3.5-1. Table l turbine 3.5-3 _ trains. total values c P2, Pz Px 3, P3, and P, for each presents the of the Unit 1 targets and the totals.for the entire unit.
Because of the symmetrical arrangement of the Limerick units, the results for Unit 2 are identical. Table 3.5-3 shows that the ;
total probability of turbine missile damage for the two turbine trains is 4.11x10-18 per unit per year.
3.5.1.3.4 Conformance with Regulatory Guide 1.115, Protection Against Low-Trajectory Turbine Missiles This guide discusses guidelines for the protection of essential systems from turbine missiles. The Limerick design :s predicated on the low probability of turbine missile damage, and therefore the discussion of Paragraph C.4 applies: the probability of damage summed over all essential systems, if there is a turbine failure, should be less than 10-3.-
As discussed above, the Limerick analysis for damage ey spalling did not calculate the probability of damage, assuming a turbine failure, to each essential system.' Rather, the upper bound values (i.e., the probability that a missile will hit a protective wall and cause spalling) were calculated. The sum of
-the upper bound. values is 8.4 x 10-2 Although this value is greater than 10-3, it should be viewed in the light of the following conservatisms:
- a. -The value includes the probabilities associated with high-trajectory missiles, while the Regulatory Guide specifically refers to low trajectory (direct shot) missiles.
Rev. 7, 06/82 3.5-16b
- b. Not-all spalling causes damage,_and damage to an essential system is not necessarily unacceptable if damage does not occur to the redundant counterpart of the damaged system.
- c. The analysis assumes only the final barrier (reactor f enclosure wall, roof, etc). No credit is given to intermediate walls, equipment, etc, which would impede missile flight.
. These considerations, coupled with the fact that the total
. probability that a turbine will fail and damage essential systems t
6 1
3.5-16c Rev. 7, 06/S2
THIS PAGE IS INTENTIONALLY BLANK
):
1 P
Rev. 7, 06/82 3.5-16d .
LGS FSAR is 4.11 x 10-1* per year, which is well below the acceptable
~
value of 10-7, demonstrates that Limerick conforms with the intent of the= guide.
3.5.1.4 Missiles Generated by' Natural Phenomena Only tornado-generated missiles have been considered. Missiles used in the design and assessment of structures and openings are
' listed in Table 3.5-4. The structures designed for these tornado missiles and the systems protected are listed in Table 3.3-2.
. Table 3.5-8 provides information on the characteristics of these
-barriers. . Additionally, emergency service water and RHR service water systems yard piping is protected by burial.and separation
.of. redundant loops.
The spray pond system is not provided with' tornado missile protection per se;'however, it is designed to accommodate fpostulated multiple failures during tornado events. Limerick has four 50% capacity networks that are independent and can~be individually damage-isolated. Only two networks.are needed for a safe' shutdown of both units. Missile failure of one network with a single active failure of another or (multiple) missile-failure of two networks still permits safe shutdown. Furthermore, the only single active failure that can_cause the loss of a spray network-is the. failure of a. network valve to open due to valve operator-failure. This is only a temporary loss, since the valve can be manually opened to that, in reality, two spray networks can be lost by tornado missiles as well as by the single active
. failure of-the valve motor operator without affecting the ability
-of-the ultimate _ heat sink to perform its safety function.
Limerick i:s in conformance with Regulatory Guide 1.117, " Tornado
~
Design Classification," regarding systems to be protected from tornado missiles except as discussed above where unacceptable damage to' unprotected spray networks is not considered crediole.
3.5.1.5 Missiles Generated by Events Near the Site The_ nearest possible train explosion accident and its consequent missiles are considered to be the most severe missile-generating
-event that could occur near the site. The postulated missiles
- resulting'from such an accident considered in the design of
' structures protecting safety-related systems are listed in
. Table 3.5-5. Missiles resulting from truck, industrial, and
- ; pipeline explosions would be less severe and therefore are not considered. -_As demonstrated in Section 2.2, there is no
-potential-for missiles from ship or barge explosions or military e installations. . Descriptions of-the-railroad, its location relative' to the plant,.the railroad explosion, and explosions
_from:other' sources are given in Section 2.2.
3.5-17 Rev. 22, 07/83 v
^ .; ;
- b. The actual' calculated loads if the non-seismic side ,
piping is designed to a conservative simplified seismic j design criteria (e.g., by simplified span methods such !
as-those used for designed of.small piping) l
- c. The loads determ'ined by the plastic capability of the piping, i
. 3. 7 '. 3.14 Seismic Analysis for Reactor Internals (NSSS)
The modeling of RPV internals is discussed in ,
Section 3.7.2.3.1.2. The damping values are given in Table 3.7-1. A comparison of seismic responses is shown in ;
Table 3.7-4.
I 3.7.3.15 Analysis Procedures for Dampino i
3.7.3.15.1 . Analysis Procedures for Damping (NSSS) {
' Analysis procedures for damping are discussed in Section
-3.7.2.15.1.
Analysis Proce' dure for Damping.(Non-NSSS)
~
3'.7.3.15.2 If the equipment damping is unknown, the response spectrum curve for 0.5%. damping is used to arrive at a conservative seismic
. loading. The damping values used for the.OBE are increased for )
the SSE, where sufficient justification is established.
3.7.'4 SEISMIC INSTRUMENTATION 3.7.4.1~ Comparison With NRC Reculatory Guide 1.12 Rev 1 The seismic instrumentation program complies with Regulatory Guide 1.12 Rev.1, except for the item listed below:
Response spectrum recorders are not. supplied as discrete instruments. A response spectrum analyzer, permanently installed in the control room, presents more complete information than that
~
-presented by response spectrum recorders. Recorded data from the triaxial time-history accelerographs are. fed'into the response spectrum analyzer to produce earthquake spectra immediately following an earthquake. All locations where response spectrum recorders are required by the regulatory guide are monitored by time-history accelerographs. This system achieves the intent of Regulatory Guide 1.12 Rev 1.
3.7.4.2 Locati.on and Description of Instramentation
.The'following instrumentation is provided for Unit 1 only, as essentially the same response is expected at Unit 2.
Rev. 20, 05/83 3.7-28
i i
- a. Seven_ triaxial time-history accelerographs -
i
- b. Three triaxial peak recording accelerographs l j
- c. One triaxial seismic switch f d.- Two triaxial seismic triggers j e .One response spectrum analyzer [
- f. A' system ~ control panel which includes seismic event i
! visual and audible annunciators l g.' Cassette recorders and a playback unit i
-All instrument characteristics meet the requirements of ANSI I
!N18.5-1974 Section 5.
l 3;7.4.2.1 Triarial~ Time-History Accelerographs (T/As) t i
I T/As produce a record of the t'ime varying acceleration at the !
sensor location. These data are used directly for analysis and !
comparison with reference information, and may tue converted to l response spectra form for spectral. comparisons with design t parameters. ;
Each T/A contains three accelerometers mounted in a mutually orthogonal array. All T/As have their principal axes oriented l
identically, with one horizontal axis parallel to the major.
horizontal _ axis assumed in the seismic analysis. T/As are
-located as follows:
( a.. Free field (plant site east!of reactor / turbine L i enclosures) e b. Primary containment' foundation
.c.- Containment structure-(diaphragm slab) t L l$ . - ~ Reactor enclosure foundation !
i e.- Reactor. piping support (in containment)
- f. Outside containment on seismic Category I equipment l' residual heat removal'(RHR)' heat exchanger in reactor l- enclosure) l i
- g. Foundation of an independent seismic Category I l
structure (spray pond pump structure) j I A triarial-seismic trigger (S/T), sensitive in north, south, and I I I vertical directions,-is provided to start the T/A sensors L 3.7-29 Rev. 20, 05/83 I i
, i
?
recording system, One S/T is' shared by items a. through f. i Labove, and a second S/T is provided for item g. above. A !
magnetic tape recording system located in the control room is >
.provided for multiple channel recording of the signals from the l T/As mounted on items a. through f. A separate (locally mounted) '
recorder is provided for the T/A mounted on item g. A single playback unit is located in the control room for playback of the
' tapes from all of the recorders.
3.7.4.2.2 Triaxial Peak Recording'Accelerographs (P/AS) i
' Triaxial P/AS are provided to record the actual peak response. ,
Each. sensing device contains three accelerographs mounted in a !
mutually orthogonal array. Data from the peak recording :
accelerographs are manually retrieved following an earthquake. I P/As'are located as follows:
i
- a. Primary containment - on reactor. vessel equipment l
- b. Primary containment - on reactor piping c.. Outside of containment - on seismic Category I equipment
-(top of RHR heat exchanger piping in reactor enclosure) j F
3.7'4.2.3
. ' Triaxial Seismic Switch L
- 0ne triarial seismic switch'is installed on the primary ,
containment foundation. It activates visual and audible .
annunciators in the control room if an OBE input acceleration i level has been exceeded. . ;
3.7.4.2.4 Reeponse Spectrum Analyzer \
The_ response spectrum analyzer is an electronic device which )
generates a peak acceleration versus frequency curve from a time-based complex waveform. The analyzer receives data from the T/A ;
playback unit and computes the spectra. These can be compared j with'the spectra generated from the mathematical model and used .
to make timely operating decisions. ,
e '3.7.4.2.5 System Control Panel [
A panel located in the control room houses the recording, !
playback, and spectrum analysis units which are used in !
conjunction with the T/A sensors'to produce a time-history and [
. frequency-amplitude record of the seismic event. The panel also ;
contains. signal conditioning and display equipment associated , ;
with the response spectrum analyzer, audible and visual annunciators associated with activation of the seismic switch, !
and the system power supply unit. )
i c
i Rev.-20, 05/83 3.7-30 ;
LGS FSAR 3.7.4.3 Control Room Operator Notification !
Activation of the two S/Ts (Section 3.7.4.2.1) causes an audible t and visual annunciation in the control room to alert the plant !
operator that the T/A recording system has been activated. The set point of the triggers will be at horizontal or vertical acceleration levels slightly higher than the expected background I level, including induced vibration from sources such as traffic, ,
elevators, people, and machinery. These initial set points may be changed once significant plant operating data have been ,
obtained which indicate that a different setpoint would provide l better system operation.
- The seismic switch is connected to audible and visual
-annunciators in the control room and will indicate if the OBE acceleration has been exceeded.
The peak acceleration level experienced on the containment base ;
slab is available immediately following a seismic event. The level is obtained by playing back the recorded T/A data from the base slab location and reading the peak value from a chart recorder or the spectrum analyzer.
Significant response spectra from the containment base slab are available in the control room immediately following a seismic )
event. These will be on readout equipment suitable for comparing the measured response spectra with the OBE and SSE response '
spectra.
3.7.4.4 Comparison of Measured and Predicted Responses l
I Initial determination of the seismic event level is performed immediately after'the event by comparing the measured response ,
spectra from the containment base slab with the calculated OBE f and SSE response spectra for the corresponding location. An outline of the order of actions to be taken after a seismic event L is provided in Figure 3.7-44.
3.
7.5 REFERENCES
3.7 N. C. Tsai, " Spectrum Compatible Motions for Design Purposes", Journal of Encineerino Mechanics Division, uSCF. , Vol. 98, No. EM2, Proc. Paper 8G07 (April 1972),
pp. 345-356. '
3.7-2 " Seismic Analyses of Structures and Equipment for Nuclear Power Plants", BC-TOP-4A, Rev. 3, Bechtel Power Corporation, San Francisco, California (November 1974).
3.7-3 Uniform Buildin*c Code (UBC), by International Conference r of But1 ding Officials, Whittier, California, 1970 Edition. :
3.7-31 Rev. 20, 05/83 -
P
i
, +
7.5.1'4.2.1.2 -~ Reactor Pressure t
niso reactor'pt. essure signals are transmitted from two independent
. pressure transmitters _and are recorded on two, 2-pen recorders.
One pen recor'ds pressure and the other pen records the wide-range l level. The range of recorded pressure is from 0 to 1500 psig.
Power is fed-from two-independent Class 1E power sources.
i O !
"7.5.1.4.2.1.3 Primary Containment Pressure ;
IWide-range primary containment signals are transmitted from two ;
pressure transmitters. One signal is recorded on pen't of a 2-pen recorder while the other signal is displayed on an ,
indicator: located in the control room. ThePower range of both is supplied from instruments is.from 10-psia to 165 psig. ;
.two independent Class 1E power sources.
~
t I One narrow-range' primary containment signal is transmitted from a pressure transmitter and-is indicated in-the control room. The range of the indicated pressure is-'from -5 to +5 psig. Power is -
supplied from a' Class 1E power source.
725.1'.4.2.1.4- Primary Containment Gas Analyzers q
'2
? !
i
.Two redundant. analyzer pack' ages,-each'containing a hydrogen
~
analyzer cell and an oxygen analyzer cell, monitor primary These analyzers are part of the n containment hydrogen and oxygen.
, containment atmospheric control system (Section-6.2.5.2.2). The !
hydrogen analyzer _has a range of 0 to 30% (by volume) and the ;
oxygen analyzer.is a dual-range devi~ce that measures the' ranges of.0-to-10% and 0 to 25%.(by volume). The hydrogen and oxygen [
concentrations are indicated at the sample _ cabinet and in the j I Control room. '
Power is supplied from two independent Class 1E power sources.
{
7.5.1.4.2.1.5 Primary Containment Radiation Monitors l t
Four-ion chamber sensors measure the gross radioactivity present
-in'the containment atmosphere and transmit their signals to j
' radiation recorders located in the control room. (Section :
7.6.1.1.6). The range of recorded radiation is from 1 to 10s l
Rev 16',~01/83 7.5-4 [
f f
- - - . ~ . _ _ _ _ -
I' s LGS FSAR rads per hour. Power is supplied from two independent Class 1E power sources.
27.5.1.4.2.1.6- Suppression Chamber Pressure One suppression chamber pressure _ signal is transmitted from a pressure transmitter and is recorded in the control room. This signal is recorded on pen 2 of a two-pen pressure recorder located in the control room. The range of recorded pressure is from 10 psia to 165 psig.
Power is supplied from a Class 1E power source.
7.5.1.4.2.1.7 Suppression Pool Temperature Two independent divisionalized microprocessors are located in the control room to monitor temperatures from 16 independent corperature sensors located in the suppression pool. Eight
. temperature sensors are dedicated to each microprocessor. Power is supplied from two independent Class 1E power sources. Each micrcprocessor has a digital display with which the operator can
- select the sensor to be displayed. Normally the display indicates-the average temperature of the eight temperature sensor inputs. -A control room alarm is generated when the average temperature' increases to 95, 105, 110, and 1200F. In addition, an alarm will be provided when any temperature loop malfunctions.
The suppression pool temperature monitoring system (SPTMS) is described-in the Design Assessment Report, Appendix I.1.
7.5.1.4.2.1.8 Suppression Pool Water Level Two suppression pool water level signals are transmitted from two independent level transmitters. Each signal is transmitted to an indicator, located in the control room.
Power is supplied from two independent Class 1E sources (Table 7.5-3).
7.5-5 Rev. 21, 06/83
LGS FSAR 7.5.2.5 General Functional Recuirements Conformance s
. Conformance_of-the' transmitter / trip unit ~ system, used for safe shutdown display, is discussed in the Licensing Topical Report NEDO-21617, Rev. A, Analog Transmitter / Trip Unit System for Engineered Safeguard, Sensor Inputs. Conformance of the other features with the Regulatory and Industry Standards is discussed in the following sections. ,
7.5.2.5.1 Specific Regulatory Requirements Conformance 7.5.2.5.1.1 Conformance to Regulatory Guides 7.5.2.5.1.1.1 Regulatory Guide 1.47-1973, Bypass and Inoperable Status Indication
. The SRDI is designed to operate continuously, and tnere is no requirement for bypass provisions. Removal of instrumentation for servicing during plant operation is administratively controlled. Refer to the individual safety system analysis discussions of Regulatory Guide 1.47 contained in Sections 7.2, 7.3, 7.4 and 7.6.
7.5.2.5.1.1.2 Regulatory Guide 1.97 - Revision 2, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident l Limerick conforms with the intent of the Regulatory Guide, which is to ensure that instrumentation systems be provided to assess equipment and plant conditions during and following an accident as required by 10CFR Part 50, Appendix A, General Design Criteria 13, 19 and 64. In general, the Regulatory Guide requirements are
-implemented except where deviations from the guide are justified technically and can be implemented without disrupting the general intent of the guide. In assessing Regulatory Guide 1.97, Limerick has drawn on information in ANSI /ANS4.5 and on data derived from other analyses and studies.
7.5-19 Rev. 16, 01/83
+ -
in :7.5.2.5.'1.'1.2.1 Design and Qualification Criteria -
~
L'imerick conformance with the design and qualification criteria !
' defined in the regulatory. positions of Regulatory Guide 1.97 is i summarized ~'in Table 7.5-2 and discussed below (the paragraph t
- numbers cited
- correspond to those in Regulatory Guide 1.97). 4 i
l a. Paragraph 1.3: Instruments used for accident monitoring I' .to meet the provisions of Regulatory Guide 1.97 have the >
proper sensitivity,-range,. transient response, and ;
accuracy to ensure that the control room operator is .
able'to perform the role of bringing the plant to, and ;
maintaining it in, a-safe shutdown condition and in l assessing actual or possible releases of radioactive' i material following an accident.
Accident-monitoring instruments that are required to be environmentally. qualified are qualified to the l
' ~
requirement of NUREG-0588 (Section 3.11.2). The seismic qualification of instruments, where. required,'is in -
accordance with Regulatory Guide 1.100 (Section 3.10).
The Limerick quality. assurance program is used to comply with the' quality assurance requirements (Section 8.1.6.1 j<
and Chapter 17). '
t Periodic checking, testing, calibrating, and calibration i verification of accident-monitoring instrument channels (Regulatory Guide 1.118) are in accordance with the Limerick Technical Specifications.
- b. . ~ Paragraph 1.4:-- ' Instruments designated as Categories 1 and 2 for variable types'A, B, and C are identified in ,
such a manner as to optimize the human factors -
engineering and presentation of information to the control room operator.- This oosition is taken to ,
clarify the intent of Regulato*y Guide 1.97, which specified that these instrure.ttr be easily discerned for i use during accident conditions. L c.. Paragraph 1.6: It is '.imerick's pcsition that Table 1 :
of Regulatory Guide 1.97 does not represent the minimum ,
number ~of variables, correct ranges, or instrumentation
' categories for accident monitoring at a BWR facility. '
The Limerick list of accident monitoring variables and [
I
- Rev. 16, 01/83 7.5-20 f 1
,, w; LGS FSAR classification of instrumentation as Category 1, 2 or 3 is-in compliance with the intent and method used in Regulatory' Guide 1.97. The Limerick position on implementation of each variable i:s presented in Section 7.5.-2.5.1.1.2.3.
- d. Paragraph.2: .Conformance with' Paragraph.1.3 described above i:s applicable to the Type D and E variables of Regulatory Guide 1.97. ,
!? 7.5.2.5.1.1.2.2 Analysis.for. Type A Variables l p
Regulatory Guide 1.97, Revision 2, designates all Type A variables as plant-specific, thereby defining none in particular.
- ' Type'A variables for Limerick have been selected in conformar
-with the definition in Regulatory' Guide 1.97. Variables associated.with contingency actions that will be identified it. I written procedures are excluded. The.following is a list of Type
-A variables specific to Limerick. Detailed description of each Type A variable is given in Section. 7.5.1.4.2.1. The variables listed here are also included in Section 7.5.2.5.1.1.2.3.
the combustible limits, initiate combustible gas control l system. .
Safety function: Prevent combustible concentrations and thus preserve containment integrity. .
[b. Variable A2. RPV Pressure l 7 5
. Operator actions (1) Depressurize RPV and maintain-safe !
cooldown rate by any of several systems, such as main !
turbine' bypass valves, HPCI, RCIC,'and RWCU; or (2) ,
manually open one SRV to reduce pressure to.below SRV setpoint if an SRV is cycling.
Safety function: (1) Core cooling; (2) maintain reactor ;
coolant system integrity.
7.5-21 Rev. 16, 01/83 i
I t
. Variable A3. .RPV Water Level s
ic.
Operator action: Restore and maintain RPV water level.
i Safety function: Core cooling d.. Variable A4. Supppression Pool Water Temperature i t
Operator action: (1) Operate available suppression pool cooling system when pool temperature exceeds normal operating limits; (2) scram reactor if temperature
' reaches limit for scram; (3) if suppression pool
. temperature cannot be maintained below the heat capacity temperature limit, maintain RPV pressure below the corresponding limit;.and (4) close any stuck-open relief ;
valve.
i i Safety function: -( 1 ) Maintain containment integrity and '
(2)' maintain reactor coolant system integrity.
i
- e. Variabl'e ' AS . Suppression Pool Water Level f I f Operator action: Maintain suppression pool-water level ;
' within normal operating limits: (1) transfer RCIC suction.from.the condensate storage tank to the suppression pool-in the event of high suppression pool level;.and (2) if suppression pool water. level cannot be
' maintained below the suppression pool load limit,
-maintain RPV pressure-below corresponding limit.
Safety function: . : Maintain containment integrity.
i f.- . Variable A6. Drywell Pressure- $
p Operator action: Control primary containment pressure ;
by any of several systems, such as containment-pressure ,
~~
controlisystems, standby gas treatment, suppression pool sprays, drywell. sprays. ' t Safety functions (1) Maintain containment integrity and (2) maintain reactor coolant system integrity.
1 i
.Rev. 16, 01/83 7.5-22 i
LGS FSAR 7.5.2.5.1.1.2.3 Plant Variables for Accident Monitoring l Limerick's implementation cf the variables listed in Table 1 of Regulatory Guide 1.97 and the fulfillment of design _ criteria and l assignment of qualification categories for the instrumentation proposed.for their measurement are summarized in Table 7.5-5.
Measurement of the five variable-types'provides the following
-kinds of informaticn to plant operators during and af ter an accident:
t
'a. Type A--plant pressure, barrier and heat sink information, on the basis of which operators can take specified manaal control actions .
. ~b. Type B--information about the accomplishment of plant safety functions c.
~
Type C--plant information about the breaching of barriers to fission product release d._ Type D--information about the operation of individual safety systems
- e. Type E--information about the magnitude of the release of radioactive materials.
The categories are also related (in Regulatory Guide 1.97) to l
" key variables." Key varicbles are defined differently for the different variable types. For Type B and Type C variables, the key variables are those variables that most directly indicate the accomplishment of a safety function; instrumentation for these ,
key variables is designated Category 1. 1
. , . Key variables that are Type D variables are defined as those
-variables that most directly indicate the operation of an emergency safety system; instrumentation for these key variables '
is usually Category 2. Key variables that are Type E variables are defined as those variables that most directly indicate the release of radioactive material; instrumentation for these key variables is also usually Category 2.
7.5-23 Rev. 23, 08/83
.4 LGS FSAR The. variables are listed in Table 7.5-5 in the same sequence as
- in Table 1 of Regulatory Guide 1.97; however, for convenience in cross-referencing entries and supporting data, the variables are designated by. letter and number. For example, the sixth B-cype
' variable listed in Regulatory Guide 1.97 is denoted in Table 7.5-5 as variable B6.
The Limerick position is shown for each variable. In general, there are three. kinds of responses, the variable and instrumentation ares ( 1.5 implemented to meet the Regulatory Guide criteria; (2) implemented with qualifying exceptions or (3)-
not implemented.
As~necessary, the positions are elaborated or substantiated in
- the supplementary analyses in Section 7.5.2.5.1.1.2.4.
References to these analyses are made in the tabulation by citir.g
' the appropriate FSAR section.
7.5.2.5.1.1.2.4 Regulatory Guide 1.97 Project-Position The issues used to substantiate deviation of the Limerick syste.t from the Regulatory Guide 1.97 criteria are presented below.
-7.5.2.5.1.1.2.4.1 Variable B1 Neutron Flux Issui Definition-
~ The measurement of neutron flux is specified as the key variable
- in monitoring the status of reactivity. Neutron flux is classified as a Type B variable, Category.1. The specified range
- is 10-* percent to.100 percent full power (SRM, APRM). The stated purpose is " Function detectioni accomplishment of mitigation."
E vussion The lower end of the specified range, 10-* percent full power, is intended to allow detection of an approach to criticality by some undefined and noncontrolled mechanism after shutdown.
Rev. 23, 08/83- ,
7.5-24
?
In attempting tc analyze the performance of the neutron flux monitoring systems, a scenario was postulated to obtain the
-required approach to criticality. Basically, it assumes an increase in reactivity from loss of boron in the reactor water after SLCS actuation. ,
i The accident scenario incorporates the following factors: l l
- a. The control rods fail (completely or par,tially) to t insert, and the operator actuates the standby liquid control system (SLCS). ,
i
- b. The SLCS shuts down the reactor. l L
- c. A slow leak in'the primary system results in an outgo of '
borated water and its replacement by water that contains no boron.
- d. A range of leak rates up to 20 gpm was considered (Table -
7.5-4).
Calculations were made to evaluate the rise in neutron population !
as a function of different leak rates. The calculations were made for a shutdown neutron level of 5 x 10-e percent of full >
power. The choice of 5 x 10-s was based on measurements at two BWR plants. The shutdown level was assumed to have a negative ,
reactivity of 10 dollars, an assumption that is representative of !
a shutdown with all rods inserted. The results of the i calculations are presented in Table _7.5-4. The numbers in the -
table refer to the time in hours required to increase the flux by 1 decade.,For example, with a leak of 5 gpm, it takes 100 he to increase the power from 5 x 10-8 percent to a 5 x 10-7 percent, and 10 he to increase it from 5 x 10-7 percent to 5 x 10-* .
percent.
-The reactor is subcritical and the neutron level is given by: l Neutron level = S x H, l P where S is the source strength and M is the multiplication, which is given by:
r f
7.5-25 Rev. 16, 01/83
- M : = 1/( 1 -k ) , .
For k =10.9, M is 10; for k = 0.99, M is 100 and so forth. For :
criticality, the denominator approaches 0, as k approaches 1.0.
.Thus, the~above equation was used to calculate relative neutron !
flux. levels for a subcritical reactor.until the reactor was near l critical; then the critical equation of power with excess ;
reactivity was used. Reactor power is directly proportional to i neutron level.
LThetincrease'in reactivity toward criticality can be turned
.around by actuating che SLCS. A second-actuation of the'SLCS would cause a-decrease in reactivity because of the high concentration of' boron in the injected SLCS fluid relative to that in the leaking' fluid (nominally 400 ppm). The sensitivity
.of..the detector must allow adequate time for.the operator to act.
For a-scaling evaluation, ten minutes was' considered sufficient
- time for operator action for' accident prevention and mitigation.
Table 7.5-4' shows that the detector sensitivity (i.e., lower r - range) requirement is a function.of leak rate and therefore of
-reactivity-addition rate. On the basis of a 20-gpm leak rate, Table 7.5-4 shows that a detector that is one scale (i.e., about s) within 3 decades.cf the shutdown power (10-8) would allow O.'18.,hr (10.8 min) for operator action before reactor power increased another decade. A total of 0.36 hr (21.6 min) would be
-availableffor operator action from the time the ndicator comes
. 4
.on-scale to the~ time reactor power reaches 0.5 percent of full power. An alarm.would be provided to warn the operator when the neutron flux reaches =some plant-specific setpoint.
The 20-gpm leak rate, which,was assumed to continue for 27.75 he, was used to define the sensitivity of the detector. It should be
.noted that the assumed leak rate, extended over the 27.75-hr period, would result-in a loss of inventory so largeMoreover, that it
- could not in reality.go undetected by the' operator.
reactivity-addition caused by this gradual boron depletion is funlikely because boron concentration is sampled and measured
-periodically. Again, the improbable 20-gpm-leak rate was used only1to.obtain:a mechanistic and conservative approach for selection of/ instrument sensitivity.
An absolute criterion for the lower range musc include t
consideration-of the neutron source level. The use of the neutron level'100 days after shutdown is censervative. There is
'high probability that conditions would be stable and controllable 2 days.after the emergency shutdown, for the core-decay heat is Rev. 16, 01/83 7.5-26 4
.. ,-em9 -. ~ ~ . , ,,.
ws -.,e- ,,.,,<,y,,,,,w.,. ,,v. --.-w--m mma,.----%,
LGS FSAR at a low level and the boron monitoring system should be functioning by that time. The actual neutron level will vary.
-with fuel design, fuel history, and shutdown control. strength.
Measurements of shutdown neutron flux (with all rods-inserted) at two.BWR reactors show readings of 30 to 80 counts /sec (1000 counts /sec corresponds to 10-8 of full power). Measurements on other.BWR' reactors and for different fuel histories would show acme variation, but those variations would be small compared with a criterion that is concerned with units of decades.
Neutron flux is the key variable for measuring reactivity control.- The degree to which this variable is important to safety is another consideration. The large number of detectors (i.e., source-range monitors and intermediate-range monitors)-
that are driven into the core soon after shutdown makes it highly probable that one or more of the existing NMS detectors will be J inserted. On the other hand, there is little probability that there would be, simultaneously, a need for this measurement (in terms of operator action to be taken) and an accident environment in which the NHS would be rendered inoperable. Further, the ope.acor can always actuate the SLCS on loss o' instrumentation.
Conclusion l Although some upgrading of the current NHS is appropriate to improve system reliability and its ability to survive a spectrum of accidents, a rigorous Category 1 requirement is not justified relative to the criterion of "importance to safety." A ,
Category 2 classification of this measurement fully meets the intent of Regulatory Guide 1.97 for neutron flux indication.
7.5.2.5.1.1.2.4.2 Variable B4, Coolant Level in Reactor l 7.5.2.5.1.1.2.4.2.1 Reactor Level Monitoring l
[
Limerick used overlapping ranges to monitor coolant level in the reactor as discussed in Section 7.5.1.4.2.1.1.
The reference leg of the wide-range water level transmitter is 5
- feet lower than the Regulatory Guide.1.97 required tap, i.e.,
centerline of the main steam lines. It was necessary to use this range to el;minate long runs of exposed sensing line tubing that contributt to erratic indication. The variable leg of the fuel zone water level is 39.12 inches lower than the bottom of the 7.5-27 Rev. 16, 01/83
LGS FSAR core support plate. These two level monitors cover the range specified by Regulatory 1.97'except as mentioned'above.
7.5.2.5.1.1.2.4.2.2 Trend Recording
-Issue Definition The purpose of addressing this issue is to determine which-variables. set forth in Regulatory Guide 1.97 need trend recording.
Discussion -
Regulatory' Guide 1.97, paragraph 1.3.2f, states the general requirements for trend recording as follows: "Where direct and
-Owners Group Emergency Procedures Guidelines (EPGs) as a basis, the only trended variables required for cperator action are reactor water level and reactor vessel pressure.
-Conclusions
'For Limerick, only reactor water level (variable B4) and reactor vessel pressure (variable B6) require trend recording.
7.5.2.5.1.1.2.4.3 Variables B8 and C6, Drywell Sump Level and Drywell Drain Sumps Level, respectively
. Issue Definition Regulatory Gu'ide 1.97 specifies Category 1 instrumentation to monitor drywell sump level (variable B8) and drywell drain sumps level (variable C6). .These designations refer to the drywell equipment' drain tank and floor drain tank levels. Category 1
. instrumentation indicates that the variable being monitored is a key variable. In Regulatory Guide 1.97, a key variable is defined as ". ' . that single variable (or minimum number of
-variables) that most directly indicates the accomplishment of a safety function . . .
The following discussion supports the BWR Owners. Croup alternative position that drywell sump level and Rev. 16, 01/83 7.5-28
drywell drain sumps levels should be qualified to Category 3
. instrumentation' requirements.
Discussion LThe Limerick drywell has two drain sumps. One drain is the equipment drain sump, which collects identified leakage; the other is the floor drain sump, which collects unidentified leakage.
i Although the level of.the drain sumps can be a direct indication of breach of the reactor coolant system pressure boundary, the indication is not unambiguous because there can be water in those sumps during normal operation. Other instrumentation specified ,
in Regulatory Guide 1.97 that would also indicate leakage in the !
I drywell are identiftsd below:
- a. Drywell pressure--variable B7, Category'1 [
- b. Drywell temperature--variable D7, Category 2
- c. Primary containment area radiation--variable C5, Category 1. ,
i The drywell sump levels signal neither automatically initiates safety-related systems nor alerts the operator to take safety- ,
related actions. Both sumps have level detectors that provide :
only the following nonsafety indications:
- a. Continuous level indication
- b. High-level alarm
- c. High-high-level alarm ,
- d. Low-low level alarm l
- e. Average sump leak rat'e l
- f. Sump leak' rate ine ease alarm l l i
i Regulatory Guide 1.97 specifies instrumentation to function i during and after an accident. The drywell sump systems are deliberately isolated at the primary containment penetration on "
receipt of an accident signal.to establish containment integrity.
This fact renders the drywell sump level signal irrelevant.
7.5-29 Rev. 23, 08/83
_.__ _ _ _ _ _ _ _ _ _ , _ . _ . ~ . . . _ _ -
_ - - . , _ _ _ _)
LGS FSAR Therefore, by design drywell level instrumentation serves no useful accident mon 8.toring function.
The Emergency Procedure Guidelines use the RPV level and the drywell pressure as~ entry conditions for the Level Control Guideline. A small line break wil; cause the drywell pressure to increase before a noticeable increase in the sump level.
Therefore, the drywell sumps will provide a " lagging" versus "early" indication of a leak.
Conclusions
? Based on the above, Limerick believes that Category 3, "high-quality off-the-shelf instrumentation" is appropriate for drywell sump level and drydell drain sumps level instrumentation.
~
7.5.2.5.1.1.2.4.4 Variable C1, Radioactivity Concentration or g Radiatior Level in Circulating Primary Coolant Issue Definition Regulatory Guide 1.97 specifies that the status of the fuel 7
-cladding be monitored. The specified variable is C1--
radioactivity . concentration or radiation level in circulating primary coolant. The range is given as "1/2 Tech Spec Limit to 100 times Tech ~ Spec Limit, R/hr." Instrumentation for measuring variable C1 is designated as Category 1. The purpose for monitoring this variable is given as " detection of breach",
referring, in.this case, to breach of fuel cladding. ,
_D_iscussion
~
The critical actions that must be taken to prevent and mitigate a gross breach of fuel cladding in a BWR are (1) shut down N.e reactor and (2) maintain water level. Monitoring varia' ale C1, as directed in Regulatory Guide 1.97, will have no influente on either of these actions. Any usefulness from this monitored
, variable falls into the category of "information that the barriers to release of radioactive material are being challenged" and " identification of degraded conditions and their magnitude, so the operator can take actions that are available to mitigate the consequences." There are no additional operator actions to
, mitigate the consequences of fuel barriers being challenged, other than those based on Type A and B variables.
Rev. 16, 01/83 7.5-30
P g
~Although the subject of concern in.the Regulatory Guide 1.97 l requirement is assumed to be an isolated NSSS, Limerick has given i consideration to events.that do not isolate the NSSS. The post- [
accident: sampling system (PASS) provides a means of obtaining ,
samples of reactor coolant and primary containoent atmosphere. !
Analyses of these samples provide information on the status of i fuel cladding integrity when the plant.is isolated. Radiation l monitors.in the steam jet air ejector and main steam lines :
provide information on the status of fuel cladding'when the plant !
is not isolated. ;
Conclusion Instrumentation for measuring variable C1 is implemented as l Category 3 because no planned operator actions are identified and -
no operator actions are anticipated based on this variable serving as the_ key variable. ;
i i
7.5.2.5.1.1.2.4.5 Variable C14, Radiation Exposure Rate ,
!P Issue Definition f i
Variable C14 is defined in Table 1 of Regulatory Guide 1.97 as ;
follows: " Radiation exposure rate (inside buildings or areas ;
which_are in direct contact with primary containment where l
. penetrations and' hatches'are located)." The reason for :
monitoring variable C14 is given as " Indication of breach". ;
i Discussion The use of local radiation exposure rate monitors to detect !
breach or leakage through primary containment penetrations is j impractical. 'In general, radiation exposure rate in the secondary containment will be largely a function of radioactivity l in primary containment and in the fluids flowing in ECCS piping, i which will cause direct radiation shine on the areas of concern !
where radioactive fluids are piped. Because of the amount of ,
piping and the number of electrical penetrations and hatches and i their widely scattered locations, local radiation exposure rate :
monitors could give ambiguous indications. Breach of containment L is more appropriately assessed by using the ncble gas effluent [
monitor provided to monitor variable E4. See also Section L 7.5.2.1.1.2.4.10. l f
. . p 1
I
/
~ Limerick is not-implementing this parameter. Other means of breach ~ detection that'are better suited to this function (as-described above), are available. Radiation exposure rate monitors as described in Section 12.3.4.1 are provided in these
-buildings for indication of habitability only.
7."5.2.5.1.1.2.4.6 Variables D3 and D8, Suppression Spray Flow and Drywell Spray Flow, respectively.
' Issue Definition Regulatory Guide 1.97 specifies flow measurements of suppression
. chamber spray (variable D3) and drywell spray (variable D8) for monitoring the operation of the' primary containment-related
' systems. Instrumentation for measuring these variables is designated Category 2, with a range of 0 to 110 percent of design flow. These floss relate to spray flow for controlling pressure and temperature of.the drywell and suppression chamber.
Discussion ,
The drywell sprays can be used to control ~the pressure'and temperature of the drywell. . The residual heat removal (RHR) system flow element is used for measuring drywell flow.
The pressure suppression enamber sprays can be used to control the pressure and temperature in the suppression chamber. 1From
- the control room, the' operator controls pressure and temperature by adjusting suppression chamber spray flow. The RHR system ~ flow
- element is used for flow i.idication. The suppression chamber-spray operates-in parallel with the drywell spray and is
. regulated-with a throttling valve. The flow is determined by RHR flow indication. The effectiveness-of' spray flow can be verified
- by pressu.e and temperature changes of the drywell and the suppression chamber as indicated in the control room.
G2gelusions The current plant equipment, in conjunction with operating practice, meets performance requirements of accuracy and Rev. 16, 01/83 7.5-32
7.5.2.5.1.1.2.4.7 Variables D13-D17, RCIC Flow (D13), HPCI Flow (D14), Core Spray System Flow (D15), LPCI ;
System Flow (D16) and SLCS Flow (D17) -
- Issue Definition !
Regulatory; Guide 1.97 specifies flow measurements of the .
following systems: reactor core isolation cooling (RCIC)
(variable D13), high pressure coolant injection (HPCI) (variable D14),--core spray (CS) (variable D15), low pressure coolant
- injection (LPCI) (variable D16), and standby liquid control (SLC)
- t. -(variable D17). The purpose is for monitoring the operation of
- individual safety systems. . Instrumentation for measuring these
- variables is designated as Category 2; the range is specified as O to 110 percent of design flow. These variables are related to L
flow into the reactor pressure vessel.
f Discussion -
The RCIC, HPCI, and CS systems each have one branch line--the l test line--downstream of-the flow-measuring element. The test line is provided with a motor-operated valve that is normally closed (two valves in series in the case of the HPCI). In
-addition, the valve.in the test line closes automatically when the emergency system is actuated, thereby ensuring that indicated flow is'not being diverted by the test.line. Proper valve *
- position can_be verified by a direct indication of valve position. ,
t Although the LPCI has several' branch lines located downstream of each flow-measuring element, each of those lines is either normally closed or automatically aligned. On initiation of the LPCI, the valves in the system automatically line up for proper operation and prevent flow diversion by branch lines. Proper ,
valve position can be verified by a direct indication of valve position.
For all of the above systems, there are valid primary indicators other than flow measurement to verify the performance of the emergency system; for example, vessel water level. The SLC system is manually ini'.iated. Flow-measuring devices were not provided for this system The pump discharge header pressure, l ,
. 7.5-33 Rev. 23, 08/83 .
,-,+mm , -w- - -c -r - -.c%w,,.m.---,--%,-- y,-c.---,-- , - . - - , - - - - - - . , - , . ,,v---.w.--------,n, -w-,,,---, - - , - . - - . - + - - - - - - - -
e LGS FSAR which is indicated in the control room, will indicate SLC pump operation. Besides the discharge header pressure observation,
- the operator can verify the proper functioning of the SLCS by monitoring the following:
- a. Decrease in,the level of the boric acid storage tank
- b. Reactivity change in the reactor as measured by neutron flux and concentration of boron
- c. Motor contactor indicating lights (or motor current);
the use of these indications is believed to be a valid altern.'tive to SLCS flow indication (some. plants have i'.dichtses for open/close positions of check valves)
~
~
1.97. Monitoring the SLCS can be adequately _done by measuring variables other'than the flow.
7.5.2.5.1.1.2.4.8 Variable D18, SLCS Storage Tank Level Issue Definition Regulatory Guide 1.97. lists standby liquid control system (SLCS)
- storage tank level as a Type D~ variable with Category 2 design and qualification criteria.
' Discussion Regarding the instrumentation category requirement for variable D18, Regulatory Guide 1.97 indicates that it is a key variable in monitoring SLC system operation. Regulatory Guide 1.97 also states.that,.in general, key Type D variables be designed and qualified to Category 2 requirements.
Rev.'16, 01/83 7.5-34
~
InLapplying these requirements of the Guide to this ;
-instrumentation, the following are noted: f
- a. The current design basis.for the SLCS~ assumes a need for i
_ an: alternative method of reactivity control without a concurrent-loss-of-coolant accident or high-energy line i break. The environment'in which the SLCS !
instrumentation must work is therefore a mild l environment for qualification purposes. j
- b. The current design basis for the SLCS is recognized as ,
considerably less than the importance tx) safety of the
=
reactor protection system and the engineered safeguards. l
- systems. Therefore, in accordance.with the graded j approach tx> quality assurance specified in Regulatory l Guide 1.97, it is unnecessary to apply a full quality l assurance. program to this instrumentation, t Conclusion l i i
SLCS-storage tank level instrumeatation will meet Category 3- l !
l design and qualification criteria as required by Regulatory Guide ' l 1.97. i i
7.~5.2.5.1.1.2.4'.9 Variables D26-D30, Main Steam Bypass Valve :
Position (D26), Condenser Hotwell Level (D27), l Condenser Pressure (D28), Circulating Water Pump Discharge Pressure (D29)'and Reactor -
i Recirculation Pump Flow (D30) [
Issue Definition l l hH I Regulatory Guide 1.97 states that "The plant designer should L select: variables and information display channels required by his .
' design to enable the control room personnel to ascertain the !
operating status of each individual safety system and other }
-systems important to safety-to that extent necessary to determine j if each system-is operating or can be placed in operation..." ';
The purpose of this analysis was to determine whether certain
.ther D-type variables should be added to Table 1, Regulatory G9ide 1.97.
i
{
f 7.5-35 Rev. 16, 01/83 (
l
_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ .__ .i
l LGS FSAR Discussion Regulatory Guide.1.97 addressed safety systems and svstems important-to safety'to mitigate consequences of an accident.
Another lict of_ variables has been compiled for the BWR in NUREG/CR-2100. That report and a companion report, NUREG/CR-1440, address. plant systems not important to safety, as well as systems that are important to safety. In particular, these reports consider the potential role of the turbin? plant in
-mitigating'certain accidents. These two reports were reviewed in determining whether the listed variables (D26 to D30) should be added to the Regulatory Guide 1.97 list.
The NUREG evaluations used a systematic approach to derive a
' variables list. The basic approach of the analysis was to focus on those accident conditions under which the operator is most
'likely to be confronted with "and/or" accident conditions that result in the most serious consequences if the operator should ;
. fail to accomplish his required tasks. This is a probabilistic event-tree-type of study, and the reports used the sequences of ,
'the Reactor Safety. Study (WASH 1440), and similar studies. The <
events in each sequence that involved operator action were i identified; also, events were added to the event tree to include p additional operator actions ^that could mitigate the accident.
The event tree defines a series of key plant states that could ;
evolve as the accident progresses and as the operator attempts to
- respond. Thus the operator's informational needs are linked to ,
these plant states.
6 NUREG/CR-2100 is a BWR evaluation undertaken to address !
appropriate operator actions, the information needed to take those actions, and the instrumentation necessary to provide the ;
required information. '
.7 L
The coquences evaluated were:
- a. . Anticipated transient followed by loss of decay heat
_ removal }
- b. Anticipated transients without scram (ATWS) i
b
>[
- d. .Large loss-of-coolant accident (LOCA) with failure of emergency core cooling systems
, e. Small LOCA with failure of emergency core cooling systems The Regulatory Guide 1.97 list is based on accidents that result in an isolated NSSS. The NUREG documents considered accidents that could be prevented or mitigated by using water inventory and the heat sink in the turbine plant.
Conclusion Five of the 15 variables. identified in the NUREG, but not in Regulatory Guide 1.97, are included as Type D, Category 3 additions to the Regulatory Guide 1.97 list. Four of these v'ariables are in the turbine plant: the main steam bypass valve position, condenser hotwell level, condenser pressure, and
-circulating water pump discharge pressure. These variables provide a primary measure of the status of a heat sink or water inventory in the turbine' plant. The turbine plant systems are not to be' classed as " safety systems" or as systems important to safety. .The reactor recirculation pump flow is also added~to the Regulatory Guide 1.97 list.
7.5.2.1.1.2.4.10 Variable E2, Reactor Building or Secondary Containment Radiation Issue Definition Regulatory Guide 1.97 specifies that. reactor building or secondary containment area radiation (variable E2) should be monitored over the range of 10-1 tc 10' R/h for Mark II containments. The classification for Mark II is Category 2.
Discussion As discussed in the variable C14 analysis (Section 7.5.2.5.1.1.2.4.5), secondary containment area radiation is not i an appropriate parameter to use to detect or assess primary containment leakage.
7.5-37 ,
Rev. 23, 08/83 i
LGS-FSAR Conclusion The specified reactor enclosure area radiation monitors are not
' required for the Limerick secondary containment.
7.5.2.5.1.1.2.4.11 Variable E3, Radiation Exposure Rate
" Issue Definition Regulatory Guide 1.97 specifies in Table 1, variable E3, that Eradiation exposure rate (inside buildings or areas where access is required to service equipment important to safety) be monitored over the range of 10-1 to 10' R/hr for detection of significant releases, for release assessment, and'for long-term surveillance.
Discussion-
'In general, access is not required'to any area of the secondary containment to service equipment important to safety in a post-accident situation. If and when accessibility is reestablished
-in the long term,'it will be done by a combination of portable radiation survey instruments and post-accident sampling of the
- secondary containment atmosphere. Lower-range area radiation monitors (typically 3 decades lower than the Regulatory Guide 1.97 range) are provided for use only in those instances in which radiation levels are very mild.
There are areas outside. secondary containment 'there access is
.needed post-accident for specific sampling, monitoring or analysis tasks.- Dose rates greater than 10 R/hr are not expected in these areas.. In the event that these areas experience dose
-rates greater than 1.0 R/hr, access could result in excessive operator exposure.
-Conclusion The Limerick design does not require access to a harsh environment area to service safety-related equipment during an accident; portable radiation monitors will be provided to reestablish accessibility. Lower range area monitors are implemented as Category 3 for areas outside secondary containment where post-accident access is needed.
Rev. 16,'01/83 7.5-38
7.5.2.5.1.1.2.4.12 Variable E8, Plant and Environs Radiation l \
i Issue Definition l r
Regulatory Guide 1.97 specifies that plant and environs radiation (variable E8) should be monitored over the range of 10-3 R/hr to 10' R/hr, photons-and 10-3 Rads /hr to 10* Rads /hr, beta radiations and low energy photons. The classification is !
Category 3.
Discussion l l The plant inventory of portable radiation survey instrumentation described in Section 12.5.2.2.3 will be supplemented with additional equipment to enhance post-accident monitoring capabilities. This additional equipment will be comprised of i low , medium , and high-range portable ion chambers (1 mR/hr to ;
20,000 R/hr gamma and 20,000 Rad /hr beta), open window alpha !
scintillation probes, pankake GM probes, energy compensated
. beta / gamma GM probes (for low energy photons), and portable i beta / gamma geiger counters. Audio speakers, alarming count rate !
meters, and extention arms will be provided for attachments to ;
the survey instruments. Airborne radioactivity levels will be determined from laboratory analysis of particulate filters and ,
iodine cartridge samples obtained with high and low volume , !
samplers. Portable instruments and equipment reserved for ;
emergency use will be located at an assembly area remote from the j main plant. r i
Conclusion l ;
A range of monitoring of 10* R/hr would not enhance plant and environs radiation monitoring. Limerick meets the intention and >
the purpose of the Regulatory Guide criteria.
7.5.2.5.1.1.2.4.13 Variable E13, Primary Coolant and Sump l i i
r Issue Definition l Regulatory Guide 1.97 requires installation of the capability for ,
obtaining grab samples (variable E13) of the containment sump, ECCS pump room sumps, and other similar auxiliary building sumps !
F i
7.5-39 Rev. 16, 01/83 w -. . _ _ . - - _ _ - _ _ _ . - _ - , - . _ _ - . _ - - . - - - - . -.-..-.- -.-____-.O
for the. purpose of release assessment, verification, and j
- analysis.
l Discussion )
e The need for sampling a particular sump must take into account i its location and the sump design. For all accidents in which radioactive material would be in the primary cont.ainment sump, it will be isolated and will overflow to the suppression pool. A ;
suppression pool sample can be obtained through-the post-accident i sampling station as described in Section 11.5.5 and this can !
" therefore be used as a valid alternative to a containment sump [
sample. !
. The analysis of ECCS pump room sumps and other similar auxiliary !
building sump liquid samples can be used for release assessment, I as. suggested'in Regulatory. Guide 1.97 only if potentially !
radioactive water can be pumped out of a controlled area to an :
area such as radwaste. If_the design does not allow sump pump- !
out on a high-radiation.aignal, a sump sample does not contribute ;
t to release assessment. The use of the subject sump' samples for verification and analysis is of little value; a sample of the ,
suppression ~ pool and reactor water, as required by other portions '
of Regulatory Guide 1.97, provides a_better measurement for these l purposes' l
1 Conclusion I
E !
! A suppression pool sample-will be used as an alternative to a
- primary containment sump sample. The. analysis of ECCS pump room
, iencicsure (e.g., pumped to radwaste). Limerick design does not y allow sump pump-out on receipt of high radiation signal. The !
L capability for sampling and analysis of ECCS pump room and [
l
-auxiliary building sumps is therefore not provided. j f
I i
.7.5.2.5.1.2 Conformance to 10 CFR Part 50, Appendix B j
. I I
The SRDIs, except the displays, are of the same type, and are subject to the same qualification testing, quality control, and !
documentation, in accordance with the recommendation of 10 CFR, Part 50, Appendix B as the safety systems' instrumentation. The i I
Rev. 16, 01/83 7.5-40 [
t :
7.6.1.1.6 Primary Containment Post-LOCA Radiation Monitoring [
System (PCPL-RMS) - Instrumentation and Controls f l 7.6.1.1.6.1 PCPL-RMS Identification The objective of the PCPL-RMS is to monitor.the total intensity !
f of gross radioactivity present inside the containment. In
' correlation with other points of consideration this information :
-provides a basis for making post-accident decisions.
The PCPL-RMS is shown in Figure 11.5.1 and specifications are given in Table 7.6-1, t 7.6.1.1.6.2 PCPL-RMS Power Sources l Power to this monitoring system is supplied from two 120 V ac l Class 1E power buses. {
PCPL-RMS Redundancy and Diversity !
. 7.6.1.1.6.3-
~
l Four physically separated sensors, two in each of two electrical !
separation channels, provide the required redundancy. [
7.6.1.1.6.4 PCPL-RMS Testability A built-in source of adjustable current is provided to simulate sensor input to each radiation monitor for test purposes. The L operability of each monitoring channel can be routinely verified l by comparing the outputs of the channels during power operation.
7.6.1.1.6.5 PCPL-RMS Environmental Considerations l This system is designed and qualified to meet the environmental o conditions under post-accident considerations. In addition, this L.
system is seismically qualified as described in Section 3.10.
7.6.1.1.6.6 PCPL-RMS Operational Considerations No trip or annunciation capability is provided for this system as it is designed for post-accident surveillance only. Continuous radiation levels are recorded on dedicated Class 1E recorders in the control room.
1 Rev. 19, 04/83 7.6-12
t
.. i I
t
?
'11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND
[
SAMPLING SYSTEMS '
s The process and effluent radiological monitoring and sampling f systems, including the primary containment radiation monitoring i system, are contained in the process radiation monitoring system '
(PRMS). The PRMS is provided to furnish information to operations personnel regarding radioactivity levels in principal plant process l and effluent streams to assist in maintaining radiation levels as !
Iow as reasonably achievable. The system is also provided to j verify compliance with applicable governmental regulations for the i
' containment, control, and release of radioactivity in liquid and [
gaseous effluents generated as a result of normal or abnormal operation of the plant. The design objectives and criteria are i primarily determined by the system designation of either:
a.- Safety-related systems
- b. Monitoring-systems for plant operation, and operator information.. j
- c. Post-accident monitoring and sampling systems !
' 'The PRMS is composed of the following process and effluent radiation monitors:
t i
- a. Safety-related systems-
- 2. Reactor enclosure ventilation exhaust radiation
- monitors 3.- Refueling area ventilation exhaust radiation monitors
- 4. Control room air supply radiation monitors )
- 5. Control room emergency fresh air radiation monitors
- 6. Residual heat removal service water radiation '
monitors l ,
- b. For plant operation and operator information r t
- 1. North stack effluent radiation monitors l f
- 2. South stack effluent radiation monitors l- f f
- 3. Radwaste equipment rooms ventilation exhaust t radiation monitor 11.5-1 Rev. 29, 02/84 .
1
_ _ _ . -. _ ~- - .._.-_.-- _ ___ _ ._ -
m .
'LG5 FSAR
- 4. Charcoal treatment system process exhaust radiation monitors'
- 5. Recombiner compartments and hydrogen analyzers exhaust radiation monitors
- 6. Steam seal effluent ~ radiation monitors
- 7. Radwaste enclosure ventilation exhaust radiation monitor
- 8. Air ejector effluent radiation monitors
- 9. Primary containment leak detection radiation monitors
- 10. Hot maintenance shop ventilation exhaust radiation monitor
- 11. Liquid radwaste discharge radiation monitor
- 12. Plant service water radiation monitors-
- 13. Reactor enclosure cooling water radiation monitors.
- c. Post-accident systems r
- 1. Primary containment post-LOCA radiation monitors
- 2. North stack wide-range accident monitoring system
- 3. Pcst-accident' sampling system 11.5.1 ' DESIGN BASES AND SPECIFIC REQUIREMENTS The radiation monitoring systems are designed to measure and record
. radioactivity levels, to alarm on high radioactivity levels, and to prevent the release of radioactive liquids, gases, and particulates. The PRMS aids in protecting the general public and plant personnel from exposure to radiation or radioactive materials in excess of'those limits allowed by the applicable regulations.
The main objectives of the PRMS for normal operation are as
.follows:
- a. To provide surveillance of radioactivity levels in process and effluent streams from minimum detectable levels to levels commensurate with Technical Specification limits by indicating and recording these levels, by' alarming at abnormal activity levels, and by initiating or causing the initiation of corrective action when applicable.
Rev. 29, 02/84 11.5-2 .
i l
- b. To provide data for estimating total released activity.
For some anticipated operational occurrences the PRMS activates I necessary isolation or diversion valves, thereby terminating releases if radioactivity levels exceed preestablished setpoints. ;
( l
! The main objective of radiation monitoring systems required for !
L safety'is to initiate appropriate protective action to limit the :
potential release of radioactive materials trom the reactor l vessel and reactor enclosure refueling area,~if predetermined i radiation levels are exceeded.
l 11.5.1.1 General Desion Criteria Specific design criteria for monitoring systems are discussed !
under the system description when applicable. General criteria !
taken into consideration in the design of the PRMS are as '
follows: ;
! a. The PRMS is' designed to monitor pathways for release of -
radioactive materials to the environment in conformance i with General Design Criteria 60, 63, and 64 of i 10 CFR 50, Appendix A and Appendix I, and Regulatory L Guide 1.21. :
- b. The PRMS provides early warning of increasing radioactivity levels indicative of equipment failure, i filter failure, system malfunction, or deteriorating L system performance by using a high alarm setpoint. ;
i l- c. The PRMS initiates prompt corrective action, either !
automatically or through operator response, on high radioactivity level by using a high-high alarm setpoint.
I d. Monitors and detectors are selected with sensitivities
! and ranges in accordance with radiation levels j anticipated at specific detector locations. ;
i ;
l e. Monitors register full-scale if exposed to radiation !
- levels exceeding full-scale indication. l
- f. Radioactivity levels are continuously indicated and }
recorded in the control room with the exception'of !
liquid radwaste effluent activity, which is recorded in !
the radwaste control room, and the hot maintenance shop l ventilation exhaust ~ activity, which is indicated and l recorded locally. i l g. Control room alarms annunciate high radioactivity levels !
and signal, circuit, or power failures. l I
11.5-3 ;
mmwe-m ew.e--~.a.rwm-se.n~ . . , - -
,v .
- h. Monitor components requiring calibration, maintenance and inspection are accessible, and spare equipment is commercially available.
- 1. Insofar as practical, self-monitoring of components is provided to the extent that power. failure or component malfunction causes annunciation and channel trip.
- j. Off-line (sampling-type) radiation monitoring systems
^
are designed to be unaffected by background radiation levels of up to 2.5 mrem / hour. Off-line monitors are located in areas where the maximum background does not increase the minimum detectable concentration above the required monitor lower. range. Accident dose rates have been considered in locating monitors that are required for safety or post-accident. In-line radiation monitoring systems will generally indicate changes in background radiation levels. Environmental qualification of safety-related components is discussed in Section 3.11.
- k. Safety-related components of the PRMS required for safe
~
I' . shutdown are protected against the effects of extreme winds, floods, tornadoes, or missiles by locating them in a structure designed to withstand such conditions I (see Chapter 3).
=1. LSafety-related monitors are designed to seismic
- requirements consistent with the seismic design of the system being monitored.
- m. Independence of redundant monitors that are safety--
related is maintained by providing adequate separation
' of detectors,. signal cabling, power supplies, and actuation circuits for isolation and diversion valves to meet IEEE 279 criteria.
11.5.1.2 Basis for Detector Location Selection Normal and potential paths for release of radioactive material are selected for monitoring as follows:
- a. Process lines that may discharge radioactive fluids to the environs, in order to indicate the radioactivity level and to alarm in the control room when established limits for the release of radioactive materials are reached or exceeded.
- b. Process lines that do not discharge directly to the environs, in order to indicate possible process system malfune: ions by detec:ing increases in radioactivity levels.
Rev. 29, 02/84 11.5-4
i i
3 l
[
Monitored processes and detector locations are listed in [
Table 11.5-1 and shown in Figure 11.5-1. !
11.5.1.3 Expected Radiation Levels The expected radioactivity concentrations in the process and effluent streams are such that radiation levels at the site boundary are a small fraction of 10 CFR 20 limits and will be as low as reasonably achievable.
Ouantity'to be Measured ;
11.5.1.4 '
The principal radionui.'_ des to be monitored are indicated in i Table 11.5-1. All' channels measure gross radioactivity. {
t L 11.5.1.5 Detector Tvoe, Sensitivity, and Rance j j The detectors are Geiger'-Mueller (GM) tubes, ionization chambers, t L or: scintillation crystals that' detect either beta or gamma radiation, depending on the application.- In general, ion i i
chambers are used in high intensity or high temperature
! applications, GM tubes are used for gross measurements, beta L
scintillators are used for relatively precise measurements of t noble-gases, and gamma scintillators are used for relatively i precise measurement of iodines and particulates. The sensitivity [
and range are selected so that_the alarm setpoint is at least an l I' order of magnitude higher than the detector threshold and so that l
the instrument reads on scale during normal operation. Detector
~
type, sensitivity, and nominal ranges of'each process and ,
effluent monitor are indicated in Table 11.5-1. [
11.5.1.6 Setootnts f i
Setpoints for effluent monitors are established to meet Technical !
i Specification limits, which encompass 10 CFR 20 limits and as low !
as reasonably achievable guidelines. Setpoints.for process [
monitors are established to provide a warning of increased system j activity and to initiate corrective action wnere appropriate. In i all cases, the setpoints are established to maintain offsite f radiological effects within applicable regulation limits. !
Changing of the setpoints is under the administrative control of 1 the plant superintendent or his authorized delegate. I r
Two independently adjustable radiation setpoints are provided for ,
most' monitors. The "high" setpoint normally activates only an ;
, ' alarm;-while the "high-high" setpoint activates an alarm and i initiates corrective action where appropriate. The alarm and trip circuits are of the latching type and must be manually' reset .
on the PRMS panels located in the control room. Setpoints are at
. least twice the background level if practicable to reduce the [
. number of spurious trips. Radiation monitoring system setpoints ;
and associated functions are provided in Table 11.5-2.
, i 11.5-5 i
. i i
~. - - - -
.?
3- f i
f LGS FSAR i 11.5.1.7- 'Annunciaters and Alarms. !
i All process and. effluent radiation monitors indicate end audibly >
alarm in the control room. A specific annunciator window is '
illuminated for each low (failure), high, or high-high rediation t alarm or low sample flow alarm, as shown in Table 11.5-2.
i e
An eperator'can acknowledge the alarm and silence the audible alarm, but he cannot clear the annunciator window until the alarm ;
condition has been cleared at the PRMS panels located in the !
control room. Radiation alarms can be cleared only if the l indication is less than the setpoint. l At the_PRMS panels in the control room, the channel that alarmed ;
.and the-type of alarm are determined by the lights associated with three types of alarms. These alarm lights are as follows: I la .
A high alarm light illuminates when the radioactivity l exceeds preset limits that have been selected to provide ,
j an early warning.
- b. A high-high alarm light' illuminates when the J radioactivity exceeds preset limits that have been selected to initiate appropriate corrective action. The l i
high-high alarm actuates the trip auxiliaries. in those )
applications where a control trip is automatic. e
- c. A low (failure) alarm light illuminates when the meter l reaches a downscale trip point which indicates that there is a detector signal, circuit, or power failure. i In certain cases, as shown in Table 11.5-2, this l downscale alarm also actuates the trip auxiliaries, s
'11.5.1.8 Calibration, Maintenance, Inspection. Decenta : nation, and Replacement
.All instruments are calibrated upon installation. Calibration l sources.are provided for periodic recalibration of the detectors l without removing them from their installed positions. )
Pubge capability is provided to all offline instrument racks. In f event of severe contamination, the modules in question or the l entire rack-may beitransferred to the hot maintenance shop for decontamination.
i i
i i
i 1
Rev. 29, 02/84 11.5-6 I
~-- - -- __-__ ___,__,_______ __
t I
r V
A schedule for periodic replacement of instrument components, such al photomultiplier tubes and detectors, is based on manuracturer's recommendations. Instrument rack installation I facilitates access for inspection, testing, maintenance, and i
' repairs.
i
-Components of the radiologicalThe monitoring systems assemblies are designed are designed for a (
for convenient replacement. !
minimum life of 40 years when maintained in accordance with the manufacturers' recommendations including periodic parts
[
replacement. .
11.5.2 SYSTEM DESCRIPTION t
Specific information on the PRMS is tabulated in Tables 11.5-1 !
and.11.5-2 and arrangements are shown in Figure 11.5-1.
11.5.2.1 Systems Reouired for Safety l ;
i 11'.5.2.1.1 Main Steam Line Radiation Monitoring System This system monitors the gamma radiation level exterior to the main steam lines. The normal radiation level is produced l
f primarily by coolant activation gases plus smaller quantities of fission gases being transported with the steam. In the event of I
a gross release of fission products from the core, this monitoring system provides channel trip signals to the reactor [
j protection system (RPS) to initiate protective action.
l L
i i
i I
r i
i
[
i Rev. 17, 02/83 l l 11,$.6a
r .
.THIS PAGE IS INTENTIONALLY BLANK Rev. 1, 09/81 11.5-6b
t
The system consists of four redundant instrument channels. Each !
channel consists of a local detector (gamma-sensitive ion .
chamber) and a control room radiation monitor with an auxiliary (
trip unit.. Power for two channels (A and C) is supplied from RPS !
bus A and che other two channels (B and D) from RPS bus B. i Channels A and C are physically and electrically independent of l channels B and D. One two-pen recorder allows the output of anv* .
two selected channels to be recorded. j The detectors are physically located near the main steam lines just downstream of the outboard main steam line isolation valves !
in the space between the primary containment and secondary l containment walls. The detectors are geometrically arranged so l that this system is capable of detecting significant increases in i radiation level with any number of main steam lines in operation. ;
Table 11.5-1 lists the sensitivity and range of the detectors. '
Each radiation monitor has four alarm circuits: two upscale (high-high and high), one downscale (low), and one inoperative.
Each alarm is visually displayed on the PRMS panels located in j the control room. A high-high or inoperative signal results in a r channel trip in the auxiliary unit that is an input to the RPS. !
An RPS logic trip from the main steam line channel input results ;
in initiation of main steam line isolation valve closure, reactor ;
ceram, mechanical vacuum pump shutdown, and mechanical vacuum i pump suction valve closure. A high alarm actuates a main steam I line high radiation annunciator common to all channels. A !
downscale alarm actuates a main steam line downscale annunciator I common to all channels. High and low alarms do not result in a I channel trip. Each radiation monitor visually displays the measured radiation level.
1:
The main steam line radiation monitoring system is safety-related and is discussed in detail in Section 7.3.
11.5.2.1.2 Reactor Enclosure Ventilation Exhaust Radiation !
Monitors s l
This system monitors the radiation level of the air in the :
reactor enclosure ventilation system exhaust duct prior to its l discharge from the structure. .
1 The system consists of four redundant instrument channels. Each channel consists of a local detection assembly (a sensor and i converter unit containing a GM tube and electronics) and a !
radiation monitor. Power for two channels (A and C) is supplied i from RPS bus A and the other two channels (B and D) from RPS bus i B. Channels A and C are physically and electrically independent of channels 3 and D. Two two-pen recorders allow the output of )
the channels to be recorded continuously. The detection }
assemblies are located inside the exhaust duct upstream of the ;
ventilation isolation valves. ,
11.5-7 l
I
-o-,--sw vm-wwe--,ve %
cxs ,
- Each radiation monitor has three alarm circuits: high-high, high, and low. Two high-high trips in channels A and B initiate closure of the reactor enclosure ventilation isolation valves and the i startup of the reactor enclosure recirculation system and standby gas treatment system (SGTS) train A. The same conditions for channels C and D initiate closure of the corresponding tandem isolation valves and startup of the reactor enclosure recirculation !
system and SGTS train B. Alternate inboard and outboard valves are
- actuated independently by the two sets of trip channels. The same !
logic configuration of the four channels also initiates closure of the primary containment outboard purge and vent valves.
An upscale trip is visually displayed on the affected radiation i monitor and actuates a reactor enclosure ventilation high-high radiation annunciator common to all channels. A downscale trip ,
is also visually displayed on the radiation monitor and actuates an annunciator common to all channels. An additional alarm for hign radiation is provided by the recorder and actuates a common reactor enclosure ventilation high radiation annunciator. Each radiation monitor visually displays the measured radiation level.
The location and monitoring characteristics of the reactor enclosure ventilation radiation monitoring channels are adequate to provide detection capability for abnormal amounts of radioactivity in the reactor enclosure ventilation exhaust and to initiate isolation. The redundancy and arrangement of channels are sufficient to ensure that no single failure can prevent isolation when required. The upscale trips meet the design requirements of IEEE 279-1971 as described in Section 7.3.2.2.
11.5.2.1.3 Refueling Area Ventilation Exhaust Monitoring System This system monitors the radiation level in the ventilation exhaust duct from the refueling area, including the area over the fuel pool. The monitoring system is identical to the reactor enclosure exhaust radiation monitoring system with the same channel trip logic and protective action initiation, with the exception of the initiation of the reactor enclosure recirculation system.
During refueling operation (including criticality tests), the monitoring system acts as an emergency safety feature against the consequences of a refueling or control rod drop accident. The response of the reactor enclosure ventilation exhaust radiation monitortng system to the refueling accident is discussed in Chapter 15.
11.5.2.1.4 Control Room Air Supply Radiation Monitors This system monitors the radiation level in the supply air to the control room. No measurable activity is espected to be present; however, in the event of a design basis accident (DBA), fission Rev. 22, 07/83 11.5-8
gases could escape from the plant structures and be drawn into the supply air intake.
There are four independent monitors, separated in accordance with
=IEEE 279 '.971, that monitor air inside the control room intake i , duct. These inline monitors respond to the gross radioactivity in the vicinity of the detectors. Each monitor provides three
- alarm conditions low, high, and high-high. The low and high alarms trip the control room annunciator. The high-high alarm trips the control room fresh air isolation valves and starts the control' room emergency fresh air supply, which provides for the
' filtration of the incoming air through HEPA/ charcoal filters.
The trip of either monitor A or 5 shuts off the control room fresh air supply, and the trip of either monitors I. and C, or 5 and D, starts the cor. trol room emergency fresh air supply. (See Section 6.4 for a more detailed discussion of control room isolation on detection of high radiation.)
11.5.2.1.5 Control Room Emergency Fresh Air Radiation Monitors Upon initiation of the control room emergency fresh air supply system, this monitor indicates radioactivity concentration levels downstream of. control room ventilation HEPA/ charcoal filters.
Radioactive noble gas. concentration is measured and continuously recorded. Tnese inline monitors detect gross radiation only.
.Two monitors, separated in accordance with IEEE 279-1971, monitor sample air from the control room emergency fresh air duct.
Each monitor provides three alarm conditions: low, high, and high-high. These alarms trip annunciators in the control room.
i' All requirements of this system are identical with those of the control room air supply radiation monitors except that no provisions for valve closure are made.
11.5.2.1.6 Residual Heat Removal (RHR) Service Water Radiation Monitors This system is comprised of four monitors for sampling the RHR service water return from the two Unit 1 and two Unit 2 RHR heat enchangers, and two backup monitors for sampling the combined RHR
. loop return to the spray pond or cooling tower. Detection of
. leakage of radioactive fission products into the service water results in the tripping of RHR service water pumps in the 11.5-9 Rev. 29, 02/84
LGS FSAR affected loop and closing.the RHR heat exchanger service water return valve.
Each monitor provides three alarm conditions: low radiation, high radiation, and low sample pump flow. These signals trip annunciators in the control room, and the high closes the service water return valve or trips the RHR pumps depending on whether the signal originates from the heat exchanger return monitor or the combined loop return monitor, respectively.
These monitors are qualified as IEEE Clasc 1E.
11.5.2.2 Svitems Recuired for Plant Operation 11.5.2.2.1 North Stack Ventilation Exhaust Radiation Monitors The north stack ventilation exhaust radiation monitoring system is comprised of two subsystems: .
- a. Normal plant operation monitoring subsystem >
- b. Wide-range accident monitoring subsystem The objectives of the normal plant operation subsystem are to indicate whether the limits of actual release of radioactive material to the environs are reached or exceeded, and to measure the quantity of release of radioactive material during normal plant operation, in compliance with 10CFR50 and Regulatory Guide 1.21.
Rev. 29, 02/84 11.5-10
LGS'FSAR The stack radiation monitoring system, including the isokinetic sampling system and the wide-range accident monttoring subsystem, is designed to carry out the following functions:
- a. To provide continuous isokinetic and representative samples of the stack flow in compliance with the requirements of General Design Criterion 64 of 10CFR50, i Appendix A, Regulatory Guide 1.21, and ANSI 13.1-1971. l
- b. To continuously record releases of radioactive particulates, todines and noble gases to the environs so that the total quantity of radioactive material released can be evaluated.
- c. To alarm, in event that specified rates of release of radioactive material are exceeded,
- d. To provide continuous real-time indications of radioactive releases during the accident and post-accident modes of operation,
- e. Provide an isolation signal to the containment purge valves in the event of high radiation in the north stack l effluent.
The nocth stack exhausts from the following systems:
- a. Unit 1 turbine enclosure exhaust
- b. Unit 1 turbine enclosure equipment compartment exhaust (including mechanical vacuum pump exhaust)
- c. Unit 2 turbine enclosure exhaust
- d. Unit 2 turbine enclosure equipment compartment exhaust (including mechanical vacuum pump exhaust)
- e. Radweste enclosure equipment compartment exhaust i
- f. Radweste enclosure fume hood exhaust I
- g. Radweste service and control area exhaust Rev.23,68/83 11.5-11 0
I .
- h. 6 Control structure battery compartment exhaust
- 1. Unit 1 steam packing condenser and effluents from the recombination system
- j. Unit 2 steam packing condenser and effluents from the recombination system
- k. Standby gas treatment enclosure exhaust
- 1. Unit 1 battery compartment exhaust m.' Unit 2 battery compartment exhaust
- n. Control structure toilet room exhaust
- o. Standby gas treatment filter exhaust
- p. Drywell purge system exhaust
- q. difgastreatmentsystemexhaust.
- r. Radweste enclosure air exhaust Ssepling stubs are provided on the exhaust ducts of most of the systems listed above for the purpose, of extracting grab samples as needed (see Figure 11.5-1 for specific locations).
Units 1 and 2 share the north stack and consequently the same radition monitoring system. Under normal plant operation, the stack flow rate varies from about 183,000 cfm to about 664,000 cim.
The expected composition and concentrations of tho' ef fluent under normal plant operation are given in Table 11.5-4. Following an accident flow may be reduced as low as 1250 cfm. Under this condition, the flow rate will be below the range capaht11ty of the isokinetic sampling system, but the wide-range accident subsystem 4111 continue to provide representative data. The north stack is ,
provided with three equally-spaced honeycomb grids that serve the purpose of stabilizing, equalizing, and co111 mating the stack flow in order that the exhaust flow rata can be measured accurately and Rev. 17, 02/83 11.5 '1
y-l l
r A flow rate sensing representative air sampling can be achieved.
array is provided, consisting of 128 uniformly-spaced totd1 pressure sensors and 32 uniformly-spaced static pressure sensors Two .
for"providing an instantaneous traverse across the stack. l independent sampling arrays, each consisting of a set of l 64 uniformly-spaced isokinetic nozzles are provided for extracting ll l representative samples at the stack cross section.
?
One array provides a sample for'the normal plant operation radiation monitoring subsystem. The stack flow rate and sampling i The sample is flow rate are recorded and indicated on demand.
spilt into parallel paths. Each half is passed through a particulate f11ter provided with a radiation detector indicating ,
the corresponding integrated measurement of the particulate effluent, an iodine filter provided with an in-place detector, and ,l a noble gas monitoring chamber. Thus, each of.the two redundant !
l monitoring racks provide the following outputs: .
I t
l [
- Sampling flow rates l l
- Particulate radioactivity, integrated l
- Iodine radioactivity, integrated -
i Noble gas radioactive concentration l l e
E t
From these data and the stack flow measurement,Readouts the total from the j radioactive effluent may readily be evaluated. !
" detectors are fed into microprocessors, which in turn provide (
outputs to readout modules in the The aus111ary equipmentare microprocessors room and to provided recorders in the control capability with memory-retention room. to preclude the loss of data in ;
event of a power failure.
l' For each detector there are one downscale The and two upscale upscale alarmsalarms indicate lj which annunciate in the control room.
high and high-high radiation,'and the downscale alarm indicates l instrument malfunction. f j
For the normal plant operation mode, the characteristics of the i isokinetic sampling system and radiation' monitoring subsystem l ;
provide plant operations personnel with complete and accurate data ;
of radioactive materials released to the environs from the north )
- stack. The system thus enables personnel to control activity f
11.5 13 Rev. 17, 02/83 I
LGS TSAR release rates. Sufficient redundancy is provided to allow maintenance and checking of one channel without losing monitoring capability.
The wide-range accident monitoring subsystem is independent of the normal plant operation monitoring subsystem and operates continuously. Two samples are available. One sample, drawn from the second 64-nozzle array described above, is passed through a particulate filter, lodine filter, and noble gas monitoring chamber. This provides redundancy to the normal plant operation monitoring subsystem. For broad-range radiation monitoring a second, much smaller sample is drawn from a separate comb-type probe located downstream of the isokinetic nozzle arrays. This sample is passed through shielded particulate and iodine filters and two extended range noble gas monitoring chambers. Detector outputs are fed into microprocessors that evaluate the total radioactive effluents. Outputs of the microprocessors are transmitted te readout modules in the auxiliary equipment room and recorders in the control room. The microproc. ars have memory retention in event of loss of power.
The particulate and iodine filters of the wide-range accident monitoring subsystem are used as grab sample modules to provide the ecpability of collecting representative samples of iodines and particulates for onsite analysis during and following an accident.
The sample lines are heat traced to preclude entrained moisture in the effluent stream that could degrade the filters. Three removable filter modules are provided in both sample flow paths to allow continuous collection. This also allows the control room operator to select a clean set of filters in order to prevent appreciable concentrations of noble gases produced by iodine decay in a loaded filter, which could be falsely interpreted by the noble gas detectors as high activity in the effluent stream. Filters on the high activity sample flow path are shielded to keep personnel exposure in sample handling and transport below the General Design Criteria 19 limits of 5 rem whole-body exposure and 75 rem to the extremities during the duration of the accident.
11.5.2.2.2 South Stack Ventilation Exhaust Radiation Monitor The objectives and functions of the south stack monitoring system are the same as those of the north stack normal plant operation monitoring subsystem. A system for post-accident monitoring is not provided because any HVAC exhaust to this stack containing accident effluents is automatically isolated ~
Rev. 17, 02/83 11.5-14
LGS FSAR The south stack encloses two independent exhaust ducts servicing the reactor enclosures for Unit 1 and Unit 2, respectively. ;
The stack exhausts ventilation air from the following systems: I Unit 1 Duct ;
t
- a. Unit 1 reactor enclosure exhaust ;
- b. Unit I reactor enclosure equipment compartment exhaust f
- c. Refueling floor Unit 1 side exhaust -
Unit 2 Duct l
- a. Unit 2 reactor enclosure exhaust
- b. Unit 2 reactor enclosure equipment compartment exhaust l
- c. Refueling floor Unit 2 side exhaust ;
Each of these two ducts is monitored by means of two redundant {
subsyhtems. Consequently, four independent sets of data are ,
obtained of stack flow rates and corresponding sampling rates. l l
Flow rates in each of the two ducts vary from about 54,000 cfm to about 234,000 cfm. The expected composition and concentrations of !
the effluent under normal plant operation are given in
- Table 11.5-4. The stack flow is collimated to provide a uniform velocity distribution over the entire cross section to assure ,
representative sampling. Stack flow rate is measured by a manifold '
containing 64 uniformly spaced total pressure sensors and 16 static ;
pressure sensors in order to provide an instantaneous ongoing velocity traverse. Sampling is done by an array of 32 uniformly ;
spaced isokinetic nozzles.
Radiation detection is done by means of a particulate filter, l iodine filter and noble gas chamber in series. Each of these items is shielded and provided with a dedicated detector. The gas chamber has a beta scintillation detector consisting of a 11.5-15 Rev. 17, 02/83 !
1
~ - - , _ _
' beta-sensitive crystal optically connected to a photomultiplier tube.
The other'two detectors are similar in construction except that the todine detector is gamma-sensitive. The output-from the preemp1 Liter is fed to a microprocessor, which in turn feeds the o -reedout module in the ausiliary equipment room and the recorder in the contro1~ room. Digitized outputs are also available.
For each detector there are one downscale and two upscale alarms that are annunciated in the control room. The downscale alarm indicates instrument malfunction and the upscale alarms indicate high and high-high radiation..
Output records of stack flow rates, sampling flow rates, and count
~ rates of particulates, iodines, and noble gases are stored in the
- radiation display monitoring computer and displayed on CRT on demand.
.11.5.2.2.3 Radweste Equipment Rooms Ventilation i Eshaust Radiation Monitor The common duct that collects exhaust air from the charcoal offgas treatment equipment compartments ventilation ducts is continually monitored for airborne radioactivity. High-high, high, and low radiation are annunciated in the control room. Although particulate /todine concentration is not anticipated, the monitors are provided with particulate / Lodine filters as backups. In the event of a high. radiation alarm, the ventilation exhaust from each compartment can be monitored separately.
11.5.2.2.4 Charcoal Treatment System Process Exhaust Radiatien Monitors Two charcoal offgas trains - Unit I and Unit 2 - exhaust the processed gases via HIPA filters to the north stack. Each of the exhaust pipes is monttored:to detect malfunction in the
'_< corresponding offgas train. Low, high, and high-high radiation are annunctated in the control room.
Rev. 29, 02/84 11.5-16
LGS FSAR 11.5.2.2.5 Recombiner Compartments and Hydrogen Analyzees t Exhaust Radiation Monitors This monitor detects airborne radioactivity in the ventilation ducts of the recombiner compartments, drain sump rooms, and l
P hydrogen analyser compartments in the control structure. Exhaust air in the common duct from the recombiner compartments and drain surp rooms and the common duct from the hydrogen analyzer compartments is continuously monitored prior to return to the turbine enclosure equipment compartment ventilation filters for eventual release through the north stack. . Low, high, and high-high l radiation are annunciated in the control room. The source of the radiation can be identified by means of a hand selector switch / solenoid valve arrangement which allows switching the sample l line to the individual return duct of potentially contaminated compartments.
11.5.2.2.6 Steam Exhauster Discharge and Vacuum Pump Exhaust l
.This monitor provides an indication of radioactivity in the discharge of,the steam packing.eshauster to the north vent stack.
The sampling arrangement allows the capahility to also monitor the l mechanical vacuum pump eshaust at the water separator discharge when this equipment is operated during startup. Low, high, and high-high radiation are annunciated in the control room.
11.5.2.2.7 Radweste Enclosure Ventilation Exhaust Radiation Monitor Monitoring of the main eshaust duct of the radwaste enclosure provides for general survsillance of gaseous effluents from equipment compartments and access areas of this structure prior to i discharge to'the north stack. The instrumentation also provides backup for the radweste equipment rooms ventilation exhaust radiation monitor. Low, high, and high-high radiation alarms annunciate in the control room.
11.5.2.2.8 Air Ejector / Holdup Pipe Discharge Radiation Monitor l r
This system monitors radioactivity in the main condenser offgas at the discharge of the holdup pipe after it has passed through the steam jet air ejector (SJM) condenser, recombiner and aftercondenser. This is representative of gaseous radioactivity released from the reactor and therefore indicates the condition o the fuel cladding.
11.5-17 Rev. 29, 02/84
)
A continuous sample is estracted from the offgas pipe via a stainless steel sample line and passed through a sample chamber and a sample panel before being returned to the suction side of the SJAE. The sample chamber is a steel pipe that is internally polished to minimise plateout. It can be purged with room air to check detector response to background radiation. The sample panel measures and indicates sample line flow..
The sample chamber is monitored by three channels. Each channel has.a gamma-sensitive ionisatten chamber mounted outside the sample chamber. Two channels have legarithmic radiation monitors with low, high, and high-high alarm outputs that annunciate in the Icontrolroom.
The third monitoring channel has linear radiation sensing capability designed to detect radioactive fragments resulting from fuel cladding deterioration. Small changes in the offess gross fission product concentration can be detected by the e.ontinuous use of the linear radiation monitor. The linear radiation monitor is not a process monitor such as the channels described above, but is utilised as an espanded scale device for aiding in measurement of small changes in the offgas radiation level. De detector is a gamma-sensitive ionisation chamber that monitors the same sample as the logarithmic-monitors. The system uses 'a linear readout with a i range switch instead of a logarithmic readout. The output from the monitor is recorded on a one-pen recorder.
Grab samples can be obtained for determining isotopic composition by using the semiautomatic vial sampler panel. To draw a sample, a serum bottle is inserted into a sample chamber, the sample lines are evacuated, and a sample valve is opened to allow offgas to enter the bottle. The bottle is then removed and the sample is analysed in the counting room with a multichannel gamma pulse height analyser to determine the concentration of the various noble gas. radionuclides. A correlation between the observed activity and the monitor reading permits calibration of the monitor.
11.$.2.2.9 primary Containment Leak Detection Monitor i The design objective of this system is to detect leakage in the primary containment. The leak detection monitor is designed to monitor and' alarm for gaseous radioactivity (section 5.2.5.2.1.4).
The monitoring system is provided with four main control room annunciated alarms lhigh-high, high, downscale, and abnormal Rev. 29, 02/84 11.5-18
LG5 TSAR .
sample flow). Ratemeter output is recorded on a recorder in the main control room.
11.5.2.2.10 Hot Maintenance Shor /entilation Exhaust Radiation Monitor Equipment serviced in the hot maintenance shop is expected to be contaminated with residual particulate radioactivity. A small quantity of radioactive iodine might also be present. No radioactive gases are anticipated.
Continuous isokinetic sampling of the maintenance shop exhaust duct, downstream of the HEPA filters, is conducted in accordance with ANSI N13.1. The flow controller in the monitor skid maintains isokinetic sample flow proportional to the constant exhaust duct flow. This sample is passed through fixed particulate and iodine filters. Separate detectors measure the gross radioactivity accumulated on these filters. The microprocessor-based monitor calculates radioactivity cencontration based on the value of sample flow in memory and measured accumulated radioactivity over a time period. A recording of this radioactivity concentration is made locally, and an annunciator alarms in the event of discharges in excess of 10 CFR 20 Appendix B limits. All instrumentation is mounted locally.
11.5.2.2.11 Radwaste Discharge Radiation Monitor l Weste liquid may be discharged on a batch basis from sample tanks in the liquid waste management system as discussed in Section 11.2.3. Prior to discharge, the liquid in the tank is recirculated, sampled, and analysed for radioactivity, and the release and dilution rates are determined. The expected compsition and concentrations of the effluent from each sample tank under normal plant operation are given in Table 11.5-5. The liquid radweste effluent discharges into the cooling tower blowdown line. The liquid radweste discharge radiation monitor measures activity in the discharge line to prevent concentration in the cooling tower blowdown line from exceeding the 10 CFR 20 Appendix 3 activity limits.
Monitoring is performed in an offline sample rack in preference to inline monitoring. This arrangement affords improved sensitivity and precludes the necessity of shutting down the radweste discharge line in order to purge accumulated radioactive sludge from the ronitoring system.
11.5-19 Rev. 29, 02/84 m
t tas rsAn The monitoring channel consists of a gamma scintillation detector /preemplifier, a ratemeter in the auxiliary equipment l area. and a one-pen recorder-in the radweste centrol room. A low l alors trip wL1' initiate discharge valve closure because the trip l circuit has been doutened to "fati safe"'in the event of a loss [
et power. The high trip alarms in the radweste control room'and )
the high-high tris initiates discharge valve closure. Since the l redweste release ls based upon batch analysis, the basis for the '
alare setpoint en the monitor is that an alarm be given en a gross release in the range 10-* to 10.a ,CL/cc as a grees check against significant operater error. The alarm setpoint may vary
~ free batch to batch depending upon the activity concentratton of the batch and the available cooling tower discharge flow that is i' used to dilute the fluid effluent prior to leaving the site boundary. .
1't.l.1.1.11 Service Water Radiation Monitor plant service water return flow discharges to the cooling tower.
The return flow is menttered by the plant service water radiatten mentter.. Ne activity attributable to reactor operation is espected to be present in this line. For radteactivity to be present, leakage would have to occur simultaneously in equipment cooled by the reacter enclosure cooling water system and in the i reacter enclosure cooling water heat enchanger. Thus, this
-mentter provides a backup for the reacter enclosure cooling water mentter.
Offline manitoring is selected to facilitate decontamination-i- without shutdown. The enannel consists of a detector, preampittter, s rateeeter in the auntliary equipment room, and a one-pen recorder in the control room. The ratemeter provides a low and high alarm in the control rees. Annunciation due to low sample flew rate aise is provided. The alarm setpoint is based upon detecting leahage'into the servtco water witn the setpoint set sufficiently ateve background to preclude spurious alarms.
11.1.1.1.13 Reacter Enclosure Coeling Water Radiation Monitor The reacter enclosure cooling water system cools components that may contain radteactive liquids, but does not normally carry any r61Leactive materials unless the cooled components leak. A radtation mentter is provided to measure radtoactivity in the system.
Rev. 19, C1/84 11.5-10
i LGS FSAR An offline monitor is employed, to facilitate decontamination without shutdown. The channel consists of a gamma scintillation detector, preamplifier, ratemeter, and one-pen recorder. The channel is provided with a low and high alarm that annunciate in the control room. The high alarm set point tsAset sufficiently low flow above background to preclude spurious alarms.
annunciator also is prvvided.
11.5.2.2.14 Process Sampling System l A process sampling system is provided to allow grab sampling for evaluation of water quality and radioactivity levels in liquid process waste streams. The sample analysis results will provide i
operators with information for taking necessary corrective actions. This system is dasigned to provide representative samples from process streams at central sample stations for use in minimizing leakage, spillage, and potential radiation exposure I
to operational personnel. Where applicable, means are provided i for sample water cooling and for maintaining a fixed or measured sample flow rate. This system is described in Section 9.3.2. l Although the process sampling system is designed to provide liquid samples from many plant process streams, radionuclide ,
sampling will be periodically perfortaed on the following process ,
systems:
- a. Fuel pool cooling and clean-up
- b. Reactor enclosure cooling water Liquid radwaste - equipment drain processing c.
- d. Liquid radweste - floor drain processing
- e. Liquid radweste - chemical and laundry processing 11.5.2.3 Post-Accident Syste?s 11.5.2.3.1 Primary Containment Post-LOCA Radiation Monitor This monitoring system is comprised of four ton chamber sensors i for the primary containment in the event of a loss-of-coolant l accident (LOCA). After such a postulated accident, the monitoring system measures the gross radioactivity present in the l containment atmosphere. This information is transmitted to control room personnel to provide ther with a casts for making 11.5-21 Rev. 29, 0 /84
't t
L
- safety-related decisions. .The primary' containment post-LOCA F radiat0.on. monitoring system provides a. trip signal to the i containment sump pumps on an upscale alarm indication. -
A downscale annunciator is provided to' indicate instrument ;
malfunction. i l
The sensors are located in separate areas of containment to !
provide independent measurements and to view large fractions of
~
i the containment:vtlume (Figure 12.3-16). Consideration to !
accessibility for maintenance and calibration was given in the !
selection of the sensor locations. The sensors are located in !
relatively open areas to prevent shielding that could impair l their' detection function. ,
- The monitoring system provides energy response from 60 kev to ;
~
'3 Mev, with unifrom response within -205.from 80 kev to 3 MeV.
Onsite calibration of the monitors will be performed with a ;
calibrated 100 millicurie Cs-137 gamma source that will provide an effective dose rate of 10 R/hr. A built-in current source is provided in the monitors to allow calibration checks through !
electronic signal substitution'for the decades above 10 R/hr.
11.5.2.3.2 North Stack Wide-range Accident Monitoring System The north stack wide-range accident monitoring system is j
- discussed in Section 1 1. 5 . 2. 2.~ 1. :
- 11.~ 5. 2. 3 . 3 Post-Accident Sampling System [
- The post-accident sampling system is discussed in Section 11.5.5. f 11.5.3 EFFLUENT MONITORING AND SAMPLING The requirements of General Design Criterion 64 of 10CFR50 !
Appendix.A are implemented with respect-to effluent discharge !
paths by means of the following monitoring stations:
. l
- a. Gaseous Effluents:
.1. Reactor enclosure ventilation exhaust (Section }
11.5.2.1.2) j
- 2. Refueling area ventilation exhaust (Section ;
11.5.2.1.3) ;
- 3. Standby' gas treatment system (Section 11,5.2.1.8)
- 4. ' North stack ventilation exhaust (Section !
11.5.2.1.9) {
t I
Rev. 21, 06/83 11.5-22 i i
?
L t Y-* gy wpci w- pw --mm.,.-,y-- p g .g- mm.- q. - - - m---,e=,-ee-m ---ee- e, ., -.-,,m ,-w-- ,z .w w-y.-.p,5--m--.=-eg4--- ei----. g- m,ww-%,y gtie-Me '---
l S. South stack ventilation exhaust (S*ction 11.5.2.2.1)
{
- 6. Charcoal offgas systam compartments ventilation exhaust (Section 11.5.2.2.2)
- 7. Charcoal offgas system effluent (Section 11.5.2.2.3)
- 9. Steam seal effluent (Section 11.5.2.2.5) 10.- Radwaste enclosure ventilation exhaust (Section 9
11.5.2.2.6)
- 11. Hot shop ventilation exhaust (Section 11.5.2.2.9).
Sampling stub's are provided on all' major exhaust ducts for the purpose of extracting grab samples as required,
- b. Liquid Effluents: -
1 ~. Residual heat removal service water (Section 11.5.2.1.7) t
- 2. Liquid radwaste discharge (Section 11.5.2.2.10)
- 3. Plant service water (Section 11. 5. 2. 2.11~) .
t 11.5.4 PROCESS MONITORING AND SAMPLING The~ requirements of General, Design Criterion 60 are implemented with respect to the automatic closure of isolation valves in
-gaseous and liquid effluent discharge paths by means of the following monitoring systems:
11.5-23 Rev. 17, 02/83
- a. Main steam line (Section 11.5.2.1.1)
- b. Reactor enclosure ventilation exhaust (Section .
11 '. 5. 2.1. 2 )
c.' Refueling area ventilation exhaust (Section 11.5.2.1.3) j
- d. Residual' heat removal service water (Section 11.5.2.1.7) .
e ,-
5
.e. Liquid radwaste. discharge (Section 11.5.2.2.10).
i The requirements of General Design Criterion 63 are implemented
.with respect to the monitoring.of radiation levels in radioactive ;
fuel and waste storage systems by means of the following ;
monitoring systems:
a.. Area radiation monitor channels 31 and 32 (Section 12.3.4.1) ,
- b. Refueling. area ventilation exhaust-(Section 11.5.2.1.3).
i
- c. Radwaste enclosure ventilation exhaust (Section 11.5.2.2.6)
- d. Liquid radwaste discharge (Section 11.5.2.2.10).
i
. The following liquid process systems are provided with grab '
sample stations for laboratory measurement of radioactive concentrations for satisfying the-requirements of General Design i . Criteria'63-and 64:
- a. Liquid radwaste systems (Section 11.5.2.2.14c, d, e) l l b. Reactor enclosure cooling water (Section 11. 5. 2. 2.14b )
- c. Spent fuel pool treatment system (Section 11.5.2.2.14a.
'Rev. 29,~ - C 2/8 4 11.5-24 .
I
I i
t l
- 11.5.5 POST-ACCIDENT SAMPLING SYSTEMS t are designed to obtain s The-post-accidert sampling systems IPASS) ;
representative .iquid and gas grab samples from the primary :
I coolant system and from within the primary and secondary containments for radiological
~
and chemical analysis under The grab samples are subsequently accident conditions.
transported to the radwaste enclosure chemistry laboratory and I
counting facility for chemical and radiosotopic analyses, or. l shipped offsite for analysis. f i
The item II.B.3.
PASS is designed to satisfy the requirements of NUREG-07 f and "in-line" instrumentation, is modular for maintenance and contamination control purposes, andThe is compact system can in size to reduce be used to i the amount of shielding required. !
provide samples under all plant conditions, ranging from normal shutdown and power operation to post-accident conditions.
i The PASS piping and instrument diagram is shown in Figure 11.5-2. l The equipment includes isolation and control valves, piping liquid ;
racks, shielded sample stations (gas and liquid),The seismic category, quality i chillers, and control panels.
group classification, and corresponding codes and standards that A ,
apply to the design of the PASS are discussed in Section 3.2. l separate PASS is provided for each unit with common demineralized water, nitrogen and tracer gas support systems. i l
11.5.5.1 System Description i,
I i 11.5.5.1.1 Sample Points I
+
l
- a. Wetwell and Drywell Atmospheres i t
Sample lines are installed to obtain atmosphere Wetwell sa: '
'Drywell samples are taken at El 291 and 242 ft.. sam containment. The sample lines tap into the containment !
sample lines outside atmospheric contrcl system (CACS) the primary containment and outboard of the secondContainment r
containment will be representative isolation valve. of conditions throughout the ;
primary containment because the containment is n !
Rev. 17, 02/83 f 11.5-25
- l w aw,+ ,*-e ,=
3 , - . , , , . - ,
.-,,y. ,,,,-vi er--w -, ret-y-, -
p.,w-,-a.,.-mg --.pu-----e,.-wm-wi-ww.-,wwi----- r ., e .c o.e w---,,i-w-,----,-e-ve-,,,rw-+-a- ++
b Q
- b. Secondary Containment Atmosphere ;
I A sample line is provided to allow sampling of secondary f containment atmosphere'to aid in determining post- ;
accident accessibility of the reactor enclosure. !
Samples are taken in the vicinity of access doors 191 i (Unit 1).and 287 (Unit 2) on El. 217 ft. ;
t c.. Reactor Coolant and Suppression Pool t When the reactor is pressurized, reactor coolant samples are obtained from a tap off the jet pump gressure i instrument system. The sample point is en a noncalibrated jet pump instrument line outside the :
primary containment and downstream of the excess flow !
l check valve. This sample point location is preferred !
- over.the normal reactor sample points on the reactor ,
water cleanup system inlet line and recirculation line !
l because the reactor cleanup system is expected to remain !
. isolated under accident conditions, and it is possible f l
that the recirculation line containing the sample line may be isolated. The jet pump pressure tap is in a ;
location protected from damage.and debris. This sample
-point provides representative. samples of reactor coolant under various reactor conditions:
j- l l I
e Normal operation /small pipe break: Reactor water l level can be maintained.at or near normal water !
l
. level. With a nearly normal water level, or at ;
least water in the upper plenum, natural '
!- circulation will occur with a large loop from the downcomer to the shroud region via the jet pumps. ,
With thermal conditions pumping water up through >
the core and back down past the tap from which the ,
PASS sample is taken, a representative relationship !
will exist which will allow the results of the sample to be related to the condition of the core. ;
- Large Pipe Break: A large pipe break, such as a recirculation pump suction line break, may occur l wherein the water level may be centrolled only by ;
the height of the jet pumps and the ability to add 3 o up water to the vessel. The sample taps are located l sufficiently low to permit sampling at a reactor !
water level even belcw the lower core support plate. As reactor pressure decays, low pressure L -coolant injection (LPCI) is initiated into the core ,
- Rev. 20, 05/83 11.5-26
~
region. This water volume supplies more coolant j than is boiled off by the decay heat. This excess water will flow down past the core, up through the !
jet pumps, and out through the postulated break, !
assuring a representative sample at the sample {;
point.
To ensure a representative liquid sample from the jet pumps at low (<1%) power conditions for small break or ,
non-break events, the reactor water level will be raised l to the level of the moisture separator when this action !
is not inconsistent with station emergency procedures, j This will fully flood the separators and will provide a !
thermally-induced recirculation flow path for mixing. [
Samples will be taken from the reactor via the jet pump !
pressure instrument lines as long as possible. This allows.a more direct and therefore faster response to !
core conditions.- Upon decay or loss of reactor !
pressure, the jet pump sample point is lost, and the RHR !
loops sample ~ points must be employed for sampling. l Reactor coolant and/or suppression pool samples may be t taken from the RHR sample lines, depending on the mode i of RHR operation. .These sample lines tap off downstream of the second system-isolation valve in the RHR system i i
sample lines at the discharge of each RHR heat exchanger. ;
e LPCI: Suppression pool water is injected into the i core, flows up through the jet pumps, and back to i the suppression pool.via the postulated break. The ;
i system will tue operated for an estimated 30 ' minutes minimum prior to sampling of the suppression pool l water to ensure that a representative sample is ;
obtained at the sample taps.
- Shutdown Cooling: -The RHR system, aligned in the [
shutdown cooling mode, provides cooling and !
circulation of reactor coolant through the core, i resulting in a representative sample at the RHR li sample taps.
Suppression Pool Cooling: The RHR system, aligned l in the suppression pool cooling mode, provides l cooling and circulation of the suppression pool !
water. The system will be operated for an i estimated 30 minutes minimum prior to sampling of l the suppression pool water to ensure that a f representative sample is obtained at the RHR sample
" taps. !
I 11.5-27 Rev. 20, 05/83 l e
-- , , , , . , . . , . _ , - . - , . - . - - . _ . _ - _ . . - . . _ - - . . . , , _ - , , . - . . - - . - - ~
i 111.5.5.1.2~ Isolation Valves and Sample Lines l Containment isolation for the drywell and wetwell gas sample I
-lines is provided by the existing CACS sample line isolation valves. Jet pump instrument line containment isolation is :
provided by the existing restricting orifice and excess flow .
check valve upstream of the sample tap. Containment isolation !
for.the RHR sample lines is provided by the existing RHR sample !
line automatic isolation valves. All automatic isolation valves -
can be overridden from the control room. PASS remote operated sample line valves are' controlled from local panels located adjacent to the-sample stations.
The gas sample' lines _are heat traced to prevent precipitation of moisture and the resultant loss of iodine in the sample lines.
Sample line routings are as direct and short as practical. ,
Recirculation flow rates in the liquid sample lines are maintained in the turbulent flow regime. -
11.5.5.1.3 Piping Rack l ,
t The piping rack, which is1to be installed within the reactor i enclosure, includes sample. coolers and control valves that ,
determine the liquid sample flow path to the sample station. The l rack provides ,a flow path to recirculate liquid samples, ,
bypassing the sample stations, untti a representative sample ;
condition is obtained. The cooling water is supplied by the- ;
reactor enclosure cooling water system. l d
11.5.5.1.4 Sample Station and Control Panels- l 1
l l .7;te sample station consists of a floor stand, frame, and sample :
l epclosures, and is mounted flush against the outside of the l secondary containment wall. Included within the sample station are' equipment trays that contain modularized liquid and gas r sImplers. The liquid sample portion of the sample station is !
shielded with 6 inches of lead brick, whereas the gas sampler has '
.i inches of lead shielding. The various sample and return lines enter the sample station enclosure through the back by way of a penetration through the reactor enclosure wall. Control
~ instrumentation is installed in two control panels mounted side ;
by side. One of these panels contains the conductivity and ,
radiation' level readouts. The other control panel contains the flow, pressure, and temperature indicators, and various valve !
controls and switches. A graphic display is provided directly !
-Rev. 17, 02/83 11.5-28 j P
- , , ~ ~ - _ _ . - - _ _ _ _ . _ _ _ , , , . _ _ , . . , _ _ . , .. _ _._.._.____-_ _ _._.
l i
i LGS FSAR f below the main control panel which shows the status of the pumps l and valves. j 11.5.5.1.4.1 Gas Sampler The gas sample system is designed to operate at pressures ranging from sub-atmospheric to the design pressures of the primary containment one hour after a LOCA. The gas samples may be passed j through a particulate filter and silver zeolite cartridge for determination of particulate activity and total iodine activity by subsequent spectroscopic analysis. A radiation monitor is :
mounted close to the filter tray to measure the activity buildup ,
on the cartridges. Alternatively, the sample flow bypasses the iodine sampler, is chilled to remove moisture, and a 10 milliliter grab sample can be taken for determination of gaseous activity and gas composition by gas chromatography. The r gas is collected in an evacuated vial using hypodermic needles. ;
When purging the drywell and wetwell gas sample lines to obtain a t representative sample,.the flow is returned to the wetwell. I During purging of the secondary containment line and'when !
flushing the sample panel lines with nitrogen, flow is returned l to secondary containment. The sample station design allows for sample gas or nitrogen flushing of the entire sample panel line !
downstream of the four-position selector valve. This capability l will minimize cross-contamination between the various samples. ,
.11.5.5.1.4.2 Liquid Sampler I
-The liquid sample system is designed to operate at pressures from
- 0 to 1150 psi.. The design recirculation flow rate of 1 gpm is i sufficient to maintain turbulent flow in the sample line and ;
serves to minimize cross-contamination between samples. The recirculation flow is returned to the suppression pool. The l liquid sampling system is designed to allow demineralized water !
flushing of the system lines from a point in the piping station 3 through the sampling needles, j e Diluted Liquid Sample l [
f All liquid samples are taken into'15 milliliter septum l bottles mounted on sampling needles. In the sampling :
lineup, the sample flows through a conductivity cell !
(0.1 to 1000 micromhos/cm) and through a ball valve l bored to 0.10 milliliter volume. After flow through the ;
sample panel is established, the ball valve is rotated 90 degrees, and a syringe is used to flush the sample [
11.5-29 Rev. 18, 03/83 ;
j
. .- -~ .- - . ,.- - -.-. - ...-.- __ __ __________-___ - .
2 i
.and a measured volume of diluent-(generally
.10 milliliters).through the valve and into the sample bottle. This provides an initial dilution of up-to :
100:1. The. sample bottle is contained in a shielded 1 cask'and remotely positioned on the sample needles !
through an opening in the bottom of the sample i enclosure.
e Non-Diluted Liquid and Dissolved Gas Samples i
Alternatively, the sample can be diverted through a 70 milliliter holdup. cylinder to obtain depressurized !
samples of primary coolant gas and liquid phases. A coolant sample is. circulated through a holdup cylinder, ,
the cylinder is then isolated and the contents circulated through a gas loop containing a measured amount of inert krypton. The gases are vented to an -
evacuated gas: collection chamber, and a fraction of the gas is expanded into a sample vial for analysis by gas chromatography. The concentration of krypton in the sample is used to calculate-the fraction of the
. dissolved gares recovered. The krypton also serves as a stripping agent at low gas concentrations. Ten milliliter aliquets of degassed liquid can then be taken
-for offsite (or onsite depending on activity level) analyses which require a relatively large undiluted
-sample. This sample is obtained remotely using the large volume cask and cask positioner through needles on the underside of the sample station enclosure.
11.5.5.1.4.3 Sample Station Ventilation The sample station enclosure will be vented to a Zone V room in the secondary containment. Ventilation is facilitated by-differential pressure between the control structure and reactor enclosure. The ventilation rate required for heat removal and proper sweep velocity during operation is about 40 scfm. A pressure gauge is attached to the sample station enclosure to
. monitor the pressure differential between the enclosure and the general' sampling area in the control structure. The pressure differential will assure the operator that airborne activity in the sample enclosure will be swept into secondary containment.
Rev. 17, 02/83 11.5-30'
i
.s i
11.5.5.1.4.4 Sample Station Sump ;
l!
The sample station is provided with a bottom sump to collect l liquid leakage. This sump can be isolated, pressurized, and discharged into the sample station liquid return line to the !
suppression pool.
, -11.5.5.1.4.5 Sample Handling Tools and Transport Containers Appropriate sample handling tools and transporting casks are i used. . Gas vials are installed and removed by use of a vial
. positioner through the front of the gas sampler. The vial is i manually lowered into a shielded cask directly from the l
. positioning tool. This allows the operator to maintain a j distance of about three feet ftpm the unshielded vial. The cask i
provides about 1-1/8 inches of lead shielding. A 1/8-inch j diameter hole is-drilled in the cask so that an aliquot can be !
withdrawn from the vial with a gas syringe without exposing the i analyst to-the-unshielded vial. .
i The particulate and iodine cartridges are removed via a drawer l arrangement.- The quantity of activity accumulated on the !
cartridge is limited by controlling the line flow using a flow !
orifice and by timing the sample duration either manually or by j use of preset timer. In addition, the radioactivity 1pvel is !
monitored.during. sampling using a radiation probe installed i adjacent to the cartridge. These samples will be limited to i activity levels that will not require shielded sample carriers. l The small volume (diluted) liquid sample cask is a cylinder with !
a lead wall thickness of.about 2 inches. The cask weighs l approximately 50 pounds and has.a handle which allows it to be l carried by one person. _
l The l'0 milliliter undiluted sample is taken in a 700 pound lead shielded cask which is transported and positioned by'a four-wheel
- dolly. The sample is shielded by about 5-1/2 inches of lead. A i
. licensed shipping cask for transport of-the undiluted samples to ,
the offsite analysis facility (Section 11.5.5.2.2) has been r procured in conjunction with a group cf other utilities. This ,
cask will be placed in a centrally-located, continuously attended ;
. warehouse facility. i j
11.5.5.1.4.6 Sample Station Power Supply i i
The PASS isolation and control valves, sa.mple station control !
panels, and auxiliary equipment are connected to an instrument ac l distribution panel which is powered from an engineered safeguard 8 system (ESS) bus. Following a loss of offsite power, the ESS bus [
l 11.5-31 Rev. 20, 05/83 i
-...,.l__ , . . . . . . _ . . _ . , . _._._,,,,_ _ _ ____.._ _ _ _ ___, _
is powered from the onsite diesel generators. The reactor enclosure cooling water system, which is needed for the sample coolers, is also powered from the emergency diesel generators ;
following a loss of offsite power.
All electrically operated components associated with the .3 ASS are ;
capable of being supplied with power and operated within '
30 minutes of an accident in which there is core degradation, r assuming loss of offsite power. -
i 11.5.5.2 Description of Samole Preoaration/ Chemistry and o Nuclear Countina Facilities
'After the samples are obtained from the sample station, they will be transported to a sample preparation / chemistry area where they
,w ill be diluted as necessary and appropriate aliquots taken for chemical and radioisotopic analysis. The radioisotopic analysis :
will be done in an area where background radiation can be kept to ;
a minimum. The primary facility for performance of these analyses is the existing radwaste chemistry laboratory and l counting room on El. 217 ft of the radwaste enclosure. In addition to these onsite facilities, which are intended to handle the gas samples and the diluted liquid samples, prior arrangements will be made with an offsite laboratory for supplemental and confirmatory analysisaof samples as required.
11.5.5.2.1 Onsite Facilities The chemistry lab is equipped to handle the gas samples and the r 0.1 ml diluted liquid samples. The maximum. activities of these samples will be 0.47 Ci and 0.29 Ci, respectively, using one-hour decay and the fractional releases of core inventory as discussed in Section 11.5.5.5.
i The laboratory will maintain a dedicated inventory of items such i as lead bricks for shielding, gas syringes, gloves, reagents for ,
analysis, etc, which will be needed in case of an accident. .The ,
laboratory wil'1 be equipped with a gas chromatograph, pH meter, conductivity meter, turbidimeter and other instrumentation needed to perform the required analyses. This equipment, however, may not be dedicated exclusively to post-accident analysis. Supplied air or self-contained breathing masks will be available in the event of high activity levels in the ventilation supply or accidental spills in the laboratory.
Rev. 17, 02/83 11.5-32
i The existing counting facility located adjacent tg the chemistry laboratory is equipped to handle the gamma spectra analyses l required for post-accident samples. The counting room is equipped with two Ge (Li) detectors with 4-inch lead shields ,
connected to a computer based analyzer system. The system has :
automatic peak search and isotope identification capabilities.
' The Ge (Li) detector and shelf assembly in the lead shield can be isolated and the_ capability to purge the volume within the shield l with' compressed gas will be provided. This will help prevent i atmospheric noble gas activity released during an accident from i swamping the detector.
I It is expected that the first set of post-accident samples will i be analyzed in the radwaste enclosure. chemistry lab / counting room facilities approximately two hours after the start of an i accident. At this time, the chemistry lab will be a Zone III !
area and therefore accessible for performing the required l chemical analyses. The lab becomes a Zone II area within !
_20 hours following an accident. The counting room is a Zone II l area _within two hours,.and a Zone I area within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following !
an accident. Shielding for the PASS is further discussed in l
- Section 1.13.2 (Item II.B.2).
The most direct route from the sample'r location to the chemistry ]
lab is through the control structure and Unit 1 turbine enclosure -
to the radwaste enclosure on El. 217 ft. However, during the {
first 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />s-following an accident, high radiation in portions j of this access path require that an alternate route from sampler ;
to' labs be taken. Operators must exit through the north of the !
turbine enclosure, travel around the west end of'the radwaste l building, and enter the lab / counting room area throgh the south ,
of the radwaste building. l l
11.5.5.2.2 Arrangements for Offsite Analyses l V ,
A part of the Limerick approach to post-accident sampling is the ;
establishment of prior arrangements with an offsite laboratory !
for confirmatory and supplemental analyses. j l
7
.11.5.5.3 Sample Collection and Transport Procedures l l
L It is anticipated that the first set of samples will be taken !
within one hour following a LOCA, with samples taken !
approximately every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the remainder of the first !
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Following day one, it is expected that three samples ,
-per day may be taken for the remainder of the first week, with 11.5-33 Rev. 20, 05/83 }
r
W l
I samples.once per day.following_the first week for the duration of ;
the accident. !
i The following-is a conservative time sequence for sampling, transport, and analysis to demonstrate that samples can be
- - obtained.and analyzed witain the specified-3-hour period
- -j l
- a. ' Recirculate sample, install sample vial /or cartridge ;
- - 15 min.
- b. ' Operate sample station - - 15 min. I
- c. Transport sample to lab - - 20 min, j
- d. Analyze sample - - 60 min.' 5 i
Atmosphere sample lines from'the CACS and liquid samples from the
- RHR system are automatically isolated following a containment ;
isolation signal. The sample station operatcr(s) must confirm r with the control room personnel that-the necessary isolation valves are opened. A telephone extension to the control room is :
. provided near the PASS control panel.
Controls for all other PASS sample line valves are located-on or f adjacent to the PASS control panel. All controls for. valves that are a part of the sampler or piping rack are located on the PASS ,
control panel.
Following a series of presampling checks-and procedures, !
including adjustment of the enclosure damper to ensure adequate ventilation, checks of demineralized water and nitrogen supplies, l flushing of system with demineralized water, draining the trap ;
and sump, etc, the system is ready for obtaining the samples. :
I 11.5.5.4 Chemical / Radiochemical Procedures ;
' The PASS provides a means of obtaining primary coolant, i suppression pool, and primary and secondary containment air ;
. . samples for radiochemical and chemical analysis following a major .
reactor accident. Because of'the extremely high radioactivity levels associated with extensive fuel damage, the PASS was !
- developed to provide the capability of obtaining the necessary ,
samples and performing analyses, as required, for immediate plant needs, or as defined by regulatory requirements. Procedures will ,
be established for shipping samples to appropriate facilities to ;
perform detailed analyses on multi-curie level samples. l .
t Rev. 17, 01/33 11.5-34 l i
-- m m , p y *,_
t j
r 11.5.5.4.1 Sample Preparation [ [
t l' All sample bottles, iodine cartridges, etc, will be identified I prior to sampling to eliminate unnecessary exposure resulting i from' handling high level samples. A centralized logging system l
will'be-developed to track sample aliquot identification, dilution factors, sample disposition, etc.
Liquid samples will be taken at the sample station in septum-type !
! bottles and transported to the analysis facility in lead :
containers.
! - Sample aliquets are taken from the septum bottles for analysis or further dilution. Aliquoting and transfer will be performed l.
using shielded containers, or behind a lead brick pile.
E . Calibrated hypodermic syringes will be used for aliquoting the
- higher activity samples. Tongs or other holding / clasping devices will be available for holding the sample bottle during the transfer.and dilutions.to reduce hand and body exposure. Unless
- prohibited by the intended analysis, dilutions will tue done using very dilute (about 0.01N) nitric acid as the diluent to minimize t
sample plateout problems.
l' Primary' coolant samples obtained from the sampling station are L diluted by a factor of 100 (0.1 ml coolant diluted to 10 ml).
p Under severe accident conditions, a calibrated syringe would be used to obtain an aliquot for this sample for further dilutions.
At the maximum expected primary coolant activity level (3 Ci/cc),
a dilution factor of 1 x los would be required foregamma spectroscopy.
Direct counting of the initial 100:1 dilution sample would allow analysis at coolant activity levels down to 1 nCi/cc. In l
l addition, the degassed, undiluted 10 ml sample available from the sample station could be used for analysis of samples in the 10-a to 10-8 nci/cc range. Thus, useful samples may be obtained from the post-accident sampling station for coolant activity levels l
ranging from design basis accident source terms tu well below the maximum level that can be tolerated at the normal reactor sample station.
I Gas samples are taken at the sample station in a 15 milliliter septum bottle. A lead carrier is furnished with a small hole at l
the septum end so that a gas sample can be withdrawn from the L carrier using a hypodermic syringe without having to handle the L bottle.
11.5-35 Rev. 17, 02/83 l
t L.
n ,
LGS FSAR Samples taken from~the gas sample bottle will either be injected into a gas chromatograph for analysis or diluted for gamma
,' spectroscopy. The dilutions will be performed in a manner
> analogous to the liquid samples. Fractional milliliter samples can be transferred to new 15 ml gas bottles without concern for sample leakage-due to pressurization. For larger volume aliquots, a gas syringe will be used to draw a partial vacuum in the bottle prior to sample transfer.
11.5.5.4.2 Chemical Analysis The chosen procedures are not necessarily the most sensitive nor the most accurate. They were chosen primarily on the basis of [
simplicity, stability and availability of reagents, minimum t radiation exposure, and least likelihood to cause major contamination problems. They have been tested for radiation !
sensitivity and are suitable for use at the PASS design basis !
source term of 2.83 CL/gm, and where applicable, with the design ;
basis 0.1 m1 to-10 ml dilution at the sample station. !
, a. Gross activity, gamma spectra analyses will be accurate f within at least a factor of two over a coolant activity l range of 10 nCi/cc to 10 Ci/cc. Additional information i t
is provided in Section 11.5.5.4.1.
- t i
- b. Boron concentrations will be. determined by the carminic t acid analysis method. Concentrations between 50 and [
1100 ppe are of interest for BWR reactivity control in j the event sufficient control rods are not inserted to i shut down the reactor. The use of the carminic acid l method with the 100:1 diluted sample will result in an ;
accuracy of 250 ppm over this range.
None of the expected post-accident chemical constituents (I, Cs+, Ba+a, La+a, Ce**, Cl , B, Li+, NO-8, NH+*, K+)
will interfere with this analysis method. Irradiation tests conducted by GE have demonstrated that the design basis source term activity level causes only a 1-ppm error in the analysis,
- c. The chloride analysis performed onsite will be limited to a scoping analysis using the turbidimetric method.
The sensitivity of this method is such that coolant concentrations must be greater than 10 ppm for detection in the diluted sample. In addition, iodine can be expected to interfere somewhat with the turbidimetric Rev. 17, 02/83 11.5-36
F :.
LG5 FSAR method by the formation of silver iodide. Tests performed by GE have verified.that irradiation has a negligible effect on the accuracy of the analysis.
Offsite provisions for chloride analysis will be accurate 210 percent over the range 0.5 to 20 ppm and
' O.05 ppm below concentrations of 0.5 ppm.
o
- d. A combination electrode will be used to measure the pH of coolant samples. Testing performed by GE has verified that expected levels of irradiation result in a
-shift of less than 0.3 pH units,
- e. The post-accident sample station is equipped with a 0.1 cm-1 conductivity cell. The conductivity meter has a_ linear scale with a six-position ~ range selector switch to give conductivity ranges of 0-3, 0-10, 0-30, 0-100, 0-300, and 0-1000 micromho/cm when using the 0.1 cm-1 cell. This conductivity measurement. system will be used to-determine the primary coolant or suppression pool conductivity. During normal operation the BWR technical L
specifications require maintaining the primary coolant below 0.1 micromho/cm, and conductivity measurements are the primary method of coolant chemical control.
~
Conductivity measurements are, of course, non-specific,
~
but they serve the important function of indicating changes in chemical concentrations and conditions.
Perhaps even more important, in the case of the BWR primary coolant, the conductivity measurements can establish upper limits of possible chemical concentrations and can eliminate the need for additional analyses.
The conductivity measurement can also be used to bound the possible range of pH values.
11.5.5.4.3 Radiochemical Analysis--Gamma Ray Spectroscopy After the samples have been brought to the chemistry laboratory and appropriately diluted, they can be carried without shielding
- to the counting room which is adjacent to the chemistry i _ laboratory. The appropriate dilution factors will be somewhat dependent on the detector and shelf arrangements available. A
[ prior determination of the maximum desirable dose rates for the 11.5-37 Rev. 20, 05/83
7_
various shelf configuration will be made.to minimize this I
, . problem. The_present high resolution, high efficiency Ge(Li) ('
detectors, coupled with.the multichannel analyzers, and computer '
data' reduction in the onsite counting room will handle the analysis of these samples within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from.the time a decision [
is made to sample. .
-The gas samples will be counted in the PASS gas sample vials, and !
, the liquid samples will be counted in the standard sample bottles used during normal operation because calibration curves for these geometries will be available and regularly updated. Calibration l curves will,also be available for the particulate filter and I todine cartridge geometries. In general, the counting of the i post-accident sample will follow the normal counting room !
procedures. The peak search and identification library will contain the principal gamma rays of the following isotopes in !
addition to the standard activated corrosion products: ~ j i
- a. Noble _ gases:
Kr-85, Kr-85m, Kr-87, Kr-88, ,
Xe-131m, Xe-133, Xe-133m, Xe-135 l
?
I
.b. Iodines: I-131, I-132, I-133, I-135 i
- d. Others:. Ba/La-140, Ce-141, Ce-144, Ru-106, Te-129, Te-129a,-Te-131, Te-131m, Np-239 If the levels of noble gases in the ambient atmosphere i surrounding the detector are high enough to cause significant !
interference or to overload the detector, a compressed air or nitrogen purge of the detector shield volume will be maintained. j I The onsite radiological and chemical laboratory-facilities are equipped with gamma spectral analysis equipment to quantify the !
radionuclides present in gas and liquid samples. Shielding is i provided for the radiation detectors to minimize the effect of background radiation. Initial dilutions are performed in the .
process of taking liquid samples at the sample stations. Any ;
- additional dilutions required will be performed in the laboratory ,
fume hood behind a lead brick pile. j i
Rev. 20, 05/83 11.5-38 ;
f
0 l
t i
LGS FSAR I 11.5.5.4.4 Gas Analysis-Gas Chromatography ;
I A gas chromatograph will be used to measure hydrogen and oxygen I j
concentrations in containment atmosphere and dissolved gas i samples. !
l
- a. Dissolved hydrogen concentrations - - An ac:uracy of +10 i percent can be expected Below over the range of concentrations 50 cc/Kg, the accuracy I from 50 to 2000 cc/Kg.
will be +0.05 cc/Kg. Gas chromatography has been l successfully demonstrated for the determination of :
i
. hydrogen in THI-2 post-accident gas samples. )
.b. Dissolved oxygen concentrations - - Dissolved ozygen i '
will be measured indirectly using the residual hydrogen method of analysis. Using this method, dissolved oxygen concentration is verified to be less than 0.1 ppm by I measurement of positive hydrogen residuals of greater !
than 10 cc/Kg. '
r l
11.5.5.~4.5 ~ Determination of Extent of Core Damage !
h l
A generic procedure to assess the extent of core damage based on !
radionuclide concentrations and other parameters has been !
prepared by the BWR Owners Group and transmitted to-the NRC1983. by '
letter from Mr. T. J. Dente to D. Eisenhut dated June 17, A Limerick procedure based on this methodology will be prepared
- prior to fuel load, i 11.5.5.4.6 Storage and Disposal of Sample l jl Short-term sample storage areas will be provided in the chemistry t laboratory'and counting room facilities. An area for long-term storage of the samples will be designated at a later date.. Low level wastes generated by the. chemistry procedures will be [
flushed to radwaste. Ultimate procedures for disposal of the [
samples will be determined later; however, after a sufficiently ;
-long decay period, the activity levels will be significantly !
. reduced. This will ease exposure problems during disposal.
l Rev. 22, 07/83 11.5-39 f r
- i L
11.5.5.4.7 . System Testing and Operator Training Equipment used for post-accident sampling and analyses will be calibrated or tested approximately every six months. At least i five members of the onsite organization will participate in !
sampling operations within any 6-month period. ,
t i
11.5.5.5 Dose Rate Analysis !
f l
The post-LOCA core inventory of fission products was calculated l assuming a three-year irradiation, 100 percent availability, and ,
reactor operation at 102 percent of rated power. Fractional releases of fission products from the fuel to the reactor water, !
suppression pool, and containment atmosphere were based on ;
Regulatory Guide 1.3 and 1.7 assumptions. The resulting source !
terms were used in the design of PASS shielding and in determining doses to operators.
The sampling and analysis provisions at Limerick have been !
designed such that it will be possible to obtain and analyze a !
sample following an accident without exceeding the criteria of GDC 19. Timo sequences and calculated dose-rates to verify ,
complianet for a sample taken one hour post-LOCA are given in !
Table 11.5-6.
t t
i i
Rev. 22, 07/83 11.5-40 l
i' !
i
(
f 12.3.4 AREA RADIATION AND. AIRBORNE RADIOACTIVITY MONITORING t INSTRUMENTATION l l
12.3.4.1 Area Radiation Monitors j 1
Fixed area radiation monitors are mounted throughout the plant at !
selected locations. Each location contains a gamma sensitive l detector, local indicator,,and local alarm (visual and audio). j Indicators, alarms, and recorders are located in the control +
room, except for the four. local area monitors (see l Section: 12.3.4.1.3). Alarm setpoints are adjusted from the l auxiliary equipment room. j The location and range for each monitor are given in l Table 12.3-7. Ranges specified are consistent with the potential- !
dose: rates for a given area. Selection of location is based on ~
the equipment.in the area and the need for personnel access. In some locations,.the detector is mounted immediatelp inside a :
cavity while associated local indicators and alarms are mounted i just-outside the access door. Personnel are thereby alerted to [
unusual radiation levels within the cavity before initiating access. Experience at Peach Bottom Units 2 and 3 is used in I determining locations.
Area monitors located'at fuel. storage areas comply with the !
requirements of 10 CFR, Part 70.24. These and area monitors in L
'the radwaste enclosure comply with requirements of 10 CFR, i Part 50,, Appendix A General Design Criterion 63. [
$ i s.
The area radiation monitors, with associated alarms, indicators }
and recorders in the main control room provide the capability to I alert supervisors and personnel in that area of abnormally high [
and unexpected radiation levels. Subsequent notification to t appropriate. plant personnel could avoid inadvertent unnecessary j exposure. Unexpected increases in radiation levels may be due to changes in operations, deposition of crud, or transport of radioactive material (e.g., sources) within the plant. Recorded ,
< area radiation monitor readings serve to define trends that may- !
result from buildup, spills, or contamination of a process fluid. l Qualifications and training of health physics and chemistry !
personnel will follow ANSI-13.1-1969 guidance. l t
(". - 12. 3. 4.1.' 1 Design Bases f The' purpose of the area radiation monitoring system is to provide ;
.. personnel protection in accordance.with the guidelines of 10 CFR, j Part 20, 10 CFR, Part 50, 10 CFR, Part 70, and Regulatory !
Guide 8.8. The area radiation monitoring system has no function t related to the safe shutdown of the plant or to the quantitative ,
E . monitoring of the release of radioactive material to the ;
f I
f 12.3-21 Rev. 6, 06/82 i
I
~
I i
environment. Consistent with this purpose, the area radiction
~ monitoring system is designed to provide the following functions: l.
-a. Warn of excessive gamma radiat' ion levels in areas where !
nuclear fuel is stored or handled, in accordance with !
.- '10 CFR, Part 70, Section 70.24 (a)(1) t
- b. ' Provide' operating personnel with a record and continuous l indication in the control ~ room of gamma radiation levels {
at selected locations within the various plant :
structures, in accordance with Regulatory Guide 8.8 l
-c. Supplement other systems in detecting abnormal [
migrations of radioactive macerial in or from process ;
streams i
-d. Provide local alarms at strategic locations throughout ;
the giant, in accordance with Regulatory Guide 8.8,
! where a substantial. change in radiation levels or loss ;
i of sensing cap' ability might be of immediate danger to ]
personnel in the area
- e._ Contribute supervisory information to the control room !
so that correct decisions can be made with respect to i deployment of personnel in the event of a radiation accident. ,
- f. ' Assist in the detection of unauthorized or inadvertent !
movement of. radioactive material in the plant including l
-the radwaste enclosure, j l
- g. Furnish information for conducting radiation surveys. l
{
'- Ihe above functions are performed under the following desinn j i
conditions:
.a .
Environmental parameters shown in Table 12.3-8 are i applicable to the design of the area radiation !
monitoring equipment except for the monitors located ;
I inside primary containment, where the detectors are !
designed for a normal operating temperature range of ;
64*F-151oF (up'to 340*F for accident conditions) and_a
> pressure range of atmospheric to 2 psig. [
t
- b. Noise from any source in the operating environment should not disturb the meter indication by more than 22% i of equivalent full scale.
- c. _The detector-indicator and trip unit should be "esponsive to gamma radiation over an energy range of [
0.08 MeV to 7 MeV. The energy dependence should not f exceed 220% of the indicated scale reading for a dose ;
D i Rev. 18, 03/83 12.3-22 l t
f I - - - . - - - - -
I
rate of appecximately 50 mrem /hr resulting from 0.1 MeV l to 3 MeV gammas. ,
f
- d. At the control room, the reading should be reproducible within 210% of the local indicated point, and drift
.should not exceed 20.2% of equivalent linear full scale for a 24-hour period or 22% for a 30-day period. j
- e. .The range of the monitors is shown in Table 12.3-7. The I ranges selected ensure readout of both the highest and i
I L lowest anticipated radiation levels, in accordance with j l Regulatory Guide 8.8.
, 'f. Operate without effective change in performance when the i ac supply voltage changes over a range of 210 percent ;
L from nominal value or has a frequency variation of- 5 ,
percent from nominal value. :
- g. Operate so that interruption or failure of the ac power supply will result in actuation of the trip circuit to l produce an alarm, in accordance with Regulatory 1 Guide 8.8.
i 12.3.4.1.2 Design Details The area radiation monitoring system is shown in the typical functional block. diagram.of' Figure 12.3-27. Each channel consists of a combined sensor / converter unit, a local auxiliary l unit (readout with visual and audible alarm), a combined !
indicator / trip unit,;a shared power supply,-and a shared i multipoint recorder.. The locations of the radiation sensors are !
indicated.in Table 12.-3-7. Further design details of the area radiation monitoring. system are as follows: l
- a. Each indicator and trip unit is provided with one upscale trip continuously adjustable over the entire j scale and one downscale trip adjustable over the lower i decade... Provision is made to permit the upscale trip to !
tue set and checked for accuracy with respect to the
~
(
indicator.
s b.- Detectors are wall-mounted and suitable for operation in l the anticipated plant environment with no additional ;
I 4 ;
y protection.
! c. Panel locations of the remote equipment are shown in i Table 12.3-9.
The following features are provided for components i
d.
located in the auxiliary equipment roem:
- 1. Radiation level indicator (meter) 12.3-23 Rev. 19, 04/83 e
i
- 2. High-radiation _ alarm. light (amber)
- 3. Downscale alarm light (white)
.4. Alarm reset (push _ button) 5; Meter.zero adjust (on.the amplifier)
~6. Alarm level adjust
- 7. . Trip' check pushbutton'
- 8. Power supply switch and " power-on" light (clear lens)
- 9. . Indicators to show power. supply voltages 1 10. . Annunciator outputs
- 11. . Recorder outputs
- e. .The radiction monitors are calibrated at regular time
< intervals in accordance with station procedures.
' Calibration-methods are covered in detail in the equipment procedures-manual.
- f. The,following annunciators are located in the control room to alert the operator:
- 1. ~ Reactor enclosure area, high-radiation (Units 1 and 2)
- 2. Refueling floor area, high-radiation (Units 1 and 2)
- 3. LTurbine enclosure. area,.high-radiation (Units 1 and 2)
- 4. Turbine. enclosure common area, high-radiation
- 5. Radwaste enclosure area, high-radiation (common)
- 6. _ Refueling hoistway common area, high-radiation
- 7. : Hot maintenance shop, high-radiation (common)
- 8. ' Unitized area, low-radiation (trouble) 9 .- Common area, low-radiation-(trouble) 12.3.4.1.3 Local Area Monitors
- ' In addition to the area radiation monitors described above, five local acea monitors-are provided, located on each of the two
. refueling bridges, the_two turbine enclosure crane cabs, and the HP&C source storage and calibration room. The essential
' differences between these monitors and those area monitors described above are as follows:
- a. No outputs to the control ~ room are provided.
b.. Alarms are local only.
- c. No recorders are provided.
- d. Local power (battery) packs are provided.in the event of external power cutoff, except for the local ARM in the HP&C. source _ storage and calibration room.
A portable alarming rate meter will be used in the drywell to warn plant personnel if temporary shielding is removed during ifuel transfer. operations. The meter-will have local and remote alarms.to ensure adequate personnel warning. Administrative controls will be used to prevent access to the upper'drywell and unnecessary high exposures to personnel.
'Rev. 27, 12/83 12.3-24
i l
r LGS FSAR r Also, accessible portions of the spent fuel transfer canal will ,
be clearly marked with signs stating that potentially lethal l radiation fields are possible during fuel transfer. !
i Post Accident Area Radiation Monitors l 12.3.4.1.4 ;
I In the event of.an accident, access will be required to certain locations-outside secondary containment to perform sampling, ;
analysis, and monitoring tasks. The identified areas include the -l
-counting room and chemistry laboratory, the main control room, the turbine eraclosure near the control room exit, the north stack :
instrument room,.the post-accident sampling station, and the operational support center. The area monitors associated with i
these locations have been designated as post-accident area i radiation monitors. In addition to having dose rates available f
through'the shared multipoint recorder in the control room, instantaneous and stored data are available on demand for display 4
i and trending in the control. room, Technical Support Center, andThe '
Emergency Operations Facility via the ERFDS computer system. i ranges of these area monitors are specified in Table 12.3-7 and incorporate the highest anticipated post-accident dose rates for ;
these locations. The design details and operating conditions are as described in Sections 12.3.4.1 and 12.3.4.2. . Compliance with i
.i Regulatory Guide:1.97-(Revision 2) is discussed in Section 7.5.
12.3.4.2 Airborne Monitorino i
Airborne radioactivity monitoring is accomplished by use of the following:
f
'1. Monitoring ventilation ducts in key locations throughout ;
the plant.
- 2. Continuous air monitors (CAMS) which are portable, cart mounted, and monitor particulate and iodine activity.
I l
i t
12.3-24a Rev. 27, 12/83 !
m ,,,..,-.,.,_-,..-,,,r.,_
.----,,-r ,.,,,-,_n,,,,w,, ___,._.,,_n-s_ . . _ _ .,,n_------,--,c-r---
~
{-
LGS FSAR THIS PAGE IS INTENTIONALLY BLANK
(
Rev. 27, 12/83 12.3-24b
e -
- 3. High- and low-volume portable samplers capable of attaching filters and charcoal cartridges for particulate and iodine monitoring.
The hentilation system monitors are located at positions which .
provide representative air concentrations and a rapid indication of abnormal conditions. Those systems which require HEPA filtration have monitors upstream of the filters. Both the inline Geiger-Muller tube and beta scintillator, and offline particulate, iodine, and noble gas monitoring configurations are utilized.- Readout and annunciation are provided in the main control room and/or radwaste control roor Emergency de power is provided in.the event of loss of offsite power. The detectors are calibrated routinely.and after any maintenance work is performed on the detector.
Continuous air monitors (CAMS) are located in freely necessible areas where airborne radioactivity is most likely to exist.
These CAMS are mobile and can be moved from area to area as deemed necessary by plant conditions or maintenance operations.
CAMS-incorporate either fixed or movable filters-for the collection of particulate activity, which.is monitored directly by a detector. Readout is recorded in CPM. The filters can be removed for further analysis using counting room instrumentation.
LAudible'and visual alarms indicate when set point levels have
- been exceeded. The detectors are calibrated routinely and after any maintenance work is performed on the detector.
The CAM's primary function is to indicate trends and sudden 4
changes in airborne activity. - Typical locations are solid waste handling areas, spent fuel pool areas, and the reactor operating floor and turbine building. The monitoring system is capable of detecting ten MPC-hours of particulate and iodine radioactivity from compartments which have a possibility of containing airborne
' radioactivity,1and which normally may be occupied by personnel.
A flexible hose can be attached to the monitor intake and inserted into a cavity or work area to detect the presence of localized airborne activity. Conformance to Regulatory Guide 8.2 is discussed in Section 12.5.1. The guidance of Regulatory Gutde 8.25 will'be followed.
Ventilation monitors and CAMS are used as trending devices and will indicate areas and times needing special samples taken.
Alarm set' points are set at low levels to ensure close respiratory controls. CAMS, however, cannot account for inversion conditions or properly identify isotopic content of the air. When a set point is reached, grab air samples are taken and
. analyzed in the counting lab. NThe MPC hours are isotopically calculated by the computer program or are done manually.
Appropriate actions can then be taken based on accurate data, s
12.3-25 Rev. 21, 06/83
+
LGS FSAR Potentially airborne accessible. walk areas.are' air sampled at regular-intervals. Areas not routinely sampled require an RWP for entry. Air samples are.taken on initial entry. The survey / sampling frequency is designed for that area and controlled by the RWP.
Low- and.high-volume samples with filter paper or charcoal cartridges are used. These.are described in more detail in Section 12.5.3.1.3.
t' Airborne monitoring provides the information necessary to
. determine stay times in given areas and applicable respiratory equipment. The information is also of valve in identifying process system leakage. Such monitoring is conducted in accordance with the guidelines of Regulatory Guide 1.21.
Qualification and training of health physics and chemistry
. personnel.will follow ANSI-13.1-1969 guidance.
12.
3.5 REFERENCES
12.3-1 J.J. Martin and P.H. Blichert-Toft, " Radioactive Atoms, Auger Electrons, and X-Ray Data," Nuclear Data Tables, Academic Press, (October 1970).
12.3-2 J.J. Martin, Radioacti've Atoms Succlement 1, ORNL 4923, (August 1973).
12.3-3 'W.W. Bowr.an and K.W. MacMurdo, " Radioactive Decays ordered by Energy and Nuclide," Atomic Data and Nuclear Data _ Tables, Academic Press, (February 1970) 12.3-4 M.E. Meek and R.S. Gilbert, " Summary of X-Ray and Gamma Ray Energy and Intensity Data," NEM-12037 (January 1970).
12.3-5 C.M. Lederer, et al., Table of Isotones, Lawrence Radiation Laboratory, Universtty of. California (March 1968).
12.3-6 D.S. Duncan and A.B. Spear, Grace 1 - An IBM 704-709 Procram Desion for Co9mutina C===a Ray Attenuation and Heatino in Reactor f,telds, Atomics International (June 1959).
12.3-7 D.S..Duncan and A.B. Spear, Grace 2 - An IBM 709 Procram for Comnuthne G===a Rav Attentuation and Heatino in cylindrica;, and Scherica; Geometries, Atomtes International (Novemoer 959).
12.3-8 W.W. Engle, Jr., "A User's Manual for ANISN: A one Dimensional Discrete ordinates Transport Code with Rev. 21, 06/83 12.3-26
. l i
LGS TSAR
[
avatlable for emergency use include air sampling equipment with !
particulate-filters and silver zeolite cartridges, portable ion ,
chambers, alpha scintillation probes, energy compensated beta / gamma GM probes (for low energy photons), and portable l beta / gamma geiger counters. Ranges of the instruments and [
compliance with Regulatory Guide 1.97 are discussed in Section ,
7.5. ,
At least some of the icn chambers and G-M probes have movable I beta shields to enable disttnguishing between beta and gamma ;
radtation. The ion cha=cers are the prime devices for dose rate !'
determinations and have beta factors specified as appropetate for the type of meter and its use.
Geiger-Mueller probes may be used for dose rate determinations, j
'Their application for this purpose is of value when the ability of ;
an ton chamoer to. respond reliably is impaired by high humidity, i high or low temperature, or very low dose rates. Certain !
instrument designs with G-M probes have extendable arms. This i j
feature allows the user to remain in a relatively lower radiation '
area while measuring high dose rates at the point of interest.
This application is a good example of an effective ALARA policy. l The ranges and numbers of instruments are adequate for their <
intended use whether routine or emergency.
12.5.2.2.4 Personnel Desimeters l Personnel monitoring is provided by the use of such devices as thermoluminescent dosimeters (TLDs), direct reading pocket i dosimeters, or calculations from area survey data and exposure r times. Personnel monitoring is provided per 10 CFR, Part 20.202, i The form of personnel monitoring depends on the type of radiation l and the expected radiation level. Types of dosimetry devices -
change as the state of the art improves, j l
Devices are available and used appropriately for determining whole body or equivalent exposure, extesmities exposure and skin [
exposure. Health Physics practices include the use of multiple badges in addition to the whole body badge when other parts of the body are exposed to het spots greater than the general area dose j rate, t
Self-reader dostmeters are issued primarily for the benefit of ,
individuals. They may read the dostmeter at any time during i their work in order to be aware of the exposure they are !
experiencing. These dostmeters are calibrated and drift checked [
on a routine basis per procedures. The use of self-reader i dosimeters will follow the guidance of Regulatory Guide 8.4. .
Rev. 18, 03/83 12.5-6 [
r
,,,-n- m- .s ,, ..n,,--w--, , m m, nn ,,n , mnm,. ...,--,e
^
.c ,
r Dosimetry devices used for determining the official exposure doses are subject to extensive quality control programs. This is i true whether the processing is by a contractor or onsite. r Currently, thermoluminescence type dosimeters (TLDs) are used for l penetrating (gamma, neutron) and non-penetrating (beta) radiation. WLth regard to. Regulatory Guide 8.14 (Rev 1), neutron exposure is ny mally insignificant. Whenever neutron exposure is of concern, the current technique is to use neutron sQrvey instruments.tb determine a. rem dose rate, then to multiply by the l ,
Y f
l i
i l
r l
l s
i i
)
I l
t t
i i
f I
(
12.5-6a Rev. 18, 03/83 i
i
u t stay time. This technique is preferred because neutron dosimetry
~
is.not as well developed as beta-gamma dosimetry. Effective ;
- state-of-the-art improvements for neutron dosimetry will be reviewed for possible future. incorporation into the dosimetry ;
program.s ;
F The persondel dosimetry program is conducted by qualified !
personne1'under the direction of the Senior Health-Physicist. l Because.TLDs are used for personnel ~ monitoring, Regulatory i Guide 8.3 (February.-2, 1973) is not applied. The applicant I utilizes the' practices described in ANSI N13.6 - 1966 (R 1972),
Practice-for Occupational Radiation Exposure Record Systems e
(invoked by.R'gulatory' Guide 8.7 (May, 1973)), as applicable to TLD monitoring systems. [
'12.5.2.2.5 Miscellaneous Instruments and Equipment i t
Other> monitoring devices are available and may be assigned to l personnel.or located in work areas for the purpose of alerting i personnel to changes in radiation condition. ;
i
-Radiat' ion. monitors 1that " chirp" at a rate proportional to gamma
~
~
dose rate or contain an audible alarming feature are assigned to personnel as deemed necessary.- These devices -are of particular ;
.value.when worn by individuals whose duties routinely involve !
entering various areas:of the plant. Changing operating i conditions or the progression of the individual into the area .i could cause sudden changes in. radiation levels. These devices i alarm at preset values to~ alert personnel of-the changed. ,
L
. conditions as the individual walks-into the area.-
Alarming rate meters may be positioned in a given work area. i
..They serve theLpurpose of alerting individuals working in that i
' area of-increases in radiation level above a pre-set value. !
Frisker or portal monitors having sensitive G-M probes are -r' positioned at'or near the various change areas and plant exit points. The purpose of these devices-is to control the spread of
' contamination. Other devices that prove to be of equal or better r sensitivity will be considered for use instead of portal monitors :
. and G-M probes. fFriskers can be used within the plant for j personnel to monitor themselves at any time, especially when ;
leaving a controlled area. Portal monitors with friskers as t backup are used primarily within the guard house as personnel i
- leave the restricted area. j Continuous airimonitors (CAMS) are used to monitor airborne l concentration at' specific' work-locations. These CAMS provide the [
-means to observe trends or sudden changes in the airborne i F 4 concentrations. They are not intended for quantitative analysis. I
-The fixed filter type can be used as a low volume grab air sample ;
r G !
)
12.5-7 Rev. 17, 02/83 [
[
s 5
s !
,1- . . . - . _ . .
O t
)
1 j
I
~
-in that the filter medium can be removed and analyzed in more !
- detail in the counting room.
Fixed area radiation monitors are mounted in selected locations.
Each contains a gamma sensitive detectcc, local indicator and i local alarm, as well as indicators, alarms, and recorders in the l control room. These monitiers will alert personnel to unexpected ;
or abnormally high. radiation levels.in these areas. Area j
- .monitiors in the fuel storage areas comply with 10CFR70.24. !
-These and-area monitors in the radwaste enclosure comply with
.10CFR50, Appendix A, General Design Criterion 63. A criticality i accident would necessarily be accompanied by an increase in the I gamma radiation in the area. The location and alarm points !
associated with these monitors allow them to serve the purpose of !
criticality accident alarms, in accordance with Regulatory i Guide 8.12.- ;
12.5.2.2.6 Bioassay l Internally deposited radioactive material is evaluated by use of whole body. counting or urinalysis. The whole body counter -
! sensitivity for gamma emitting isotopes of interest will be equivalent to small fractions of isotopic organ burdens.- The !
whole body counter is leased from a contractor; however, purchase !
.will be considered if deemed beneficial. The contractor is responsible for maintenance and calibration of the instrument.
Urinalysis is used primarily for the detection of tritium. This !
analysis-is performed with the onsite liquid scintillation i detector or samples are shipped to a contractor for analysis.
o .
-The bioassay program will incorporate features of Regulatory f Guide 8.9 (September, 1973) and Regulatory Guide 8.26.
Conservative investigation levels are established. When :
investigation. levels are exceeded, an investigation will be i performed and, as necessary, will include consultation with RMC l for further evaluation of internal exposure consistent with !
L Regulatory Guide 8.9 and International Commission on Radiological l Protection (ICRP) criteria. l 12.5.2.2.7 Personnel Protective Equipment Special protective equipment such as coveralls, plastic suits, i shoe covers, gloves, head covers, and respirators are available, t and are stored in various plant locations and clothing change .
areas. This equipment is used to prevent deposition of !
radioactive material internally or on body surfaces. Most of the plant will be kept free of contamination so that no special ;
protective equipment will be needed. Contaminated areas are :
identified with posted signs. Radiation signs and Radiation Work l Permits (RWP) are the primary mechanisms for defining the i equipment required to enter these contaminated areas. !
, f Rev. 6, 06/82 12.5-8 l .
. = . - . .
6 i
- t
v A-variety of combinations of protective equipment may be ,
prescribed depending on the nature and level of the contamination. For example, cotton clothes may be adequate -
normally; but in wet areas plastic rain suits or bubble suits may -
be prescribed. Respirators would be required if airborne hazards exist or if surface contamination could cause an airborne hazard as defined.in the implementing procedures. Sufficient quantities ,
of NIOSH approvo,d respiratory equipment will be provided to i adequately protect workers in general and emergency conditions. l The' guidance of Regulatory Guide 8.15 will be followed to ensure i the~ proper selection,_ care, fitting, and use of respiratory protective equipment. Regular sampling and survey programs will ?
be provided to determine the need for the appropriate respiratory i protective' devices. Bioassays, record keeping, and medical evaluations will be incorported into the Limerick respiratory '
protection program. _ Written procedures will be developed with ;
the help of the operating experience at Peach Bottom Atomic Power Station to provide an effective and acceptable respiratory ';
protection program. Section 12.5.3.5.3 also addresses compliance with Regulatory Guide 8.15. '
12.5.3 PROCEDURES ~AND PRACTICES t
Health physics procedures are classified according to specific !
concerns. .Thus, there are operating procedures, analytical procedures,. emergency procedures, and surveillance test procedures. Health physics is a consideration in other plant ,
- procedures as well. This section describes the fundamental :
procedures applicable to health physics oper&tions. The '
procedures-and methods of operation which, in combination with !
PECo policy and training, ensure that radiation exposures will be i ALARA are those which implement the controls and prescribe the use of equipment described in this section and in Section 12.5.1. l o ~ 1 :2. 5. 3.1 Radiation Surveys (Area Surveys) - I i
Area survey procedures describe the purpose and techniques of }
- detecting the~ presence of and measuring the level of radiation ,
and' contamination. Contamination may be on surfaces or airborne. i Area surveys are conducted throughout the plant. Such surveys >
may be~ routine or~may be related to specific jobs upon request.
An area survey may be performed before, during, and after various !
work activities. Area surveys are performed by health physics !'
technicians. ,
t 12.5.3.1.1 Radiation Detection on The primary instrument for beta-gamma doce rate measurements is i un ion chamber. Circumstances may require the use of other )
instruments to determine dose rates. G-M proben may be used for low radiation levels or where environmental conditions f
12.5-9 Rev. 6, 06/82 l f t I !
I
.- _ _ ~ , ~ . . _ _ - ~ _ . . . , - _ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ - . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . , ~ . _ _
w LGS FSAR (temperatures, humidity) cause erratic responses from ion chambers. -
I Surveys:for neutrons are performed by instruments designed for that purpose. A rem counter or equivalent that has the ability !
.to measure neutron dose rate in rem per hour is the preferred 4
. neutron. measurement instrument. However,-other instruments or devices for determining neutron energies may also be used. For :
- example, moderated and unmoderated boron triflouride (BF 3) probes.
l may be used to detect the presence of low and high energy ;
j neutrons. Some instruments are designed with extension arms. [
These types of instruments.may be used in high dose rate areas to i
!: provide. distance from the source to the technician, thus reducing '
f personnel dose. .
12.5.3.1.2 Surface Contamination Detection l l A variety of techniques are necessary to detect and measure ;
radioactive contamination. Procedures describe the use of smears ;
- (e.g. , small paper discs) and swipes (e.g., paper towels) to wipe !
a surface to pick up removable contamination. The smears are :
. normally_ analyzed using counting room equipment. Swipes are ;
normally analyzed using portable survey meters. Fixed ;
contamination is determined by scanning a surface with portable j survey meters. G-M probes are used for beta-gamma measurements ;
-and alpha detectors are us,ed to distinguish the alpha component. l 12.5.3.1.3 Airborne Contamination ,
t Airborne contamination is determined by using air samplers to !
draw a known volume of air through a filter paper or charcoal !
cartridge. A charcoal cartridge is used with the filter paper [
where iodine-is of concern. The filter paper is analyzed by }
L gross beta-gamma count and gamma spectrometry. Gamma I
spectroscopy is also performed on'the charcoal cartridges. The !
gamma spectremetry identifies the-particulate and iodine isotopic r
, . activity. Gross beta-gamma count data is used to judge the need ;
for filter paper gamma spectrometry. High volume air samplers ,.
and low volume air samplers having nominal sample rates of 25 .
scfm and 3 scfm, respectively, are available. The high volume i I: air sampler is used primarily to obtain grab samples rapidly ,
i before, during, and after work activities. The low volume air L sampler is used primarily to obtain the average air concentration for the work period.
12.5.3.1.4 Survey Frequency and Techniques The frequency and extent (scope) of area surveys is a function of I
. dose rate in the area, accessibility to the area, and nature of i the area in plant operations, For example, , contamination checks i
.in control' rooms and eating areas are performed nominally on a l daily basis; seldom entered high radiation areas are surveyed as j Rev. 6, 06/82 12.5-10 [
_ _ -- ~ _. _ _._..__ _ _ , _ _ _ _.. _ _ _ ________ _______.____J
i infrequently as monthly, or only as needed for entries. As part !
of the ALARA program, the performance of area surveys are i coordinated so that, wherever feasible, routine area survey data i will-be applied to specific RWP requests. This practice avoids l
- j. duplication and reduces exposure to technicians. ;
Area survey data are usually obtained or known before work starts i in an area. Re-surveys may be performed in the area during the -
job if the work activity is prolonged. Other conditions for re- i surveys would be if the work activity or other plant operation caused changed conditions. Surveys may also be performed at the l completion of the work activity. j Depending on the survey results, the area surveyed is roped off !
and posted appropriately to alert personnel to the radiation !
conditions and requirements for entry. (
' Procedures for area surveys describe the use of instruments, !
effective survey techniques, and documentation of data. The ;
specifications for clean, radiation and high radiation areas are defined. Various levels of surface and airborne contamination
. are established as guidelines for prescribing protective
, equipment such as clothing and respirators. Procedures include l consideration for potential as well as actual radiation hazards. i r
l 12.5.3.2 Radiation Work Permits j i Where radiation-dose rates, airborne concentrations, or surface I contamination levels exceed procedural limits, an RWP is issued l l prior to scheduled work. This is accomplished by the submittal t of a RWP request form to health Physics. Health Physics will l evaluate the radiological conditions associated with the work.to !
be performed. Health Physics will specify the appropriate I protective clothing, equipment, and monitoring required including :
dosimetry. Area survey' frequency is established by Health !
Physics. A11' personnel performing work under a particular RWP .
must.be familiar with the permit conditions and must sign in on !
,- the sign-in sheet. The sign-in sheet identifies the individual, l
! his time in and out, and his self reader dosimeter value in and l L out. Health Physics may terminate an RWP if radiation conditions l L change. '
o !
Health Physics supervision selectively reviews completed RWPs
, which are then filed. RWPs serve as a source of data for dose !
i -comparison on repeat jobs. They can be used to determine the i
[ . effectiveness of ALARA efforts. j 12.5.3.3 Handlinc and Storace of Radioactive Material i Health Physics personnel are notified of the receipt of l
! radioactive material, and of intended shipment of radioactive I material._ This is done so that' required surveys can be performed i i
y 12.5-11 Rev. 6, 06/82
!: l l I l
__J
bl I
~
and to verify that. correct.' labeling and placarding has been
. accomplished. i Calibration sources for. radiation instrumentation and sources !
.used to prepare secondary standards are stored in a source vault. I
~
This vault i:s. capable.of being locked. The lock is under the ,
control of'the Senior Health Physicist. :
Small-quantities ~of sealed or unsealed sources may'be stored for.
convenience in shielded cabinets, caves, or safes. Such sources are used locally in the chemical laboratories, counting room, or !
when response-checking instruments throughout the plant. j Spare ~ instruments containing built-in sources and slightly contaminated equipment intended for reuse may be stored at times :
in the warehouse. Such items are clearly identified as ',
containing radioactive material, 12.5.3.4 Controllino Access and Stav Time ,
The plant.is surrounded by security fencing. Entrance must be I via the guardhouse.- Security and health physics procedures are
. applied by the guards at this point to identify each individual, .
to determine.their radiaticn exposure and health physics training !
history, and toidetermine their purpose for entry. Security and ;
dosimetry badges are assigned. Escorts are provided to satisfy ;
procedural' requirements. ,
r p 12.5.3.4.1 Access to Radiation and High Radiation-Areas i
. _ Radiation areas are identified by posted radiation signs. Signs are'used to define requirements for entry. Where' appropriate, ;
yellow and magenta rope or tape is used as a barrier to prevent ~
i access or to divert personnel to a specific control point for i
-access. RWPs are:used to describe work activities in an area, to i prescribe radiation protection clothing and equipment, and to 7 _ document entry and exit of each individual. Procedures describe the purpose and application of the RWPs. Administrative guides ;
for personnel-exposure are established by procedure. These guides are set at values less than the exposure dose limits in 10 CFR, Part 20, Variations from these guides must be requested. i Procedures describe'the steps for approval of dose extensions. ,
Additionally,' positive control is exercised over entries into i high radiation areas by using locked doors or gates. Keys for these hich. radiation doors must be cbtained from the shift .
. supervision. Each key used and the individual using the key are recorded. Certain werk activities having high dose rates and short stay time may be monitored directly by Health Physics !
Technicians'to prevent persennel from inadvertently exceeding the j prescribed stay time. Personnel are advised to observe the !
reading on their pocket dosimeter frequently. :
r l
Rev. 17, 02/83 12.5-12
_..,,_.____________________,,____.___.,____I
3r L
'12.5.3.4.2 Contamination Control j i
Usually, radioactive contamination exists in radiation areas. i Access to these areas is confined to a specific control point. !
Paper floor coverings called. step-off-pads define these control !
points. Personnel wishing to enter at a control point must i review the RWP to determine the prescribed radiation protection f clothing required. This may include respiratory protection '
equipment as-well. The RWP or related survey data sheets attached contain stay time information based on actual or l
. potential airborne contamination. Where respiratory protection i factors are applied, maximum permissible concentration (MPC)- !
hours-are maintained for each individual involved. Appropriate procedures give guidance for the selection and application of
' protective. clothing and respirators under various specific i conditions. Procedures'are established also for inspection, !
cleaning, and maintenance of such protective equipment. !
i ~ !
As personnel leave the work area, they remove the protective i
. clothing and respirators before stepping across the step-off-pad. ;
Personnel must frisk themselves.to assure that no contamination has been transferred to their bodies. . !
Contaminated' tools and equipment to be transported from a !
contaminated area must be placed in a clean container (e.g., a !
plastic bag), surveyed, and tagged at the step-off-pad. The -
tools may be placed in storage, may be taken to another '
contaminated area for reuse, or they may be designated for l
- ' . appropriate storage, disposal, or decontamination if their level ;
of contamination exceeds procedural limits. i
{
The presence of radioactive contamination, whether surface or ;
airborne,-inhibits mobility of personnel around the plant. '
Protective equipment that must be worn creates inconveniences and i introduces other factors that affect performance. For these i reasons, plus the obvious external and internal radiation !
, hazards, decontamination is initiated judiciously to confine the ';
contamination to as small an area as practicable and/or to reduce the contamination levels to minimize protective requirements. i Special coatings that aid in decontamination are applied to walls ;
and floors. The ventilation ficw pattern is from clean areas to '
contaminated areas. Process equipment is isolated in various !
cavities or cells. These cells are vented in a controlled !
manner, usually through filters to effluent stacks that monitor
. flow rate and radioactivity. Highly contaminated equipment ;
drains are piped to sumps to avoid the use of floor drains and .
attendant spillage of fluids on the floor.
l i
l l
12.5-13 Rev. 6, 06/82 !
- + - . . . _ _ . _ re ,_.._ .-.- -,-..__,,,,-,% ,_ ,.,_,_ -, w_, r,. _ _ _ ,,,_,.~ . - _ - _
4' t
i 12.5.3;5 Trainino- ,
Thef training administered ta) personnel is a function of their responsibilities, the areas of the plant they. require access to, and the radiation conditions in those areas. There are a variety :
of. training programs.- These training programs are periodically reviewed and modified.to reflect changes in practices, ,
procedures, and regulations. l
\
12.5.3.5.1 ' General Employee Training !
z Eachuindividual, whether employee, contractor, or visitor must ;
receive General Employee Training (GET) in order to be eligible j for unescorted access to the controlled area within the security
' fence. (Security clearance is also required.) 'This training (
provides knowledge of basic radiation protection, emergency ;
procedures, security, and quality assurance and will conform to !
the provisions of Regulatory Guides 8.27 and 8.29~. Documentation !
of this training is transmitted to the guardhouse to assist i guards in processing personnel. The type of security badge :
issued is. based on this documentation. Re-training is !
administered annually; however, this may be waived for people :
whose work routinely involves them in radiation areas and who i pass a written' examination on the subject. .
A supplement to the GET training program describes the ,
- application of anti-contamination clothing. Trainees are ;
instructed in the purposes served by the different types of !
clothing, how tus properly' don and remove the clothing, and how to l dispose of it.
Female employees working in or frequenting any portion of the i controlled area, their immediate supervisor, and those specifically identified as co-workers of the female employee will be given the instruction concerning prenatal radiation exposure '
as defined in Regulatory Guide 8.13 (Rev 1). .- Female visitors who will enter.the controlled area will also receive the instruction. {
12.5.3.5.2 Respiratory: Equipment Training Certain individuals may be required to wear respiratory equipment l
.in the discharge of their responsibilities. A separate training ;
class'is conducted for them in the purpose, use, and limitations i of specific respiratory protective equipment used at the site. l To be e.1.igible for duty which requires the use of a respirator !
and the application of protection factors, an individual must l pass the Respiratory Fit Test, must have received the respiratory !
equipment training, and must be medically certified as being ?
capable of working safely while wearing a respirator.
i I
Rev. 6, OG/82 12.5-14 I
i
-- . .- _ . _ ._. - . _ .__. _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ O
c: ,
12.5.3.5.3 Respiratory Fit Test >
A functioning Respiratory Protection Program, which meets the i requirements of 10 CFR, Part 20.103 and Regulatory Guide 8.15 !
(October, 1976), permits the application of protection factors to [
select the appropriate-type of respirators. This can only be !
, done forLindividuals who have received training in the use of the
. respirators, have been tested for the validity of fit for the type of respirators intended for their use, and who have been i medically certified. Test equipment such as a shroud or booth is I used to validate fit. A test environment of NaC1' aerosol or an
! equivalent is used to quantitatively prove a satisfactory fit.
Procedures are established that describe the technique and define
(
i acceptance criteria. !
i 12.5.3.5.4 Health Physics Technician Training A training program for Health Physics Technicians is established.
-This program serves the purpose of instructing new' technicians in :
-the operating and analytical Health Physics procedures and l t orienting them in plant layout and systems. 1 i
12.5.3.5.5 Compliance with Regulatory Guides j o .
i l- Reculatory Guide 8.2 --Section 12.1 describes the general ~ALARA ;
I program and policies. As specific procedures and the Limerick l ALARA plan are developed, Regulatory Guide 8.2 will be used for [
guidance. t Reculatory Guide 8.7 - Exposure records will be kept in accordance with 10CFR20 and maintained until the NRC authorizes
, their disposition. As the procedures are developed, the guidance i
! of ANSI N13.6-1969 as endorsed by Regulatory Guide 8.7 will be '
i followed.
Reculatory Guide 8.8 - FSAR Sections 13.1, 13.2, 13.5, 12.1, and 12.5 address the management commitment, organization responsibilities, authority, training, procedures, and review techniques which implement Regulatory Guide 8.8 for Limerick's i operation.- l Reculatorv Guide 8.9 - As stated in Section 12.5.2.2.6, the l Bioassay program wi;,1 incorporate features of Regulatory Guide 8.9. When conservative investigative levels are exceeded, l
Radiation Management Corporation will assist in a more detailed evaluation of internal exposure. These procedures and practices ,
will follow the guidance of. International Committee of Radiation !
Protection publications and Regulatory Guide 8.9.
Reaulato ry Guide 8.1.2 - The development and implementation of a forma; ALARA program will follow the guidance of Regulatory [
l 12.5-15 Rev. 6, 06/82 l
i er LGS FSAR Guide 8.8 which is a nuclear power plant specific reference of Regulatory Guide 8.10. ,
Reculatory Guide-8.13 - Instruction to workers concerning ;
prenatal radiation exposure will be given as part of the General i Employee Training, Program. This program will provide all employees with the information that is identified in Regulatory ,
Guide 8.-13.
Reculatory Guide 1.8'- Limerick is in compliance with this guide to the extent discussed in Sections 13.1 and 13.2. PECo has
-implemented ANSI /ANS 3.1-1978 Section 5. This standard is a revision to ANSI N18.1-1971, which is endorsed by Regulatory Guide 1.8.
Reculatory Guide 1.16 - Limerick Technical Specifications will be based on NUREG-0123 Revision 2 " Standard Technical Specifications for General Electric BWRs", as described in Chapter 16.
, Reporting of operating information will be in accordance with the Technical Specifications, Reaulatory Guide 1.33 - Limerick wil1 follow the guidance of Regulatory Guide 1.33 which endorses / modifies ANSI N18.7-1976
-with certain alternate. approaches which are addressed in detail in Section 17.2A. .
Reculatory Guide 1.39 - Limerick is in. general conformance to the guidance of Regulatory Guide 1.39, which endorses / modifies ANSI N45.2.3-1973. Limerick will comply with Regulatory Guide 1.39 with certain alternate approaches which are outlined
-in detail in Sections 17.2 and 17.2B.
Rev. 6, 06/82 12.5-16 Lo
y ,
5%7 -
DGCKLT NUMBER l' ADD. & UTIL FAC.k b:g. LGS TSAR
( Sd OD b L that the larger of the two lines (20") ruptures at the point $$#' 0 where the pipeline passes closest to the Unit 2 reactor (approximately 3000 feet). It is further. assumed to be a N thly y P4:i; double-ended rupture (complete separation of the pipe at the point of rupture). ry;y y _C%
DX- . L.4
&~$F .
A portion of the cloud downwind within flammable limits is assumed to ignite and deflagrate. The radiant heat load at the Unit 2 reactor enclosure is calculated to be about 70 Btu per square foot per hour (Ref. 2.2-5) for a short time. This level would cause a slight warming of the outer layer of concrete.
2.2.3.1.3 Exposure to Hazardous Chemical Releases Exposure of control room personnel to hazardous chemical vapors could potentially result from an accident involving a chemical spill. Such spills could occur on the rail line, one of several highways close by, nearby industrial facilities, or from onsite chemical storage. A chemical is considered a potential hazard if it is stored or transported nearby in such quantities that its concentration at the control room air intake following a spill could exceed the toxic incapacitation concentration. Acceptable toxic incapacitation levels were baced on compliance with the Regulatory Guide 1.78 requirement of 2 minutes for operator protective action, NUREG/CR-1741 incapacitation models (Ref. 2.2-8), OSHA exposure limits, and ACGIH concentration criteria.
Potential chemical hazards vere identified by first compiling a list of toxic chemicals that could pose a vapor hazard based on Regulatory Guide 1.78, NUREG-0570, and other sources.* Surveys were conducted to determine which of these are actually stored or shipped within 5 miles of the Limerick site, with what frequency, and in what quantities. For the railroads, Conrail provided information on which of these are shipped. Shipment frequency and quantity for those chemicals determined to be a hazard to control room operators are indicated in Table 2.2-6. Per Regulatory Guide 1.78, chemicals shipped less than 30 times per year are disregarded. For the highways, no centralized information source exists to determine what' chemicals are shipped. A manufacturers and users survey was therefore conducted to ascertain potential shippers and receivers of hazardous chemicals. Various directories were used to identify such manufacturers in Pennsylvania and the surrounding states and users in the local area. Based on geographic location, competing highways, and direct routes, these manufacturers and users who
{ would reasonably use the three highways near the site were 2.2-7 Rev. 18, 03/83
s 4
4 4
4
's q_
s 9
(
i 4
4 g
1
-ass.
1% ,
k,j I
\.
\
e l
I ,h...,.
N ,. .
llhi
.'1 . 9. (
LGS FSAR contacted r*6arding chemicals shipped or received, routes, and ;
container sises. An analysis was then conducted to determine which of these cl.'aicals, if spilled, could escoed toxic. :
incapacitation levels in the control room. These are listed in Table 2.2-6, along with container sizes.
The analysis assumed complete. release of the contents of a single ,
container or tank. In accordance.with Regulatory Guide 1.78, it l was assumed that after an initial puff of vapor, any remaining !
liquid spreads over the ground and evaporates. The methodology i
'of Regulatory Guide 1.78 and NUREG-0570 was used to model the !
initial
~
control,putf roomand air subsequent intake. The plume controltransport and dilutionwere room concentrations to the determined using'the following control room parameters:
- a. control room envelope volume of 126,000 fta, as defined f in Section 6.4.2.1. j
- b. 2100'cfm of incoming / outgoing air, based on the design outside air. flow rate supplied by the normal control room MVAC system, as described in Sections 6.4.3.1 and 9.4.1.1.
- c. air. intake 36.5 meters above ground, as indicated in ,
Figures 1,2-27 and 6.4-2. t
- d. inleakage rate of 0.25 air changes per hour, during isolation, as discussed in Section 6.4.2.3. l
- e. 40 seconds time delay in the ductwork between the detectors at the control room intake plenum and the isolation valve at the entry into the control room air space, based on the air velocity'in the duct.during normal operation. [
l The consequences of an accidental release of phosgene gas, a !
combustion product of vinyl chloride, resulting.from a fire in l conjunction with an accident involving spillage of vinyl chloride were also evaluated. The phosgene concentration in the control !
room was calculated using the models of NUREG-0570 and the heat !
rise models of J.A.' Briggs'(Ref.2.2-9). I i
l Chemicals stored onsite include carbon dioxide, chlorine, L nitrogen, and sulfuric' acid, in quantities and at-locations !
listed on Table 2.2-5. Analysis of accidents involving onsite f chemicals resulted in identification of chlorine spillage as l potentially hasardous to control room personnel. ;
k !
Rev. 22, 07/03 2.2-8
As a eesult of the analyses, sin potentially hazardous cbsmicals L requiring monitoring were identified, as listed in Table 2.2-6. ,
A brief description of each chemical and its effects on numans !
and laboratory animals ate presented belove t
Ammonia, NH 3 l .
Ammonia is a colorless gas with sharp, intensely irritating odor, f It has ao odor threshold of 46.8 ppm for humans (Ref. 2.2-13).
Complaint levels of 20 to 25 ppm were first observed. Human effects such as eye irritation, sometimes with lacrimation, nose, throat, and chest irritation (coughing, edema of lungs), were [
found at concentrations up to 700 ppm, depending on exposure time '
(Ref. 2.2-10, 2.2-11 & 2.2-12). The chemical then becomes lethal '
starting at 2,000 ppm concentration even for exposures at very short duration (Ref. 2.2-10).
Chlorine, C1, l Chlorine in its gaseous form is greenish-yellow in color. It has a disagreeable, suffocating and irritating odor readily detectable at 3 to 5 ppm. Its effects on humans depend on the concentration. Irritant effects to eyes, nose, throat and/or face were noted at low concentrations. Effects on the upper and lower respiratory tracts and pulmonary edema were reported on exposures at high concentrations. It becomes highly dangerous to be exposed for 30 minutes at 40-60 ppm, fatal at concentrations of 833 ppm if breathed for 30-60 minutes, and rapidly fatal after a few breaths at 1,000 ppm (Ref. 2.2-10). There were reports on effects of concentrations around 5 ppm causing respiratory complaints, corrosion of teeth, inflammation of mucous membranes of nose, and increased tuberculosis susceptability ( Seference 2.2-14).
Ethylene Oxide, C,H.0 l Ethylcae Oxide, a suspected carcinogen, is a colorless gas, sickening and nauseating at moderate concentrations and irritating at high concentrations. Humans exposed even to low concentrations showed delayed nausea and vomiting and at continued exposure, numbing of the olft.ctory sense. Inhalation at high concentrations resulted in general anesthetic effects as well as coughing, vomiting, and irrit ation of , eyes and respiratory passages leading to emph)sema, bronchitis and pulmonary edema (Ref. 2.2-10). The lowest toxic concentration in humans through inhalation is 12,500 ppm for 10 minutes with only irritant effects observed (Ref. 2.2-12). Odor threshold is
(- 50 ppm for this chemical (Ref. 2.2-13).
2.2-8a Rev. 22, 07/83
Formaldehyde, a suspected carcinogen, is detectable by most ;
people at levels below 1 ppm from References 2.2-11 and 2.2-14, ;
and at 0.8 ppm from Reference 2.2-13. Humans experienced irritant effects on the eyes, nose, throat, and upper respiratory tract at concentration ranges of less than 1 ppm to 12 ppm. At ,
high concentrations, a severe respiratory tract irritation which lead to death was reported on humans (Ref. 2.2-14). Inhalation study on rats and mice showed that formaldehyde has a carcinogenic effect on rats. Rats developed nasal cavity squamous cell carcinomas after 12 to 24 months of exposure to ;
15 ppm, with deaths occurring ducing this period. Fatalities on rats were also observed at exposures to 81 ppm concentration (Ref. 2.2-14).
Vinyl chloride is a colorless, toxic, highly flammable gas at room temperature and atmospheric pressure, with a pleasant-, sweet odor at high concentrations (Ref. 2.2-10). Evidence has shown it "to be a carcinogen to persons exposed over extended periods of time (Ref. 2.2-10). Exposure through inhalation at 200 ppm for 14 years showed occurrence of tumors on humans, carcinogenic effects at 500 ppm for 5 years (Ref. 2.2-12). At concentrations above 1,000 ppm, vinyl chloride was reported to slowly affect a mild disturbance in humans such as drowsiness, blurred vision, staggering gait, and tingling and numbness in the hands and feet (Ref. 2.2-10). The odor threshold for this chemical is 260 ppm (Ref. 2.2-13).
Phoscene, COC1, .
Phosgene is a colorless, nonflammable, highly toxic gas at ordinary temperature and pressure, with a musty hay-like odor detectable 6t 0.5 to 2 ppm. It is a strong lung irritant and causes damage to the alveoli of the lungs. Inhalation of phosgene produces catching of breath, choking, immediate coughing, tightness of the chest, lacrimation, difficulty and pain in breathing, and cyanosis (Ref. 2.2-10). Humans experience throat irritation at 3 ppm, immediate eye irritation at 4 ppm and coughing at 4.8 ppm. Brief exposure at 50 ppm may be rapidly fatal (Ref. 2.2-11).
To ensare adequate protection of control room personnel, control room operators will be trained and periodically tested on their ability tc put on breathing apparatus within 2 minutes after initiation of the toxic chemical alarm. Subsequently, the operators will manually isolate the control room as described in (
l Rev. 22, 07/83 2.2-8b i
- . . _ . . . --=
f LGS FSAR Section 6.4.3.2.3. If chlorine is detected, automatic isolation !
of the control room will occur as described in Section 6.4.3.2.1.
The Limerick toxic chemical analysis complies with the intent of Regulatory Guide 1.78. The analysis goes beyond the methodologies outlined in this guide in the following areas:
- a. In addition to the chemicals listed on Table C-1 of Regulatory Guide 1.78, other chemicals were investigated I to determine if potential hazards existed. A total of 153 chemicals were evaluated. ,
- b. The models of NUREG-0570 were used to determine the ,
concentrations of hazardous chemicals in the control !
room.
- c. The more stringent TLV levels were initially used
- instead of the Regulatory Guide 1.78 Table C-1 toxicity limits to determine which chemicals were potentially hazardous. Table C-2 of Regulatory Guide 1.78 was not used to determine which chemicals were hazardous. !
i r
b
(.
r 2.2-8c Rev. 22, 07/83
~~,,-~w.- ,,.,--,-.-r-.-,..-. , - - , --.e _ . . .- , r.--,,w, - - - , -- , - - - - - --,-,,-.,--g,,-n----, ,c,-- --- r
(
THIS PAGE IS INTENTIONALLY BLANK t
Rev. 22, 07/83 2.2-8d
Il s
- d. Potentially hazardous chemicals were re-evaluated using the incapacitation models of NUREG/CR-1741 (Ref. 2.2-8) to determine if control room operations would be incapacitated. This analysis is an amplification of Regulatory Position C.4 of Regulatory Guide 1.78.
2.2.3.1.4. ,Fh In addition to the flammable vapor clouds discussed earlier, fire hazards may also exist due to a burning tank car on the railroad, a fire subsequent to a ruptured pipeline, or a nearby forest / brush fire.- Potential adverse effects of such fires are radiant heat load on plant structures and smoke generation.
To estimate the effects of a railroad fire, an accident is hypothesized in which a railroad tank car derails, ruptures,' and releases a cargo of 62 tons of liquified propane. A 62-ton car is typically the-largest size used for propane, and from a fire standpoint liquified propane-represents one of the most severe materials transported by rail. The site of the hypothetical derailment is'the closest point of approach to the Unit I reactor enclosure, about 600 feet. The tank car propane is assumed to be
' released into the drainage ditch alongside the eastern side of the'right of way, where it pools and is subsequently ignited.
The vapor pressure of liquid propane is sufficiently high at ambient conditions that there will be an adequate supply of
- gaseous propane for ignition, after which the fire is self-propagating. The fire duration is assumed to be 20 minutes, based on experience with this material.
i Assuming 19,600 Btu per pound of propane and 62 tons being consumed in 20 minutes, the radiant heat load on the reactor
' enclosure may be calculated using the relationship (Ref 2.5-5):
l 1/2
- D =_(FQ/12.57K) (2.2-1)
'where: D= distance, feet F= fraction of heat that is radiant O= heat release, Btu per hour 7
K= radiation load, Btu per square foot per hour i-
- The result of this calculation indicates a radiant heat load of
! approximately 500 Btu per square foot per hour for 20 minutes at J-l 2.2-9 Rev. '8, 03.'93
i f
l T ABLF '2. 2- 1 :
I HOOKER CIMMI"At COMPANY l
- ?
RELIFF ELEMTIOtt MLVE 5 FORA 3R I MAXIMUN OF TANKS CAPA0ITY TEMPERATURE :
GENIEM SE&t!IIII .I.f f f i l _ (peiql__ 3lfg_ggj53QRg Vinyl chloride 3,000,000 lb 12 100 30 psig-ambient !
Butadiene 500,000 lb 12 100 20 psig-ambient l l
Tri- flurs- Portable ;
chloro-enthylene 2,000 lb cylinder 315 68 psig-ambient I
Tri- flurs-chlor-ethylene 1,000 lb In process None Ar.bient i Tceca15ehy3e 50 Dranis Warehouse --
Ambient i I
%tbanol 10 Drums Warebouse -- Ambient Nitrogen 139,000 SCF 3 350 -32S*F ,
Toltene 13,000 gal. 12 (100) Ambient-vent ;
Gasoline 52,000 gal. Underground -- Ambient vent i
Styrene 50,000 gal. 12 (100) Ambient-vent ;
L Vinyl a::etate 25,000 gal. 12 (10 0) Ambient-vent Tri-chia rs- .
ethylene 25,000 gal. 12 (100) Ambient-ve nt Vinyl pyridine 10,000 gal. 8 --
40*F i ___.___ _____.__________________...___________ ____ ._________.________ ;
! I I l ,
I i
l l
Pev. 8, 07/32 l
w _
(?~ .
LGS FSAR TABLE 2.2-5 ONSITE CHEMICAL STORAGE Stored Volume Number of Chemical- (Standard cubic feet) Tanks Location Carbon Dioxide 171,275 .' yp Turbine Enclosure El. 239 (Common)
Carbon Dioxide 47,100 2 Turbine Enclosure El. 217 (Unit 1 and Unit 2)
Chlorine 580,000 1 Circulating Water Pumphouse (Common)
Nitrogen 539,150 2 West of Radwaste Enclosure E. 218 (Common)
Sulfuric Acid 1,337 2 Adjacent to Cooling (10,000 gallons) Towers (Unit 1 and
- Unit 2)
Sulfuric Acid 535 1 Water Treatment (4,000 gallons) Building (Common) e l
A %
I
)
POTENTIALLS MAXIMUM CALCULATED CONCENTRATION MONI'IOR (No Control Room SET POINT Isolation)
GliMIEMt _lEEBL_ ____1995l__
Anmonia 25 2700 Chlorine 0.5 200/730C*3 Ethylene oxide 50 1200 Formaldehyde 5 48 Vinyl Chloride 10 12,000/3300(*3 I Phosgene 0.4 320 (13 Rail snipments are average weights. No ad tons / carload was considered.
cal Phosgene is a combustion product of vinyl (33 For chlorine,, data presented are based on
(*) First value is for storage /second value is (53 Incapacitation model types are taken from 1
l
i IAS FSAR TABLE 2. 2-6 BAEARDOUS CHEMICALS REQUIRING MONITORING } l
-- - - - -__ _____.____________----- - ---- ____________ l i
ENITOR INCAPACITATION l 4 DELAY TIME MODEL(s> SHIPMENT ; FREQUENCY l imeci JeiDL_______________ __L4 ppg __ i jserloadS/YEl AtlQWIlll i
<40 2.7 A Rail 500-1000 54 tons / carload I 8 9.4/4.3(3)(*) A Storage / Rail 500-1000 74 tons / carload I (40 2.6 B Rail 500-1000 75 tons /caricad i
<40 9.~ 2 A Rail ; 30-99 87 tons / carload i
<40 17.4/25(*3 D Storage / Rail 500-1000 92 tons /caricad I
<40 15.3/2.9(*) B cm) __ __ g tional chemical hazards were identified when the maximum weight of 90 ilorido. I Itomatic isolation of control room. I or railroad. I fREG/CR-1741.
- 4 I
IT[
APERTugg CAgg i
1 him Aullable On Aperture Card I 8406270315-01 e
}
nov. 22, 07/83 i
1
_ , . . - , , , ~ , , _