ML20011D759
| ML20011D759 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/01/1989 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20011D723 | List: |
| References | |
| NUDOCS 8912280374 | |
| Download: ML20011D759 (29) | |
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TABLE OF CONTENTS CHAPTER 11: RADIOACTIVE WASTE KANAGEMENT 11.1 SOURCE TERMS 11.1.1 ACTIVATION PRODUCTS 11.1.2 FISSION PRODUCTS 11.2 LIQUID WASTE MANAGEMENT SYSTEM 11.2.1 DESIGN BASES 11.2.2 SYSTEM DESCRIPTION 11.2.3 RADIOACTIVE RELEASES 11.3 CASEOUS WASTE MANAGEMENT SYSTEM 11.3.1 DESIGN BASES 11.3.2 SYSTEM DESCRIPTION 11.3.3 RADIOACTIVE RELEASES 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4.1 DESIGN BASES 11.4.2 SYSTEM DESCRIPTICN 11.4.3 RADIOACTIVE SHIPMENTS 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 11.5.1 DESIGN BASES 11.5.2 SYSTEM DESCRIPTION i
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i 8912280374 9932,155 fDR ADOCK 05000 PDC l
O 11.1 SOURCE TERMS Radioactive material from the operation of the plant arises from two sources. First, the products of nuclear fission are generally radioactive. Some may escape from the fuel from time to time. A small number of fission reactions also occur outside of the fuel from uranium as an impurity existing on or in the components near the reactor core and the cooling water flowing through the core. Second, a small fraction of the neutrons available from the fissioning process are captured by various materials near the reactor core including impurities in the circulating primary coolant. Many of these products of neutron capture become radioactive. With L.a exception of times of steam leakage they generally remain in the circulating coolant, and hence, add to the source of liquid effluents and solid wastes. The products of fission entering the coolant are generally soluble and can be volatile allowing some to primarily add to the source of gaseous waste.
The initial licensing process 'for the plant included a criterion that gaseous radioactive effluents be limited to permit no more than 500 millirads per year to be delivered at any point in the off-site environment.
(This is equivalent to the 10 CFR 20 limit of 500 millirem for whole body doses to a member of the general public in continuous residence.) Liquid releases were, likewise, to be limited so that it was unlikely that any individual would be exposed to O-radiation in excess of that permitted by regulations.
(At that time continuous exposure at MPC levels of 10 CFR 20.)
The liquid and gaseous waste systems installed at the plant provide flexibility and processing capability to meet this criterion. The
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operational history of the plant, particularly during the period of significant fuel failures, has validated the design in that effluent limitations based upon the limiting dose criterion were never approached.
On June 4, 1976 Consumers Power Company submitted the necessary information to permit the USNRC to evaluate the effectiveness of the radioactive waste treatment systems at the plant in accordance with the then, newly established Appendix I to 10 CFR 50. This submittal was supplemented by submittals of December 3,1979, August 28, 1980, June 7, 1982 and September 29, 1982. On May 15, 1981 the NRC published their evaluation and concluded that the installed systems are capable of maintaining releases within the design objectives of Appendix I.
The requirements of Appendix I much more severely limit radioactive effluents than the original limits imposed on the plant. These limits, called Design Objective Annual Quantities, are defined in the Off-Site Dose Calculation Manual and establish the long term upper limit for radioactive materials concentrations in effluent streams.
Maximum permissible radionuclide concentrations in the primary coolant system have not been specifically calculated, which, when applied to the available liquid and gaseous waste processing would result in releases to the environment approaching Appendix I to 11.1-1 MIO389-0125A-BX01 I
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10 CFR 50. However, over many years numerous measurements of radio-nuclide concentrations in primary coolant have been made under various plant operating conditions which define the actual source term and can be used to approximate the maximum permissible source term for sustained plant operation.
11.1.1 Activation Products Camma spectral analyses of activation products in the primary coolant at a steady state power level of 216 Hwt several months af ter a ref ueling outage and in the absence of significant fuel defects, have recently been made on unfiltered samples. The results appear in Table 11-1 below and provide an appropriate partial source term for input to the liquid and gaseous waste management systems. Where the "less than" symbol (<) appears, the nuclide was identified as present in the primary coolant only occasionally at most. Its minimum detectable concentration was utilized when its presence was not measurable to calculate an appropriate average primary coolant concentration. The ( $ ) symbol is used when the nuclide appeared between one fourth and three fourths of the time. Again its minimum detectable concentration was used for those samples, when absent, to calculate an appropriate value. Where no symbol appears, the nuclide was generally measurable in all samples. All values in the table have been multiplied by 240/216 to correct the measured data for 240 Hwt operation.
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O TABLE 11-1 PRIMARY COOLANT CONCENTRATIONS OF ACTIVATION PRODUCTS Primary Coolant Radionuclide Concentration (pCi/ml)
Na-24 8.1E-4 Cr-51 3.4E-3 Hn-54 5 3.7E-4 Mn-56 1.4E-3 Co-58 1 8.5E-5 Fe-59
< 3.8E-4 Co-60 3.3E-4 Zn-65
< 2.7E-4
/
Ag-110H
< 9.1E-5 lig-203
< 5.0E-5 Np-239
< 2.6E-4 Concentrations of activation products, with the exception of Np-239 are independent of the amount of fuel defects. 'The measurements resulting in the values presented in Table 11-1 were taken in February and March, 1989 and represent equilibrium values achieved after many fuel cycles of sustained operation near the plant's licensed thermal power limit with no significant changes in primary coolant system l
metals or chemistry occurring during this period.
In addition to the activated metals shown in Table 11-1 activation gases are also produced within the coolant surrounding the core.
They are generally either very short lived or small in abundance.
Their presence is masked by the larger amounts of gaseous fission products and hence they are not normally measurable. The only activation gas nuclide of significance in terms of gaseous vaste release is nitrogen-13. Its value is calculated based upon measure-ments made in 1963 which resulted in a calculated release rate, after hold-up in the gaseous waste system, of 50 pCi/sec at a power level of 49.5 MWe.
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Finally, tritium exists in the primary coolant both as an activation and fission product.
It is only occasionally measured and has remained relatively constant at about 6.0E-3 pci/mi in the absence of significant fuel defects.
11.1.2 Fission Products Camma spectral analyses performed on samples of primary coolant have i
not revealed measurable levels of many fission products during periods of operation with no significant fuel cladding defects. Over the last several fuel cycles, with one exception in 1984, the plant has experienced no significant fuel cladding defects. The resulting average fission gas release rate to the environment (after hold up) during such operation has been approximately 300 pC1/sec. Primary coolant concentrations of solid and halogen fission products and release rates of gaseous fission products to the off gas hold-up system under these conditions are presented in Table 11-2.
Again where the symbols (<) and ( $ ) appear they have the same meaning as in Table 11-1.
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TABLE 11-2 PRIKARY COOLANT SYSTEM FISSION PRODUCTS (NO FUEL DEFECTS)
Radionuclide Primary Coolant Concentration pCi/mi Br-82 1E-4 Sr-91 5.3E-4 Sr-92 1.4E-3 2r-95
< 8.3E-5 Nb-95
< 5.0E-5 Mo-99 6.3E-3 Ru-105
< 2.1E-4 Sn-113 2 5.9E-5 I-131 6.9E-5 I-132 1.9E-3 1-133 8.6E-4 1-134 5.2E-3 I-135 2.0E-3 Dose Eq I-131 6.3E-4 Ba-133
< 6
'--5 Cs-134
< a...-5
(T Cs-136
< 5.1E-5
' 'Q Cs-137
< 5.7E-5 Cs-138
< 8.7E-4 Ba-140
< 2.0E-4 La-140
< 3.0E-5 Ba-141
< 7.9E-3 Ce-141
< 5.2E-5 La-142
< 2.9E-4 Ce-144
< 3.0E-4 Radionuclide Release Rate From Air Ejector pCi/sec Kr-85M 5.1 Kr-87 29 Kr-88 17 Xe-133 2.0 Xe-135M 117 Xe-135 24 Xe-138 530 11.1-5 MIO389-0125A-BX01
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O During early 1984 under steady state power operations fuel cladding defects began to appear and grew until May when, during the entire month at steady state power operation, releases (after hold-up) to the environment remained essentially constant at 26,000 pCi/sec.
Concentrations of Individual radionuclides in primary coolant except for the halogens, Cs-134 and 137, and Np-239 were not nessured during this period. Table 11-3 presents the results of these measurements and where parentheses appear the concentration was inferred from the value given in Table 11-2 multiplied by 26,000/300. Comparison of I
radioactive gaseous releases with the number of failed fuel rods found during this period of operation and several earlier periods in which some fuel defects were noted consistently has shown that a single defective rod contributes approximately 1000 pCi/sec of noble i
gas release to the environment after hold-up.
Tritium concentrations in the primary coolant were not measured during this period of operation with fuel defects in 1984, but release data from effluent reports can be used to calculate an approximate concentration of tritium reached by the primary coolant during this time. The resulting value is 1.8E-2.
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TABLE 11-3 PRIMARY COOLANT SYSTEM FISSION PRODUCTS (WITH FUEL DEFECTS)
Radionuclide Primary Coolant System Concentration pCi/ml j
Br-82
(<8.7E-3)
Sr-91
( 4.6E-2)
Sr-92
( 1.2E-1)
Zr-95
(<7.2E-3)
Nb-95
(<4.3E-3)
( 5.5E-1) l Ru-105
(<1.8E-1)
Sn-113
(<5.1E-3)
I-131 2.5E-2 1-132 4.0E-2 I-133 3.5E-2 I-134 3.0E-2
+
I-135 3.2E-2 Dose Eq I-131 3.9E-2 Ba-133
(<5.5E-3) t Cs-134 3.7E-4 Cs-136
(<4.4E-3)
Cs-137 5.aE-4 Cs-138
( 7.5E-2)
Ba-139
( 2.3E-1)
Ba-140
(<1.7E-2)
La-140
(<2.6E-3)
Ba-141
(<6.8E-1)
Ce-141
(<4.5E-3) e La-142
(<2.5E-2)
Ce-144
(<2.6E-2) t Np-239 (activation product from fuel) 8.8E-4 Release Rate From Air Ejector pCi/sec Kr-85M 1500 Kr-87 3740 Kr-88 3150 Xe-133 2910 Xe-135 5810 Xe-138 19200 1
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Additional Nuclides Calculated from Measured Releases and Known Fission Yields Kr-83H 740 Kr-85 95 Kr-89 8200 Kr-90 87 Xe-131H 5
Xe-133M 95 Xe-135M 14500 Xe-137 13200 l
Xe-139 500 i
During this period, Hay, 1984, releases to the environment from both the gaseous and liquid waste system were such that long term sustained power operation under these conditions would have resulted in the approach to the limits of Appendix I to 10 CFR 50. llence the values in Table 11-3 coupled with those of Table 11-1 represent an approximate maximum permissible long term source term for the plant. The values in Table 11-1 coupled with those of Table 11-2 on the other hand represent typical operational values consistently achieved over the last several fuel cycles.
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M10389-0125A-BX01
t 11.2 LIQUID WASTE MANACEMENT SYSTEM Radioactive materials in liquid waste arise from the activation of corrosion products formed in the nuclear steam supply system and the possible escape of fission products from fuel element cladding defects.
11.2.1 DESIGN BASES The liquid waste system is of sufficient design such that expected input source terms from section 11.1 can be processed so that release of ef fluents is kept within the numerical guidance of Appendia I to 10 CFR 50. System design capacity allows for temporary short term waste volume increases and the ability to store some liquids for reasonable periods of time to permit decay of short lived radioactive material. The use of both filt ration and demineralization prior to release, when necessary, is provided. Recycle of liquids which can be made sufficiently pure to meet the specifications for primary coolant make-up is included. The system is also designed with the waste tanks behind a shield wall which separates the bulk of the waste from the valves and other controls normally manipulated to process the liquid. The capability to recycle collected liquids to obtain a representative sample is provided.
Influent waste streams are carefully segregated to maximize the reuse of collected liquids as primary coolant make-up. When a predetermined concentration is O
exceeded, alarms are automatically received allowing for the manual termination of the release.
Finally, the system is designed to operate in a batch discharge mode to insure careful control of all liquids to be released.
11.2.2 SYSTEM DESCRIPTION The liquid waste management system consists of collection sumps, receiver tanks, hold-up tanks, tank mixing eductors, strainers, filters, a demineralizer, pumps, interconnecting piping and instrumentation. The system is designed to be capable (all pumps operating continuously), to process approximately 70,000 gallons per day.
Data from the last four fuel cycles reveal that during steady state power ot-ration approximately 3800 gallons of liquids are processed per d4y with an average 200 of these gallons per day discharged and the rest processed for reuse. During shutdown with the reactor vessel head removed approximately 37,300 gallons are processed per day with approximately 1000 of these discharged.
Drawing 0740040132 presents a block collection diagram for the liquid waste management system. Liquids to be processed are segregated based upon total solids content. Waste water which normally has a low solids content is collected in a " clean" sump in the containment building and routed to one or both.of the 5000 gallon clean waste receiver tanks. Clean waste is almost always processed for reuse in
-O the plant though provisions exist to mix, sample, analyze and discharge
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the collected liquids.
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Waste water arising from sources in the containment which potentially has a high solids content is collected in a " dirty" sump. A turbine building sump also exists which collects liquids from the condensate feedwater train. Provision exists to route water collected by these two sumps to either the " dirty" or " clean" waste receiver tanks.
Normally, the liquids are of sufficiently high purity so that they can be routed to the clean waste tanks.
i A fourth sump, the "radwaste" sump, exists in the liquid radioactive waste area which collects water overflow from the spent resin or other liquid waste tanks as well as liquids from various floor and equi :nent drains. The collected liquids are routed to the dirty l
waste receiver tanks. Two 5000 gallon tanks exist with one tank undergoing processing while the second is filling. Data from the last several fuel cycles reveals that approximately one half the liquids collected in the " dirty" receiver tanks is able to be processed for reuse in the plant while the remainder is discharged. The liquids to be discharged are mixed to obtain a representative sample, then sampled, analyzed and normally filtered prior to being released at controlled rates to assure effluent limits are not being exceeded.
A single 5000 gallon " chem" waste receiver tank is also provided which collects liquids from radioactive sinks, emergency showers, the plant laundry, chemistry laboratory and equipment decontamination area drains. Though provision exists to recycle these liquids, they O
contain a high solids and organic content, much of which may be in i
colloidal suspension and cannot be processed for reuse, hence these liquids are normally all discharged af ter mixing, sampling, analyses and filtration similar to that conducted on all " dirty" waste releases.
Finally, provision exists to regenerate resins from the reactor clean-up, condensate and liquid radioactive vaste system demineralisers.
Because of the volume of liquids involved and the potentially high-radioactive content, such resin regeneration has not been conducted for well over ten years.
All resins, once exhausted, are collected in either of two spent resin tanks, one of 10,000 gallon and one of 5,000 gallon capacity and, when sufficient volume accumulates, they are disposed of as solid radioactive waste.
prawing 0740C40108 shows a schematic diagram of the liquid waste management system including all ta ks, pumps, valves, instruments and interconnecting piping.
Clean waste, after collection in a clean waste receiver tank is typically routed through a strainer / filter, a cartridge filter, the radwaste demineralizer to one of two waste hold t ar.k s.
It is then pumped to the cond r.. ate storage tank when needed for reactor coolant makeup. Dirty waste, after collection in a dirty waste receiver tank, is either processed to a clean vaste receiver tank and further processed as clean waste if its solids content is sufficiently low or after the strainer / filter it is normally further cleaned by a second filter prior to discharge.
Chem vaste after collection is normally filtered prior to discharge.
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M10389-0125A-BX01
The construction of the liquid waste management system is consistent with conventional construction practice. No seismic design specifi-cations or criteria were applied to equipment and components within the system. The construction of the structure, an underground concrete vault housing the system, was designed consistent with the Uniform Building Code which includes a specification for a 0.025g static horizontal load. A 1981 analysis of the structure by D'appolonia Associates concluded, however, that the structure has an adequate safety margin at USNRC Regulatory Culde 1.60 spectra anchored at 0.12g zero period acceleration.
t.ny overflows, leaks, spills or component breakage is thus expected to be retained by the structure under credible seismic events.
Even so, should radioactive liquids be released to the underground strata their travel is limited by the low horizontal velocity (0.05 feet / day) toward Lake Michigan.
11.2.3 RADIOACTIVE RELEASES Releases of radioactive liquids to the environment have remained consistent over the past several years due to long periods of steady state operation of the plant and the absence of significant long term fuel defects. Table 11-4 presents an annual average summary of discharged liquids since 1985.
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TABLE 11-4 i
TYPICAL ANWUAL LIQUID RADI0 ACTIVE EFFLUENT DATA (No Significant Fuel Defects)
Variable Annual Quantity (gallons for I
volume and curies for radio-l active material)
[
Volume released 1.2E5 I
" clean" waste 2.4E3
" dirty" waste 2.0E4 "chen" waste 9.8E4 Radioactivity Total alpha 7.2E-6 Tritium 4.2E-1 Fission and Activation Products 1.8E-1 i
Radionuclides in Fission and Activation Products Cr-51 7.4E-3 i
Mn-54 3.8E-2 Co-58 7.9E-5 sO.
Fe-59 4.6E-3 Co-60 4.0E-2 Zn-65 2.2E-3 Sr-89 5.0E-4 L
1.5E-4 Sr,.
1.0E-5 Nb-95 4.6E-5 Ho-99 9.7E-5 i
Ag-110H 4.2E-4 i
Sb-124 1.8E-4 4
I-131 1.0E-3 I-133 2.0E-4 I-135 9.4E-5 Cs-134 5.0E-3 Cs-137 2.6E-2 Unid. Beta 5.0E-2 I
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11.2 H10389-0125A-BX01
O 11.3 CASEOUS WASTE MANACEMENT SYSTEM Radioactive materials in gaseous discharges are composed of activation and fission gases and tritium discharged primarily through the condenser air ejector and turbine gland seal system, and radioactive halogens and particulates released to the environment primarily from the plant ventilation system during the presence of primary coolant steam leakage.
11.3.1 DESIGN BASES The gaseous waste system is of sufficient design such that input.
source terms from Section 11.1 can be processed so that release of effluents is kept within the numerical guidance of Appendix I to 10 CFR $0.
System design capacity for fission and activating gas releases from the condenser air ejector allows a small hold-up time for short lived products to decay. High efficiency particulate air (llEPA) filtration is also provided to minimite releases of radio-active particulates via this pathway. The hold up line is designed to remain intact should the Stoichiometric hydrogen oxygen mixture making up the largest portion of the gaseous volume ignite. The turbine gland seal system utilizes about 0.1% of the total steam flow, hence releases of gases from this pathway is limited to a small fraction of that from the air ejector.
Plant ventilation system and the gland seal system exhausts are not filtered or processed prior to discharge. Ventilation flow rate controls are provided to permit appropriate flow distribution to minimize the spread of airborne radioactive contamination within the plant, Drawing 0740040119 shows the ventilation arrangement for the plant.
All ventilation except for a small chemistry laboratory fume hood exhaust is directed to a stack where continuous sampling, monitoring and an alarm feature exists if releases exceed a pre-determined setpoint. Refer to Chapter 9, Subsection 9.4 of this Updated FHSR for a description of the Ventilation System.
Provisions also exist to automatically terminate discharge of the condenser of f gases (normally the largest source of release) if a predetermined release rate is exceeded.
11.3.2 SYSTEM DESCRIPTION The gaseous waste system consists of ventilation fans, ducting dampers, louvers, filters, hold-up piping a 240 foot high stack, controls and instrumentation tot 1) Provide suf ficient decay time for the short lived activation gases produced in the reactor so that releases to the environment are negligible) 2) Provide for the controlled release (below regulatory limits) and dispersion of the noble fission gases which can be released in significant quantities during times of fuel element failuret 3) Provide sufficient ventilation to minimize airborne contamination within the plant; and 4) Keep 1
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limits.
Drawica 0740040108 shows the condenser off gas and turbine seal steam exhaust portion of the gaseous waste system. The system consists of hold-up piping, a HEPA filter for the condenser off-gas releases, the stack and associated controls and instrumentation. The volume of the hold-up piping provides for typical hold-up periods of between 25 and 30 minutes for the condenser off gases and about 90 seconds for the measured at approximately 0.05 f t* production at the plant has been gland seal gases. Radiolytic gas 1
/NWt min at typical condenser i
outlet temperature and at atmospheric pressure or about 12 ft*/ min at t
240 MWt. The off gas hold-up pipe has a free volume of 340 ft* which provides about a 28 minute hold-up time for the gases in the absence of condenser air inleakage. Air inleakage is periodically measured i
and appropriate hold-up time corrections are made to the release calculations. The capacity of the air ejectors is 24.5 ft*/ min which, when coupled with the known volume of the off gas hold-up line, produces a minimum decay period of about 14 minutes below which l
condenser vacuum would begin to be adversely affected.
The gases are continuously monitored by lon-chamber radiation monitors l
preset to terminate the release by closing the hold-up pipe at a preset release rate. The gases are passed through a HEPA filter post hold-up and directed to the discharge side of stack fans.
Two full
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capacity fans are provided to dilute the radioactive gases and provide sufficient induced draft for plant ventilation needs. Their 8
flow rate is kept at approximately 40,000 ft / min each. After dilution the off gases as well as the ventilation flow is continuously r
sampled by an isokinetic sampler within the stack.
The gland seal gases are also directed to the stack fan discharge where they become part of the diluted stack flow.
Drawina 0740040124 and 0740040125 chow the plant ventilation scheme.
With the exception of the containment building ventilation all are induced draft flows with the stack fans providing the energy source.
The containment building system also contains inlet supply fans which provide a forced draft contribution to ventilation. Flow rates are varied consistent with heating requirements but remain sufficient to t
minimize build-up of airborne contamination. Stack flow is kept constant at about 30,000 cfm by the use of louvers in the stack base.
Two full capacity fans are provided. Ventilation flows begin in radioactively clean areas and are directed to potentially more highly contaminated areas then exhausted to the stack at the inlet of the fans.
The construction of the gaseous waste system is consistent with conventional construction practice. No seismic design specifications or criteria were applied to equipment and components within the O\\
system. The construction of the stack, however, was designed consistent with the Uniform Building Code which includes a specification for a 11.3-2 HIO389-0125A-BX01
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i 0.0253 static horizontal load. A 1981 analysis of the stack by D'appalonia Associates concluded however that it has an adequate safety margin using USNRC Regulatory cuide 1.60 spectra anchored at 0.12g zero period acceleration.
f 11.3.3 RADIOACTIVE RELEASES Releases of radioactive materials from the gaseous waste system l
including plant ventilation have remained consistent over the past several years due to long periods of steady state operation of the plant and the absence of significant long term fuel defects.
Table 11-5 presents an annual average summary of discharges through the plant stack since 1985.
TABLE 11-5 l
TYPICAL ANNUAL ATHOSPilERIC RADIOACTIVE RELEASES (No Significant Puel Defects)
Radionuclides Annual Quantity (Ci)
Fission Activation Cases 4.0E4 Radiolodines 4.7E-2 Particulates (T) >8 days) 9.6E-3 Tritium 11.9 Radionuclides in Fission and Activation Cases Kr-83H 1410
- Kr-85H 1390 Kr-85 5.7
- Kr-87 4180
- Kr-88 5560 Kr-89 700 Kr-90 343 l
Xe-131H 7.9 l
Xe-133M 125 l
- Xe-133 2480 l
- Xe-135H 3500
- Xe-135 6350 Xc-137 1210
- Xe-138 10300 Xe-139 500 N-13 1560 l
- Actually measured.
Other nuclide release totals are calculated from known ratios between measured nuclide and nuclide of concern.
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11.4 SOLID WASTE MANAGEMENT SYSTEM i
Solid radioactive wastes (other than spent nuclear fuel) are composed of ion exchange resins, equipment that is in contact with the primary coolant, radioactive waste systems or neutron field from the reactor l
core, filters that filter radioactive gaseous or liquid streams and worn out protective clothing, plastic sheeting, tape, absorbant paper and the like which are used in working on, storing or isolating radioactive equipment. Storage of these solids is accomplished in three separate areas of the plant.
11.4.1 DESICW BASES The solid waste management system is designed to accept radioactive solids and store them safely in sufficient volume to accommodate several shipments.
Shielding is provided to minimize radiation exposure to workers while handling and shipping the wastes. System l
design capacity allows the accumulation of several years of wastes so that decay of the shorter lived material can occur prior to shipments, and to permit continued plant operation in the event shipment and i
disposal is temporarily not po s si bl e.
11.4.2 SYSTEM DESCRIPTION g
The solid waste management system consists of a water filled pool that accepts relatively highly radioactive components from the reactor such as fuel channels, control rod blades and other vessel internalst a 10,000 galion tank located in an underground vault containing the liquid waste management system that accepts spent resins from the reactor clean-up, condensate and liquid radioactive warte system demineralizerst and a separate building used for compacting and storing lower level solid wastes.
The spent fuel pool is utilized to. temporarily store, underwater, highly radioactive solid reactor components, Components are transferred to the pool from the reactor by use of a shielded fuel transfer cask.
When sufficient volume accumulates to accommodate at least one shipment, a transport cask is lowered into the pool, filled with i
waste, removed from the pool, decontaminated and loaded onto the shipping vehicle.
Resins, when exhausted, are sluiced to the 10,000 gallon tank, and, when close to full, shipments are made by pumping the resins from their underground location to casks on waiting shipping vehicles.
The Radwaste Processing Building located within a security fenced area of the plant (reference Chapter 2, Figure 2.3), has a shipping bay, crane, a compactor to compress compactible waste into 55 gallon drums, shielded areas for higher level wastes and sufficient storage capacity for several years accumulation of plant produced radioactive b
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Compactible wastes are normally accumulated in plastic bags, wastes.
transported to and stored in the compacting area prior to compression.
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O Other non compactible solids are simply boxed and stored until i
suf ficient volume accumulates to ef ficiently ship them to off-site disposal locations.
11.4.3 RADIOACTIVE SHIPMENTS Shipments of radioactive salids to off-site disposal locations have consistently averaged five per year over the last several years.
Table 11-6 provides a summary of these shipments since 1985.
t TABLE 11-6 TYPICAL SOLID WASTE SHIPMENTS e
Annual Average 3
Solid Tyoe Class Volume (ft]
Radioactivity (Cl)
Spent Resins and A
145 12 Filter Cartridges C
36 160 Dry compressible A
1890 2.8 Waste Irradiated Components B
9 290 C
74 500 1
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11.5 AREA. PROCESS AND EFFLUENT MONITORING AND SAMPLINC SYSTEMS The area, process and effluent monitoring systems installed at Big Rock point provide indications of the presence of radiation and radioactive material in areas, ventilation and liquid streams.
Monitors are provided to measure radiation fields and the presence of radioactive materials for normal operations and under accident conditions.
11.5.1 DESIGN BASES The area monitoring system detects, indicates and records gamma radiation in selected areas throughout the plant primarily for personnel protection. Two monitors, the new and spent fuel storage monitors, (reference Section 15.7.1 of this Updated FHSR) provide a safety actuation function closing the containment building ventilation valves (reference Section 6.2.4 of this Updated FHSR), when either of their setpoints are exceeded. Their setpoints are established between 5 and 20 mr/hr in accordance with 10 CPR 70.24. CPCo letter dated October 2, 1973 requested an exemption from the requirements of 10 CFR 70.24 which permits temporarily raising the alarm set points on these monitors above the 20 mR/ hour allowed. The exemption was granted by the Atomic Energy Commission (AEC) by letter dated February 26, 1974. The settings and alarm trip points along with the setpoint exemption criteria are reflected in the Technical Specifications.
Operability requirements during refueling are addressed in the Technical Specifications and in Section 9.1.4.3 of this Updated FHSR.
The area monitoring system includes a 20 point recorder and a common alarm indication in the control room when a predetermined radiation level is exceeded. Four selected monitors also have local alarms and indicating lights. Each monitor is capable of responding to gamma radiation levels over three orders of magnitude. The operable three decade range for each monitor is chosen to correspono to expected operational occurrences. An additional " accident monitor" is included which is capable of measuring radiation levels in the containment building during accidents including those which severely damage the reactor core and release large amounts of radioactive material to the containment. Its range covers seven orders of magnitude. Table 11-7 identifies each monitor, location, range, and function. The process monitoring system detects, indicates and records levels of radioactive materials in plant liquid and gaseous' effluent streams and other selected liquid streams. The system is designed to be able to detect radioactive materials in effluents below the continuous exposure concentration limits of 10 CFR 20.
For those monitors in non-effluent streams detection capability is provided to warn plant operators of changes in radiation levels. The system includes a common alarm indication in the control room when a predetermined level is exceeded.
Each monitor is capable of responding to concentrations of radioactive material over seven orders of magnitude. The operable range for each i
monitor is chosen to correspond to the expected level of radioactive O
material in each stream. An additional " accident monitor" is included which is capable of measuring radioactive materials releases from the 11.5-1 H10389-0125A-BX01
plant's ventilation stack during severe accident conditions. Its range covers twelve orders of magnitude. Table 11-8 identifies each monitor, location, range and function.
The design and construction of the area and process monitoring systems is consistent with the Uniform Building Code which includes a specification for a 0.025g static horizontal load. The containment and ventilation stack accident monitors are designed and constructed to the seismic classification of USNRC Regulatory Guide 1.60.
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5 TABLE 11-7 AREA MONITORING-SYSTEM Containment Local Vent Valve Indicator.
Clorure When Detector Range Alarm and Setpoint
' Element Detector Location ' Aux Function)'
mRfhr Alarm Light Exceeded RE-8261 Personnel Lock 0.1-100 No RE-8259 Spent Fuel Storage Pool 1-1000 Yes Yes (Criticality Monitor)
RE-8253 Condenser - Access Area 0.1-100 No RE-8268 Office Corridor 0.01-10 No RE-8264 Air Compresscr Room 0.01-10 No RE-8258 New Fuel Storage Area 1-1000 Yes Yes (Criticality Monitor)
RE-8254 Condensate Demineralizer Entrance 0.1-100 No RE-8269 Shop Area 0.1-100 No RE-8252 Control Room 0.01-10 No RE-8260 Sphere Laydown Area 0.1-100 No RE-8266 NW Wall Sphere Elevation 573' O.1-100 No RE-8257 Condenser 1-1000 Yes RE-8256 Laundry Room-0.01.10 No RE-8276 Exhaust Plenum Elevation 592' O.1-100
- No RE-8262 Locker Room 0.01-10 No RE-8255 Turbine Shield Wall 1-1000 No RE-8263 Rad Waste Area Behind Panel 1-1000 ves RE-8265 Access Control Entrance 0.01-10 No RE-8277 Emergency Condenser Vent - East 0.1-100 No (Effluent.via Condenser Vent)
RE-8270 Emergency Condenser Vent - West 0.1-100 No (Effluent via Condenser Vent).
7 RE-8280 Exterior Containment Penetration Rm 1-10 R/hr No 7
RE-8281 Exterior Containeeri Penetration Rm 1-10 R/hr (These last two are Containment High
' Range Accident Monitors)'
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O TABLE 11-8 PROCESS MONITORING SYSTEM Action Detector Monitor When Set Element Detector Location and Function Type Range Point Exceeded RE-RLO7 Off-Gas Monitor at the Outlet Ion Chamber 10-350C1/Sec Control Room Alarm RE-8251 of the Condenser Air Ejector and Off-gas Isol-ation after Pre-selected Delay RE-8282 Stack Gas Monitor at Entrance-Scintillation (Low Range) 4x10-6-4 C1/Sec Alarm i
RE-8284 to Liq Radweste Valve Gallery Ion Chamber and 6
Intrinsic Germanium (Hi Range).02-2x10 C1/Sec 7
RE-8275 Radioactive Waste Effluent to Scintillation 0-10 Cpm Alarm Discharge Canal in Liquid Redwaste Valve Gallery 7
RE-8272 Reactor Enc Cooling Water Scintillation 0-10 cpm Alarm in Reactor Cooling Water Heat Exchanger Room 7
RE-8273 Service Water from Reactor Scintillation 0-10 cpu Alarm in the Station Power Room 7
RE-8271 Main Condensate Demineralizer.
Scintillation 0-10 cpm Alarm Influent in Condensate Pump Room 7
RE-8274 Circulating Water Discharge Scintillation 0-10 cpm Alarm in the Lower Level of the Screen House:
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O 11.5.2 SYSTEM DESCRIPTION l
The area radiation monitoring system is made up of equipment capable of monitoring gamma levels at 20 locations throughout the Plant. Two monitors, Numbers 20 and 21, are dedicated to emergency condenser vent monitoring and are covered later. The location of the 19 remaining monitors functioning as area monitors are listed on Table 11-7.
i The monitor units are divided into 3 groups determined by their level monitoring rangest 0.01 mR/hr to 10 mR/hr; 0.1 mR/hr to 100 mR/hr; 1 mR/hr to 1000 mR/hr. The detectors are then positioned within the Plant where radiation levels are expected to be within these ranges.
Drawing 0740F30762 provides an elementary diagram of the system. The detection chain is made up of; a detector unit, located in a' specific area of the plant; area radiation monitor unit, located in the control room; recorder, located in the control room.
Four units.are also provided with local meters, alarms and alarm lights. These'four.
(4) units are identified on Table 11-7.
The scintillation detector uses a plastic scintillator as a gamma sensitive device. The scintillator is comprised'of a terphenyl-impregnated polystyrene base.
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The detector units are housed in wall mounted aluminum cases, which
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are water and vapor proof. The detector units are equipped with thermoelectrical coolers and forced air fan heat exchangers, along with a de power supply to provide current for thermoelectric coolers.
There are four area radiation monitor units located in the Main Control Panel.
Each unit contains amplifier channels (one for each of the five detectors connected to the unit). Each unit also contains i
the necessary de power supplies and trip units.
Mounted on the front panel of each unit are' indicators for each channel, indicating _ lights for power and fuse failures; alarm trip reset switch with corresponding alarm indicating light; trip test switch; trip adjustment and high-voltage test switch.
The amplifiers have solid-state circuitry except for the electrometer l
input tube. Sensitivity of the amplifiers is different for the radiation flux level range being monitored.
A trip on any detector channel results in a red indicating lamp l
lighting below its respective indicator.
The trip circuit must be reset manually.
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Trip set adjustments are located on the front panel of each monitor for each respective channel.
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Annunciation of trip signals is common for all but one channel, and' N-is indicated by the station annunciator labeled _" Area Monitoring Hi Radiation" on the Main Control Panel. The East Emergency Condenser Vent Monitor is indicated on a separate annunciator on the same panel.
Area radiation flux levels are recorded on a Strip Chart Recorder which is located adjacent to the area radiation monitors on the Main Control Panel.
The flux level recordings are primarily for trend indications and '
information.
The emergency condenser vent monitoring system is composed to two Detector Units, two area monitor units and two points on the. area monitor recorder. These monitors cause annunciation in the control room when gamma radiation levels exceed a-preselected level in the vent from the shell side of the emergency condenser (reference Chapter 6, Subsection 6.8.3 of this Updated FHSR).
Continuous monitoring of the air ejector (refer to Section.10.4.2 of this Updated FHSR), off gas radioactivity is provided by two independent, but identical, ion chamber systems. The systems are designed to detect noble gas fission products indicative of a fuel element rupture. One of the two channels provided should be in service and g
capable of initiating closure of the off-gas isolation valves during power operation if the condenser is receiving steam from the reactor.
Drawing 0740C44016 provides a line diagram of the off gas flow paths into the exhaust. stack.
The sampling system is designed to hold up the gas sample to allow' time for the decay of Nitrogen-16 and other short-lived activation
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gases. Samples are taken from the air ejector outlet to the stack.
The gas sample flow is controlled through a needle valve and accumulated l
in a volume chamber. After passing through the volume chamber,,the l
gas sample returns to the gas (suction) side (first stage) of the air l
ejectors. The main condenser vacuum " pulls" the continuous gas i
sample through the system.
l The off gas activity is monitored by two ion chamber detectors which
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are mounted on the outside of the volume chamber. The detector output l
is fed into a Logarithmic Radiation Monitor (LRM) where it is amplified and is applied to a Meter on the LRM, to a recorder located'in the main control room panel, and to the input of:both valve trip units.
i The trip circuit in the air ejector off-gas monitor has an alarm L
which annunciates in the control room if the off gas radioactivity-l reaches a predetermined level. At a higher predetermined level the i
air ejector off gas monitor trip circuit initiates action of the time i
delay switch, which will alarm another annunciator in the control
! [~ T room and trip the off gas shutoff valves closed after a preselected
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delay adjustable up to 15 minutes.
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HIO389-0125A-BX01 a
The off gas flow rate to the stack is measured indicated, and recorded.
The off gas monitoring equipment is calibrated to read curies per unit volume so the flow rate is necessary in order to determine the release rate in curies per second.
Power supply for Channel 1 is from the Reactor Protection MC Set Number 1.
Power supply for Channel 2 is from Reactor Protection MC Set Number 2.
A voltage regulator provides a 115 volt, ac stabilizing voltage to each of the units.
The adjustable time delay relay can be set up to 15 minutes. A preset radiation level will initiate the time celay switch. At the end of the set time, if the release rate has not dropped below the preset level, valve trip units will be activated to prevent the gases from entering the stack.
4 The Stack Cas Monitoring System (depicted on Drawings 0740040108 Sheets 1 and 2), also called the Radioactive Caseous Effluent Monitoring System (RGEM), receives a representative isokinetic sample of plant gaseous effluent from the plant stack during normal and accident conditions. The sample is monitored during normal conditions for particulates/radioiodine by continuous sampling and laboratory analysis and noble gases by a scintillation Beta Detector. During abnormally high levels of activity in the effluent; the system bypasses the normal range detectors, automatically takes a grab sample, switches to a high range Ionization Gamma Detector for noble gases activity and a high range Intrinsic Germanium Detector for particulate /radioiodine filter activity. Controls, indication, and recorders are located in the Control Room, at the main' equipment skid located at the entrance to the liquid radwaste area,~ and in the Chemistry Laboratory.
4 A back-up C.M. detector exists on the stack gas sampling discharge line to permit an alternate monitoring method for high noble gas release rates, in the event of the inoperability or the loss of the high range ionization gamma detector.
It is made operable as specified by the Technical Specifications.. Readout indication and a high level alarm exists in the plant air compressor room (the Operations Support Center during emergency plan implementation).
Sample air flow is provided by one of two 100% capacity diaphragm vacuum pumps mounted on the skid located at the entrance to the liquid radwaste area. The second pump operates automatically following a failure of the running pump. The pumps are positive displacement 0.5 scfm to 3.5 scfm.
The normal range particulate /radioiodine fi' tat is sized to handle airflow of 5 scfm maximum, 3.1 1 3 scfm nominal. The filter is cartridge type, made of activated charcoal to trap iodine, and paper i
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MIO389-0125A-BX01
to trap particulates, The filter elements are replaced weekly and taken to the Chemistry Laboratory for analysis.
The normal range NaI Camma Detector is located to monitor activity of the normal range particulate /radiolodine filter, The detector readout in the Control Room contains high and high-high radiation alarm points and an equipment failure alarm point. The high and high-high radiation setpoints are not used,'as all they indicate is-high filter activity, and the filter is removed and analyzed weekly no matter what the activity. The equipment failure alarm causes the green FAIL / RESET pushbutton on the readout to extinguish and activates the control room stack gas system trouble annunciator. This alarm can be triggered by loss of power, circuit failure, detector failure, or low background noise.
ThenormalrangescintillationghamberBetaDetectorhasanindication range of approximately 4 x 10 - Ci/see to 4 Ci/sec. High radiation, i
high-high radiation, and equipment failure alarms are provided with the scintillation chamber Beta Detector readout located in the j
Control Room. The high and high-high radiation setpoints are adjustable and located internal to the readout. The high alarm is normally set at approximately 30,000 pCi/sec. This limit is based on 10 CFR 50 Appendix I criteria. The high-high alarm is normally-set at approximately 3.3 Ci/sec. This limit is based on 10 CFR 20 criteria..
Trip test pushbuttons, internal to the readout, allow test of the alarm and control functions. The equipment failure alarm causes the
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green FAIL / RESET pushbutton on the readout to extinguish and activates the Stack Cas System Trouble Annuticiator. Again, the alarm-can be triggered by loss of power, circuit failure, detector failure, or low background noise.
A 150 mL Grab Sample Bottle is provided in the normal range monitoring loop. The bottle is normally isolated by an inlet solenoid valve and outlet solenoid valve. Upon a high radiation level alarm from the normal range noble gas detector isolation valves open while the normal flow valve closes to direct flow through the grab sample 1
bottle for approximately 30 seconds.. After 30 seconds, the isolation valves close to isolate the grab sample bottle for future analysis 1
while the normal flow valve opens to reestablish normal flow path conditions. The grab sample bottle is supplied with manual shutoff-valves for isolation of the sample and quick disconnects for bottle removal and replacement.
The Automatic Accident Filter is a filter change mechanism that is i
electric motor operated with a cam drive allowing step operation, changing one filter at a time. Actuation of filter change is made normally from the remote station in the Control Room.- The filter can also be changed from the local normal skid control station or from the filter panel pushbutton. A maximum of 45 filter cartridges can be stacked on the feeder to the filter change wheel. Upon filter change, the spent filter is dropped into a lead shielded cask for count verifi' cation at a later date or ultimate disposal.
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ThehighrangeIonizationChamberhaganindicationrangeof approximately 0.02 Ci/see to 2 x 10 Ci/sec. The chamber is shielded from background radiation by a lead shield and has a particulate /
radiolodine charcoal prefilter (bulk filter) to allow only noble gases to enter the chamber. The readout for the noble gas activity is located in t.he Control Room.
It contains high and high-high radiation alarm points and an equipment failure alarm point. The high and high-high alarms only alarm at the read out module. The alarm setpoints are based on Protective Action Guidelines (PAC) for Site Area Emergency and General Emergency respectively. The equipment failure alarm causes the green FAIL / RESET pushbutton to extinguish and activates the stack gas system trouble annunciator. The alarm
. can be triggered by loss of power, circuit failure, detector failure, or low background noise.
The Isokonetic Sample Nozzles in the plant vent stack are used fer the gaseous effluent monitoring system. The sample is drawn through two half inch stainless steel tubing runs down-the outside of the stack and over into the turbine building where the two lines combine into one 3/4 inch line. An isokinetic probe heater prevents freezing of the sample nozzles.
The sample flew tubing is heat traced and insulated from the' isokinetic nozzle to the entrance to the turbine building. Each line has a redundant heat tape installed in case of failure, The tapes are powered through a control box in the stack base.
Filter differential pressure indication for.both the normal range particulate / iodine filter and high range bulk filter is provided by a single Differential Pressure Indicator at the skid. This indicator j
actuates a filter high D/P alarm light (yellow) at both controllers,
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(at the skid and in the control-Room).
l System (pump discharge) pressure is indicated at the sample pumps..
discharge. This pressure indicating switch also provides indication of a high discharge pressure condition by illeminating a yellow alarm light.
Either of these alarms also annunciates the stack gas system trouble alarm in the Control Room.
j A flow Element, flow Indicating Transmitter, and flow Controller are located on the main equipment skid at the entrance to radwaste and control the cample flow rate through the system at 3.1 i.3 cfm. The transmitter provides local indication of sample flow rate and will actuate a yellow alarm light, start the standby sample pump, and stop the running pump at approximately 2.25 cim. If the low flow condition continues for a preset time delay, the system shuts itself down and annunciates the Stack Cas System Trouble Alarm in the Control Room.
Another flow Element, flow Indicating Transmitter and flow Controller are located on the main equipment skid at the entrance to radwaste 1
MIO389-0125A-BX01
and control the sample flow rate through the Automatic Accident Pilter at approximately 780 1 80 seca. The Transmitter provides local indication of sample flow rate and actuates a yellow alarm light at 700 seem low flow, which actuates the stack gas system
. trouble annunciator in the Control Room.
The process Liquid Monitor System, employing gamma scintillation detector channels, is provided to give indication of radioactivity trends in procest liquid streams normally containing radioactive liquids. It also serves to warn the operator of radioactivity in those process liquid streams that do not normally contain radioactive liquids.
The process liquid streams which shall be monitored aret Radioactive Waste System Effluent to Canal a.
b.
Reactor Enclosure Cooling Water I
c.
Main Condensate Demineralizer Influent d.
Circulating Water Discharge-e.
Service Water Return From Reactor Enclosure Alarms on Monitors a and d are set so as to warn the control room operator via a common annunciator when concentrations are present which exceed predetermined levels corresponding to technical i
specification limits. The alarms for b, e and e also annunciate on this same common annunciator. The set points for b,- e and e are set based on experience, and the alarms are to alert operators to_ unexpected changes of radioactivity levels in these process streams. The setpoint for e is normally set to detect approximately.3 x 104 microcuries/ milliliter of Cs-137.- Monitors on streams c, d and e are capable of detecting a concentration as low as 3-x 10~ microcuri:s of Cs-137 per milliliter of fluid. The reactor enclosure cooling water monitor is capable of detecting levels of chrome consistantly experienced, while the radioactive wast monitor is capable of detecting 5 x 10'g system effluent to canal microcuries per milliliter of Cs-137.
Radiation levels of all five (5) process liquid streams are recorded on a multipoint recorder which is located in the control room.
Monitors for the five (5) process liquid streams are of the fixed gamma type consisting of a scintillation detector mounted in a lead-lined stainless steel pig, a high-voltage power supply and a Linear Count Rate Meter (LCRM).
This detector has a scintillator (sodium-iodide thallium activated crystal), a light pipe, a photocathode and a photomultiplier tube.
Each is located in a detector shield.
The shielding consists of a large tee (commonly referred to as a
" pig") made of stainless steel and lined with lead. The detector is 11.5-10 i
HIO389-0125A-BX01 j
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mounted through the top of the tee.
The sample stream passes through this tee and around the detector well. The shielding tee reduces background radiation pickup by the detector.
The LCRM is designed to provide a direct indication of the average rate of arrival of random radiation pulses.
It provides a de voltage output proportional to the rate of occurrence of the pulses. The count rate is indicated on a direct-reading meter located on the rate meter panel. The count rate signal is also available for remote-7 meter and recorder connections. The range of measurement is 0 to 10 counts per minute in ten overlapping ranges. The meter is dual-scaled from 0 to 1, and from 0 to 2.5 to correspond to alternate positions of the range switch. Thelowestrange9easuresfrom0to250cpmand the highest range measures from 0 to 10 cpm.
A var'able alarm set point, which trips when a preset point on the meter scale is exceeded, can be_ adjusted by a potentiometer located on the front panel of the instrument. Also located on the front panel is a pulse height discriminator adjustment. The discriminator.
provides for rejection of, background noise and pulses of very low
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amplitude. Input sensitivity can be set to i 50 millivolts. An adjustable time constant variation, designated'" percent probable error" is also furnished on the front panel. The percent probable error control is an adjustment of the integrating time constant of the count-rate meter circuit. It has a range of 1% to 10% probable l
- error, l
l The linear count-rate meter is provided with several output connectors i
on the rear panel for remote indicating instruments and auxiliary-contacts for control purposes. It is designed for continuous operation.
The high voltage power supplies are designed for' continuous operation-and provide excellent regulation, low ripple content and-a high-degree of drift stability.
A recorder located in the control Room provides recorded information for radiation counts per minute from each process stream. The recorder scale is from o% to 100%. A full-scale recorder indication corresponds to the full-scale indicator cpm on the particular count-rate meter of the channel being monitored.
For the five inputs to the.
recorder, each intermittently recorded channe1~ reading must be-interpreted by referring to the channel range switch setting.
Drawing 0740044021 provides & schematic diagram of each process-monitoring system.
In the. event of a failure of the normal power source to either or both the area and process monitoring systems, electrical power is supplied.through a separate transformer fed from the on-site emergency diesel generator.
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