ML20148J825
| ML20148J825 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/22/1988 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20148J831 | List: |
| References | |
| DPR-50-A-137 NUDOCS 8803300399 | |
| Download: ML20148J825 (58) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _,
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%,1 UNITcD STATES NUCLEAR REGUL ATORY COMMISSION j
j VVASHING TON, D. C. 20555 e
e t ETRO,PO,LJT,A,h,,EDI_ SON COMPANY j
J ER SEY CENTRA,L,,P,09,E,R,,8,,LJ GHT,,C,0,M,P,A,NY PENNSYLVANIA ELECTRIC _ COMPANY GPU NUCLEAR CORP,0,R,ATJ,0N pp,C,KET, Np0.,50-289 THREE HILE, J,5,L App,pp,C,L,E,AR,,SJAJJpp,,,pNJJ,,Np,,,1
.AME NDM,E N,T, J p,,F,A,CJ,LJJ Y, _0 P E R A,TJ NG,,LJ C E NS E Amendment No.137 License No. DPR-50 1.
The Nuclear Regulatory Cornission (the Comission) has fcund that:
A.
The application for amendnent by GPU Nuclear Corporation, et al.
(the licensee) dated July 28, 1987, ccaplies with the standards and requirenents of the Atomic Energy Act of 1954, as amended (the Act),
and the Comissicn's rules and regulations set forth in 10 CFR Chapter I; C.
The facility will operate in conformity with the applicaticn, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this anendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be ccnducted in compliarce with the Comission's regulations; D.
The issuance of this amendment will ret be inimical to the common defense and security or to the health ano safety of the public; and E.
The issuance of this amendaent is in accordance with 10 CFR Part 51 of the Comission's regulaticns and all applicable requirea.ents have been satisfied.
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I 0803300399 080322 l
PDR ADOCK 05000209 P
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Accordingly, the license is amended hv channes to the Technical Speci#ications as indicated in the attachment to this license amendment, and paragraph P.c.(?) of Facility Operating License No. DPR-50 is hereby amended to read as follows:
(*) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.137, are herebv incorporated in the license.
GPlj Nuclear Corporation shall operate the facility in accordance with the Technical Snecifications.
3.
This license amendment is effective as of its date of issuance and shall implemented within 60 days of receipt.
FOR THE NtfrLEAR DEGl'lATORY COMMISSION r
oh
.StIolz, Director ro ect Directorate I-
- ision of Qeactor Pro,1ects I/II Office of Nuclear Peactor Requlation
Attachment:
Chances to the Technical Specifications Date of Issuance:
March 22, 1988
s ATTACHMENT TO LICENSE AMEhCMENT NO.137 FACILITY CPERATING LICENSE N0. OPR-50 00CKET 30,.,,5p,-2,89 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by anendment
- r. umber and contain vertical lines indicatin9 the area of change.
Perove Insert i,iii iv,vi 1,iii,1v,vi 1-3 1-3 1-4 1-4 1-4a 1-6, 1-6a 1-6 1-7, 1-7a 1-7 1-8 1-8 3-13 3-13 3-22 3-22 3-96 3-96 3-99 3-99 3-100 3-100 3-102 3-102 3-103 3-103 3-104 3-104 3-105 3-105 3-105a 3-105a 3-106 3-106 3-107 3-107 3-108 3-108 3-111 3-111 3-112 3-112 3-113 3-113 3-115 3-115 3-117 3-117 4-3 4-3 4-4 4-4 4-7a 4-7a 4-87 4-87 4-90 thru 4-116 4-90 thru 4-108 (see note l
atbottomof4-108) 1 5-1 5-1 6-18 6-18 6-19 6-19 6-19a l
l l
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TABLE OF CONTENTS Section Page TECHNICAL SPECIFICATIONS 1
DEFINITIONS 1 -1 1.1 RATED POWER 1 -1 1.2 REACTOR OPERATING CONDITIONS 1 -1 1.2.1 Cold shutdown 1 -1 1.2.2 Hot Shutdown 1 -1 1.2.3 Reactor Critical 1 -1 1.2.4 Hot Standby 1 -1 1.2.S Pcwer Operation 1 -1 1.2.6 Refueling Shutdown 1 -1 1.2.7 Refueling Operation 1-2 1.2.8 Refueling Interval 1-2 1.2.9
$tartup 1-2 1.2.10 8 Avg 1-2 1.2.11 Heatup-Cooldown Mode 1-2 1.2.12 Station. Unit, Plant, and Facility 1-2 1.3 OPERABLE 1-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument Channel 1-2 1.4.2 Reactor Protection System 1-2 1.4.3 Protection Channel 1-3 l
1.4.4 Reactor Protection System Logic 1-3 1.4.5 Engineered Safety Features System 1-3 1.4.6 Degree of Redundancy 1-3 1.5 INSTRUMENTATION SURVEILLANCE l-3 1.5.1 Trip Test 1-3 l
1.5.2 Channel Test 1-3 l
1.5.3 Channel Check 1-4 l
1.5.4 Channel Calibration 1 -4 l
1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER DISTRIBUTION 1-5 1.6.1 Quadrant Power Tilt 1-5 1.6.2 Reactor Power Imbalance 1-5 1.7 CONTAINMENT INTEGRITY l-5 1.8 FIRE SUPRESSION WATER SYSTEM 1-5 l
1.12 005E EQUIVALENT I-131 1-6 1.13 SOURCE CHECK l-6 l
1.14 SOLIDIFICATION 1-6 1.15 0FFSITE DOSE CALCULATION MANUAL l-6 1.16 PROCESS CONTROL PROGRAM l-6 1.17 UA3EOUS RADWA5TE TREATMENT SYSTEM 1-6 1.18 VENTILATION EXHAUST TREATMENT SYSTEM 1-6 1.19 PURGE-PURGING l -7 1.20 VENTING l -7 1.21 REPORTABLE EVENT l-7 I
1.22 MEMBER (5) 0F THE PUBLIC 1 -7 l
1 An. admen t No. M', Jf, 12(, 137 i
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-n
TABLE OF CONTENTS Section Page 3.16 SHOCK SUPPRESSORS (SNUBBERS) 3-63 3.17 REACTOR BUILDING AIR TEMPERATURE 3-80 3.18 FIRE PROTECTION 3-86 3.18.1 Fire Detection Instrumentation 3-86 3.18.2 Fire Suppression Water System 3-88 3.18.3 Deluge / Sprinkler Systems 3-89 3.18.4 CO2 Sy stem 3-90 3.18.5 Halon Systems 3-91 3.18.6 Fire Hose Stations 3-92 3.18.7 Fire Barrier Penetration Seals 3-94 3.19 CONTAINMENT SYSTEMS 3-95 3.20 SPECIAL TEST EXCEPTIONS 3-95a 3.20.1 Low Power Natural Circulation Test 3-95a 3.21 RADIOACTIVE EFFLUENT INSTRUMENTATION 3-96 l
3.21.1 Radioactive Liquid Effluent Instrumentation 3-96 3.21.2 Radioactive Gaseous Process and Effluent Monitoring 3-100 Instrumentation 3.22 RADI0 ACTIVE EFFLUENTS 3-106 3.22.1 Liquid Effluents 3-106 3.22.2 Gaseous Effluents 3-111 3.22.3 Solid Radioactive Waste 3-118 3.22.4 Total Dose 3-119 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING 3-120 3.23.1 Monitoring Program 3-120 3.23.2 Land Use Census 3-125 3.23.3 Interlaboratory Comparison Program 3-127 4
SURVEILLANCE STANDARDS 4-1 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4-11 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4-28 4.4 REACTOR BUILDING 4-29 4.4.1 Containment Leakage Tests 4-29 4.4.2 Structural Integrity 4-35 4.4.3 Deleted 4-37 4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY 4-39 CORF. COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM FEhIODIC TESTING 4.5.1 Emergency Loading Sequence 4-39 4.5.2 Emergency Core Cooling System 4-41 4.5.3 Reactor Building Cooling and Isolation System 4-43 4.5.4 Decay Heat Removal System Leakage 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 Control Rod Drive System Functional Tests 4-48 4.7.2 Control Rod Program Verification 4-50 fit Amendment No. Jf, Af,186, J21,137
a TABLE OF CONTENTS Section Page 4.8 MAIN STEAM ISOLATION VALVES 4-51 1.9 DECAY HEAT REMOVAL CAPABILITY - PERIODIC TESTING 4-52 9.9.1 Emergency Feedwater System - Periodic Testing 4-52 (Reactor Coolant Temperature Greater Than 250*F) 4.9.2 Decay Heat Removal Capability - Periodic Testing 4-52a (Reactor Coolant Temperature 250*F or Less) 4,10 REACTIVITY ANOMALIES 4-53 4.11 REACTOR COOLANT SYSTEM VENTS 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 Emergency Control Room Air Treatment System 4-55 4.12.2 Reactor Building Purge Air Treatment System 4-55b 4.12.3 Auxiliary and Fuel Handling Building Air Treatment System 4-55d 4.12.4 Fuel Handling Building ESF Air Treatment System 4-55f 4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 DELETED 4-56 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS (SNUBBERS) 4-60 4.18 FIRE PROTECTION SYSTEMS 4-72 4.18.1 Fire Protection Instruments 4-72 4.18.2 Fire Suppression Water System 4-73 4.18.3 Deluge / Sprinkler System 4-74 4.18.4 CO2 System 4-74 4.18.5 Halon Systems 4-75 4.18.6 Hose Stations 4-76 4.18.7 Fire Barrier Penetration Seals 4-76a 4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 Steam Generator Sample Selection and Inspection Methods 4-77 4.19.2 Steam Generator Tube Sample Selection and Inspection 4-77 4.19.3 Inspection Frequencies 4-79 4.19.4 Acceptance Criteria 4-80 4.19.5 Reports 4-81 4.20 REACTOR BUILDING AIR TEMPERATURE 4-86 4.21 RADIOACTIVE EFFLUENT INSTRUMENTATION 4-87 l
4.21.1 Radioactive Liquid Effluent Instrumentation 4-87 4.21.2 Radioactive Gaseous Process and Effluent Monitoring 4-90 Instrumentation j
4.22 RADIOACTIVE EFFLUENTS 4-95 4.22.1 Lfquid Effluents 4-95 4.22.2 Gaseous Effluents 4-101 4.22.3 Solid Radioactive Waste 4-107 4.22.4 Total Dose 4-108 4.23.1 Monitoring Program 4-117 4.23.2 Land Use Census 4-1 21 l
4.23.3 Interlaboratory Comparison Program 4-122 iv Amendment No. pr, y,36, fr, p't, g, X, p!r, y, pr, 99, pr, J21,137
LIST OF TABLES TABLE TITLE PAGE 1.2 Frequency Notation 1-8 2.3-1 Reactor Protection System Trip Setting Limits 2-9 3.1.6.1 Pressure Isolation Check Valves Between the Primary 3-15a Coolant System and LPIS 3.5-1 Instruments Operating Conditions 3-29 3.5-2 Accident Monitoring Instruments 3-40c 3.5-3 Post Accident Monitoring Instrumentation 3-40d 3.18-1 Fire Detection Instruments 3-87 3.21 -1 Radioactive Liquid Effluent Monitoring Instrumentation 3-97
- 3. 21 -2 Radioactive Gaseous Process and Effluent 3-101 Monitoring Instrumentation 3.23-1 Radiological Environmental Monitoring Program 3-122 3.23-2
.9eporting Levels for Radioactivity Concentration 3-126 in Environmental Samples 4.1 -1 Instrument Surveillances Requirements 4-3 4.1 -2 Minimum Equipment Test Frequency 4-8 4.1-3 Minimum Sampling Frequency 4-9 4.1 -4 Post Accident Monitoring Instrumentation 4-10a 4.19-1 Minimum Number of Steam Generators to be 4-84 Inspected During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 4.21 -1 Radioactive 8.iquid Effluent Monitoring 4-88 Instrumentation Surveillance Requirements l
4. 21 -2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements 4.22-1 Radioactive Liquid Waste Sampling & Analysis Program 4-96 4.22-2 Radioactive Gaseous Waste Sampling & Analysis Program 4-102 4.23-1 Maximum Values for the Lower Limits of Detection (LLD) 4-118 vi Amendment No. ff, J(, 340', Mti, JAT,137
m l
foMhs the automatic system that protects the reactor by control rod trip.
It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod drive control protection trip breakers, and activating relays or coils.
1.4.3 PROTECTION CHANNEL A PROTECTION CHANNEL as shown in Figure 7.1-1 of the updated FSAR (one of three or l
l one of four independent channels, complete with sensors, sensor power supply units, l
amplifiers, and bistable modules provided for every reactor protection safety l
parameter) is a combination of instrument channels forming a single digital output to the protection system's coincidence logic.
It incluc'es a shutdown bypass circuit, a protection channel bypass circuit and a reactor trip module.
1.4.4 REACTOR PROTECTION SYSTEM LOGIC This system utilizes reactor trip module relays (coils and contacts) in all four of the protection channels as shown in Figure 7.1-1 of the updated FSAR, to provide l
reactor trip signals for de-energizing the six control rod drive trip breakers.
The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic.
Each element of the one-out-of-two-times-two logic is controlled by a separate set of two-out-of-four logic contacts from the four reactor protection channels.
1.4.5 ENGINEERED SAFETY FEATURES SYSTEM This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as shown in Figure 7.1-4 of the updated FSAR.
The logic sub-system is wired to provide l
appropriate signals for the actuation of redundant engineered safety features equipment on a two-of-three basis for any given parameter, 1.4.6 DEGREE OF REDUNDANCY The difference between the number of operable channels and the number of channels which, when tripped, will cause an automatic system trip.
1.5 INSTRUMENTATION SURVEILLANCE 1.5.1 TRIP TEST A TRIP TEST is a test of logic elements in a protection channel to verify their associated trip action.
1.5.2 CHANNEL TEST A CHANNEL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practical to verify OPERABILITY, including alam and/or trip functions.
1-3 i
1.5.3 CHANNEL CHECX A CHANNEL. CHECX shall be the qualitative assessment of channel behavior during operation by observation.
This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.
1.5.4 CHANNEL CALIBRATION An instrument CHANNEL CALIBRATION is a test, and adjustment (if necessary),
to establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values.
Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include the channel test.
1.5.5 HEAT BALANCE CHECX A HEAT BALANCE CHECK is a comparison of the indicated neutron power and core
{
thermal power.
1.5.6 HEAT BALANCE CALIBRATICN A HEAT BALANCE CALIBRATION is an adjustment of the power range channel amplifiers l;
output to agree with the core thermal power as defined by a weighted primary and secondary heat balance considering heat losses.
The weighting factor, a is shown in the figure below as a function of power level.
The equations below define the value of a as a function of power level and the use of a in determining the core thermal power.
1.0 o
58 50e 50 0
i t
0 POWER 14
Core Thermal Power = a
'Qsec + (1-a ) Qprim for, POWER less than or equal to 15%, a=0 POWER greater than 15% AND less than 50%
POWER - 15
=
a 85 WHERE:
POWER =
Oprim 100 QmaX POWER greater than 50% AND less than 100%
POWER - 15 WHERE:
POWER = Osec 100 a =
85 Qmax POWER greater than or equal to 100%,
a=1 l
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e
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yy
(Definitions 1.9 - 1.11 have been deleted).
1.12 OOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.
The thyroid dose conversion f actors used for this calculation shall be those listed in Table III of TID 14844, "Calculation of Distance Factors for Power and Test Reactor Sites".
[Or in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.]
1.13 SOURCE CHECX A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
1.14 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid treatment systems to a uniformly distributed, monolithic immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (f ree-s ta ndi ng).
1.15 0FFSITE DOSE CALCULATION MANUAL (00CM)
An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used in the calculation of off-site doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints, and in the conduct of the environmental radiclogical monitoring program.
1.16 PROCESS CONTROL PROGRAM (PCP)
The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes f rom liquid systems is assured.
1.17 GASEOUS RA0 WASTE TREATMENT SYSTEM The GASEOUS RADWASTE TREATHENT SYSTEM is the system designed and installed to l
reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the erNironment.
1.18 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATNENT SYSTEM is any system designed and installed to reduce gaseous radiofodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing f odines or particulates from the gaseous exhaust system prior to the release to the envircrrnent.
Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATNENT SYSTEMS.
1-6 Amendment No. JE,137
1.19 PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is required to purify the confinement.
1.20 VENTING YENTING is the controlled process of discharging air as gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is not provided.
Vent used in system name does not imply a VENTING process.
1.21 REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
1.22 MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant.
This category does not include employees of the GPU System, GPU contractors or vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
1 1 -7 Amendment No. J(,137
TABLE 1.2 FREQUEMCY NOTATION NOTATION FREQUENCY S
Shif tly (once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) 0 Oaily (once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
W Weekly (once per 7 days)
M Monthly (once per 31 days)
Q Quarterly (once per 92 days)
S/A Semi-Annually (once per 184 days)
R Refueling Interval P S/U Prior to each reactor startup, if not done during the previous 7 days P
Completed prior to each release l
N/A (NA)
Not applicable E
Once per 18 months i
The Surveillance Requirements shall be performed within the specified l
time interval with:
A.
A maximum allowable extension not to exceed 25% of the surveillance interval, and l
B.
A total maximum combined interval time for any 3 consecutive tests not to exceed 3.25 times the specified surveillance in terval.
l l
i 1-8 Amendment No. g.137 a
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3.1.6.9 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas vent header and from which coolant can be returned to the reactor coolant system shall not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1.6.1, 3.1.6.2, 3.1.6.3, 3.1.6.4, 3.1.6.5, 3.1.6.6 or 3.1.6.7, except tha t such losses when added to leakage shall not exceed 30 gpm.
If leakage plus losses exceeds 30 gpm, the reactor shall be placed in HOT SHUT 00WN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. l 3.1.6.10 Operating conditions of POWER OPERATION, STARTUP and HOT SHUT 00WN apply to the operational status of the high pressure isolation valves between the primary coolant system and the low pressure injection system.
During all operating conditions in this specification, all pressure a.
isolation valves listed in Table 3.1.6.1 that are located between the primary coolant system and the LPIS shall function as pressure isolation devices except as specified in 3.1.6.10.b.
Valve leakage shall not exceed the amount indicated in Table 3.1.6.1.(a) b.
In the event that integrity of any high pressure isolation check valves specified in Table 3.1.6.1 cannot be demonstrated, reactor operation may continue provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition. (b)
If Stacification 3.1.6.10.a or 3.1.6.10.b cannot be met, an orderly c.
i shutdown shall be accomplished by achieving HOT SHUTDOWN witnin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTOOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Bases l
Any leak of radioactive fluid, whether from the reactor coolant system primary boundary or not, can be a serious pr:blem with respect to in-plant radioactive contamination and required cleanup or, in the case of reactor coolant, it could j
develop into a still more serious problem and, therefore, the first indications of such leakage will be followed up as soon as practical.
The unit's makeup system has the capability to makeup considerably more than 30 gpm of reactor coolant leakage.
i Water f rwentory balances, monitoring equipment, radioactive tracing, boric acid crystalline deposits, and physical inspections-can disclose reactor coolant leaks.
i (a)
For the purpose of this specification integrity is considered to have been t
demonstrated by meeting Specification 4.2.7.
l (b)
Motor operated valves shall be placed in the closed position and power supplies deenergized.
Amendment No. g, 137 3-13 Order dtd. 4/20/81 l
l l
~.
Core flood tank (CFT) vent valves CF-V3A and CF-V3B shall be closed e.
and the breakers to the CFT vent valve motor operators shall be tagged open, except when adjusting core flood tank level and/or pressure.
Specification 3.0.1 applies.
3.3.1.3
_ Reactor Building Spray System and Reactor Building Emergency
_ Cooling System The following components must be OPERABLE:
j.
Two reactor building spray pumps and their associated spray a.
nozzles headers and two reactor building emergency cooling fans and associated cooling units (one in each train).
Specification 3.0.1 applies.
b.
The sodium hydroxide (NaOH) tank shall be maintained at 8 ft.:
6 inches lower than the BWST level as measured by the BWST/NaOH tank differential pressure indicator.
The NaOH tank concentration shall be 10.0
.5 weight percent _(%).
c.
All manual valves in the discharge lines of the sodium hydroxide tank shall be locked open.
3.3.1.4 Cooling Water Systems - Specification 3.0.1 applies, Two nuclear service closed cycle cooling water pumps must be a.
l b.
Two nuclear service river water pumps must be OPERABLE.
l c.
Two decay heat closed cycle cooling water pumps must be OPERABLE.
l d.
Two decay heat river water pumps must be OPERABLE.
e.
Two reactor building emergency cooling river water pumps must be OPERABLE.
l 3.3.1.5 Engineered Safeguards Valves and Interlocks Associated with the Systems in Specifications 3.3.1.1, 3.3.1.2, 3.3.1.3, 3.3.1.4 are OPERABLE.
Specification 3.0.1 applies.
3.3.2 Maintenance shall be Otowed during power operation on any I
component (s) in the makaup and purification, decay heat, RB l
emergency cooling water, RB spray, CFT pressure instrumentation, CFT level instrumentation, BWST 1evel instrumentation, or cooling water systems which will not remove more than one train of each system from service.
Components shall not be removed from service so that the affected system train is inoperable for more than 72 i
consecutive hours.
If the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor i
shall be placed in a COLD SHUTDOWN condition within twelve hours.
3-22 Amendment No. J6 )(, JN(,137
3.21 RADI0ACTfVE EFFLUENT fNSTRUMENTATION 3.21.1 RADI0 ACTIVE LIOUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3,21.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.21-1 shall be OPERABLE with their alarm / trip setpoints set to ensure j
that the limits of Specification 3.22.1.1 are not exceeded.
The alarm / trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CM).
APPLICABILITY:
At all times
- ACTION:
With a radioactive if quid effluent monitoring instrumentation channel a.
alam/ trip setpoint less conservative than required by the above specification, irrrn?diately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperabl e.
b.
With less than the minimum number of radioactiw liquid effluent monitoring instrumentation channels OPERARLE, take the ACTION shown in Table 3.21-1.
Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.
- For FT-84, and RM-L6, operability is not required when discharges are positively controlled through the closure of WDL-V257
- For RM-L12 and associated IWTS/IWFS flow interlocks, operability is not required when discharges are positively controlled through the closure of IW-Y72, 75 and IWW280, 281.
- For FT-146, operability is not required when discharges are positively controlled through the closure of WDL-Y257, IW-Y72, 75 and IW-Y280, 281.
BASES The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases.
The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alam/ trip will occur prior to exceeding the limits of 10 CFR Part 20.
3-96 Amendment No. Jf,,M,137 t
TABLE 3.21-1 (continued)
TABLE NOTATION ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue, provided that prior to initiating a release:
1.
At least two independent samples are analyzed in accordance with Specifications 4.22.1.l A & B and; 2.
At least two technically qualified members of the Unit staff independently verify the release rate calculations and verify the discharge valve ifneup.
3.
Operations and Maintenance Director Unit i shall approve each release.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 20 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may commence or continue provided that grab samples are collected ard analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least lx10-/ microcuries/ml, prior to initiating a release and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during release.
ACTION 21 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, radioactive effluent releases via this pathway may continue, provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
Pump curves may be used to estimate flow.
3-99 Amendment No. J2, JHf,137
l l
3.21.2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.21.2 The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.21-2 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.22.2.1 are not exceeded.
The alarm / trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (00CM).
APPLICABILITY:
As shown in Table 3.21-2.
ACTION:
a.
With a radioactive gaseous process or effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive effluents monitored by the affected channel or declare the channel inoperable, b.
With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.21-2.
Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-Annual Effluent Release Report why the inoperability was not corrected in a timely manner.
BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases.
The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The low range condenser offgas noble gas activity monitors also provide data for determination of steam generator primary to secondary leakage rate.
Channel operability requirements are based on an ASLB Order dated October 31, 1984.
3-100 Amendment No. X, 101, 137
TABLE 3.21-2 (Continued)
!I l
RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION a
z
?
MINIMUM CHANNELS se i !*
INSTRUMENT OPERABLE APPLICABILITY ACTION
!k 3.
Containment Purge Monitoring System
- C a.
Noble Gas Activity Monitor 1
27
{
(RM-A9) 4 b.
Iodine Sampler (RM-A9) 1 31 c.
Particulate Sampler (RM-A9) 1 31 d.
Effluent System Flow Rate 1
26 Measuring Device (FR-148) e.
Sampler Flow Rate Monitor 1
26 i
l l
f i
i n
3 TABLE 3.21-2 (Continued)
[
RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION O'
m MINIRIM I
CHANNELS
{
INSTRUE NT OPERABLE APPLICABILITY ACTION 4.
Condenser Vent System l
g a.
Low Range Noble Gas Activity Monitor 2(l) 32
]
(RM-ASLo and Suitable Equivalent)
'l
)
NOTE (1): For one of the channels, an operable chanr.el may be defined for purposes of this specification and 4.21.2 only as a suitable equivalent monitoring system capable of being placed in service within one hour.
A suitable equivalent system shall include instrumentation with comparable sensitivity and response time to the RM-ASLo monitoring channel. When the equivalent monitoring system is in service, indication will be continuously available to the operator, either through indication and alarm in the control room or through communication with a designated individual continuously observing i
g local indication.
l i
l l
i i
l l
I i
TABLE 3.21-2 (Continu:d) g RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION I
3
,3 MINIMUM CHANNELS j
5 INSTRUE NT OPERAELE APPLICABILITY ACTION M
5.
Auxiliary and Fuel Handling i~
Building Ventilation System e*
a.
Noble Gas Activity 1
27 i
Monitor (RM-A8) or (RM-A4 and RM-A6) i O
b.
Iodine Sampler (RM-A8) or 1
31 (RM-A4 and RM-A6) l c.
Particulate Sampler 1
31
]
(RM-A8) or (RM-A4 and RM-A6) d.
Effluent System Flow Rate 1
26 Measuring Devices (FR-151, or FR-149 and FR-150) j g
e.
Sampler Flow Rate Monitor 1
26 i
j 6.
Fuel Handling Building ESF Air Treatment System i
a.
Noble Gas Activity Monitor 1(2) 27, 33 j
(RM-A14 or Suitable Equivalent) b.
Iodine Cartridge N/A(3) 31, 33 c.
Particulate Filter N/A(3) l 31, 33 d.
Effluent System Flow (UR-1104A/B) 1 26, 33 i
e.
Sampler Flow Rate Monitor 1
26, 33 l
NOTE 2: Until the beginning of the 7R refueling outage, a suitable equivalent OPERABLE channel may l
be defined for item 6.a of this specification and specification 4.21.2 (Table 4.21-2, item 6.a) as a system capable of alerting the Control Room b alarm or voice communication and ci.antconditions.pable of measuring the full range of normal and calcu ated accident releases for existing p
NOTE 3: No instrumentation channel is nrovided However for determining operability, the equipment named must be installed and functional,or the ACTION applies.
TABLE 3.21 -2 (Continued)
TABLE NOTATION
- At all times.
- During waste gas holdup system operation.
- Operability is not required when discharges are positively controlled through the closure of WDG-V47, and RM-A8 and FT-151 are operable, s
- During Fuel Handling Buf1 ding ESF Air Treatment System Operation.
- At all times during containment purging.
H At all times when condenser vacuum is established.
ACTION 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release:
l 1.
At least two independent samples of the tank's contents are analyzed in accordance with Table 4.22-2, Item A, and l
2.
At least two technically qualified members of the Unit staff independently verify the release rate calculations and verify the discharge valve lineup.
l 3.
The Operations & Maintenance Director, Unit 1, shall approve each j
release.
Otherwise, suspend release of radioactive effluent via this pathway.
ACTION 26 With the number of channels OPERABLE less than required by the Minimum l
Channels OPERABLE requirement, effluent releases via this pathway may l
continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the initial samples are analyzed for gross activity (gamma scan) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the channel has been declared inoperable.
If RM-A9 is declared inoperable, see also Specification 3.5.1, Table 3-5.1, Item C.3.f.
l ACTION 30 1.
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, a grab sample shall be I
collected and analyzed for the inoperable gas channel (s) at least I
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both channels inoperable, a grab sample shall be collected and analyzed for the inoperable gas channel (s):
(a) at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations.
(b) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations (e.g. Feed and Bleed).
3-105
TABLE 3.21 -2 (Continued)
TABLE NOTATION ACTION 30 2.
If the inoperable gas channel (s) is not restored to service within (CONT'D.)
14 days, a special report shall be submitted to the Regional Administrator of the 14RC Region I Office and a copy to the Director, Office of Iaspection and Enfort.9 ment within 30 days of declaring the channel's) inoperable.
The report shall describe (a) the cause of the monitor inoperability, (b) action being taken to restore the instrument to service, and (c) action to be taker, to prevent recurrence.
ACTION 31 With the number of channels OPERABLE less than reautred by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that within four hours after the channel has been l
declared incperable, samples are centinuously collected with auxiliary sampling equipment.
ACTION 32 With the number of channelk' OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathvay may continue for up to 28 days, provfded that one OPERABLE channel remains in service or is placed in service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
After 28 days, or if one OPERABLE channel does not remain in service or is not placed in service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the provisions of 3.0.1 apply.
ACTION 33 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable channel to OPERABLE status within 7 days, or prepare and submit a special report within 30 days outlining the action (s) taken, the cause of the inoperability, and plans and schedule for restoring t% system to OPERABLE status.
3-105a Amendment No. Jf, 383, %, % 137
3.22 RADI0 ACTIVE EFFLUENTS 3.22.1 LIQUID EFFLUENTS 3.22.1.1 CONCENTRATION LIMITING CONDITION FOR OPERATION i
3.22.1.1 The concentration of i;dioactive material released at anytime from the unit to unrestr cted areas (see Figure 5-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II Column 2 for radionuclides other than dissolved or entrainad noble gases.
For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10,> uCf /cc total activity.
APPLICABILITY: At all times ACTION:
a.
With the concentration of radioactive material released from the unit to unrestricted areas exceeding the above limits, immediately restore concentration within the above limits, b.
If action "a" cannot be met, then be in:
1.
At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2.
At least HOT FHUTOOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 3.
At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
BASES This specification is provided to ensure that the concentration of radioactive materials rel2ased in liquid waste effluents from the unit to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II.
This limitation provides additional assurance that the levels of radioactive material in bodies of water outside the site will not result in exposures with (1)
...e Section II.A design objectives of Appendix I,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106 (e) to the population.
The concentration limit for noble gases is based upon the assumption the Xe-135 is the contrciling ridioisotope and its MPC in air (submersion) was conver ted to an equivalent concentration in water using the methods described in Internaticaal Commission on Radiological Protecticn (ICRP)
Publication 2.
3-106 Amendment No. J4,137
--m~
RADI0AC*IVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.22.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from the unit to the site boundary (see Figure 5-4) shall be limited:
a.
During any calendar quarter to < 1.5 mrem to the total body and to s 5 mrem to aliy organ, b.
During any calendar year to < 3 mrem to the total body and to 1 10 mrem to any-organ.
APPLICABILITY:
At all times ACTION:
a.
With the calculated dose from the release af radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administratn within 30 days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar quarters so that the cumulative dose or dose commitment to any individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ.
This Special Report shall also include (1) the result of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.
BASES This specification is provided to implement the requirements of Sections II.A.
III. A. and IV. A of Appendix I,10 CFP. Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable asaurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in exess of the requirements of 10 CFR 20.
The dose calculations in the ODCM implement 3-107 Amendment No. J(, %,137
the requirements in Section III. A. of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the act" ' exposure of a MEMBER OF THE PUBLIC l
through appropriate pathways is unlikely to be substantially underestimated.
The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liouid effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"
Revision 1, October,1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April,1977.
NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113, 3-108 Amendment No. Jf, j)f,137
r RADIOACTIVE EFFLUENTS 3.22.2 GASEOUS EFFLJENTS 3.22.2.1 DOSE RATE LIMITING CONDITION FOR OPERATIONS 3.22.2.1 The dose rate due to ev.joactive materials released in gaseous effluents from the site (see Figure 5-3) snt11 be limited to the following:
a.
For noble gases: less than or equal to 500 mrem /yr to the total body and less than or equal to 3030 mrem /yr to the skin, and b.
For I-131, I-133, tritium and all radionuclides in particulate form with half lives greater than 8 days:
less than or equal to 1500 mrem /yr to any organ.
, APPLICABILITY: At all tims.
ACTION:
a.
With the release rate (s) exceeding the above limits, immediately decrease the release rate to comply with the above limit (s).
b.
If action "a" cannot be met, then be in:
1.
At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2.
At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 3.
At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
BASES The specification is provided to ensure that the release rate at anytime at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II.
These limits provide reasonable assurance that radioactive material discharged in gaseous effluents v:ill not result in the exposure of a MEMBER OF THE PUBLIC in an unrastricted area, either within or outside the site boundary, to annual average concentrations excecding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).
For MEMBERS OF THE PUBLIC who may at times be within the site boundary, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.
The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the site boundary to less l
than or squal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin.
These relesse rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the
- ow-milk-infar.t pathway to less than or equal to 1500 mrem / year for the nearest cow to the plant.
3-111 Amendment No. Jf.137
4 RADIOACTIVE EFFLUENTS GASEOUS EFFLUENTS DOSE-NOBLE GASES LIMITING CONDITION FOR OPERATION 3.22.2.2 The air dose due to noble gases released in gaseous effluents from the unit to areas at and beyond the site boundary (see Figure 5-3) shall be limited to the following:
a.
During any calendar quarter:
< 5 mrad for gamma radiation and 5 10 mrad for beta radiati7n and, b.
During any calandar year:
< 10 mrad for gamma radiation and 5 20 mrad for deta radiatioii.
APPLICABILITY:
At all times.
ACTION:
a.
With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the abo'te Ifmits.
BASES This specification applies to the release of radioactive materials in gaseous eff'. ants from TMI-1.
This specification is provided to implement the requirements of Section II.8, III. A and IV. A of Appendix I,10 CFR Part 50.
The limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I.
The l
ACTION statements provide the required operating flexibility and at the same l
time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept
(
"as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through the l
appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actu:: release rates of radioactive noble gases in I
gaseous effluentc are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man frem Routine Release of j
Reactor l
3-112 Amendment No. J2, % 137
RADIOACTIVE EFFLUENTS DOSE - 10 DINE-131,10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.22.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, l
tritium, and all radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents released from the unit to areas at and beyond the site bound 6ry (See Figure 5-3) shall be limited to the following:
a.
During any calendar quarter: 5 7.5 mrem to any organ, and b.
During any calendar year: 5 15 mrem to any organ.
APPLICABILITY:
At all times.
ACTION:
With the calculated dose from the release of fodine-131, fodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
BASES This specification applies to the release of radioactive materials in gaseous effluents from THI-1.
l This specification is provided to implement the requirements of Section II.C, III. A and IV. A of Appendix I,10 CFR Part 50.
The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.
The ACTION statement provides the required operating flexibility and at the same time I
I implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III. A of Appendix I that confonnance with the guides of Appendix ! be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to l
be substantially underestimated.
The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of 3-113 Amendment J{ 12I,137 i
RADf0 ACTIVE EFFLUENTS 3.22.2.4 GASEOUS RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.22.2.4 The GASEOUS RA0 WASTE TREATMENT SYSTEM and the VENTILATICN EXHAUST TREATMENT SYSTEM shall be OPERABLE.
The appropriate portions of the GASECUS RA0 WASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in the gaseous waste prior to their discharge when the r:cnthly projected gaseous l
effluent air doses due to untreated gaseous effluent releases from the unit (see Figure 5-3) would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation.
The appropriate portions of the VENTILATICN EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the monthly projected doses due to gaseous effluent l
releases f rom the site (see Figure 5-3) would exceed 0.3 mrem to any organ.
APPLICABILITY:
At all times.
ACTION:
a.
With the GASEOUS RA0 WASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the NRC Region !
Administrator within 30 days, a Special Report which includes the following information:
1.
Identification of the inoperable equipment or subsystems and the reason for inoperability, 2.
Action (s) taken to restore the inoperable equipment to CPERABLE status, and 3.
A summary description of action (s) taken to prevent a recurrence.
BASES The use of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATICH EXHAUST TREATMENT SYSTEM ensures that gaseous effluents are treated as appropriate l
prior to release to the environment.
The appropriate portions of this system provide reasonable assurance that the releases of radioactive materials in l
gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirenents of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous ef fluents, l
l l
3-115 AmendmentNo.J/,137 l
RADIOACTIVE EFFLUENTS 3.22.2.6 WASTE GAS DECAY TANK,5 LIMITING CONDITION FOR OPERATION 3.22.2.6 The quantity of radioactivity contained in each waste gas decay l
tank shall be limited < 8800 curies noble gases (considered as Xe-133).
APPLICABILITY:
At all times.
ACTION:
With the quantity of radioactive material in any waste gas decay a.
tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.
BASES Restricting the quantity of radioactivity contained in each waste gas decay
(
tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest exclusion area boundary will not exceed 0.5 rem.
This is consistent with Standard Review Plan 15.7.1, "Waste Gas System Failure."
1 l
l
[
3-117 Amendment No. g 137
k TABLE 4.1-1 o
[
INSTRUENT SURVEILLANCE REQUIREENTS
~
?o l
f CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS R
1.
Protection Channel NA M
NA Coincidence Logic 2.
Control Rod Drive Trip NA M
NA (1)
Includes independent testing Breaker and Regulating of shunt trip and undervoltage Rod Power SCRs trip features.
C 3.
Power Range Amplifier D(1)
NA (2)
(1)
When reactor power is greater than 15%.
(2)
When above 15% reactor poter run a heat balance check once per shif t.
Heat balance J
calibration shall be performed whenever heat balance exceeds indicated neutron power by more than two percent.
4.
Power Range Channel S
M M(1)(2)
(1)
When reactor power is greater j
than 60% verify imbalance i
using incore instrumentation.
1 (2)
When above 15% reactor power calculate axial offset upper and lower chambers af ter each
~
startup if not done within the j
previous seven days.
S.
Intermediate Range Channel S(l)
PS/U NA (1)
When in service.
6.
Source Range Channel S(1)
PS/U NA (1 )
When in service.
7.
Reactor Coolant Temperature S
M R
Channel 4
i
i 3
TABLE 4.1-1 (Continued) i y
CHANNEL DESCRIPTION CilECK TEST CALIBRATE REMARKS if 8.
High Reactor Coolant S
M R
~
Pressure Channel k
9.
Low Reactor Ccolant S
M R
C Pressure Channei
~
10.
Flux-Reactor Coolant Flow S
M R
Comparator
- 11. Reactor Coolant Pressure S
M R
Temperature Comparator
- 12. Pump Flux Comparator S
M R
- 13. High Reactor Buf1 ding S
M R
Pressure Channel
- 14. High Reactor Injection NA Q
NA Logic Channel a
1.
- 15. High Pressure Injection
]
Analog Channels a.
Reactor Coolant S(l)
M R
(1)
When reactor coolant system is Pressure Channel pressurized above 300 psig or Tav is greater than 200*F o
i l
- 16. Low Pressure Injection NA Q
NA Logic Channel i
j 17.
Low Pressure Injection NA Q
NA Analog Channels 1
a.
Reactor Coolant S(1)
K R
(1)
When reactor coolant system is Pressure Charnel i
pressurized above 300 psig or Tav is greater than 200*F 18.
Reactor Building Emergency NA Q
NA Cooling and Isolation System l
Logic Channel l
TABLE 4.1-1 (Continu:d) 3 S
CHANNEL DESCRIPTIbN CHECK TEST CALIBRATE REMARKS E
49.
Saturation Margin Monitor S(1)
M(1)
R (1) When Tave is greater than 525'F.
"ea l
50.
Emergency Feedwater Flow NA M(1)
R (1) When Tave is greater than 250*F.
g Instrumentation
- 51. Heat Sink Protection System a.
EFW Auto Initiation (1) Includes logic test only.
k Instrument Channels 1.
Loss of Both Feedwater Pumps NA Q(1)
R C
2.
Loss of All RC Pumps NA Q(1)
R 3.
Reactor Building Pressure NA Q
R J
4.
OTSG Low Level W
Q R
b.
WW Isolation OTSG Low Pressure NA Q
R
)
I c.
EFW Control Valve Control System 1.
OTSG Level Loops W
Q R
2.
Controllers W
NA R
?
- i;'
d.
HSPS Train Actuation Logic NA Q(1)
R
- 52. Backup Incore Thermocouple Display M(1)
NA R
(1) When Tave is greater than 250*F.
53.
Chlorine Detection System W
M R(1)
(1) Calibration is a one concentration Instrumentation point check (need not be traceable to NBS standards) l l
1 4.21 RA0iOACT!VE EFFLUENT TNSTRLNENTATf 0N 4.21.1 RA0!0 ACTIVE LIOUID EFFLUEN1 INSTRlNENTATION SURVEILLANCE REQUIREMENTS 4.21.1 Each radioactive liquid effluent monitoring instruantation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations during the MODES and at the frequencies shown in Table A.21-1.
4 87 Amndant No. J/,137 i
1 4.21.2 RA0f 0ACTfVE GASEOUS PROCESS AND EFFLUENT MOHfiORf NG fNSTRUMENTATf 0H SURVEILLANCE RE0VIREMENTS 4.21.2 Esch radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations at the frequencias shown' in Table 4.21-2.
e 4-90 knendnent No JE,137
-_-----------J
g TABLE 4.21-2 h
RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MON'TORING INSTRlMENTATION SURV
$n g
CHANNEL SOURCE CHANNEL CHANNEL INSTRtMENT CHECK CHECK CALIBRATION TEST APPLICABILITY k-1.
Waste Gas Holdup System a.
Noble Gas Activity Monitor P
P E(3)
Q(1)
(RM-A7 )
~3 b.
Effluent System Flow Rate P
N/A E
Q Measuring Device (FT-123) 2 Waste Gas Holdup System Explosive Gas Monitoring System a.
Hydrogen Monitor D
N/A Q(4 )
M b.
Oxygen Monitor D
N/A Q(S)
H g
3.
Containment Purge Vent System a.
Noble Gas Activity Monitor D
P E(3)
M(1) i (M-A9 )
b.
W N/A N/A N/A t
c.
Particulate Sampler (RM-A9)
W N/A N/A N/A d.
Effluent System Flow Rate D
N/A E
Q Measuring Device (FR-148) e.
Sampler Flow Rate Monitor D
N/A E
N/A 4.
Condenser Vent System a.
Noble Gas Activity Monitor D
M E(3)
Q(2)
(RM-A5 and Sultable Equivalent -
See Table 3.21-2, Item 4.a)
TABLE 4.21-2 (Ccrtinuzd) k RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREE NTS I
I z
?
CHANNEL SOURCE CHANNEL CHANNEL INSTRUENT CHECK CHECK CALIBRATION TEST APPLICABILITY yn I
5.
Auxiliary and Fuel Handling k
Building Ventilation System a.
Noble Gas Activity Monitor D
M E(3)
Q(1)
(RM-A8) or (RM-A4 and RM-A6) b.
Iodine Sampler (RM-A8) or (RM-A4 and RM-A6)
W N/A N/A N/A c.
Particulate Sampler (RM-A8) or (RM-A4 and RM-A6)
W N/A N/A N/A
}
]
d.
System Effluent Flow Rate l
Measurement Devices D
N/A E
Q (FR 151, or FR-149 and FR-150)
?
e.
Sampler Flow Rate Monitor D
N/A E
N/A 8
d 6.
Fuel Handling Building ESF Air Treatment System a.
Noble Gas Activity Monitor (RM-A14) 0 M
R(3)
Q(2) b.
System Effluent Flow Rate (UR-1104 A/B)
D N/A R
Q j
c.
Sampler Flow Rate Measurement Device D
N/A R
Q i
I
TABLE 4.21-2 (Continued)
TABLE NOTATION
- At all times.
During waste gas holdup system operation.
Operability is not required when discharges are positively controlled through the closure of WOG-Y47, and RM-A8 and FT-151 are operable.
During Fuel Handling Building ESF Air Treatment System Operation.
- At all times during containment purging.
H At all times when condenser vacuum is established.
(1 )
The CHANNEL TEST shall also demonstrate that automatic isolation of this pathway for the Auxiliary and Fuel Handling Building Ventilation System, the supply ventilation is isolated and control room alarm annunciation occurs if the following conditio, exists:
1.
Instrument indicates measured levels above the high alam/ trip setpoint.
(Includes circuit failure) 2.
Instrument indicates a down scale failure.
(Alarm function only)
(Includes circuit failure) 3.
Instrument controls moved from the operate mode.
(Alam function only)
(2)
The CHANNEL TEST shall also demonstrate that control room alam annunciation occurs if any of the following conditions exist:
1.
Instrument indicates measured levels above the alarm setpoint.
(includes circuit failure) 2.
Instrument indicates a down scale failure (Includes circuit failure) 3.
Instrument controls moved from the operate mode.
l (3)
The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation shall be performed using one or more of the reference standards certified by i
the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.
These standards should pemit calibrating the system over its intended range of energy and measurement range.
For subsequent CHANNEL CALIBRATION, l
sources that have been related to the initial calibration should be used.
(Operating plants may substitute previously established calibration procedures for this requirement.)
(4)
The CHANNEL CALIBrtATION shall include the use of standard gas samples containing a nominal:
l 1.
One volume percent hydrogen, balance nitrogen, and 2.
Four volume percent hydrogen, balance nitrogen.
4-93 Amendment No M, 122,137
TABLE 4.21-2 (Continued)
TABLE NOTATION (5)
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1.
One volume percent oxygen, balance nitrogen, and 2.
Four volume percent oxygen, balance nitrogen.
O f
i l
l 4-94 Amendment No. g, 137
a 4.22 RADIOACTIVE EFFLUENTS 4.22.1 LIQUID EFFLUENTS SURVEILLANCE REQUIREMENTS 4.22.1.1 CONCENTRATION 4.22.1.1.A The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 4.22-1.
The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.22.1.1.
4.22.1.1.8 Post-release analysis of samples composited from batch releases shall be performed in accordance with Table 4.22-1.
The results of the previous post-release analysis shall be used with the calculational metnods in the 00CM to assure that the concentrations at the point of release were maintained within the limits of Specification 3.22.1.1.
4.22.1.1.C The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 4.22-1.
The results of the analysis shall be used with the calculational methods of the 00CM to assure that the concentration at the point of release is maintained within the limits of Specification 3.22.1.1.
4-95 Amendment No. JE',137
t
- g TABLE 4.22-1 RADI0 ACTIVE LIQUID WASTE SAW LING AND ANALYSIS PROGRAM M
l Sampling i
Minimum l
l Lower Limit g
Liquid Release Type l
Frequency l
Analysis l
Type of Activity I
of Detection l
Frequency l
Analysis 1
(LLD) i I
l (uC1/ml)(Note a)
I I
I I
l P
l P
l H-3 l 1 x 10-5 A.1 Batch Waste i
Each Batch i
Each Batch l Principal Gamma i 5 x 10-7 Release Tanks l
l l Emitters (Note f) l (Note d) l l
l l
l l
1 I-1 31 l 1 x 10-6 l
1 l
l I
I I
I l
P l
l l 1 x 10-4 l
One Batch /M i
M l
Dissolved and l
l l
l Entrained Gases l
?
l I
I (Gamma Emitters) l E
l l
1 1
I I
l l
1 I
I I
l P
l Q
l Gross alpha l 1 x 10-7 I
Each Batch l
Composite l
l l
l l
(Note b) l Sr-89, Sr-90 l 5 x 10-8 I
i l
l L
l l
l Fe-55 l 1 x 10-6 I
I l
1
t
- s TABLE 4.22-1 (Continued)
[
RADI0 ACTIVE LIQUID WASTE SAWLING AND ANALYSIS PROGRAM l
3 l
Sampling i
Minimm i
I Lower Limit g
Liquid Release Type l
Frequency I
Analysis l
Type of Activity I
of Detection
.A l
l Frequency
]
Analysis I
(LLD) 1 l
l l(uCf/ml) (Note a) t i
I W
I I
A.2 Continuous l
Continuous l
Composite i Principal Gamma 1 5 x 10-7 Releases I
(Note c)
I (Note c) l Emitters (Note f) l (Note e) l l
l l
l l
l I-131 l 1 x 10-6 I
I I
I I
I I
I l
M i
l i 1 x 10-5 i
I Dissolved and i
I l
l Entrained Gases I
A l
l l
(Gamma Emitters) l I
I I
I I
l I
l
.a l
I I
I l
l d
l H-3 l 1 x 10-5 N
i Continuoas I
Composite i
I l
(Note c)
I (Note c) i Gross alpha l 1 x 10-7 I
I I
I I
I I
I l
l Q
l Sr-89, Sr-90 1 5 x 10-3 I
Continuous l
Composite i
l l
(Note c)
-l (Note c) l Fe-SS I 1 x 10-6 I
I I
I I
I I
I
}
e O
e
a Table 4.22-1 (Continued)
TABLE NOTATION The LLD is defined, for purposes of this specification, as the smallert a.
concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with SS probability of falsely concluding that a blank observation repre;r9ts a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4' S b LLO =
E x Y x 2.22 x 106 x Y x exp (- AAt)
LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume),
so is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per disintegration),
Y is the sample size (in units of mass or volume),
2.22 x 106 is the number of disintegrations per minute per microcurie, l
Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between midpoint of sample collection and time of counting.
}
l Typical values of E, V, Y and at shall be used in the calculation.
It should be recognized that the LLD is defined as an "a priori" (before l
the fact) limit representing the capability of a measurement systein and not as an "a postcriori" (after the fact) limit for a particular measurement.
b.
A composite sample is one in which the quantity of liquid sampled is I
proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.
I l
4-98 Amendment No. Jf.137
a To be representative of the quantities and concentrations of radioactive c.
materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream.
Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release, d.
A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and the thoroughly mixed, by a method described in the ODCM, to assure representative
- sampling, e.
A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow during the continuous release, f.
The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 This list does not mean that only these nuclides are to be considered.
Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report pursuant to TS 6.9.4.
1 4
4-99 Amendment No. Jf.137
4.22.1.2 DOSE CALCULATIONS Cumulative dose contributions f om liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least once a month.
4.22.1.3 LIOUID WASTE TREATMENT 4.22.1.3.1 Doses due to liquid releases shall be projected at least once a month, in accordance with the ODCM.
4.22.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 60 minutes quarterly unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.
4.22.1.4 LIQUID HOLDUP TANKS The quantity of radioactive material contained in each of the tanks specified in Specification 3.22.1.4 shall be determined to be within the limit by analyzing a representative sample of the tank's content weekly when radioactive materials are being added to the tank.
4-100 Amendment No. Jf,137
4.22.2 GASEOUS EFFLUENTS SURVEILLANCE RE0VIREMENTS 4.22.2.1 OOSE RATES 4.22.2.1.1 The dose rate due to noble gases in gaseous ef fluents shall be determined to be within the limits of Specification 3.22.2.1.a in accordance with the methods and procedures of the 00CM.
4.22.2.1.2 The dose rate of radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the limits of Specification 3.22.2.1.b in accordance with methods and procedures of l
the 00CM by obtaining representative samples and performing analyses in accordance with the saepling and analysis program, specified in Table
- 4. ?2 -2.
4-101 Amendment No. J2',137
g TABLE 4.22-2 i
a ra g
RADI0 ACTIVE GASEOUS WASTE SAWLING AND ANALYSIS PROGRAM
~
~
M j
l Sampling l
Minimum i
I Lower Limit i
Gaseous Release Type l
Frequency i
Analysis l
Type of Activity I
of Detection l
l Frequency
'l Analysis l
(LLD)
I l
l l
l(uCi/ml) (Note a) k l
I I
I I
l 1
I I
I g
l P
l P
l l
J A.
Waste Gas l
Each Tank l
Each Tank l
Principal Gamma l
1 x 10-4
~
l Decay Tank i
Grab l
l Emitters (Note g) l
'{
l Sample l
l l
I l
l l
I I
I I
B.
Containment l
P (Note b) l P (Note b) l H-3 l
1 x 10-6 Purge i
Each Purge l
Each Purge l
Principal Gamma l
1 x 10-4 l
Grab l
l Emmitters I
I Sample l
l (Note g) l l
l 1
1 I
i
?
I I
I I
i 5
C.
Auxiliary and l
l l H-3 l
1 x 10-6 i
Fuel Handling l
M (Notes c, e) l M
l Principal Gamma l
1 x 10-4 r*
j Building Air l
Grab l
l Emitters l
Treatment System l
Sample 1
i (Note g) 1 l
l l
l 5
I I
I I
l D.
Fuel Handling l
M (during l
l l
l Building ESF Air I
system opera-1 M (during l
H-3 1
1 x 10-6 Treatment System I
tion)
I system l Principal Gamma l
1 x 10-4 j
i Grab Sample 1
operation) l Emitters l
l l
l (Note g) l I
I I
I 1
E.
Condenser Vacuum l
M (Note h) l M
i H-3 l
l Pumps Exhaust i
(Note h) l Principal Camma l
1 x 10-6 (Note h) l l
l Emitters l
1 x 10-4 4
I l
l l
(Note g) l l
l I
i
~
1
E h
TABLE 4.22-2 5"
RADI0 ACTIVE GASEOUS WASTE SAW LING AND ANALYSIS PROGRAM E.
.h l
Sampling i
Minimum l
l Lower Limit Gaseous Release Type l
Frequency 1
Analysis l
Type of Activity I
of Detection
{
l l
Frequency i
Analysis 1
(LLD) l l
l l(uCi/ml) (Note a) l I
I I
I I
I I
F.
All Release l
Continuous l
W (Note d) l I-1 31 l
1 x 10-12 C
Types as Listed I
(Note f) l Charcoal l
l Above in A, B, C l
l Sample l
I and D (during system i
l i
I operation) i l
l I
l l
l W (Note d) l Principal Gamma l
1 x 10-11 l
Continuous l
Particulate l
Emitters (Note g) l l
(Note f) l l
(I-131, Others) l I
I l
l 1
I I
I l
l 1
0 I
l 1 x 10-11 l
Continuous l
Composite l
Gross alpha l
w
- .L I
(Note f) l Particulate l
8 l
l Sample i
i l
l I
I I
I I
I l
Continuous I
Q l
l l
(Note f) l Composite i
Sr-89, Sr-90 l
1 x 10-11 j
l l
Particulate i
l i
l i
Sample I
I I
I I
1 1
I I
I G.
Condenser Vent Stack l
l W (Note d) l I-1 31 l
1 x 10-12 1
Continuous Iodine i
Continuous l
Charcoal l
l Sampler (Note j) l (Note k) l Sample l
1 1
1 I
I I
e a
Table 4.22-2 (Continued)
TABLE NOTATION The LLO is defined, for purposes of this specification, as the smallest a.
concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with 5%
probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4*
S b LLO =
E x Y x 2.22 x 1 9.t Y x exp (- AAt)
LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per disintegration),
V is the sample size (in units of mass or volume),
2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and A t is the elapsed time between midpoint of sample collection and time of counting.
Typical values of E, Y, Y, and A t shall be used in the calculation.
It should be recognized that the LLD is defined as an "a priori" (before the fact) Ifmit representing the capability of a measurtment system and not as an "a posteriori" (after the fact) limit for a particular measurement.
b.
Sampling and analysis shall also be perfomed following shutdown, startup, or a THERMAL p0WER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.
c.
Tritium grab samples from the spent fuel pool area shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.
4-104 q
Amendment No, g.137
Table 4.22-2 (Continued)
TABLE NOTATION d.
Charcoal cartridges and particulate filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or af ter removal from sampler),
Tritium grab samples shall be taken weekly from the spent fuel pool area e.
whenever spent fuel is in the spent fuel pool.
f.
The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.22.2.1, 3.22.2.2, and 3.22.2.3.
g.
The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emmissions and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-137, Ce-141 and Ce-144 for particulate emissions.
This list does not mean that only these nuclides are to be considered.
Other gama peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semi-Annual Radioactive Effluent Release Report pursuant to TS 6.9.4 h.
Applicable only when condenser vacuum is established.
Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.
f.
Gross Alpha, Sr-89, and Sr-90 analyses do not apply to the Fuel Handling Building ESF Air Treatment System.
j.
If the Condenser Vent Stack Continuous Iodine Sampler is unavailable, then alternate sampling equipment will be placed in service within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, k.
Applicable only when condenser vacuum is established.
4-105 Amendment No. JE, JM, 136, 137
4.22.2 GASEOUS EFFLUENTS SURVEILLANCE REQUIREMENTS 4.22.2.2 OOSE, NOBLE GAS Cumulative dose contributions from noble gas effluents for the current calendar quarter and current calendar year shall be determined in accordance with the OFFSITE 00SE CALCULATION MANUAL (00CM) monthly.
4.22.2.3 OOSE, IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIOES IN PARTICULATE FORM Cumulative dose contributions from Iodine-131, Iodine-133 Tritium, and radionuclides in particulate form with half lives greater than 8 days for the current calendar quarter and current calendar year shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM) monthly.
4.22.2.4 GASEOUS WASTE TREATMENT Doses due to gaseous releases from the unit shall be projected monthly in accordance with the 00CM.
4.22.2.5 EXPLOSIVE GAS MIXTURE The concentrations of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the limits of TS 3.22.2.5 by monitoring the waste gases in the Waste Gas Holdup System with the hydrogen and oxygen monitors covered in Table 3.21-2 of Specification l
3.21.2.
s I
4.22.2.6 WASTE GAS DECAY TANK The concentration of radioactivity contained in the vent header shall be determined weekly.
If the concentration of the vent header exceeds 10.7 uCi/ce, daily samples shall be taken of each waste gas decay tank
[
being added to, to determine if the tank (s) is 18800 Cf / tank.
4-106 Amendment No, M.137
4.22.3 SOLIO RADIOACTIVE WASTE SURVEILLANCE REQUIREMENTS 4.22.3.1 SOLIO RA0 WASTE SYSTEH The solid radwaste system shall be demonstrated CPERABLE quarterly by:
Operating the solid radwaste system at least once in the previous a.
92 days in accordance with the PROCESS CONTROL PROGRAM or; b.
Verification of the existence of a valid contract for SOLI 0IFICATION to be performed by a Contractor in accordance with a PROCESS CONTROL PROGRAM.
4.22.3.2 PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of radioactive waste required to be solidified by the Process Control Program, If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION a.
of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the Process Control Program, and a subsequent test verif fes solidification.
Solidification of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.
b.
If the initial test specimen from a batch of waste fails to verify SOLIO!FICATION, the PROCESS CCNTROL PROGRAM shall provide for the collection and testing of representative Cest specimens from each consecutive batch of the same type of wet waste until 3 consecutive initial test specimens demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required, to assure SOLIDIFICATION of subsequent batches of waste.
4-107 Amendment No. g,137
4.22.4 TOTAL DOSE SURYEILLANCE REQUIREMENT 4.22.4.1 OOSE CALCULATION Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with TS 4.22.1.2, 4.22.2.2 and 4.22.2.3 and in accordance with the 00CM.
2 l
l l
4-108 (Pages 4-109 to 4-116 intentionally blank)
Amendment No. 72.137 l
5.0 DESfGN FEATURES 5.1 SITE Applicability Applies to the location and extent of the exclusion boundary, restricted area, and low population zone.
Objective To define the above by location and distance description.
Specification 5.1.1 The Three Mile Island Nuclear Station Unit 1 is located in an area of low population density about ten miles southeast of Harrisburg, PA.
It is in Londonderry Township of Dauphin County, Pennsylvania, about two and one-half miles north of the southern tip of Dauphin County, where Dauphin is coterminal with York and Lancaster Counties.
The station is located on an island approximately three miles in length situated in the Susquehanna River upstream from York Haven Dam.
Figure 2.1-3 of the updated TMI Unit 1 l
FSAR is an extended plot plan of the site showing the plant orientation and immediate surroundings.
The exclusion area as defined in 10 CFR 100.3, is a 2,000 ft. radius, including portions of Three Mile Island, the river surface around it, anc' a portion of Shelley Island, which is owned by Met Ed.
The minimum distance of 2,000 f t. occurs on the shore of the mainland in a due easterly direction from the plant as shown on Figure 2.1-3 of the FSAR.
Figure 1.1-1 of the FSAR is a plot plan showing the l
physical location of the fence which defines the "Restricted Area" surrounding the plant.
The minimum distance of the "Restricted Area" is approximately 560 feet and is from the center!ine of the TMI Unit 2 Reactor Building to a point on the westerly shoreline of Three Mile Island.
T.S.
l Figure 5-1 is the Extended Plot Plan for Three Mile Island and includes the Exclusion Area and the meteorological tower locations.
The minimum distance to the outer boundary of the low population zone is two miles as shown on T.S. Figure 5-2.
For discharge points for gaseous effluents, see T.S. Figure 5-3 and for liquid effluents, see T.S. Figure 5-4.
l 5-1 l
l Amendment No. )(,137 i
j 6.9.4 SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT Note: A single submittal may be made for the station.
The submittal should combine those sections that are common to both units at i
the station however, for units with separate radwaste systems, the submittal shall specify the release of radioactive material from each unit.
6.9.4.1 poutine Radioactive Effluent Release Reports covering the operations l
of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year.
6.9.4.2 The following information shall be included in both Radioactive Effluent Release Reports to be submitted each year:
6.s.4.2.1 The Radioactive Effluent Release Reports shall include a sumary of l
the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", Revision 1 June 1974, with data sumarized on a quarterly basis following the format of Appendix B thereof.
6.9.4.2.2 The Radioactive Effluent Release Reports shall include the following l
infonnation for each type of solid waste shipped offsite during the report period:
a.
container volume, b.
total curie quantity (specify whether determined by measurement or estimate),
c.
principal radionuclides (specify whether determined by measurement or estimate),
d.
type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),
e.
type of shipment (e.g., LSA, Type A, Type B) and f.
solidification agent (e.g., cement).
6.9.4.2.3 The Radioactive Effluent Release Reports shall include a summary of l
unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents made during the reporting period.
6.9.4.2.4 The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (00CM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.23.2.
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6.9.4.2.5 The Radioactive Effluent Release Reports shall include the instrumentation not returned to OPERABLE status within 30 days per TS 3.21.1.b and TS 3.21.2.b.
)
6.9.4.3 The following information shall be included in the Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year.
6.9.4.3.1 The Radioactive Effluent Release Report to be submitted 60 days af ter l
January 1 of each year shall include an annual sumary of hourly meteorological data collected over the previous year.
This annual sumary may be either in the fom of an hour-by-hour listing of wind speed, wind direction, atmosphere stability, and precipitation (if l
measured) on magnetic tape, or in the fom of joint frequency distribution of wind speed, wind direction, and atmospheric stability.
6.9.4.3.2 The Radioactive Effluent Release Report to be submitted 60 days after l
January 1 of each year shall include an assussment of the radiation doses due to the radioactive liquid and gaseous effluents released 1
from the unit or station during tha previous calendar year.
6.9.4.3.3 The Radioactive Effluent Release Report to be submitted 60 days after l
January 1 of each year shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the site boundary (Figures 5-3 and 5-4) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports.
The meteorological conditions concurrent with the time of release of radioactive matorials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses.
The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).
l 6.9.4.3.4 The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases and other nearby uranium fuel cycle sources including doses from primary effluent pathways and direct radiation for the previous 12 consecutive months to show conformance with 40 CFR 190 "Environmental Radiation Protection Standards for Nuclear Power Operation".
Acceptable methods for calculating the dose contributiens from Liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1.
6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
a.
Records of normal station operation including power leYels and periods of operation at each power level.
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b.
Records of principal maintenance activities, including t
inspection, repairs, substitution, or replacement of principal items of equipment important to safety, c.
All REPORTABLE EVENTS.
d.
Records of periodic checks, tests and calibrations, e.
Records of reactor physics tests and other special tests important to safety, f.
Changes to operating procedures important to safety, g.
Records of solid radioactive shipments.
6-19a l'
heendment No. g g g,137
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