ML20028B947
| ML20028B947 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 12/17/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Northern States Power Co |
| Shared Package | |
| ML20028B948 | List: |
| References | |
| DPR-22-A-015 NUDOCS 8301040070 | |
| Download: ML20028B947 (78) | |
Text
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o UNITED STATES h.h y,( (}3 8
NUCLEAR REGULATORY COMMISSION
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.15 License No. DPR-22 1.
The Nuclear Regulatory Commission (the Co mission) has found that:
A.
This application for amendment by Northern States Power Company (the licensee) dated May 1,1979 and revised by letter dated July 23, 1982, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specific'ations contained in Appendix A as revised through Amendment No. 15 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
8301040070 821217 I
PDR ADOCK 05000263
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This license amendment is effective on January 1,1983.
l FOR THE NUCLEA REGULATORY COMMISSION Ad Domenic B. Vassallo, Chief i
Operating Reactors Branch #2 Division of Licensing i
Attachment:
l Changes to the Technical Specifications Date of Issuance: December 17, 1982 l
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ATTACHMENT TO LICENSE AMENDMENT NO.15 l
FACJLITY OPERATING LICENSE NO. DPR-22 i
DOCKET NO. 50-263
.1. Replace the following pages of Appendix A Technical Specifications, with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of changes.
REMOVE INSERT i
i iii 111 iv iv l
v v
vi vi vii 5
5 Sa 46 46 47 47 48 48 48a 62 62 68 68 124 124 192 192 193 193 194 194 195 195 196 196 197 197 198 198 198a to 198z 229g to 229s 239 239
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242 242 244-244 246 246 246a to 246c 249 249 249a 250 250 251 251 251a 252 252 2'53 253 253a 254 254 2.
Delete Appendix B in i ts entirety.
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TABLE OF CONTENTS
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Pace 1.0 DSFINITIONS 1
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 6
1 2.1 and 2.3 Fuel Cladding Integrity 6
2.1 Bases 10 2.3 Bases 14 9
2.2 and 2.4 Reactor Coolant System 21 l
2.2 Bases 22 j
2.4 Bases 24 1
3.0 LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILLANCE REQUIREMENTS 26 3.1 and 4.1 Reactor Protection System 26 3.1 Bases 35 l
4.1 Bases 41 I
3.2 and 4.2 Protective Instrumentation 45 A.
Primary Containment Isolation Functions 45
.i [
B.
Emergency Core Cooling Subsystems Actuation 46 s
C.
Control Rod Block Actuation 46 D.
Deleted E.
Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation' 47 F.
Recirculation Pump Trip Initiation 48 G.
Safeguards Bus Voltage Protection 48 l
3.2 Bases 64 4.2 Bases 72 3.3 and 4.3 Control Rod System 76 A.
Reactivity Limitations 76 B.
Control Rod Withdrawal 77 C.
S c r a.2 Insertion Times 81 l
D.
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E.
Reactivity Anomalies 83 F.
Scram Discharge Volume 83 G.
Required Action 83A i
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3.3 and 4.3 Bases 84 i
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3.8 and 4.3 Radioactive Ef fluents 192 i
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Liquid Effluents 192 B.
Gaseous Effluents 197 s
C.
Solid Radioactive Waste 198e i
D.
Dose from All Uranium Fuel Cycle Sources 193f 3.8 and 4.8 9ases 198u 1
3.9 and 4.9 Auxiliary Electrical Systems 199 A.
Operational Requirements for Startup 199 3.
Operational Requirenents for Continued Operation 200 1
1.
Transmission Lines 200 j
2.
Reserve Transformers 201 3.
Standby Diesel Generators 201
'l 4.
Station Battery System 202 i
i 3.9 Bases 204 4.9 Bases 205 3.10 and 4.10 Refueling 206 i (-
A.
Refueling Interlocks 206 j
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B.
Core Monitoring 207 i
C.
Fuel Storage Pool Water Level 207 J
D.
Movement of Fuel 207 E.
Extended Core and Control Rod Drive '!aintenance 208 3.10 and 4.10 Bases 209
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3.11 and 4.11 Reactor Fuel Assemblies 211 i
A.
Average Planar Linear Heat Ceneration Rate 211 1
9.
Linear 9 eat Generation Rate 212 C.
?tinimun Critical Power Ratio 213 l
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3.11 Bases 215 r
i; 4.11 3ases 217
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3.12 and 4.12 Sealed Source Contamination 219 A.
Contamination 219 J
B.
Records 221
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3.12 and 4.12 Bases 222 1
1 111
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Page 3.13 and 4.13 Fire Detection and Protection Systems 223 J
A.
Fire Detection Instrumentation 223 l
B.
Fire Suppression Water Systen 224 C.
Mose Stations 226 D.
Yard Hydrant Mose Mouses 227 E.
Sprinkler Systems 227a F.
Halon Systens 227b G.
Penetration Fire Barriers 227b l
3.13 3ases 228
- l 4.13 Bases 229 t
I 3.14 and 4.14 Accident ?fonitoring Instrumentation 229a 3.14 and 4.14 Bases 229d j
j 3.15 and 4.15 Inservice Inspection and Testing 229e 3.15 and 4.15 Bases 229f l
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3.16 and 4.16 Radiation Environmental ?!onitoring Program 229g i
- l{s A.
Sampling and Analysis 229g B.
Land Use Census 2291 C.
Interlaboratory Comparison 229j 4.16 Bases 229s n
5.0 DESIGt! FEATURES 230 i
a 5.1 Site 210
- 3J 5.2 Reactor 4
j 5.3 Reactor vessel 230 i.
5.4 Containment 230
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5.5 Fuel Storage 231
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5.6 Seismic Designs 231 1
6.0 AD?!INISTRATIVE CONTROLS 232 J
6.1 Organization 232 6.2 Review and Audit 237 6.3 Special Inspection and Audits,
243 6.4 Action to be taken if a Safety Linit is Exceeded 243 j
6.5 Plant Operating Procedures 244 6.6 Plant Operating RecorMs 246 l
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6.7 Reporting Requirenents 243 f
6.8 Environmental Qualification 254 o
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4.1.1
'M' Factor - Graphical Aid in the Selection of an Adequate 44 Interval Between Tests e
4.2.1 System Unavailability 75 3.4.1 Sodium Pentaborate Solution Volume-Concentration Requirements 97 I
3.4.2 Sodium Pentaborate Solution Temperature Requirements 98 1
3.6.1 Change in Charpy V Transition Temperature versus Neutron Exposure 133 3.6.2 "inimum Temperature versus Pressure for Pressure Tests 134
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l 3.6.3 Minimum Temperature versus Pressure for Mechanical Heatup or 135 ij Cooldown Following Nuclear Shutdown 3.6.4 flinimum Temperature versus Pressure for Core Operation 136 i
4.6.2 Chloride Stress Corrosion Test Results @ 500 F 137 3.7.1 Differential Pressure Decay Between the Drywell and Wetwell 191 3.8.1 Monticello Nuclear Generating Plant Site Boundary for Liquid 198g Effluents 3.8.2 Monticello Nuclear Generating Plant Site Boundary for Gaseous 198h Effluents l
3.11.3 K Factor versus Percent of Rated Core Flow 218 g
i 6.1.1 NSP Corporate Organizational Relationship to On-Site Operating 234 Organization I
6.1.2 Monticello Nuclear Generating Plant Functional Organization for 235 On-Sit 3 6perating Group Amendment No. 15 6
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i LIST OF TABLES t
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3.1.1 Reactor Protection System (Scram) Instrument Requirements 28
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4.1.1 Scram Instrument, Functional Tests - Minimum Functional 32 1
Test Frequencies for Safety Instrumentation and Control Circuits j
4.1.2 Scram Instrument Calibration - ttinimum Calibration 34
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Frequencies for Reactor Protection Instrument Channels I
3.2.1 Instrumentation that Initiates Primary Containment 49 Isolation Functions 3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 3.2.3 Instrumentation that Initiates Rod Block 57 3.2.4 Instrumentation that Initiates Reactor Building Ventilation 59 Isolation and Standby Gas Treatment System Initiation 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip 60 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and 60a Loss of Voltage Protection 3.2.7 Trip Functions and Deviations 70 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation 61 3.6.1 Safety Related Snubbers 131 L
3.7.1 Primary Containment Isolation 172 f
3.8.1 Radioactive Liquid Ef fluent Monitoring instrumentation 1891 3.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198k 4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 198m Surveillance Requirements 4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198n Surveillance Requirements l
4.8.3 Radioactive Liquid tJaste Sampling and Analysis Program 198p 3
4.8.4 Radioactive Gaseous !!aste Sampling and Analysis Program 198s s.3.11.1 Maximum Average Planar Linear Heat Generation Rate 214 9
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3.13.1 Safety Related Fire Detection Instruments 227c l
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!!inimum Test and Calibration Frequency for Accident 229c I,
Monitoring Instrumentation 4
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4.16.1 Radiation Environmental Monitoring Program (REMP) 229k Sample Collection and Analysis 4
4.16.2 REMP - Maximum Values for the Lower Limits of Detection 229p 4.16.3 RE!!P - Reporting Levels for Radioactivity Concentrations 229r in Environmental Samples 6.1.1 Minimum Shift Crew composition 236 j
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Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the d
shutdown mode position and no core alterations are being perforced. In this condition, a i
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reactor scram is initiated and a rod block is inserted directly f rom tha mode switch. The sc ram can he reset after a short time delay.
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Ilot temperature greater than 212, Shutdown means conditions as above with reactor coolant f
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Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212"F.
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- 7.. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated d
'I signal to the sensor to actuate the circuit in question.
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" AA. Transition noiling - Transition hotling means the boiling regime between nucleate and film f
boiling, also referred to as partial nucleate boiling. Transition boiling is the regime in which both nucleate and film holting occur intermittently with neither type being completely,
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AB.
Prensure Boundary Imakage - Pressure boundary leakage shall be leakage through a non-isolable rault in the reactor coolant system pressure boundary.
4 AC.
Identified leakage - Identified leakage shall be:
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- 1) Reactor coolant leakage into drywell collection systems, such as pump seal or valve packing
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leaks, that is captured and conducted to a nump or collecting tank, or
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- 2) Reactor coolant leakage into the drywell atmosphere from sources which are spectrically located and known not to be Pressure Boundary leakage or.which do not significantly impair the methods used to detect reactor coolant leakage.
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Unidentirled Imakage - Unidentified leakage shall be all reactor coolant leakage which is not l'
Identified Icakage.
_ _AE.
Process Control Program (PCP) - The PCP is the manual containing the current formula, sampling, l
l analysis, test and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet so11d wastes will be accomplished in such a way as to assure compliance with IOCFR20, 10CFR71, j
and Federal and State regulations and other requirements governing the disposal of radioactive i
e wastes.
I
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Solidification - Solidification is the conversion of wet radioactive wastes into a form that il meets shipping and burial ground requirements.
A AG. Offsite' Dose Calculation Manual (ODCH) - The ODCH is the manual containing the methodology and parametern to be used in the calculation of offsite doses due to radioactive Itquid and ganeoun effluents, in the calculation of liquid and gaseous effluent monitoring instrumentation l
alarm and/or trip setpoints, and in the conduct of environmental radiological environmental monitoring.
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AH. Source Check - A Source Check la the qualitattve assessment of channel response when i
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Purging - Purging is the controlled process of discharging air or gas frors a confinement'to N
maintain temperature. pressure, humidity, concentration, or other operating condition, in i
seh a manner that replacement a!r or gas is required to purify the confinement.
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venting - venting is the controtled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required.
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Dose Equivalent 1-131 - Dose Equivalent I-131 is the concencration of I-131 (microcuries/graia) f which alone would produce the saine thyroid dose as the quantity and isotopic mixture of I-131, 8 I-132, I-133,1 -134 and 1-135 actually present. The thyroid dc.se conversion factors used 'Far this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance 3
Factors for Power and Test Reactor Sites" or in NRC Regulatory Culde 1.109, Rev 1, October, 19H. ;
I AL. of f gas Treatment system - The of fgas Treatment system is the system designed and installed to q
reduce radioactive gaseous effluents by collecting primary coolant system offgases from the
- j primary system and providing for delay or holdup for the purpose of reducing the total radio-j activity prior to release to the environment.
i AJ AM.
- Members of the Public - Means all persons who are not occupationally associated with the plant.
This category does not include employees of the utility, its contractors, or its vendors. Also excluded frois this category are person,a who enter the site to service equipment or to make j
i deliveries. This category does include persons who use portions of the site for recreational, 3
oc,cupational, or other purposes not associated with the plant.
AN.
site noondary - Means a line within which the land is owned, leased, or otherwise controlled by the licensee. The site boundary for liqi.ild releases of radioactive material is defined in j
Figure 3.8.1.
The site boundary for gaseous releases of radioactive matertal is defined in
.j Figure 3.8.2.
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linrestricted Areas, - Heans any area at or beyond the site boundary to which access is not j
controlled by the licensee for purposes of protection of individuals f rois exposure to radiation
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an.1 radioactive materials or any area within the site boundary used for residential quarters or r
j industrial, commercial,' institutional and/or recreational purposes.
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Emergency Core Cooling Snbsystems Actuation
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a tJhen irradiated fuel is in the reactor vessel
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I and the reactor water temperature is above 212*F, the limiting conditions for operation for the lastrumentation which initiates the emergency core cooling subsystems are given 8
F in Table 3.2.2.
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Control Rod lilock Actuation r
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The limiting conditions of operat ton for i
i the instrumentation that initiates control rod block are given in Table 3.2.3.
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Reactor Building Ventilat ton Isolation and Standby Cas f
Treatment System Initiation 1.
a.
Except as specified in 3.2.E.1.b below, four radiation monitors shall be operable at all times.
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One of the two reactor building vent isolation 2
monitors aad one of the two radiation monitors on the refueling floor nuy be inoperable provided l
inoperable radiation monitor instrument channels are tripped. Upon loss of both channels, the j
Reactor nullding ventilation system shall be e
isolated and the standby gas treatment system operated until repairs are complete.
2.
The radiation nonitors shall be set to trip as l
folfows:
- I (a) reactor hullding vent - refer to Specification 4.8.H.I.h i
( h) refuelinn floor f100 mr/hr i
4 3
3.
When irradiated fuel is in the reactor vessel and the reaetor water tempe ra tu re is a bove 212"I',,
the limiting conditions for operation for the instrumentation listed in Table 3.2.4 shall be met.
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1.0 1.ltitTING CotiutTIOtiS FOR OPERAT10ll 4.0 SURVE11.1.ANCE REQUIREi!ENTS i
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Itecirculation Pump Trip Initiation t
1.
Wiienever the reactor is in tlie RUN !! ode, t lie i
limiting conditions for op"cra tion for the inst ru-mentation listed in Table 3.2.5 shall be met.
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Safeguarilu Ilus Voltage Protection I
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tihenever the safeguards auxiliary electrical power system is required to be operable by j
Specification 3.9, the Limiting Conditions for Operation for the instrumentation listed in i
Table 3.2.6 shalt be met, j
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3.2 For ef fective emergency core cooling for the small pipe break the IIPCI or Automatic Pressure Relief system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either cure spray or LPCI to operate in time.
The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are a~dequate to assure the above criteria is met.
Reference Section 6.2.4 and 6.2.6 FSAR. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent ope ra t ion; 1.e., only one instrument channel out of service.
I Four radiation monitors are provided which initiate isolation of the reactor building and operation I
of the. standby gas treatment system. The monitors measure radioactivity of the reactor building ventilation exhaust and on the refueling floor. Any one upscale trip will cause the desired action. Trip settings for the ventilation exhaust isolation monitors are based upon initiating normal ventilation isolation and Standby Gas Treatment System operation prior to exceeding the maximum release rate limit for the reactor building vent.
Trip settings of 100 mR/hr for the 4
j monitors on the refueling floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor hullding via the normal ventilation stack but that all the activity is processed by the stahdby gas treatment system.
The recirculation pump trip description and performance analysis is discussed in Topical Report NEDO-25016, September 1976, " Evaluation of Anticipated Transients Without Scram for the Monticello Nuclear Cencrating Plant".
(See September 15, 1976 letter f rom Mr L 0 Maye'r, NSP, to Mr D L 7.icmann, llSNRC. ) The pump trip is provided to minimize reactor pressure in the highly unlikely event of a 2
plant transient coincident with the failure of all control rods to scram. The rapid flow reduction J
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p 3.0 1.tHiTING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQllIRDIENTS T
i analyzed for radioactive iodines of i
1-131 through I-135 during power ope ra t ion.
In addition, where steam I
Jet air ejector monitors indicate an increase in radioactive gaseous ef fluents of 25 percent or 5000 uC1/sec, whichever is greater, during steady state reactor i
operation a reactor coolant sample shall he taken and analyzed for radioactive f
8 a
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( b)
Isotopic analysis of reactor coolant
'i samples shall be made at least once i
per month.
(c)
Whenever the steady state radioiod ine concentration of prior operation is greater than I percent but less than 10 percent of Specification 3.6.C.1, 1
a sample of reactor coolant shall he taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of any reactor I
startup and analyzed for radioactive iodines of 1-131 through I-135.
4 (d)
Whenever the steady state radioiodine concentration of prfor operation is greater than 10 percent of Section 3.6.C.1, a sample of reactor coolant
..]
shall be taken daily and prior to any l
q reactor startup and analyzed for radio-t active iodines of I-131 through I-135 as well as the coolant sample and
'J analyses required by Specification
.j 4.6.C.1.(c) above.
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3.0 LlHITING CONDITIONS 14)H OPERATION 4.0 SURVEiLI.ANCE REQUIREllENTS 1
d.
Radioactive liquid wastes shall be sampled
(
ami analyzed according to clic sampling and analysis program of Table 4.8.3.
t
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The results of radioactive analysis shall be
- J used in accordance with the methods of the ODCM to assure that the concentrations at the J
point of release are maintained within 'the
- I limits of Specification 3.8.A.I.a.
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All. Source Check - A Sonrce Check is the qualitative assessment of channel response when the channel sensor is caposed to a radioactive source.
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Purging - Purging is the controlled proccus of discharging nte or gas from a confinement to l
5 maintain temperature pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
- ?
-AJ.
venting - venting to the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in m
such a manner that replacement air or gas is not provided or required.
,i
'AK.
Dose Equivalent I-131 - Dose Equivalent 1-131 is the concentration of I-131 (microcuries/ gram) '
l2 which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, ll I-132, I-133, I-134 and I-135 actually present. The thyrold dose conversion factors used "for l
this calculation shall be those llsted in Tahle III of TID-14844, " Calculation of Distance l~
Factors for Power and Test Reactor Sites" or in NRC Regulatory Guide 1.109, Rev 1, October,1977.
AL, Of fgas Treatment System - The Of fgas Treatment System is the system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the
-j 4
primary system and providing for delay or holdup for the purpose of reducing the total radio-acttvity prior to release to the environment.
AM.' Members of the Public - Heans all persons who are not occupationally associated with the plant.
This category does not include employees of the utt11ty, its contractors, or its vendors. Also LL excluded from this category are persons who enter the site to service equipment or to make f
deliveries. This category does includ'e persons who use portions of the site for recreational, g
oc,cupational, or other purposes not associated with the plant.
- 3
,j AN.
Site Boundary - Heans a line within which the land is owned, leased, or otherwise controlled by y
the lleensee. The alte boundary for Itquid releases of radioactive material is defined in j
Figure 3.8.1.
The site boundary for gaseous releases of radioactive matertal is defined in d
Figure 3.8.2.
s A0.
linrestricted Areas - Heans any area at. or beyond the site boundary to which access is not
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controlled by the Itcensee for purposes of protection of individuals f rom exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial,' institutional and/or recreational purposes.
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a 3.0 LDt1 TING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
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j 2.
Dose 2.
Dose is L]
The dose or dose commitment to an individnal a.
Cumulative dose contributions for the current 1
L a.
I f rom radioactive materials in liquid effluents calendar quarter and year from liquid effluents released from the site (Figure 3.8.1) shall be shall be determined in accordance with the ODCM l
1 limited:
monthly.
1.
During any calendar quarter to { l.5 mrem l1 to the total body and to 4 5 mrem to any j'
organ, and l
q 2.
During any calendar year to43 mrem to the 4
total body and Lo f 10 inrem to any organ.
'f h.
tilth the calculated dose f rom the release of a
radioactive materials in liquid effluents
~
jy exceeding any of the above limits, within'30 l
days submit to the Commission in lieu of any ogher report, a special report which i,
j identifies the cause(s) for exceeding the
,;.]
limit (s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure the d
subsequent releases will be within the above
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limits. This special report shall also include (1) the results of radiological analyses of the drinking water source, and (2) the radio-logical impact on finished drinking water supplies with regard to the requirements of 40CFR141, Safe Drinking Slater Act.
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3.8/4.8 194 t
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3.0 1,IMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.
Liquid Radwaste System 3.
Liquid Radwaste System a.
The liquid radwaste treatment system shall be used a.
Doses due to liquid releases shall be pro-l used to reduce the radioactive materials in liquid jected at least once each month in accordancel wastes prior to their discharge when the projected with the ODCH.
doses due to the liquid effluent from the site (Figure 3.8.1) when averaged over one month would I
exceed 0.06 mrem to the total body or 0.2 mrem to I
any organ.
3 b.
With radioactive 11guld waste being discharged b
without treatment in excess of the limit in (a) above, within 30 days submit to the Commission, in lieu of any other report, a special report which l
includes the following information:
1; Identification of the inoperable equipment or e
subsystems and the reason for inoperability, a
2.
Action (s) to be taken to restore equipment j
j to operable status, and l
3.
Summary description of action (s) taken to prevent a recurrence.
i 1
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e 3.8/4.8 195 Amendment No.15 3
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3.0 f.IrlITING CONDITIONS FOR OPERATION 4.0 StlRVF.II.l.Ar4CE RIMlJIRIO!ENTS 4.
hiquid lloidop Tanks 4.
hiquid Holdup Tanks l
a.
The quan,tity of radioactive material a.
The quantity of radioactive material contained in each outside temporary tank contained in each outside tempor.iry shall be limited to $10 curies, ex-tank shall he determined to be within
. q cluding tritium and dissolved or the limit in 3.8.A.4.a by analyzing a catrained noble gases.
repres'ntative sample of the contents s
e
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of the tank at least weekly when I
h.
litth the quantity of radioactive radioactive materials are being adileil to J
material in any outstile temporary the tank.
. tank exceeding the limit in (a) above.
I immediately suspend all additions of radioact ive materla t 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 j
contents to within the limit.
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Amendment No. 15 g
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3.0 1.IlllTING CONDITIONS FOR OPERATION 4.0 SURVEILI.ANCE REQUIREtlENTS B. Caseous Effluents 15. Caseous Effluents s
- 1. Dose Rate
- 1. Dose Rate 4
- a. The dose rate at any time due to radioactive
- a. Radioactive material in gaseous effluents materials released in gaseous effluents from released from the site shall be continuously i
a the site (Figure 3.8.2) shall he limited monitored in accordance with Table 3.8.2.
to the following:
1 l
b.
The noble gas effluent monitors having
- 1. For noble gases to s500 mrem / year provisions for the automa tic termination
?
to the total body and 6.3000 mrem / year to of gaseous releases, as listed in Table 3.8.2 the skin, and shall be used to limit offsite dose ra tes to j
the values established in Specification d
Setpoints shall be determined i
i j-(.
ticulates, with half-lives greater than in accordance with the ODCM.
j i
cinht days to {l500 mrem / year to any
'/
- organ, c.
Surveillance of gaseous effluent nonitor-I ing instruments shall be performed as required l
il
- f by Table 4.8.2.
j i
i b.
With the dose rate (s) exceeding the limits
in (a) above, immediately dec. ease the re-particulates with half-lives greater than eight lease rate within ' acceptable limits, days shall be determined by obtaining represent-ative samples and performing analyses in accordance
]
with the sampling and analysis program speci-fled in Table 4.8.4.
Following each analysis i
the dose rate due to I-131, t ri t ium, and radio-I j
l active particulates with half-lives greater than eight days shall be determined to be less tl3an j
the limit in Specifica t ion 3.8.ll. l.a.2 in j
j accordance with the ODCH.
j I
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3.8/4.8 197 t
i Amendment No. 15 l
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3.0 LilllTIf1G CO!1DITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
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- 2. Dose from Noble Cases
- 2. Dose from Noble Cases l
- a. The air dose due to noble gases released in gasenus
- a. Cumulative dose contributions for the current 1
effluents from the site (Figure 3.8.2) shall calendar quarter and year from noble gases in I
be limited to the following values:
gaseous ef fluents shall be determined in accord-j i
ance with the ODCM monthly.
I n
i i
l.
During any calendar quarter, to $5 mrad for gamma radiation and $10 mrad for beta
.radia t ion, and
- j 2.
During any calendar year, to s 10 mrad for gamma radiation and $20 mrad for beta (4
radiation.
s
.I
- h. With the calculated air dose from radioactive i
noble gpses in gaseous effluent exceeding any 3
of the above limits, within 30 days submit to the Commission, in lieu of any other report, a special report which identifies the cause(s) i for exceeding the limit (s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within
?
the above limits.
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3.0 LilllTING CONDITIONS FOR OPERATION 4.0 SllRVEILLANCE REQllIREt!ENTS S
I
- 3. Duse from I-131, t ritium, and radioactive pa r ticula tes
- 3. Dose from I-131, tritisim, and radioactive particulates.
l j
with half-lives greater than eight days.
with half-lives greater than eight days.
p i
- n. The done to any organ of an initividual due to
- a. Cumulative dose contributions for the current calendar I-131, tritium, and radioactive particulates quarter and year from I-131, tritium, and radioactive I
with hal t'-lives greater than eight days released particulates with half-lives greater than eight days in from the site (Figure 3.8.2) in gaseous effluent gaseous ef fluents shall be determined in.iccordance with.
y unal t be limited to the following:
the ODCH monthly.
8
,..p i"
1.
During any calendar quarter to $7.5 mrem, and
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D 2.
During any calendar year tof 15 mrem.
- h. With the calculated done from the release of I-131, t r i t ium, and rad ioac t ive particulates with i
half-lives greater than eight days exceeding any of the,ahove limits, within 30 days
?
submit to the Commission, in lieu of any
.I other report, a special report which i
identifies the cause(s) for exceeding the i
limit (s) and defines the corrective actions
)
taken to reduce the releases and the pro-posed corrective actions to be taken to j
assure the subsequent releases will be within the above limits.
ij l
3.H/4.8 198a Amendment No. 15
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3.0 I.IlllTING CONN!TIONS FOR OPERAT10ri 4.0 SilRVEll.l,ANCE REQUIREMENTS i
j
- 4. Offgas Treatncnt System
- 4. Offgas Treatment System
\\
- a. The offgas treatment system,shall be in.
a.
Following each isotopic analysis of a sample of i
operation whenever the main condenser air gases from the steam jet air elector required by 1
l ejector system is in opera tion. Components 4.8.R.S.c, verify that the maximum storay,e tank of the system shall he operated to provide activity limit specified in 3.8.B.4.e cannot he d.
the maximum holdup time obtainable except exceeded using the method in the ODCti,
{
p sharing periods of equipment maintenance.
]
- h. Uith gaseous waste being discharged for more than 7 days with a holdup time of less than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, within 30 days submit to the Commission, in lieu of any other report, a special report which includes the l
following information.
4 i
1.
T;tentification of equipment or sub-systems not functional and the t
reason.
j 2.
Action (s) taken to restore equipment i
to functional status.
..I 3.
Sunmia ry ilescript ton of act lon(s) taken to prevent a recurrence.
2 r
3.8/4.8 198h Amendment No. 15 l
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i l'a 3.0 1,IHITING COrlDITIONS FOR OPERATIDH 4.0 SURVEll.i.ANCE REQUIREt!ErlTS I,
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- c. The concentration of hydrogen in the offgas 1
treatment system.shalt he limited to $ 2% hy a
41 volume downstream of the recombiners. IJith 4
the concentration of hydrogen in the offgas l
Is treatment system > 2% hy volume, but $ 4% by
[:f l'l volume, res tore the concentration of hydro-
}
gen to < 2% by volume within t.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or suspend operation of the compressed storage y.
l2 subsystem.
t.
i.F.. ;
t
,C matically close the recombiner inlet valves 4
l'a at $ 4% hydrogen by volume.
1 The quantity of radioactivity af ter 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
)
e.
holdup conta,ined in each gas storage tank shall
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he limited to f 22,000 curies of noble gases (considered as dose equivalent Xe-133).
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3.0 IIMITING CONDITIONS FOR OPEHATION 4.0 SURVEILLANCE REQUIREMENTS i,
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- 5. Itain Condenser Of fgas Activity
- 5. Main Condenser Offgas Activity t
a.
The gross gamma radioactivity release
- a. The activity of radioactive material in gaseous form rate ocasured at the stegm jet air ejector shall removed f rom the main condenser shall be continuously be limited to g 2.6 x 10 uci/sec following monitored by the steam jet air ejector monitors a 30-minute decay.
in accordance with Table 3.8.2.
j I
b.
When the limit in (a) above is exceeded,
- b. The steam jet air ejector monitors shall be set to l
restore the gross gamma radioactivity release automatically terminate offgas flow'within 30 minutes
+
rate to within the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or at the limit established in Specification 3.8.B.S.a.
be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. The gross gamma radioactivity of noble gases from the main condenser air ejector shall be determined to be within the limit specified in 3.8.B.5.a at 4
{
the following times by performing an isotopic analysis of a representative sample of gases:
j 1.
Once every month.
2.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an increase j
in the continuous monitor reading of 50% af ter factoring out increases due to power level.
1
- 6. Containment Venting'and Purging
- 6. Containment Venting and Purging a.
Except for inerting operations following
- a. Except for inerting operations following startup
[
startup and deinerting prior to shutdown, and deinerting prior to shutdown, the containment containment venting and purging above cold shall be determined to be aligned for venting or shutdown shall be via the 2-?nch bypass purging through the Standby Gas Treatment Syst,em flow path using the Standby das Treatment within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to start of and at least once System.
per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during venting or purging of the containment above cold shutdown.
j f
- b. Containment inerting following startup and
- b. Prior to containment venting or purging, the I
deinerting prior to shutdown shall be via sampling and analysis requirements of Table 4.8.4 d
the Reactor Building plenum and vent, shall be met.
3.8/4.8 198d s
j Amendment No. 15
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3.0 LilliTING CONDITIONS FOR OPERATION 4.0 SURVElLLANCE REQUIRFJIENTS y
C. Solid Radioactive Waste C. Solid Radioactive Waste I
1.
Solid Radwaste System
- 1. Solid Radioactive System
- l A solid radwaste system shall be Operable a.
The Process Control Program (PCP) shall be used a.
i and used, as applicable, in accordance wLth to verify the Solidification of at least one a Process Control Program for solidi
- tcation representative test specimen from at least every j
and packaging of radioactive wastes to ensure tenth batch of each type of wet radioactive meeting the requirements of 10 CFR Part 20 waste (e.g. filter sludges, spent resins, and of 10 CFR Part 71 prior to shipment of and chemical solutions).
T l
radioactive wastes from the site.
i l'
- b. With the packaging requirements of 10 CFR
- b. If any test specimen fails to verify Solidifica tion, Part 20 or 10 CFR Part 71 not satisfied, the Solidification of the batch under test shall he
- p
]
suspend shipments of defectively packaged suspended until such time as additional test specimens ' '
]
solid radioactive wastes from the site.
can be obtained, alternative Solidification parameters.,
j can he determined in accordance with the PCP, and a sub -
sequent test verifles Solidification. Solidification i,.
1 of the hatch may then be resumed using the alternative Solidification pa.cameters determined by the PCP.
s I
c.
If the initial test specimen from a batch of waste fails to verify Solidification, the PCP shall provide for the collection ard testing of representative test s pecimens from each consecutive batch of the same type 4
of wet waste until at least three consecutive initial test specimens demonstrate Solidification. The PCP shall be modified as required, as provided for in 7
Section 6 of the Technical Specifications.
g i
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3.8/4.8 193e Amendment No. 15 k
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E 5
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i 4.0 SURVEILLANCE REQUIREMENTS 3.0 LilllT[NG CONDITIONS FOR OPEllATION D. Dose from All Uranium Fuel Cycle Sources D. Dose from A'.1 Uranium Fuel Cycle Sources
- 1. Cumulative dose contributions from all liquid and
- 1. The dose or dose commitment to,any member of the general public from all uranium gaseuus effluents shall be determined in fuel cycle sources is Limited tos25 mrem accordance with Specifications 4.8.A.2.a.
to the total body or any organ (except for 4.8.!!.2.a. and 4 ' 8.B.3.a and in accordance wi th the cliyrold, which is limited to < 75 mrem) the 00CH.
8 i
over a period of 12 consecutive months.
i a
- 2. With.the calculated dose from the release of i
radioactive noterials in lignid or gaseous effluents exceeding twice the limits of Specifications 3.8.A.2.a.1, 3.8.A.2.a.2,
.'l 3.8.H.2.a.1, 3.8.II."4.a.2, 3.8.B.3.a.1, or 3.8.ll. 3.a.2, prepare and submi t wi thin 30 I
days of special report, in lieu of any other report, to the Commission which defines corrective actions and calcu-lates the highest radiation exposure to
-l any member of the general public f rom ali u ran tum fuel cycle sources (including all effluent p.athways anil direct radiation).
linless this report shows that exposuren are less than the 40 CFR Part 190 standard, either apply to the Commission for a variance to continue releanes which exceed the 40 CFR
.]
190 standerd or reduce subsequent releanes Pa r t to pe rm i t the standard to be met.
-i
,i 1.
I 4
198f
'l 1.8/4.8 5
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.O TAHLE 3.8.1 - RADIOACTIVE LIQUID EFFLUENT !!ONITORING INSTRUMENTATION l
(Page 1 of 2)
Hinimum Channels Instrument Operable Applicability Action if !!inimum Channels not Operable i
Liquid Radwaste Effluent Line 1
During release of Effluent releases may continue for up to 14 days
[
Cross Radioactivity Honitor*
liquid radwaste provided that prior to initiating a release:
1 a.
At least two independent samples are analyzed j
l in accordance with Specification 4.8.1.d b.
At least two technically qualified members of the Facility Staf f independently verify the release rate calculations and discharge line valving;
}
Otherwise, supsend release of radioactive ef fluents I
via this pa thway.
l.iquid Hadwaste Ef fluent Line 1
During release of Effluent releases via this pathway may continue for Flow Instrument liquid radwase up to 30 days provided the flow rate is estimated at least once every four hours during actual releases.
Pu.up curves may be used to estimate flow.
- i i
Discharge Canal Flow Effluent releases via this pathway may continue for up
, Heasurement to 30 days provided the flow rate is estiroted at least t
y Open Cycle flode 1
During release of once every four hours during actual releases.
Pump l
q Closed /llelper Cycle Mode i
liquid radwaste curves may be used to estimate flow.
l Discharge Canal Cross 1
At all times Liquid radwaste releases may continue for up to 30 days Vj Radioacttvity Monitor
- provided that at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> a grab sample i
i j
shall be collected and analyzed for gyoss beta and j,
gamma radioactivity at an I.LD of 10 uC1/mt.
9 Service Water Discharge Pipe At all times Service water discharge may continue for up to 30 days
{
Cross Radioactivity Monitor
- provided that at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> a grab g
sample is collected and analyzed foggross beta,and I
gamma radioactivity at an LLD of 10 uC1/ml.
l
{
3.8/4.8 1981 g
a o
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i i
Amendment No. 15
{
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l 9
TAHI.E 3.8.1 - RADI0 ACTIVE LIQUID EFFLtlErrf !!ONITORING INSTRUt!ENTATION
^
]
(Page 2 of 2)
Hinimum Channels Instrument Ope rable Applicability Action if Minimum Channels not Operable 1
i Turbine IIuilding Normal 1
At all times
. Liquid sump releases may continue for up to 30 Drain Sump flonitor*
days provided that at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> a grah sample shall be collected and analyzed for j
grogs beta and gamma radioactivity at an I.LD o f '
i 10 uCi/ml.
i.evel !!onitors for 1
When tants are Liquid additions to a tank may continue for up Temporary Outdoor in use to 30 days provided the tank liquid level is Tanis lloiding estimated during all lignid additions.
Radioactive I,tquid a
I l
L il I
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- - Indicates monitor provided with autonutic alarm.
1 3.8/4.8 198j t
Amendment No. 15 1
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i-Tant.E 3.8.2
- RADIOACTIVE CASEOUS EFFLUlMr flotlIToltlNG illSTRUt!EllTATIOil j
(Page 2 of 2) i i
!!inimum Channels Instrument Operable Applicability Action if ttinimum Chnnuels not Operable I
)
lleactor liu11 ding Vent (inclinics Turbine Building I. Itadwaste nullding releases) tioble Gas Activity flonitor 1
At all times Releases via this pathway may continue for up j
to 30 days provided grab samples are taken and 8
analyzed at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
i i
i iodine Sampler Cartridge 1
At all times Releases via the pathway may continue for up to 30 i
days provided within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> samples are continuously
[
collected with auxiliary sampling equipment as t
required by Table 4.8.4.
i
'1 Particulate Sampler Cartridge 1
At all times Releases via the pathway may continue for up to 30 days provided within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> samples are continuously collected with auxiliary sampling equipment as required by Table 4.8.4.
Duct Flow tionitors 1
At all times Releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Sample Flow Instrinnents I
At all times Releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Notes:
- - Indica tes number of channels required per opera ting recoml !ner train.
I
- - Indicates noble gas ef fluent monitor having provision for automatic ternination of gaseous release.
i 3.8/4.8 198-1 4
Amendment No. 15 I
s h
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' /:4 Y'I' IMAGE EVALUATION
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TEST TARGET (MT-3)
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I TAltLE 4.8.1
- RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUttENTATION SURVEILLANCE REQUIREt1ENTS Source Sensor Check Check Functional Test Instrument Frequency Frequency Fregt.eacy Calibration Frequency I
hiquid Radwaste Effluent Daily during Immed ia tely Within 3 months prior Within 12 months prior to j
bine Cross Radioactivity release Prior to to making a release making a release.*
t Honitor Each s
Release t
Liquid Radwaste Effluent Daily during Within 3 months prior Within 12 months prior to
(
Line Flow Instrument release
.to making a release making a release, j
n a
instruments used in Daily during Within 3 months prior Within 18 months prior to Determination of Discharge release to making a release making a release.
l j)
Canal Flow l[
i.
Service Water Discharge Daily 11onthly Quarterly Each Operating Cycle
- i Pipe Cross Radioactivity Honitor g
l
' Discharge Canal Grosn Daily Monthly Quarterly Each Operating Cycle
- Radioactivity Honitor a
Turbine Itailding Daily Monthly Quarterly Each Operating Cycle Normal Drain Sump ilonitor Level Honitors for Daily when Quarterly when Each Operating Cycle Temporary Outdoor Tanks in use in use when in use j
llolding Radioactive Liquid g
1
- - The initial Instrument Calibration shall be performed using one or more of the reference standards certified by the National Ilureau of Standards (NHS) or using sources traceable to NBS standards. These standards shall permit calibra t tug the sys tem over its intended range of energy and measurement range.
For subsequent calibration sout;ces that have been related to the initial calibration shall he used.
..)
3.t1/4.8 198m Amendment No.15 l-
,y t
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2 TABLE 4.8.2
- RADIDACTIVE CASE 00S EFFLUENT HONITORING INSTRIRIENTATION SURVEILLANCE REQUIREMENTS (Page 1 of 2)
Source Sensor Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequency 4
1 A
Noble Cas Activity Monttors air ejector operation Quarterly Once each Operating Cycle Main Condenser At r Ejector Daily during
}
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. Itain Condenser Offgas Treat-Daily during Monthly Quarterly I i
'i ment System Ilydrogen Honttors air ejector operation N
,[
Plant Stack Noble Gas Daily Monthly Quarterly Once cach Operating Cycle
- j Activity Honitors
'5
.f Plant Stack Todine and Weekly Particulate Samplers r
i Plant Stack Flow (lonitor Daily Once cach Operating Cycle
(
l Plant Stack Sample Flow Daily Once each Operating Cycle j
4 Instruments J
l l
1
/
.J 1
l1
.1 f:
3.8/4.8 198n a
p Amendment No. 15 fa "1
.. ~,
s i
TABLR 4.8.2
- RADI0 ACTIVE CASROUS EFFLIIEtiT HONITOtING INSTRUMENTATI0tl j
SURVEILLANCE REQUIREMENTS l
(Page 2 of 2)
Source Sensor Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequency i
Reactor Building Vent Daily Honthly Quarterly once each Operating Cycle
- Noble Cas Activity Honitors Reactor Building Vent Daily Honthly Quarterly once each Opeiating Cycle #
1 Isolation Noble Cas Monitors b
Reactor Building Vent Weekly Iodine and Particulate Samplers Reactor Building Vent Duct Daily Once each Operating Cycle Flou Honitors Reactor Building Tent Daily Once each Operating Cycle Sample Flow Instruments b
i.
- - The initial Instrument Calthration shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using sources traceable to NBS standards. These standards shall l
3 permit calibrating the system over its intended range of energy and measurement range. For subsequent I
calibration sources that have been related to the initial calibration shall be used.
'i f - The Calibration shall incit.de the use of standard gas samples containing a nominal four volume percent
- hydrogen, i
3.8/4.8 198o
[
i Amendment No. 15 g
4 L
1 t
L
. D.
TABLE 4.8.3
- RADI0 ACTIVE LIQUID WASTE SAMPLINC AND ANALYSIS PROGRAN (Page I of 3)
I J
Sampling Min timam Type of Activity Lower Limit of Liquid Release Type Frequency Analysis Frequency Analysis Detection,g.LD)
(uci/ml)
-7 Batch Waste Release Each datch Each Batch Principal Camma 5 x 10 Tanks,'
d Emitters
-6 l
l-131 1 x 10 One Ba tch one Batch Dissolved and I x 10-Each Honth Each Month Entrained Casca i
-5 t
Each Batch Monthly 11 - 3 1 x 10 Composite" 1
t i
l Cross alpha 1 x 10~
I
-8 Eacle Batch Quarterly Sr-89, Sr-90 5 x 10 Composite l
-6 Fe-55 1 x 10 3.8/4.8 198p
[
3' Amendment No.15 j
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- 1 i
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b I
' CMZ'___. A.i_ J. ; ~O- ' - s m ? L,.
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m.
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v t
l TABLE 4.8.3
- RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l
(Page 2 of 3) b Notes The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95%
a t
ib probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
{
4:]
3 For a particular measurement system (which may include radiochemical separation):
-i 9
4.66 s b
~
=
4 E. V. 2.22. Y. exp(-Nt) t 8
4 where:
I t
LLD is the a priori lower limit of detection as defined above (as picocurie per unit mass or volume),
s is the standard deviation of the background counting rate or of the counting rate of a blank sample
}'
s b
,q r
as appropriate (as counts per minute).
Typical values of E, V, Y and At shall be used in the caldulations.
E is the counting efficiency (as counts per transformation),
i i,
V is the sample size (in units of mass or volume),
s F
j 2.22 is the number of transformations per minute per picoeurie, i.
Y is the fractional radiochemical yield (when applicable),
t; his the radioactive decay constant for the particular radionuclide, and L
is the elapsed time between midpoint of sample collection and time of counting.
At b
l I
' I tl E
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?
3.8/4.8 t
1989
[
't Amendment No. 15 t4 D
t i
O;;
i i.
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t
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s 1
p TABl.E 4.8.3 - RADIOACTIVE I.IQUID WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) o (Page 3 of 3)
I i
b.
A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling.
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid
[
E c.
waste discharged and in which the method of sampling employed results in a specimen which is representative of 1
the liquids released.
o I
t d.
The principal gamma emitters for which the LI.D specification will apply are exclusively the following radio-1; unclideu:
Hn-54, Fe-59 Co-58, Co-60, 2n-65. Ho-99, Cs-134. Cs-137, cc-141, and Ce-144.
This lis't does not mean tiust only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
[
t Nuclides which are below the LLD for the analyses shall be reported as "less than" the LLD of the nuclide and u.
1 should not he reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Semiannual Radioactive Ef fluent Release Report.
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i TABLF. 4.8.4 - RADI0 ACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (Page 1 of 2) i Sampling Minimum Type of Activity Lower Limit of j
Caseous Release Type Prequency Analysis Frequency Analysis Detection {LgD) j.
(uct/ml)"'
l' i
I Containment Purge Each Purge Each Purge Principal Camma 1 x 10 l
I Grah Emitters (f)
I Sample h
-6 i
11 - 3 1 x 10 s
1 I
Monthly Honthly Principal Camma
,4 l
Plant Stack anal e
Reactor Building Grah Emitters (f) 1 x 10 I
i Sample I
l Vent g
-6 11 - 3 l
1 x 10 t
i 6
+
f ContinuousE Weekly" I-131 1 x 10-l Charcoal
-10 Sample I-133 l
1 x 10 i
E
-II continuous i
Weekly" Principal Camma 1 x 10 i
Particulate Emitters (I-131, I
Sample others)
(
Continuous Monthly Composite! Gross alpha 1 x 10 "
f.
E Particulate P
Sample a
i E
-11 Continuous Quarterly Sr-89, Sr-90 1 x 10 j
Composite j
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- Particulate t
Sample i
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-6 1
Continuous Continuous Cross gamma 1 x 10 monitor or gross beta noble gas activity i.
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- 3. 8 /4. 8 198s q
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Amendment No. 15 6'
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5 TABI.E 4.8.4 - RADIDACTIVE GASEntlS WASTE SAtlPl.ING AND ANAL.YSIS PROGRAll (continued) i j
(Pane 2 of 2)
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Notes:
I a.
The I.I.n is the smallest concentration of radioactive material in a sample that will be detected with 95%
I prohahtlity with 5% probability of falsely concluding that a blank observat ton represents a "real" signal.
Note (a) of Table 4.8.3 is applicable.
b.
Grah samples taken at the discharge of the plant stack and reactor building vent are generally below minimum detectable levels for most nuclides with existing analytical equipment.
For this reason, isotopic analys ts data, corrected for holdup time, for samples taken at the steam jet air ejector may he used to calculate
{
nohle gas ratlos.
,1 c.
Uhenever the steady state radiolodine concentration is greater than 10 percent of the limit of Specification
[
3.6.C.1, da'ily sampling of reactor coolant for radtoactive todines of I-131 through I-135 is required. Whenever a change of 25% or more in calculated Dose Equivalent I-131 is detected under these conditions, the iodine and particulate collection devices for all release points shall be removed and analysed daily until it is shown j
that a pattern exists which can he used to predict the release rate.
Sampling may then revert to weekly.
When samples collected for one day are analyzed, the corresponding I,I.D's may he increased by a factor of 10.
Samples shall be analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after removal.
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d.
To be representative of the average quantities and concentrations of radioactive materials in particulate form 8
l In gaseous effluents, samples should he collected in proportion to the rate of flow of the ef fluent streams, t
The principal gamma emitters for which the I.LD specification will apply are exclusively the following radio-c.
nuclides:
Kr-87, Kr-88, Xc-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Hn-54, Fe-59, Co-58,
[
p.
Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for part iculate emissions. This list does not mean l4 that only these nuclides are to be detected and reported. Other peaks which are measurabic and identifiable,
[
]
together with the above nuclides, shall also be identified and reported.
'1
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f.
Nuclides which are below the LI.D for the analyses shall be reported as "less than" the I.I.D of the nuclide and
[
q should not he reported as being present at the I.I.D level for that nuclide. The "less than" values shall not he c
3 used in the required dose calculations. When unusual circumstances result in I.I.D's higher than reported, the 4
I reasons shall be documented In the semlannual effluent report.
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The ratio of the sample flow rate to the sampled stream flow rate shall he known for the time period sampled.
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I li. 11 analysis shall not he required prior to venting if the Ibetts of 3.8.B.1 are satisfied for other nuclides.
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The analysis sliall he completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter sampling, however.
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1.
In lieu of grah samples, continuous monitoring ulth bl-weekly analysis using silica-Jet samplers may he provided.
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3.8 and 4.8 Bases:
A.
I.iquid Effluents i
specification 3.8.6 1 is provided to ensure that the concentration of radioactive materials released in liquid waste e f fluents from the site will be less than the concentration levels specified in 10CFR Part
- 20. Appendix B. Table II.
This limitation provides additional assurance that the levels of radioactive materials in the llississippi River will not result in exposures exceeding (1) the Section II. A design object ives of Appendix I, 10 CFR Part 50, to an Individual and (2) the limits of 10'CFR Part 20.106(c)
+
to the population. The concentration limit for noble gases is based upon the assumption that Xc-135 is the controlling radioisotope and its flPC in air (submersion) was converted to an equivalent concentration i
in water using the methods described in International Commission on Radiological Protection (ICRP)
Publication 2.
.ipec i f ica t ion 3.8. A. 2.a is provided to implement the requirements of Sections II. A.
III.A and IV.A of Appendix 1, 10 CFR Part 50.
The 1.imiting Condition for Operation implements the guides set forth in Section II.A of Appendix 1.
Action required by Specification 3.8.A.2.b provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material 'in liquid ef fluents will be kept "as low as is reasonably achievable".,Considering that the nearest drinking water supply using the receiving water is 0
33 river miles downstream, there is reasonable assurance that the operation of the facility will not,
result in radionuclide concentrations in the finished drinking water that are in excess of the require-ments of 40 CFR 141. The dose calculations in the ODCtl Implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculati,onal procedures based on i
models and data such that the actual exposure of an individual through appropriate pathways is unlikely l
to be substantially underestimated. The equations specified in the ODCtl for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be consistent with the a
methodology provided in Regulatory Guide 1.109, "Calcula tion of Annual Doses to Plan 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 1, Revision 1," April 1977.
NUREC-0133, October, 1978, provides methdds for dose calculations consistent with Regulatory Culdes i
j 1.109 and 1.113.
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3.8/4.8 BASES 198u h
Amendment No. 15
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r 3.8 and 4.8 !!ases: (continued)
Spec ificat ion 3.8. A.3 provides assurance that the liquid radwaste treatment system will be available for use whenever liquid ef fluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system he used when specified provides assurance that the releases of radioactive materials in 11guld ef fluents will be kept "as low as is reasonably achievable." This speci f ica t ion implements the requirements of 10 CFR Part 50.36a, General Design criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix I to 10 CFR Pa r t 50.
The specified limits governing the use of appropriate portions of the liquid radwast-treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR Pa r t 50, for liquid effluents.
l Restrictions on the quantity of radioactive liquid material contained in tanks are required only for temporary tanks.
All exterior permanent tanks are diked to prevent release of their contents in the l
cvent of leakage. Restricting the quantity of radioacti.ve material contained in the specified tanks j
provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would he less than the limits of 10 CFR Part 20, Appendix B. Table II, Column 2, a t the nearest potable water supply and the nearest surface water supply in an unrestricted area.
g Surveillance requirements for continuous liquid release points are not provided since all Monticello releases are "b5tch" type releases.
]
Radioactive liquid effluent instrumentation is provided to monitor and control, as appitcable, the
~
releases of radioactive materials in liquid ef fluents during actual or potential releases of liquid effluents. The alarm setpoints for these instruments will he calculated in accordance with NRC approved t
methods in the ODOI to ensure that
~
the alarm will occur prior to exceeding the limits of 10 CFR Part 20.
The operability requirements for instrumentation are consistent with the requirements of General Design J
Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.
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3.H/4.8 198v E
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Amendment No.15 i
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l 3.8 and 4.8 Bases: (continued)
B.
Caseous Ef fluents t
i Specification 3.H.B.I.a is provided to ensure that the dose rate at anytime at the site boundary from gaseous ef fluents 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 4
provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix B Table 11 of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within l
the restricted area boundary, the occupancy of the individual will be suf ficiently low to compensate for ay 8
Increase in the atmospheric dif fusion factor above that for the restricted area boundary. The specified release O
rate limits. restrict, a t all times, the corresponding gamma and beta dose rates above background to an individual j
at or beyond the restricted area boundary to 500 mrem / year to the total body or to 3000 mrem / year to the skin.
These releane rate limits also restrict, a t all times, the corresponding thyroid dose rate above background
(
Ig to sl500 mrem / year at the site boundary, f
Specification 3.8.B.2.a is provided to implement the requirements of Sections II.8, III.A and IV. A of Appendix I, y
The I.imiting Conditions for Operation implement the guides set forth in Section II.B of Appendix 3
I.
Action requtred by Specification 3.8,.B.2.h provides the required operating flexibility and at the same time implement the g,uides set forth in Section IV.A of Appendix I assure that the releases of radioactive material in
)
'j gaseous ef fluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the 3
requ i remen ts 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 actual exposure of an Individual through the M
appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the
)h ODCil for calculating the doses due to the actual release rates of radioactive noble gases in gaseous ef fluents
+
will be consistent with the methodology provided in Regulatory Culde 1.109, " Calculation of Annual Doses to Man f
]
f rom Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix l
1," Revision 1, October 1977 and Regulatory Guide 1.111, "Hethods for Estimating Atmospheric Transport and i
Dispersion of Caseous Ef fluents in Routine Releases from hight-Water-Cooled Reactors," Revision 1. July 1977.
{
f The OIR'M equations provided for determining the air doses at. the restricted area boundary may be based upon the Q
historical average atmospheric conditions.' NUREC-0133, October, 1978 provides methods for dose calculations i
consistent with Regulatory Guides 1.109 and 1.111.
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Amendment No. 15 i
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3.8 and 4.8 Bases: (continued)
Specification 3.8.B.3.a is provided to implement the requirements of Sections II.C. 11[. A and IV. A of Appendix I, 10 CFR Pa r t 50.
The I.imiting Conditions for Operation are the guides set forth in Section II.C of Appendix 1.
The release rate specifications for I-131, tritium, and radioactive particulates with half-lives greater than
[
eight days are dependent on the existing radionuclide pathways to man in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, j
- 2) deposition of radionuclides onto green leafy vegetatica with subsequent consumption by man, 3) deposition onta grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by nun,
and 4) deposition on the ground with subsequent exposure of man.
t l
1 Spec tf tcation 3.8.H.4 provides assurance that the offgas treatment system will be in operation whenever i
main condenser offgas is released to the environment. The requirement that the appropriate portions of this 3
system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous
[
ef fluents will be kept "as low as is reasonably achievable."
'this specification implements the requirements of 10 CFR rart 50.36a, Ceneral Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section 110 of Appendix I to 10 CFR Pa rt 50.
The specified limits governing the use of appropriate portions of the systems we're specified as a suitable f raction of the guide set forth in Sections II.B and II.C of Appendix 1, 10 CFR Part j
50, for gaseous effluents.
t
'4 Tf Radioactive gaseous ef fluent instrumentation is provided to monitor and control, as applicable, the releases of f
P radioactive materials in gaseous ef fluents during actual or potential releases of gaseous ef fluents. The alarm / trip setpoints for these instruments will be calculated in accordance with NRC approved methods in the ODCH to ensure j
that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The operability requirements for i
4 this instrumentation are consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A h
l to 10 CFR Part 50.
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.. i 3.8 and 4.8 Bases: (continued)
Speci f ication 3.8.B.4.c is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen.
k Automa tic cont rol features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. Maintaining the concentration of hydrogen below the flammability limit provides assurance th,a t the releases of radioactive materials will be controlled in conformance with the require-
}
ments of Ceneral Design Criterton 60 of, Appendix A to 10 CFR Part 50.
f Specification 3.8.8.4.e is provided to limit the radioactivity which can be -stored in one decay tank.
Restricting i
the quantity of radioactivity contained in each gas storage tank provides assurance that. in the event of an h
uncontrolled release of the tanks contents, the resulting total oody exposure to an individual at the site 1
]
restricted area boundary will not exceed 20 mrem.
A flow restrictor in the discharge line of the decay tanks t
[
g prevents a tank from being discharged at an uncontrolled rate.
In addition, interlocks prevent the contents of a i
i tank from being released with less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of holdup.
i a
7 Spectf tcation 3.8.B.5 establishes a maximum activity at the steam jet air ejector. Restricting the gross radio-l activity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure
[
to an individual at the restricted area boundary wiLL not exceed the limits of 10 CFR Part 20 in the event this effluent is inadvertently discharged directly to the' environment with minimal treatment. This specification i
y Implements the requirements of Ceneral Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50
{
7 d
i Specification 3.d.B.6 requires the containment to be purged and vented through the standby gas treatment, system
]
except during inerting and deinerting operations. This provides for iodine and particulate removal from the L
containment atmosphere. During outages when the containment is opened for maintenance, the containment ventilation j
exhaust is directed to the immitored reactor building vent.
Use of the 2-inch flow path prevents damage to the
[
standby gas treatment system in the event of a loss of coolant accident during purging or venting. Use of the
[
reactor building pienum and vent flow path for inerting and deinerting operations permits the plenum monitors to automatteally terminate releases in the event that release rate limits are exceeded.
.i 1
C.
Solid Radioactive Waste Specifica tion 3.8.C. I provides assurance that the solid radwaste system will be used whenever solid b
radwastes require processing and packaging prior to being shipped offsite. Th it., specification implements the
[
requirements of 10 CFR Part 50.36a and Cencral Design Criteria 60 of Appendix A to 10 CFR Part 50.
lI 3.8/4.8 HASES 198y Amendment. No. 15 t
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D. Dose from All Ilranium Fuel Cycle Sources f
Specification 3.8.D is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation
[
and subietttal of a special report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix 1.
Subielttal of the report is considered a timely request and a variance is granted until Staff action on the requent is complete. For sites containing up to 4 reactors, it is highly unlikely that the resultant Jose to a real individual will exceed 40 CFR 190 if the individual reactors remain with the reporting require-(
ment level. For the purpose of the special report it may be assuised that the dose comaltment to the real individual from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear feel cycle r
q facilities at the same site or within a radius of 5 miles must be considered.
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3.8/4.8 BASES 198z I
Amendment No.15 E
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ei 3.0 1.IrllTir1C Car!DITint!S FOR OPERATIOtl 4.0 SIMVEll.l.ANCE REQlllREtlENTS
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4.16 RADI ATInfi ENVIR0felENTAI, MO!1ITORTNG PROGRAtt
+
1
'y Applicability Applies to the periodic monitoring anil recording of 4
radioactive ef fluents found in the plant environs.
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Objective f
3 To provide for measurement of radiat ton levels anil i
radioactivity in the site environs on a contlntilng l
a hasis.
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.g Specification
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3 A.
Sample Collection & Analysis
- 1. The Raillation Environmenta.1 Ploni toring i
Program given in Table 4.16.1 shall he conducted.
Radioanalysis shall be conducted meeting the requirements of
[
Table 4.16.2.
(
A map and a table identifying the locations
[
of the sampling points shall be provided in the Offsite Dose Calculation !!anual (00Ctl).
2.
11henever the Radiation Environmental
~
Ilonitoring Program is not being con.Inctei1 T
as specified in Table 4.16.1, th.t Annual,
(
j Radiation Environmental Monitoring Report h
1 shall include a description of the teasons l
for not conducting the program as required f
and plans for preventing a recurrence.
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V 3.16/4.16 229g i
4 Amendment No. 15 f
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3.0 f.ItitTING CONDITintiS Folt OPl?RATlut!
4.0 SllRVEII.I.AllCE REQt!IREt1EtiTS
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-l 3.
Devia tions are permitted f rom the requ ired j
1 sampling schedule if samples are unohtainable
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due to hazardous condittons, seasonahle unavatla-
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htlity, or to malfunction of automatic sampling i
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equipment.
If the latter occurs, every ef fort shall be made to complete corrective action
}]
prior to the end of the next sampling period.
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4.
tilth the levet of radioactivity in an environ-l mental sampling medium exceeding the reporting Y
, levels of Table 4.16.3 when averaged over any calendar quarter, in lieu of any other report,
j prepare and subalt to the Commission within 30
(
days from the end of the affected calendar quarter j
a Report guirsuan t to Specif ication 6.7.C.2.a.
11 hen more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report l
shall be submitted if:
j i
concentration (1)
, concentration (2)
) g*g ilmit level (1)
Limit level (2)
I l
11 hen radlonnelides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential j
annual dose to an. individual is equal to or greater j
than the calendar year limits of Specifications
- 3. 8. A. 2, 3. 8. n. 2, o r 1. 8.H. 3.
This report is no t l
required if the measured icvel of radioact ivity. was not the result of plant effluents; however, in such
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an event, the condition shall be reported and deserthed
?
In the Annual Radiation Eavironmental ifonitoring Report.
3.16/4.16 229h l
- I Amendment No. 15 J
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3.0 LIttITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREllENTS g
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5.
Although deviations from the required sampling schedule are permitted under Item 3, above,
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. 1 whenever milk or leafy green vegetables can no 1
longer be obtained from the designated sample locations required by Table 4.16.1, the Semi-3' annual Radioact'ive Effluent Release Report for this period shall explain why the samples can
{
no longer be obtained and will identify thet locations which will be added to and deleted f rom the monitoring program as soon as practicable.
R.
Land Use Census t
I 1.
A land use census shall be conducted and shalI t
Identify the location of the nearest milk animal, I
the nearest residenge, and the nearest garden of greater than 500 ft producing fresh leafy vegetables, in each of the 16 meteorological I
sectors within a distance of five miles. The census shall also identify the all milk animals and all 500 f t}ocations of or greater J
gardens producing broad leaf vegetation in each h
I of the meteorological sectors within a distance of three miles. This census shall he conducted l
at least once per year between the dates of Hay 1 and October 31 by door to door survey, aerial survey, or by consulting local agricultural a
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authorities associations.
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3.16/4.16 2291 a
Amendment No. 15 I
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3.0 LIMITING CONDITIONS FOR OPERNFION 4.0 SURVEILLANCE REQUIREMENTS
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I 2.
With a land use census identifying a location which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently l
heing obtained in accordance with Specification 4.16.A.1, the Semiannual Radioactive Effluent j
Release Report for this period shall identify the new location. The new location shall be added 8 to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted f rom this mont-toring program after October 31 of the year in which I
this land use census was conducted, t
J C.
Interlaboratory Comparison Program 7
8 1.
Analyses shall be performed on radioactive materials
- j l
supplied as part of an NRC approved Interlaboratory comparison program as described in the ODCH.
~
2.
The results of analyses performed as a part of the J
l above required program shall be included in the Annual i
i Radiation Environmental Honitoring Report. When required analyses are not performed, corrective lj action shall be reported in the Annual Radiation Environmental Honitoring Report.
3.16/4.16 229j i
l' ij
!q Amendment No. 15
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.li MONTICELLO NUCLEAR CENERATING PLANT f
L RADIATION ENVIRONMENTAL ts)NITORINC PROCRAM I
SAllPLE COLLECTION AND ANALYSIS e
Number'of Sansples Exposure Pathway and Sampling and Type and Frequency i
J and/or Sample Sample Locations **
Collection Frequency of Analysis i
1.
Airborne Radiotodine &
Sainples fro s 5 locations:
Continuous Sampler Radioiodine analysis Particulates 3 samples from offsite operation with sampler Weekly for I-131 locations (in different collection weekly.
q sectors) of the highest Par ticula te :
[
calculated annual average Cross beta activity ground level D/Q, I sampl.e, on each filter weekly *.
f J
from the vicinity of' a com-Analyses shall be per-munity having the highest formed more than 24 f
calculated annual average hours following filter ground-level D/Q, and change., Perform gamma 1 sample froin a control isotopic analysis on location 8-20 miles dis-composite (by location) tance and in the least sample quarterly.
prevalent wind direction 2.
Direct i
Radiation 37 TLD stations established Quarterly Camma case quarterly
,1 with dupitcate dosimeters placed at the following locations:
i e
If gross beta activity in any indication sample exceeds 10 times the yearly average of the control sample, a gannua isotopic analysis is required.
- Sample locations are given on the figure and table in the ODCH.
3.16/4.16 229k a
s Amendment No. 15
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,= n, _a., ~~
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.l Table 4.16.1
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l (Page 2 of 5)
MONTICEI.LO NUCLEAR CENERATING PLANT RADIATION EllVIRONMENTAL HolllTORING PROCRAM SAMPLE COLLECTION AND ANALYSIS Number "of Samples i
Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations **
Collection Frequency of Analysis 1
l 2.
Direct Radiation (Con't.)
1
- 1. Using the 16 meterological j
wind sectors as guidelines, G
an inner ring of station in the general area of the site boundary is estabitshed l
i and an outer ring of stations i
at 4 to 5 mile distance frois the plant site is established.
i 1
Ilecause of inacessibility, two sectors in the inner and I
outer rings are not covered.
i
- 2. Seven dosimeters are established
- j at special interest areas and a control statlon.
-]
3.
tlaterhorne a.
Surface Upstream & downstream Monthly composite of Canwna Isotopic analysis locations weekly samples (water of each monthly composite j
& ice conditions j
permitting)
Tritium analysis of quarterly composites 1
of monthly composites
's
- a Sample locations are given on the figure and table in the ODCH.
i Y
3.I6/4.16 229-1 J.
Anlendment No.15 e
4 s
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(Page 3 of 5) f t
710tTrlCEl.l.0 NilCl. EAR GENERATING Pl. ANT RADIATION ENVIRONilENTAI t10NITORING PROGRAtt SAMPl.E Col.LECTION AND ANAL.YSIS
[
f I
Number of Samples
}
Exposure Pathway and Sampling and Type and Frequency I
and/or Sample Sample 1.ocations**
Collection Frequency of Analysis s
3.
llaterhourne (con' t.)
8 I
b.
Ground Three samples from wells Quarterly Camma Isotopic and within 5 miles of the tritium ar.alyses of plant site and one sample each sample from a weLL greater than o
10 miles from the plant j
site.
i i
c.
nrinkigg one sample f rom the City of
!!onthly compostte of I-131 Analysis and
!!!nneapolis water supply weekly samples.
Gross beta and Gamma isotopic analysis of each monthly composite 1
Tritium analysis of
- qua rterly compos'ltes of monthly composites
)
i j
d.
Sed imen t from One sample upstream Semlannually Camma isotopic analysis 1
Shoreline of plant, one sample of each sample j
downst ream of plant, j
and one sample from shoreline of recreational j
a rea 1
g
- Sample locations are given on the figure and table in the ODCit.
1 3.16/4.16 L.$m W
Amendment No. 15 1
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s Table 4.16.1 (Page 4 of 5) t b
MONTICELLO NilCLEAR CENERATINC PLANT RADIATION ENVIRONilENTAL HONITORING PROGRAM h
SAMPLE COLLECTION AND ANALYSIS k
j, a
Number of Samples i
Exposure Pathway and Sampling and Type and Frecluency f
i and/or Sample Sample Locations **
Collection Frequency of Analysis d
4.
Ingestion f
a.
Hilk One sample from dairy Monthly or biweekly Camma isotopic and farm having highest D/Q, if animals are on 1-131 analysis of i
one sample from each r>f pasture each sample I
three dairy farms cal-culated to have doses from
.)
1-131 > l mrem /yr, and 1
-t one sample from 10-20
'l miles i
1 b.
Fish and
- One sample of one game Samples collected Camma Isgtopic
(
Inver t ebra tes specie of fish located semi-annually analysia on each
}
upstreana and downstream sample (edible of the plant site.
portion only on fish).
Q One sample of Invertebrates 1
upstream and downstream of the plant site.
1 z
b
- Sample locations are given on the figure anti table in the Ol>CH.
.i 1
6 t
3.16/4.16 229n si
'd Amendment No. 15
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HAXUlUti VALUES FOR Tile LOWER LDi[TS OF DETECTION (1.LD)"'
h i
i n"
Airborne Particulate l
4, 3
of Cag Fish Milk Food Products Sediment l
t l4 Water l
Analysis (pCi/1)
(pC1/m )
g (pCi/kg wet)
(pC1/1)
(pCi/kg, wet)
(pCi/kg, dry) b 1 x 10-2
.)
gross beta 4
s i
3
'2000(1000 )
g 54 15 130 lin ii 59 30 120 Ye i
1
- i 58, 60 15 130 Co
,r 65 30 260
- i
.}
7.n 6
95 15" Zr-tih I
-2 d
4 131 I '*
1 x 10 g
60
[i I
,r I
-2 134,137,
15(10'), 18 1 x 10 130 15 60 150 g
t 140 15 15 i,
Ha-La is
'i.16/4.16 229p f
Amendment No. 15 i!
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I TAllLE 4.16.2
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(Page 2 of 2) i; o
TAllLE NOTATION y
'I a - The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
l c
For a partIcular neasurement system'(which may include radiochemical separation):
F h
4.66 s LLD
=
8 E.V. 2.22. Y. exp(-A At) where LI.!) is the a priori Iower limit of detection as defined above (as picocurie per unit mass or volume),
s is the standard deviation of the background counting rate or of the counting rate of a blank sample ah apg!ropriate (as counts per minute).
Typical values of E, V, Y and at shall be used in the calculati'ns.
}
o q
.. o _
k
,a E is the counting efficiency (as counts per transformation) y li V is the sample size (Jn units of mass or volume) h
'j 2.22 is the number of transformations per minute par picocurie
[
r:
t Y la the fraction radiochemical yield (when applicable)
[
j h is the radioactive decay constant for the particular radionuclide At is the elapsed t t:ne between sample collection (or end of the sample collection period) and t,ime of counting p
U
[
h - I.I.D for drinking water.
4 c - Total *for parent and daughter, h
I d - Applies to specific isotope analysis-not to gamma spectrum analyses.
e - Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.15.2, shati be ident i fied and reported.
li
').16 /4.16 229q s
p Amendment No. 15 6
y mm ;nm_m~.w- - ~wm~
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.?
a a
j.
Table 4.16.3 i,,';
!t I!El'URTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONt! ENTAL SA)P.*LES y
r Reporting 1.cVels orCag Fish tillk
, Vegetables
- f Airborne Particulate Water i
1 Analysis (pCi/1)
(pCL/m )
(pC1/kg, wet)
(pCi/l)
(pCi/kg, wet)
~
11 - 3 2 x 10 ("}
l s
l lin-54 1 x 10 3 x 10 h
Fe-59 4 x 10 1 x 10
'k 3
4 co-58 a x 10 3 x 10 2
4 l1 Co-60 3 x 10 1 x 10 i
l t
n 2
0 2n-65 3 x 10 2 x 10 l
l a
- r-Nis-95 4 x 10 ( }
f
/.
t 2
h I-131 2
0.9 3
1 x 10 3
1 3
Cs-134 30 10 1 x 10 i
60 1 x 10
- g 3
3 M
^
Cs-137 50 20 2 x 10 70 2 x 10 I
3 x 10 (b) b Ba-La-140 2 x 10 (b) 2 2
bb l
1:
p a - For d rinking wa ter samples j[;
l b - Total for parent and daughter t
3.16/4.16 229r b
}. '.
l Amendment No. 15
- .i N
l t
- r i
O' 4
3.16 and 4.16 liASES i
A. Sampic Collec tion & Analysis The Radiation Environmental llonitoring Program required by this specification provides measurements of radiat ion and of I
radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential l
rad ia t ion exposures.of individuals resulting f rom the plant operation. This program thereby supplements the rad io-logical effluent monitoring by verify,ing that the measurable concentrations of radioactive materials and levels of a
radiation are not higher than expected on the basis of the ef fluent measurements and modeling of the environmental exposure pa thways. Af ter a specific program has been in ef fect for at least three years of operation, program
~
changes may be Initiated baned on this experience.
i The detection capabilitien required by Table 4.15.2 are state-of-the-art for routine environmental measurements in i
indust rial labora tories. The I.LI)'s for drinking water meet the requirement of 40CFR 141.
1 j
H.
1.and Use Census t
This specification is provided to ensure that changes in the use of off site areas are identified and that I
mod i f ica t ions to the monitoring program are made if required by the results of this census. The best survey information from door-to-door, aerial or consulting with local agricultur 1 authorities shall be used. This 2
census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.
Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy 0
vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of Icafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was l
}
used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of
.j g
2 kg/ square meter.
I C.
Interlaboratory Comparison Program j
t The requirement for participation in an intertaboratory comparison program is provided to ensure that independent t
checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as a part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonable vaild.
i I
i 229s I
3.16/4.16 I
J Amendment No. 15
[
[
1
i a t i
4 i
f.
Invest 1gation of all events which are required by regulation or technical specifications to be reported to I
a
-{
.lR' in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
g.
Revisions to the Facility Emergency Plan, the Facility Security Plan, and the Fire Protection Program.
t Committee involve unreviewed or l-h.
Operations Committee minutes to determine if matters considered by that l
unresolved safety questions.
- i. Other nuclear safety matters referred to the SAC by the Operations Committee, plant management or company management.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-
[
J.
l' i
related structures, systems, or components.
Reports of special inspections and audits conducted in accordance with specification 6.3.
k.
i 1.
Changes to the Of fsite Dose Calculation Hanual (ODCH).
s Review of investigative reports of unplanned releases of radioactive material to the environs.
m.
i 6.
Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to j
I 4
assure safe facility operation.
i
\\
Audits of selected aspects of plant operation, as delineated in Paragraph 4.4 of ANSI N18.7-1972, shall he a.
i performed with a frequency commensurate with their nuclear safety significance and in a n.anner to assure I
that an audit of all nuclear safety-related activities is completed within a period of two years. The l.
l audits shall be performed in accordance with appropriate written instructions and procedures.
b.
Audits of aspects of plant radioactive effluent treatment, and radiological environmental monitoring shall j
t be performed as follows:
i i
Implementation of the Of fsite Dose Calculation Hanual and quality controls for effluent monitortug at i
1.
J least once every two years.
2.
Implementation of the Process Control Program for solidification of radioactive waste at least once every t
i-two years.
E 3.
The Radiological Environmental !!onitoring Program and the results thereof, including quality controls, at i
I least once every year.
i i
I c.
Periodic review of the audit program should be performed by the SAC at least twice a year to assure its i
adequacy.
d.
liritten reports of the audits shall be reviewed by the Vice President - Power Production, by the SAC at l
}
a scheduled meeting, and by members of management having responsibility in the areas audited.
I 239 l
6.2 I
t P
[
[
ivnendment No. A 15 l
__.7-c n
r r
- e o
f.
All events which are required by regulation or technical specLfications to be reported to NRC in writing I
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
g.
Drills on emergency procedures (including plant evacuation) and adequacy of communication with of f-site support groups.
h.
All procedures required by these Technical Specifications, including impicmenting procedures of the Emergnecy i
Plan and the Security Plan,shaLL he reviewed with a frequency coimeensurate with their safety significance
,]
but at an interval of not more than two years.
i.
Perform special reviews and investigations, as requested by the Safety Audit Committee.
I J.
Review of investigative reports of unplanned releases of radioactive material to the environs.
6 l'
k.
All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (01) Cit).
i 5
Authority t
1 The DC shall he advisory to the Plant flanager.
In the event of disagreement between the recommendat ions'of the OC and the Plant flanager, the course determined by the Plant Manager to be the more conservative will he followed.
.j i
A written summary of the disagreement will he sent to the Cencral Manager Nuclear Plants and the Chairman of jl J
the SAC for review.
l i
s q
fi. Reco rds
('
llinutes shall he recorded for all meetings of the DC and shall identify all documentary materla t reviewed. The l
minutes shall he distributed to each member of the 00, the Chairman and each member of the Safety Audit Committee, l
the Cencra t Manager Nuclear Plants and otliers designated by DC Chairman or Vice Chairman.
7.
Procedure,.
A written charter for the DC shall he prepared that contains:
a.
Responsibility and authority of' the group.
b b.
Content and method of submission of presentations to the Operations Committee.
- i (3. 2 242 Amendment No. A 15 l-i i
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i-6.5 P1 ant opera t ing procedures
}]
{[
l l<
Detailed written procedures, including the applicable check-of f lists and instructions, covering areas listed below shall be prepared and followed. These procedures and changes thereto, except as specified in 6.5.F shall be l
'l reviewed by the Operation Committee and approved by a member of plant management designated by the plant !!anager.
'l l
t l<
A.
plant Operations I
1.
Integrated and system procedures for normal startup, operation and shutdown of the reactor and all systems f!
l[
[
and components involving nuclear safaty of the facility.
i
\\
in t!
2.
Fuel handling operations.
8 l
')
j 3.
Actions to be taken to correct specific and foreseen potential or actual malfunction of systems or components l
including responses to alarms, primary system leaks and abnormal reactivity changes and including follow-up it act ions required af ter plant protective system actions have initiated.
li 4.
Surveillance and testing requirements that could have an ef fect on nuclear safety.
O li 5.
Implementing procedures of the security plan.
[l
,5
't 6.
Implenjenting procedures of the emergency plan, including procedures for coping with emergency conditions
{
IlI Involving potential or actual releases of radioactivity.
.! i 7.
implementing procedures of the fire protection program.
l[
I 8.
Implementing procedures for the Process Control program and Of fsite Dose Calculation llanual including
- [
quality control measures.
Drills on the procedures specified in A.3 above shall be conducted as a part of the retraining program. Drills on the procedures specified in A.6 above shall be conducted at least semi-annually, including a check of communications with offsite support groups.
a i;
11. Radiological f!
Radiation control procedures shall be maintained and made available to all plant personnel. These procedures shall slu)w permissible radiation exposure and shall be consistent with the requirements of 10 CFR 20.
Thlu I
radiation protection program shall be organized to meet the requirements of 10 CFR 20.
1 6.5 244 i!
Amendment No. 15 i!
i t:
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w w.- w w
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_ ~-~-
7
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C.
Maintenance and Test
')
The following maintenance and test procedures will be developed to satisfy routine inspection, preventive maintenance prograins, and operating license requirements:
1 Routine testing of enc ncered Safeguards and equipment as required by the facility license and the Technical j
0 1.
i Specifications.
f 5
l
!)
2.
Routine testing of standby and redundant equipment.
!b'[
3.
Preventive or corrective maintenance of plant equipment and systeas that could have an effect on nuclear l
,l safety.
)
i 4.
Calibration and preventive maintenance of instrumentation that could affect the nuclear safety of the plant.
f l
5.
Special testing of equipment for proposed changes to operational procedures or' proposed system design changes, j
I t
t
)
I I
6
{:.
4 1
I k-e 6.5 246 Amendment No. 15
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D.
I The PCP sluill be approved by the Commission prior to initial implementation. Changes to the PCP shall satisfy i
[
the following recluirements:
'hi 1.
A description of changes shall be submitted to the Commission with the Semi-Annual Radioactive Ef fluent Release Report for the period in which the changes (s) were made. This submittal shall contain:
a.
sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
- 4 b.
a determination that the change did not reduce the overall conformance of the solidified I
waste product to existing criteria for solid wastes; and c.
documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.
2.
Shall become effective upon review and acceptance by the Operations Committee.
f, e
'I i
f, l
l i
s i
5 i
6.5
'246a l
i i.
Amendment No.
15
m-
____,s
's E. Of fsite Dose Calculation Hanual (ODCH) 9' The ODCH shall be approved by the Commission prior to initial implementation. Changes to the ODCH shall satisfy the j
following requirements:
1.
Shall be sghmitted to the Commission with the Semi-Annual Radioactive Effluent release report for the period in which the change (s) were made effective. This submittal shall contain:
l L
a.
sufficiently detailed information to totally support the rationale for the change without l'
henefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCH to be changed with each page numbered and pro-vided with a revision date, together with appropriate analyses or evaluations i
justifying the changes (s).
[
b.' a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and i
i" c.
documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.
2.
Shall become ef fective upon review and acceptance by the Operations Committee.
h I
F.
Temporary Changes to Procedures I
E Temporary changes to procedures described in A R, C, D, and E above, which do not change the intent of the l
original procedures may he made with the concurrence of two individuals holding senior operator licenses, i
Such changes shall be documented, reviewed by the Operations Committee and approved by a member of plant i'
management designated by the Plant Manager within one month.
i l
6.5 246b Amendment No. 15 I
tl
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6.6 plant Operating Records A.
Records Retained f or Five Years
}
l Recor.ls and logs rela tive to the following items aliall be retained for a minimum of five years:
1.
Normal plant operation including such Ltems as power level, periods of operation at each level, fuel exposure and shutilowns.
2.
Written shif t supervisory and reactor logs.
3.
periodic chocks, inspections, tests and calibrations of components and systems, as rel..;ed to these Technica l Specificat ions.
8 4.
Reviews of changes made to procedures or equipment and reviews of tests and experiments.
s i
i 6
1 I
h.h 24hc i
Amendment No.15 i
L b
a --
ya._ - - - - - - - - -
- _ - =
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i i
2.
Occulutional Exposure Report.(1) An annual report of occupational exposure covering the previous calendar year shall be subettted prior to liarch I of each year.
The report should tabulate on an annual basis the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and Job functions, e.g., reac tor opera tions and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste proce'ssing, and refueling. The dose assignment to various duty functions may bc estimates based on pocket dosimeter, TI.D, or film hadge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, at least 80% of the total whole body dose L
received from external sources shall be assigned to specific major work functions.
F f
Y 3.
Monthly Operating Report. A monthly report of operating statistics and shutdown experience covering the previous month shall be submitted by the 15th of the following month to the Of fice of Management Information and Program t
Control, il S Nuclear Regulatory Commission, Washington, DC 20555.
}
4.
Semiannual !!adioactive Effluent Release Report.
Routine radioactive effluent release reports covering the operation of the unit during the previous six months of operation shall be submitted within 60 days af ter January 1st and July 1st of each year.
j i
i The radioactive effluent release reports shall include a summary of the riuantities of radioactive liquid and gaseous ef fl/ents as outlined in Appendix 8 of Regulatory Guide 1.21, Revision 1, June, 1974, with dgta summarized I
on a quarterly basis.
I The report to be submitted 60 days af ter January 1st of each year shall include an assessment of the radiation doses f rom radioactive ef fluents released from the plant during the previous calenda'r year. This same report shall also l
loclude an assessment of the radiation doses from radioactive liquid and gaseous effluents to individuals due to i
their activities inside the site boundary (Figures 3.8.1 and 3.8.2) during the report period. All assumptions used in making these assessmer.ts (i.e., specific activity, exposure time and location) shall be included in these l
reports. The assessment of radiation doses shall he performed in accordance with the Of fsite Dose Calculation Hanual (ODCH) or standard NRC computer codes.
i e
1/ This report supplements the requirements of 10CFR20, Section 20.407.
If 10CFR20, Section 20.407 is revised to i
3 include such information, this Specification is unnecessary.
t f
6.7 249 b
Amendment No. y 15
.. u s
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l The radioac t ive ef fluent release report to be submitted 60 days af ter.ianuary 1 of.cach year shall also include an I
assesnment of radiat ion doses to the likely most exposed member of the general public from reactor releases and other
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nearby uranium fuel cycle sources (including doses f rom primary ef fluent pathways and direct radiation) for the previous 12 consecut ive months to show conformance with 40 CFit 190, Environmental Radiation Protection Standards for Nuclear
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Power Operation.
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The radioactive ef fluent release reports shall include the following information for solid waste shipped offsite during N
the report period.
a, container volume, h.
total curic quantity (specify whether determined by measurement or estimate).
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principal radionuclides (specify whether deteruined by measurement or estimate),
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type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),
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i type of container (e.g., LSA, Type A, Type B, Large Quantity), and e.
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solidtftcation agent (e.g., cement, urea formaldehyde),
i The radioac tive e f fluent release reports shall include unplanned releases from the site of radioactive materials in gaseous and liquid ef fluents on a quarterly basis, changes to the ODC!l, a description of changes to the PCP, a report of when milk or vegetable samples can not be obtained as required by Table 4.16.1, and changes in land use resulting in significant inc reases in calculated doses.
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- 5. Annual Summaries of Meteorological Data An annual summary of noteorological data shall be uubmitted for the previous calendar year in the form of joint f requency distributions of wind speed, wind direction, and atmospheric stahtlity at the request of the Coimnission.
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Reportable occurrences u
J Reportable occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC.
Supplemental reports may be required to fully describe final resoletion of occurrence.
In case of corrected or supplemental reports, a licensee event report shall be completed and reference shalI be nude to the original report date.
Unless explicitly stated, the requirements of this section do not apply to the ftre protection systems and measures contained in Section 3.13/4.13, the radiological ef fluent limitations and measures in Section 3.8/4.8, or, the radiological environmental monitoring program in sections 3.16/4.16.
Reporting requirements have been separately specified in those sections.
1.
Prompt Notification Uith Written Followup The types of events listed below shall le reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> bf I
telephone and confirmed by telegraph, ma11 gram, or facsimile transmission to the appropriate NRC Regional i
Administrator or designate no later than the first working day following the event, with a written followup repcrt within two weeks. The written followup report shall include, as a minimum, a completed copy of a licensee event repor t form.
Information provided on the licensee event report form shall be i
supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
I Failure of the reactor protection system or other systems subject to limiting safety system settings to a.
in,itiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to compipte the required protective function.
I Note:
Instrument drift discovered as a result of testing need not,be reported under this item but uuy t
he reportable under items 6. 7.B. I.e. 6. 7.B. I. f. 6. 7. B.2.a below.
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- h. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition li is less conservative than the least conservative aspect of the limiting condition for operation established is in the technical specifications, f
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Note:
[f specified action is taken when a system is found to be opera ting between the most conservative j
and the least conservdtive aspects of a limiting condition for operation listed in the technical
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need not be reported under this item, but it may be reportable under item 6.7.B.2.h below, b
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- c. Ahnormal degradation discovered in fuel cladding, reactor coolant pressure houndary, or primary containment Flo t e:
I.cakane of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not he reported under this term.
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- d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation, greater than or equal to 1% Ak/k; a calculated reactivity balance indicating a shutdown, margin less conservative than specified in the technical specifications; short-term react ivity increases that correspond to a reactor period of less than 5 seconds or, if sub-critical, an unplanned reactivity insertion of more than 0.5% ak/k or occurrence of any unplanned c ri t ica li t y.
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Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfilli
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ment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.
- f. Permnnel error or procedural inadequacy which prevents or could prevent, by itself, the fullfillment of 4
the functional requirements of systems required to cope with accidents analyzed in the SAR.
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flote:
For items 6.7.II. l.e and 6.7.ll. l.f reduced redundancy that does not result in a loss of system function need not he reported under 'this section but may be reportable under items 6.7.II.2.h. and 6.7.H.2.c. helow.
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e 3 Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety systems, or other protective measures required by technical specifications.
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- h. Errors discovered in the transient or accident analyses or in the method.s used for such analyses as
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deserthed in the safety analysis report or in the hat.es for the technical specLfications that have or t
j could have permitted reactor operation in a manner less conservative than assumed in the analyses.
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- i. Performance of structures, systems, or components that iequires remedial action or corrective measures to prevent ope ra t ion in a manner less conservative than assumed in the accident. analyses in the safety
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analysis report or technical. specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial
-j action or corrective measures to prevent the existence or development of an unsafe condition.
j Hote: This item is intended to provide for reporting of potentially generic problems.
i J. Release of radioactive material in liquids from the site to the unrestricted areas in excess of the concentrations specified in 10 CFR Part 20 Appendix B, T.gle II, Column 2 for radionuclides other than dissolved or entrained noble gases or in excess of 2 x 10 uCi/ml for total dissolved and entrained f
noble gases.
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- k. Release of radioactive material in gases from the site to unrestricted areas at a rate which exceeds the following dose rates:
For noble gases 500 mrem / year to the total body or 3000 mrem / year to the skin For radiotodines 1500 mrem / year to any organ and particulates with half-lives f
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- l. Execeding the Limits for the storage of radioactive materials in outside temporary tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce h
the contents. to within the specified limits.
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Thirty Day Written Reports J
The reportable occurrences discussed lyttow shall be the subject of wri: ten reports to the appropriate NRC Regional I
Administrator or designate within thirty days of occurrence of the event. The written report shall include, I
as a minimum, a completed copy of a licensee event repor t form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation i.
of the circums.tances surrounding the event.'
a.
Reactor protectton system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment l
of the functional requirements of affected systems.
b.
Conditions Icading to operation in a degraded mode tv2rmitted by a limiting condition for operation or gAant shutdown required by a limiting condition for opera tion.
tio t e:
Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 6.7.B.2.a and 6.7.B.2.b need not be reported except i
where test results themselves reveal a degraded mode as described above.
a c.
Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature 3
systems.
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- 1 d.
Abnormal degradation of systems other than those.specified in item 6.7.B. I.c above designed to contain ll radioactive material resulting from the fission process.
I Note: Scaled sources or calthration e.aurces are not included under this item.
Leakage of valve packink or gaskets within the limits for identified leakage set forth in technical specifications need not he reported under this item.
6t An unplanned offsite release of 1) more than one' curie of radioactive material in liquid effluents, 2) e.
3 more than 150 curies of noble gas in gaseous effluants, or 3) more than 0.05 curies of radiotodine in gaseous effluents.
- llie report of an unplanned offsite telease of radioactive material shall include the following infornu tion:
- 1. A description of the event and equipment involved.
- 2. cause(s) for the unplanned release.
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- 3. Ac t ions taken to prevent recurrence.
- 4. Consequences of the unplanned release.
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6.7 252 f
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Environmental Reports r
The reports listed below shall be submitted to the Administrator of the appropriate Regional Of fice or designate:
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_ Annual Radiation Environmental }lonitoring Report i
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- a. Annual Radiation Environmental flonitoring Reports covering the operation of the program during the I
previous calendar year shall be submitted prior to May I of each year.
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- b. The Annual Radiation Environmental Honitoring Reports shall include summaries, o
interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report L
period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an n
assessment of the observed impacts of the plant operation on the environment.
The reports f.
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harmful effects or evidence of irreversible damage are detected by the monitoring, the report
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shall provide an analysis of the problem and a planned course of action to alleviate problem.
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- c. The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated J
results in the fornust of Regulatory Guide 4.8, Decembet 1975 of all radiological environmental j
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samples taken during the report period.
In the event that some results are not available for l
p inclusicn with the report, the report shall be submitted noting and explaining the reasons for
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P the missing. results.
The missing data shall be submitted as soon as possible in a supplementary i;
report.
- d. The reports shalt also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions frons the reactor; and the results of licensee participation in the Interlabora tory Comparison Program, required by Specifica tion 4.16.C. I.
2.
August 1 of the subsequent year.An Annual Environmental Monitoring and Ecological Studies Progra evaluaticm of the results of the non-radiological environmental surveillance activities.The rep j.
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Amendment No. 15 t;
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Special Reports i
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When radioactivlty levels in samples exceed limits specified in Table 4.16.3 a Special Report shall i
be subiettted within 30 days frois the end of the af fected calendar quarter.
For certain cases j
involving inng analysis time, determination of quarterly averages may extend beyond the 30 day period.
In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Special Report as soon as practicable.
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L 6.8 ENVilhnitlENTAL QUAI.IFICATION
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A. By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be s
qualified in accordance with the provisions of Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Cuidelines); or, NUREC-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment". December 1979. Copics of th. sc documents are at tached to Order for flodification of License DPR-22 dated October 24, 1980, p
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I B. By no later than December 1, 1980. complete and auditible records must be availabic and maintained
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at a central location whicle describe the environmental qualification nethod used for all safety-l' nj related electrical equipment in sufficient detail to document the degree of compliance with the
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DOR Cuidelines or HUREC-0588. Thereaf ter, such records should be updated and maintained current ll as equipment is replaced, further tested, or otherwise further qualified.
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