ML19274D922
| ML19274D922 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 02/16/1979 |
| From: | Reed C COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19274D923 | List: |
| References | |
| NUDOCS 7902270235 | |
| Download: ML19274D922 (91) | |
Text
"
Commonwealth Edison One First National Plaza, Chicago, Illinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 February 16, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Zion Station Units 1 and 2 Proposed Change to Facility Operating License Nos. DPR-39 and DPR-48 NRC Docket Nos. 50-295 and 50-304 References (a) :
July 11, 1978 letter from Brian K. Grimes to "All Power Reactor Licensees."
(b) :
November 15, 1978 letter from Brian K.
Grimes to "All Pressurized Water Reactor l
Licensees."
Dear Mr. Denton:
Pursuant to References (a) and (b) and 10 CFR 50.59, Ccmmonwealth Edison company hereby requests a change to Operating License Nos. DPR-39 and DPR-48, Appendix A, Technical Specifica-tions.
The purpose of this mnendment is to revise the Zion Station Technical Specifications to conform to the requirements of 10 CFR 50, Appendix I,
" Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the criterion 'As Low As Practicable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents."
The proposed changes to the Zion Station Technical Specifications are enclosed in. contains an Offsite Dose Calculation Manual (ODCM) for Zion Station as required by Reference (b).
In addition, a discussion of Commonwealth Edison's principal differences from NRC Guidelines for the development of the ODCM is contained in Attachment 3.
The proposed Zion Station Technical Specifications of Attachment 1 follow as closely as practicable the model technical specifications of Reference (a).
However, since thi model technical specifications were not prepared explicitly for Zion Station, some changes were necessary to adapt them to the Zion design and nomenclature.
/
f QO O
f//
Commonwealth Edison NRC Docket Nos. 50-295 50-304 Mr. Harold R. Denton: February 16, 1979 The proposed technical specification changes of and the ODCM of Attachment 2 have been reviewed and approved by the Commonwealth Edison On-Site and Off-Site
- N Review with the conclusion that there are no unreviewed safety questions.
In conjunction with Reference (a), Commonwealth Edison has determined that this proposed amendment is a combined Class III and Class I Amendment per 10 CFR 170.22.
As such, Commonwealth Edison has enclosed a fee remittance in the amount of $4,400.00 for this proposed amendment.
Please address any additional questicas that yon might have to this office.
Three (3) signed originals and thirty-seven (37) copies of this letter are provided for your use.
Very truly yours, O,
on Cordell Reed Assistant Vice-President attachments (3)
SUBSCRIBED and SWO to before me this //ft, day of. l t /, A i:A tr L
, 1979.
A nti1 l}
d t ( %'s Nghary Public O
s
ZION STATION UNITS 1 AND 2 NRC DOCKET NOS. 50-295 AND 50-304 ATTACHMENT 1, Proposed Technical Specification Changes The following pages lave been revised:
i 135 iii through v 222 through 240 vil through x 250 through 253 5
275 through 280a 36 301 through 333 The following pages lave been added:
Sa 242a through 242j 280b through 280d 6a 278a The following page numbers have been deleted:
225a 275a 330 through 333 226a 328a
TABLE OF CONTENTS 1.0 Definitions Page A.
Reportable Occurrence 1
B.
Alteration of Reactor Core 1
C.
Containment Integrity 2
D.
Inadvertant Trip 2
E.
Operating Modes 3
F.
Protection Instrumentation and Logic 3
1.
Protection System 4
2.
Instrument Channel 4
3.
Protective Logic (or Centrol) Channel 4
4.
Actuation Device 4
5.
Actuated Equipment 4
6.
Cliannel Calibration 5
7.
Channel Functional Test 5
8.
Channel check 5
9.
Logic Channel Functional Test 5
10.
Degree of Redundancy 5
l 11.
Source Check Sa G.
Quadrant Power Tilt 6
II.
Rated Thermal Power 6
I.
Reactor Pressure 6
J.
Refueling Outage 6
K.
Operable 6
L.
Operating 6
6 M.
Operating Cycle 6
N.
Surveillance Interval l
O.
Surveillance Interval Notation 6a Safety Limits Limiting Safety System Setting 1.1 Reactor Core 2.1 Protective Instrumentation Bases Setpoints 1.2 Reactor Coolant System Pressure 2.2 Protective Equipment Setpoints Bases i
Table of Contents (Continued)
- s SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENT PAGE 3.7 Steam Generator Emergency Heat Removal 4.7 156 3.7.1 Steam Line Safety and Relief Valves 4.7.1 156 3.7.2 Auxiliary Feedwater Pumps 4.7.2 157 Bases 3.8 Emergency Core Cooling and Core Cooling Support 4.8 164 3.8.1 Centrifugal Charging Pump System 4.8.1 164 3.8.2 Safety Injection Pump System 4.8.2 168 3.8.3 Residual Heat Removal Pump System 4.8.3 170 3.8.5 Accumulators 4.8.4 174 3.8.6 Component Cooling System 4.8.5 175 3.8.7 Service Water System 4.8.6 178 3.8.8 Hydrogen Control System 4.8.7 180 3.8.9 Equipment for Evaluating Post LOCA 4.8.9 184 Bases 3.9 Containment Isolation Systems 4.9 197 3.9.1 Isolation valves seal Water System 4.9.1 197 3.9.2 Penetration Pressurization Systems 4.9.2 198 3.9.3 containment Isolation Valves 4.9.3 199 3.9.4 Main Steam Isolation Valves 4.9.4 200 3.9.5 Containment Integrity 4.9.5 201 Bases 3.10 Containment Structural Integrity 4.10 212 3.10.1 Containment Integrated Leak Rate fest 4.10.1 212 3.10.2 Containment Tendons 4.10.2 215 3.10.3 End Anchorages and Adjacent Concrete Surfaces 4.10.3 217 3.10.4 Containment Liner 4.10.4 218 3.10.5 Containment Temperature and Pressure 4.10.5 219 Bases iii
Table of Contents (Continued)
'N SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENT PAGE 3.11 Liquid Effluent Releases 4.11 222 3.11.1 Liquid Effluent Release Limitations 4.11.1 223 3.11.2 Liquid Effluent Instrumentation 4.11.2 225 3.11.3 Liquid Radwaste Treatment System 4.11.3 227 3.11.4 Outside Liquid Tanks 4.11.4 228 3.11.5 General Specification 4.11.5 229 Bases 3.12 Gaseous Effluent Releases 4.12 237 3.12.1 Gaseous Ef fluent Release Limitations 4.12.1 238 3.12.2 Gaseous Effluent Instrumentation 4.12.2 242 3.12.3 Gas Decay Tanks 4.12.3 242a 3.12.4 General Specification A.12.4 242b Bases 3.13 Refueling Operations 4.13 243 3.13.1 Core Reactivity 4.13.1 243 3.13.2 Fuel Drop Protection 4.13.2 244 3.13.3 Containment Status 4.13.3 245 3.13.4 Radiation Monitoring 4.13.4 246 Bases 3.13.5 Refueling Eauipment Checkout 246 3.13.6 Refueling Equipment Operability 246 3.13.7 Spent Fuel Pit Cooling Systems 246 3.13.8 Fuel Inspection Program 247 3.14 Plant Radiation Monitoring 4.14 250 3.15 Auxiliary Electrical Power System 4.15 255 3.16 Environmental Radiological Monitoring Program 4.16 275 Bases iv
N
'N Table of Contents (Continued)
~x l
SURVEILLANCE LIMITING CONDITION FOR OPERATION REQUIREMENT PAGE 3.17.1 Ventilation 4.17.1 281 Bases 3.17.2 Aircraft Fire Detection 4.17.2 283 3.18 Steam Generator 4.18 289 Bases 3.19 Failed Fuel Monitoring 4.19 292 3.21 Fire Protection 4.21 295D 5.0 Design Features 296 5.1 Site 296 5.2 Reactor Coolant System 296 5.3 Reactor Core 296 5.4 containment System 296 5.5 Fuel Storage 298 5.6 Seismic Design 299 6.0 Administrative Controls 300 6.1 Organization, Review, Investigation and Audit 300 6.2 Plant Operating Procedures 307 6.3 Actions to be Taken in the Event of a Repo.~ table Occurrence in Plant Operation 315 6.4 Action to be Taken in the Event A Safety Limit is Exceeded 315 6.5 Plant Operating Records 315 6.6 Plant Reporting Requirements 317 l
l 6.7 Offsite Dose Calculation Manual 323 V
LIST OF FIGURES (CONTINUED)
Figure Page 3.3.2-1 Reactor Coolant System Heatup Limitations 82 3.3.2-2 Reactor Coolant System Cooldown Limitations 83 3.3.2-3 Effect of Fluence and Copper on Shift 84 for Reactor Vessel Steels of RTNDT Exposed to 550 F Temperature 3.3.2-4 Fluence at 1/4T and 3/4T as a Function of Full Power Service Years 85 3.3.2-5 Reactor Coolant System Hydrostatic 86 Testieg Limitations 3.3.2-6 Radiation Induced Increase in 87 Transition Temperature for a 302-B Steel 230 3.11-1 Restricted Area Boundary 3.16-1 Locations of Fixed Environmental Radiological 278a Monitoring Stations 6.1.1.
Commonwealth Edison Corporate Organization 325 326 6.1.2 Station Organization Chart 327 6.1.3 Zion Shift Manning Chart vii
LIST OF TABLES Table Page 3.1-1 Reactor Protection System-Limiting Operating Conditions and Setpoints 30 3.1-2 Reactor Protection System Instrument Numbers 33 3.3.2-1 RT e
ng sults 88 NDT 3.3.4-1 In Service Inspection Program 106 3.3.5-1 Reactor Coolant System Chemistry 122 3.4-1 Engineered Safeguards Actuation Syst:rm-Limiting Conditions of Operation 129 and Setpoints 3.4-2 Engineered Safeguards System Instrument Numbers 132 3.7-1 Neutron Flux High Trip Points with Steam Generator Safety Valves 160a Inoperable - Four Loop Operation 3.7-2 Neutron Flux Ifigh Trip Points with Steam Generator Safety Valves 160b Inoperable - Three. Loop Operation 3.11.2 Radioactive Liquid Effluent Monitoring Instrumentation 233 3.12.2 Radioactive Gaseous Effluent Monitoring Instrumentation 242e 3.14.1 Radiation Monitoring Surveillance 252 3.15-1 Equipment Requirement with Inoperative 4KV E.S.S. Bus 268 3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus 269 3.16-1 Zion Standard Radiological Monitoring Program 279 l
viii
LIST OF TABLES (CONTINUED)
Table Page 4.1-1 Reactor Protection System Testing and Calibration Requirements 35 4.4-1 Engineered Safeguards System Testing and Calibration Requirements 134 4.4-2 Engineered Safety Equipment Actuation Test.
136 4.5-1 Containment Fan Cooler Components 148 4.6-1 Containment Spray System Components 153 4.7-1 Steam Generator Safety Valves, Set Pressures, Orifice Sizes and 166 Steam Flows 4.7-2 Auxiliary Feedwater Pumps 161 4.8-1 Centrifugal Charging Pump System 105 4.8-2 Safety Injection Pump System 186 4.8-3 Residual Heat Removal Pump System 187 4.8-4 Accumulator Tanks 188 4.8-5 Component Cooling Pump System 189 4.8-6 Service Water Pump System 190 4.8-7 Ilydrogen Control System 192 4.9-1 Isolation Seal Water System 203 4.9-2 Penetration Pressurization System 204 4.9-3 containment Isolation valves 205 4.9-4 Main Steam Isolation Valves 208 ix
LIST OF TABLES (CONTINUED)
Table Page 4.11.1 Radioactive Liquid Effluent Sampling and Analysis Program 231 4.11.2 Radioactive Liquid Effluent Monitoring Surveillance 233 4.12.1 Radioactive Gaseous Waste Sampling and Analysis Program 242c 4.12.2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance 242e 4.14.1 Radiation Monitoring Surveillance 252 4.15-1 4160 Volt Engineer Safeguard Bus Feeds 270 l
4.16-1 Practical Lower Limits of Detection 280a 4.17-1 Charcoal Filters 284 4,17-2 IIEPA Filters 285 4.19-1 Failed Fuel Monitoring Instruments 295 6.3.1 Boundary Doors for Flood Conditions 332 l
6.6.2 Special Reports 322 X
S.
Actuated Equipment a.
Analog channels - the injection of a simulated signal into the channel as A component or assembly of components close to the sensor as practicable to that performs or directly contributes verify OPERABILITY including alarm and/
to the performance of a protective or trip setpoints.
function such ac reactor trip, containment isolation, or emergency b.
Bistable channels - the injection of a coolant injection.
The following simulated signal into the sensor to are examples of actuated equipment:
verify OPERABILITY including alarm and/or an entire control rod and its trip setpoints.
release mechanism, a containment isolation valve and its operator, 8.
Channel Check ~
and a safety injection pump and its prime mover.
A CHANNEL CHECK shall be the qualitative assessment of channel behavior during 6.
Channel Calibration operation by observation.
This determination shall. include, where possible, comparison A CHANNEL CALIBRATION shall be the of the channel indication and/or status with adjustment, as necessary, of the other indications and/or status derived from ctennel output such that it responds independent instrumentation channels measuring with the necessary range and the same parameter.
accuracy to known valves of the parameter which the channel 9.
Logic Channel Functional Test monitors.
The CHANNEL CALIBRATION shall encompass the entire channel The application of input signals, or the including the sensor and alarm operation of relays or switch contacts, in all and/or trip functions, and shall the combinations required to produce the include the CHANNEL FUNCTIONAL required decision outputs including the TEST.
THE CH ANNEL CALIBRATION operation of all actuation devices.
Where may be performed by any series of practicable, the test shall include the sequential, overlapping or total operation of the actuated equipment as well, channel steps such that the entire ie, pumps will be started, valves operated, channel is calibrated.
etc.
7.
Channel Functional Test 10.
Degree of Redundancy A.
CHANNEL FUNCTIONAL TEST shall be:
With reference to redundant instrument or logic channels, the degree of redundancy is the difference between the number of operable channels and the minimum number of these channels which, when tripped, will cause an automatic system trip.
5
~
11.
Source check A SOURCE CHECK shall be the qualitative assess-ment of channel response when the channel sensor is exposed to a radioactive source.
~
Sa
0.
Surveillance Interval Notation NOTATION FREQUENCY S
At least once per shift D
At least once per day W
At least once per week.
M At least once per month.
O At least ~ once per quarter.
R At least once per refueling shutdown.
S/U Prior to each reactor startup.
P Completed prior to start of release.
N.A.
Not Applicable EFPM At least once per effective full power month.
6a
Reactor Trip Channel Channel Channel channel Description check calibration Function Test Remarks
- 17. Low Steam Generator Level S
R M
in coincidence with Feed Flow Steam Flow Mismatch
- 18. Low-Low Steam Generator S
R M
Level
- 19. Safety Injection N.A.
N.A M
(1) Manual SI function check at R only
- 20. Turbine Trip N.A.
N.A.
M II)
(1) Including Reactor Trip
- 21. Automatic Reactor Trip N.A.
N.A.
M Logic Breaker Opening PERMISSIVES
- 22. P-6 N.A.
R M
- 23. P-7 N.A.
R M
- 24. P-8 N.A.
R M
- 25. P-10 N.A.
R M
NOTE:
Specified intervals may be adjusted 125% to accommodate test schedules.
R - Once Per Refueling Shutdown - calibration of these instruments may be done as much as six months prior to the start of refueling outage and still satisfy this requirement.
- Applies to Unit 1 and Unit II TABLE 4.1-1 (Sheet 2 of 2) 36 Reactor Protection System Testing and Calibration Requirements
- Actuation Channel Channel Channel Channel Description Check Calibration Function Test IV. STEAMLINE ISOLATION
- 1. Manual Actuation N.A.
N.A.
R
- 2. Automatic Actuation N.A.
N.A.
M
- 3. High-High Containment Pressure See Item II Above
- 4. High Steam Line Flow in coincidence with Low-Low See Item I Above Tavg or Low Steam Pressure PERMISSIVES
- 1. P-ll N.A.
N.A.
M 2.
P-12 N.A.
N.A.
M NOTE:
Specified intervals may be adjusted 125% to accommodate test schedules R - Once per refueling shutdown - calibration of these instruments may be done as much as six months prior to the start of refueling outage and still satisfy this requirement.
Engineered Safeguards Systems Testing and Calibration Requirements Table 4.4-1 (Sheet 2 of 2) 135
LII'ITII;G CONDITION FOR OPERATION SURVEILLAI!CE REQUIREMENT 3.11 LIQUID EFFLUENT RELEASES 4.11 LIQUID EFFLUENT RELEASES Applicability Applicability Applies to liquid effluent releases from Applies to liquid effluent releases from the site to the unrestricted area the site to the unrestricted area.
(see Figure 3.11-1).
Objective Objective To establish a method to control liquid To establish surveillance frequencies for effluent releases from the site.
equipment in the liquid effluent systems.
e 222
LIMITING CONDITI ~ N FOR OPERATION SURVEILLANCE REQUIREMENT 3.11 4,11 Specification Specification 1.
Liquid Effluent Release Limitations 1.
Liquid Effluent Release Limitations A.
The concentration of radioactive A.
The concentration of radioactive material released at any time from material in liquid effluents the site to unrestricted areas (see released from the site shall be Figure 3.11-1) shall be limited to continuously monitored in accordance the concentrations specified in with Table 4.11.1.
The co ncentra-10 CFR Part 20, Appendix B, Table tions at the point of release are II, Column 2 for radionuclides limited to the values specified in other than dissolved or entrained 3.ll.l.A.
For routine reporting noble gases.
For dissolved or requirements, see Sec+. ion 6.6.3a entrained noble gases, the concen-Semiannual Radioactive Effluent tration shall be limited to those Release Report.
concentrations specified in the Of fsite Dose Calculation Manual B.
Not Applicable
( O DCM ).
B.
If the concentration of radioactive material released from the site to unrestricted areas exceeds the limits specified in 3.11.1.A, im-mediately decrease the release rate of radioac tive materi'.ls and/or increase the dilutiou flow rate to restore the concentration to within the above limits and provide prompt notification to the Commission pursuant to Specification 6.6.2a.
223
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11.1 4.11.1 C.
The dose or dose commitment to an C.
Dose Calculations Cumulative dose individual above background from contributions from liquid effluents radioactive materials in liquid shall be determined in accordance with effluents released from the site to the Offsite Dose Calculation Manual unrestricted areas (see Figure 3.11-1)
(ODCM) at least once per 31 days.
For shall be limited:
routine reporting requirements see section 6.6.3.b Environmental Radiological 1.
During any calendar quarter to 5 3 Monitoring.
~
mrem to the total body and to 6 10 mrem to any organ and a
2.
During any calendar year to 6 6 mrem t o the total body and to 6 20 mrem to any organ.
D.
Not Applicable D l.
With the calculated dose from the release of radioactive materials in liquid effluents exceeding the specified limits in 3.ll.l.C prepare and submit to the Commission within 30 days, pursuant to Specification 6.6.3 c.h. a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose or dose commitment to an individual from such releases during these four calendar quarters is within 6 mrem to the total body and 20 mrem to any organ.
224
o LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
- 3. ll. D 4.11.1 2.
With~ the calculated dose from the release of radioactive materials in liquid effluents exceeding twice the limits specified in 3.ll.l.C prepare and submit a Special Report to the Commission pursuant to Specification 6.6.3.c.h.
and limit the subsequent releases such that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to 6 25 mrem to the total body or any organ (except' thyroid, which is limited to 6 75 mreia) over 12 consecutive months.
This Special Report shall include an analysis which demonstrates that radiation exposnres to all real individuals from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are less than the 40 CFR Part 190
- Standard.
3.11.2 Liquid Ef fluent Instrumentation 4.11.2 Liquid Effluent Instrumentation A.
The radioactive liquid effluent monitor-A.
ing instrumentation channels shown in Table 3.11:.2 shall be OPERABLE with their 1.
The setpoints shall be determined alarm / trip setpoints set to ensure that in accordance with procedures as the limits of 3.ll.l.A are met.
described in the ODCM.
- Revision date of July 1, 1977 i
i 225
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.11.2 3.11.2 2.
Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of,the OIANNEL QIECK, SOURCE GIECK, GIANNEL CALIBRATION and GIANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.11.2.
3.
Records - Auditable racords shall be.
~
maintained, in accorocace with procedures in the ODCM, of all radioactive liquid effluent monitoring instrumentation alarm / trip setpoints.
Setpoints and setpoint calculations shall be available for review to ensure that the limits of Specification 3.11.1 are met.
B.
With a radioactive liquid effluent B.
Not Applicable monitoring instrumentation channel trip setpoint less conservative than the val e necessary to prevent u
violating 3.11.1.A, immediately suspend the release of radioactive liquid effluents monitored by the offected channel or declare the C.
Not Applicable channel inoperable.
C.
With one or more radioactiva liquid effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 3.11.2.
Provided the action is taken, this will not be considered a degraded mode for report-purposes.
(6.6.2.A.2) 3 226
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11.3 Liquid Radwaste Treatment System 4.11.3 Liquid Radwaste Treatment System A.
The liquid radwaste treatment system shal l A.
be OPERABLE, except as specified in
~
3.ll.3.C.
The systems shall be used to reduce che radioactive materials in liquid wastes prior to their discharge 1.
Doses due to liquid releases from the when the projected dose due to liquid site to unrestricted areas shall be effluent releases from the site to un-projected at least o,nce per month,
~
restricted areas (see Figure 3.11-1) when averaged over 31 days would exceed 2.
The appropriate liquid radwaste system 0.5 mrem to the total body or 1.67 mrem shall be demonsprated OPERABLE at least once per quarter, unless the appropriate to any organ.
liquid radwaste subsystem has been B.
With radicactive liquid waste bcing utilized to process radioactive liquid discharged without treatment and in exces ef fluents during the previous quarter..
s of the above limits, prepare and submit to the Commission within 30 days, B.
Not Applicable pursuant to Specification 6.6.3.c.h.
a Special Report which includes the following information:
1.
Identification;of equipment or sub-systems not OPERABLE and the reason for Inoperability.'
2.
Action (s) taken to restore.'the In-operable equipment to OPERABLE status.
3.
Summary description of action (s) taken to prevent a recurrence.
C.
Equipment in the liquid Radwaste Treatmeni.
C.
Not Applicable System may be removed from service for maintenance provided the dose limits specified in 3.11.3.A are not violated.
227
o.
LIMITING' COIIDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11.4 Outside Liquid Tanks 4.11.4 Outside Liquid Tanks A.
The quantity of radioactive material A.
The quantity of radioactive material' excluding tritium and noble gases, excluding tritium and noble gases contained in each of the specified contained in each of the listed tanks shall be determined to be within the tanks shall be limited to provide
~
limit by analyzing a representative assurance that in the event of an un-controlled release of the tanks' sample-of the tank's contents at least contents that the resulting average an-weekly when radioactive materials are nual concentrations would be less than being added to the tank.
These con-the limits of 10 CFR Part 20, Appendix centrations will be determined as B, Table II, Column 2 at the nearest prescribed in the ODCM.
surface water supply in an unrestricted PrharyWaterStorageTank (2)
B.
Not Applicable Secondary Water Storage Tank,(2)
B.
With the quantity of radioactive material, excluding tritium and noble gaces in any of the above listed tanks exceeding the above limits, immediately suspend all additions of radioactive material to the tank and witnin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> either reduce the tank contents to within thelimit or provide prompt notificstion to the Commission pursuant to Specification 6.6.2.a.
The written followup report shall include a description of activities planned and/or taken to reduce the tank contents to within the above limit.
O 228
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11.5 General' Specification l4.11.5 Not Applicable In the event a limiting condition for operation and/or associated action require-ments cannot be satisfied because of circumstances in excess of those addressed in their specifications, no changes are required to the operational conditions of the plant nor does this prevent changes -
'to the operational mode.
e e
4 229
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~
J s
Property Line l
N k
I q
kssel Exclusion Area Boundary D
NLNN
's s
s N
s b\\ D f'
Restricted Area
/
s N
/
I j
N N
/
I N
}g g -
t i
Liquid Effluent Discharge Points t
i
/
N I
\\\\\\\\\\
N
\\
I
'N N
N
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N N
N s
N
\\
l 3
N l
s N
s N NN
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's '_-
x y
g R
ywxwwxNNwNNwN't' S
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5 N
s N
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M M'
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m 3
Figure 3.11 Restricted Area Boundary 230
Table 4.11.1 Radioactive Liquid Effluent Sampling and Analysis Program (a)
Release Type Sampling Frequency Type of Activity Analysis Prior to 1.
Lake Discharge Tank each release 1.
Each release shall be analyzed for gamma emitters.
2.
A sample proportional to the volume discharged (b) shall be saved along with proportional samples from all releases via this pathway, and shall be analyzed for tritium once/ month.
3.
A sample proportional to the volume discharged (b) shall be saved along with proportional samples from all releases via this pathway, and shall be analyzed for Sr-89, 90 once/ quarter.
2.
Turbine Building Continuous during Fire Sump releases 1.
A continuous composite sample shall be analyzed for gamma emitters once/ week.
2.
A sample portional to the calculated volume discharged (b) shall be saved along with proportional samples from all releases via this pathway, and shall be analyzed for tritium once/
month.
3.
A sample proportional to the calculated volume discharged (b) shall be saved along with proportional samples from all releases via this pathway, and shall be analyzed for Sr-89, 90 once/ quarter.
231
TABLE 4.11.1 (Continued) a.
The lower limit of detection ( LLD) should be 5 10% of the concentration specified in 10CFR Part 20, Appendix B, Table II, Column 2.
The lower limit of detection (LLD) is defined in Table Notation a.
of Table 4.16-1.
b.
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in.
which the method of sampling employed results in a specimen which is representative of the liquids released.
232
Table 3.11.2 Table 4.11.2 Radioactive Liquid Effluent Radioactive Liquid Ef fluent Monitoring Instrumentation Monitoring Surveillance Minimum Channel Channels Charinel Source Channel Functional Instrumen,t Applicability Action Operable Check Check Calibration Test 1.
Gross Activity Monitors Providing Automatic Isolation A.
Lake Discharge Tunk 1)
ORT - PRO 4 18 See Action 18 P
P R
Q (1) 2)
ORT - PROS 18 See Action 18 P
P R
O (1)
B.
Turbine Bldg 1)
ORT - PR 25 20 1
D*
M R
Q (1) 2.
Continuous Composite Sampler A.
Turbine Bldg. Fire 20 1
N.A.
N.A.
N.A.
N.A.
Sump 3.
Flow Rate Monitors 21 1
N.A.
N.A.
R N.A.
B.
Lake Discharge Tank OP-WD 63 OF-UD 67
- During releases via this pathway 233
Table 3.11.2 (Continued)
Table 4.11.2 (Continued)
TABLE NOTATION TABLE NOTATION ACTION 18 - If one of the LDT moniturs is inoperable, (1) The Channel Functional test shall all LDT releases shall be made through also demonstrate the automatic the operable monitored pathway.
If both isolation of this pathway and monitors are inoperable, effluent re-control room alarm annunciation leases from the tank may continue for up occurs if the instrument indicates to 14 days provided that prior to measured levels above the alarm /
initiating the release:
trip setpoint.
- 1. At least two independent samples of the tank's contents are analyzed,and
- 2. At least two technically qualified numbers of the Facility Staff in-dependently verify the release rate calculations and discharge line valving; otherwise, suspend release of radioactive effluents via this pathway.
ACTION 20 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days pro-vided that at least once per nhift grab samples are analyzed fc.-
q,oss radioactivity (beta or gamma) at a limit of detection of at least 10-7 uCi/ml.
ACTION 21 - With the number of channels operable less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided the 234 234 flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Basis:
Specification 3.11.1A is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II Column 2.
This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures ex-ceeding (1) the Section II. A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population.
The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water.
Specification 3.11.1 C is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable".
The dose calculations in the ODCM implement the requirements in Section II.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 appropriate pathways is unlikely to be substantially underestimated.
This specification applies to the release of liquid effluents from each reactor at the site.
For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system.
235
The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive caterials in liquid effluents during actual or potential releases.
The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatuent prior to release to the environment.
The requirement that the appropriate portions of this system be used when specified provides assur-ance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable".
Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting average annual concentrations would bc less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.
These concentrations will be determined as prescribed in the ODCM.
s
.36
'\\
o LI?i1 TING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l3.12.
Gaseous Effluent Releases 4.12 Gaseous Effluent Releases Applicability Applicability Applies to gaseous effluent releases Applies to gaseous effluent releases from the site to the unrestricted from the site to the unrestricted area.
(See Figure 3.11-1) area.
Objective objective To establish a method to control To establish surveillance frequencies gaseous effluent releases from the for equipment in the gaseous effluent site.
systems.
Specification Specification e
237
LIl41 TING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT
_ _ _ =
3.12.
4.12.
- 1. Gaseous Effluent Release Limitations
- 1. Gaseous Effluent Release Limitations A. The instantaneous release rate of radio-A.
active materials released in gaseous effluents from the site shall be.such that in the unrestricted areas the:
- 1. The dose rate for noble gases shall be
- 1. The release rate, at any time, of noble
$500 mrem /yr to the total body and gases in gaseous effluents shall be
- 3000 mrem /yr to the skin, and controlled by the offsite dose rate.s a
established in Specification 3.12.1.
- 2. The dose rate for all radiciodines and 2.~The release rate of radioactive materi-for all radioactive materials in par-als, other than noble gases, in gaseous ticulate form and radionuclides other effluents shall be determined to be than noble gases shall be 4.1500 within the limits by obtaining repre-mrem /yr, to any organ.
sentative samples and performing anal-
~~
yses in accordance with the sampling and analysi's program, specified in Table 4.12.1.
B.
If the release rates exceed the limits of 3.12.1.A. then immediately decrease the B.
Por routine reporting requirements, see release rates to within the limit and Section 6.6.3.a. of the Semi-annual provide notification to the Commission Radioactive Effluent Release Report.
pursuant to Specification 6.6.2a.
C. The air dose in unrestricted areas due to C. Cumulative dose contributions for the total noble gases released in gaseous effluents time period shall be determined in accord-shall be limited to the following:
ance with the Offsite Dose Calculation a.
During any calendar quarter, to Manual (ODCM) at least once every 31 days.
- 10 mrad for gamma radiation and For routin^ reporting requirements, see 620 mrad for beta radiation; Section 6.6.3.a. of the Semi-annual
- b. During any calendar year, to 620 mrad Radioactive Effluent Release Report, for gamma radiation and 640 mrad for beta radiation.
238
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l 3.12.1.D.
4.12.1.D.
Not Applicable
- 1. With the' calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the limits of Spec. 3.12.1.C. prepare and submit to the Commission within 30 days, pursuant to Specification 6.6.3.c.h.,
a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to re-duce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quar.
ter and during the subsequent three calendar quarters so that the average dose during these four calendar quarters is within 20 mrnd for gamma radiation and 40 mrad for beta radiation.
- 2. With th'e calculated air dose from radio-active noble gases in gaseous effluents exceeding twice the limits of Specifi-cation 3.12.1.C., prepare and submit a Special Report to the Commission pur-suant to Specifications 6.6.3.c.h. and limit the subsequent releases such that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to 425 mrem to the total body or any organ (except thyroid, which is limited to #75 mrem) over 12 consecutive months.
This Special Re-port shall include an analysis which demonstrates that radiation exposures to all real individuals from all uranium fuel cycle sources (including all efflu-ent pathways and direct radiation) are less than the 40 CFR Part 190* Standard.
- Revision date of July 1, 1977 239
LIMITIliG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12.1 4.12.1 E. The dose to an individual in unrestricted E. Cumulative dose contributions for the total areas from radiciodines, radioactive mater-time period shall be determined in accordance ials in particulate form and radionuclides with the ODCM at'least once every 31 days.
other than noble gases released in gaseous For routine reporting requirements, see effluents from the site shall be limited to section 6.6.3.a. of the Semi-annual Radio-the following:
active Effluent Release Report.
- 1. During any calendar quarter to 615 mrem to any organ.
- 2. During any calendar year to 530 mrem to any organ.
~
240
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l4.12.1.F.
Not Applicable 3.12.1.F.
- 1. With the calculated dose from the release of radiciodines, radioactive materials in par-ticulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the limits of Spec.
3.12.1.E., prepare and submit to the Commission within 30 days pursuant to Spec.
6.6.3.c.h.
a Special Report which identifies the cause(s) for exceeding thc
' limit and defines the corrective actions to be taken to reduce the releases of radio-iodines, radioactive materials in particulate form, and radionuclides other than noble gases in gaseous effluents during the re-mainder of the current calendar quarter and during the subsequent three calendar quarters so that the average dose or dose commitment to an individual from such releases during these four calendar quarters is within 30 mrem to any organ.
- 2. With the calculated dose from the release of radiciodines, radioactive materials in parti-culate form, or radionuclides other than noble gases in gaseous effluents exceeding twice the limits of Spec. 3.12.1.E. prepare and submit a Special Report to the Commission pursuant to Spec.
6.6.3.c.h. and limit the subsequent releases such that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to
$25 mrem to the total body or any organ (except thyroid, which is limited to '75 mrem) over 12 consecutive months.
This Special Report shall include an analysis which demonstrates that radiation exposures to all real indivi-duals from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are less than the 40 CPR Part 190*
Standard.
y
- Revision date of July 1, 1977 241
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12.2 Gaseous Effluent Instrumentation 4.12.2 Gaseous Effluent Instrumentation A. The radioactive gaseous effluent monitoring A.
Instrumentation channels shown in Table 3.12.2
- 1) The setpoints shall be determined in accord-shall be operable with their alarm / trip set-ance with procedures as described in the points set to ensure that the limits of 3.12.1 ODCM.
2).Each radioactive gaseous effluent monitoring are met.
instrumentation channel shall be demonstrated OPERABLE by performance o.f the CIIANNEL CIIECK, SbORCE CIIECK, CIIANNEL CALIBRATION and CIIANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.12.2.
- 3) Auditable records shall be maintained of the calculations made, in accordance with proce-dures in the OCDM, of all radioactive process and effluent monitoring instrumentation alarm / trip setpoints.
Setpoints and netpoint calculations shall be available for review to ensure that the limits of Specification 3.12.1.'are' met.
B. With a radioactive gaseous effluent monitoring B.
Not Applicable instrumentation channel alarm / trip setpoint less conservative than the value necessary to prevent violating 3.12.1.A., declare the channel inoperable.
C. With one or more radioactive gaseous effluent C.
Not Applicable monitoring instrumentation channels inoperable take the ACTION shown in Table 3.12.2.
Pro-vided the action is taken, this will not be considered a degraded mode for reporting purposes (6.6.2.a.2).
242
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT 3.12.3 Gas Decay Tanks 4.12.3 Gas Decay Tanks The quantity of radioactiviby contained in A. The quantity of radioactive material contained A.
each gas decay tank shall be limited to 22000 in each gas decay tank shall be datermined to curies noble gases (considered as Xe-133).
be within the above limit at least weekly when radioactive materials are being added to the tank.
B. With the quantity of radioactive material in B.
Not' Applicable any gas decay tank exceeding the above limit, immediately suspend all additions of radio-active material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> either reduce the tank contents to with-in the limit or provide prompt notifination to the Commission pursuant to Specification 6.6.2.A.
The; written followup report shall include a description of activities planned and/or taken to reduce the tank contents to within the above limit.
S 242a
LIMITII;G CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12.4.
General Specification In the event a limiting condition for operation 4.12.4 Not Applicable and/or associated action requirements cannot be satisfied because of circumstances in excess of those addressed in the specifications, no changes are required to the operational con-ditions of the plant nor does this prevent changes to the operational mode.
e 242b
TABLE 4.12.1 Radioactive Gaseous Waste Sampling and Analysis Program Release Type Sampling Frequency Type of Activity A,nalysis 1.
Gas Decay Tank A.
Prior to each release 1.
A.
Each release shall be analyzed for gaseous gamma emitters.
B.
During,each release B.
Each release shall be analyzed for parti-culate gamma emitters, tritium and I-131 and I-133.
C. The particulate filter from the release shall be saved along with all particulate filters from this point for a quarterly composite analysis for Sr-89,90.
2.
Containment Vent or Prior to each release (b)
- 2. A.
Each release shall be analyzed for gas-Purge eous and particulate gamma emitters, tritium and I-131 and I-133.
B. The particulate filter from the release shall be saved along with all particulate filters from this release point, for a quarterly composite analysis for Sr-89,90.
3.
Continuous Release Continuous (c) 3.
A.
Each release point (Section 3.A through Points 3.C) shall be analyzed for gaseous gamma A. Air Ejector for emitters and tritium once/ month.
both (2) Units B.
Auxiliary Building B.
Each release point (Section 3.A through Ventilation for 3.D) shall be analyzed for I-131 and I-133 both (2) Units once/ week.
C. Miscellaneous Vent-ilation Stack C.
Each release point (Section 3.A through D.
Service Building 3.D) shall be analyzed for particulate Ventilation gamma emitters once/ week.
D.
The particulate filter from each release point (Section 3.A through 3.D) shall be saved along with all particulate filters from the respective release point, for a quarterly composite analysis for Sr-89,90.
242c
Table 4.12.1 (Continued)
Table Notation Radioactive Gaseous Waste Sampling and Analysis Program a.
The lower limit of detection (LLD) should be 610% of the concentration specified in 10 CPR Part 20, Appendix B, Table II Column 2.
The LLD is defined in Table Notation (a) of Table 4.16-1 of Specification 4.16.
b.
Should a shutdown, startup or power change grehter than 50% which could alter the mixture of radionuclides occur af ter sampling, another analysis shall be performed prior to release.
The ratio of the sample flow rate to the sampled stream flow rate shall c.
be known for the time period in Specification 4.12.
242d
Table 3.12.2 Table 4.12.2 Radioactive Gaseous Effluent Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation Surveillance Minimum Channel Channels Channel Source Channel Functional Instrument Applicability Action Operable Check Check Calibration Test 1.
Gas Decay Tank 1.
Gas Decay Tank A. Gas activity monitor A.
- 1. ORTPR10A-low range 25 1
- 1. P P
R Q (1)
- 2. ORTPR10B-high range 25 1
- 2. P P
R Q (1)
B.
Iodine Sampler 31 1
B.
NA NA NA NA C.
Particulate Sampler I*
31 1
C.
NA NA NA NA 2.
Air Ejector 2.
Air Ejector A.
Gas activity monitor A.
1.
1RE0015 27 1
- 1. D M
R Q (2) 2.
2 RE0015 27 1
2.
D M
R Q (2) 31 1
B.
NA NA NA NA B.
Iodine Sampler 31 1
C.
NA NA NA NA C.
Particulate sampler 3.
Containment Purge or Vent 3.
Containment Purge or Vent A.
Gas activity monitor A.
- 1. 1RTPR09C 27 1
- 1. D P
R Q (1) +
2.
2 RTPR09C 27 1
- 2. D P
R Q (1) +
B.
Iodine sampler 31 1
B.
NA NA NA NA C.
Particulate sampler 31 1
C.
NA NA NA NA
- During releases via this pathway
+ Automatic isolation applies only to containment purge pathway.
242e
Table 3.12.2 (Continued)
Table 4.12.2 (Continued)
Radioactive Gaseous Effluent Radioactive Gaseous Ef fluent Monitoring Instrumentation Monitoring Instrumentation Surveillance Minimum Channel Channels Channel Source Channel Functional Instrument Applicability Action Operable Check Check Calibration Test 4.
Auxiliary Building ventilation 4.
Auxiliary Building Ventilation A.
A.
Gas activity monitor 1.
7.RTPR25 or ORE 0014 27 1
1.
D M
R O (2) 2.
2RTPR25 or ORE 0014 27 1
2.
D M
R O (2) 31 1
B.
NA NA NA NA B.
Iodine sampler 31 1
C.
NA NA NA NA C. Particulate sampler 5.
Miscellaneous Ventilation Stack 5.
Miscellaneous Ventillation Stack A.
A.
Gas activity monitor 27 1
- 1. D M
R O (2)
- 1. ORTPR18B 31
'l B.
NA NA NA NA B.
Iodine sampler 31 1
C.
NA NA NA NA C.
Particulate sampler 6.
Service Building Ventilation' 6.
Service Building Ventillation 31 1
A.
NA NA NI NA A.
Iodine sampler 31 1
B.
NA NA NA NA B. Particulate sampler S
242f
TABLE NOTATION Table 3.12.2 (Continued)
Table 4.12.2 (Continued)
Radioactive Gaseous Effluent Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation Surveillance ACTION 25 - With the number of channels OPERABLE (1) - The CHANNEL FUNCTIONAL TEST shall also less than required by the Minimum demonstrate that automatic isolation of Channels OPERABLE requirement, the this pathway and control room alarm contents of the tank may be released-annunciation occurs if the instrument w
to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> indicates measured levels above the provided that prior to initiating the alarm / trip setpoint.
release:
1.
At least two independent samples of the tank's content are analyzed, and 2.
At least two technically qualified members of the Facility Staff in-(2) -
The CHANNEL FUNCTIONAL TEST shall also dependently verify the release rate demonstrate that control room alarm an-calculations and discharge line nunciation occurs if any of the follow-valving; ing conditione exist:
otherwise, suspend release of radioactive 1.
Instrument indicates measured levels effluents via this pathway.
above the alarm / trip setpoint.
ACTION 27 -
With the number of channels OPERABLE less 2.
Instrument indicates a downscale than required by the Minimum Channels failure.
OPERABLE requirement, effluent releases via this pathway may continue for up to 28 days provided grab samples are taken at least once per shif t. and these samples are analyzed for gross activity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
During releases via the pathway 242g
TABLE NOTATION Table 3.12.2 (Continued)
Table 4.12.2 (Continued)
Radioactive Gaseous Effluent Radioactive Gaseous Effluent Monitoring Instrumentation Monitoring Instrumentation Surveillance ACTION 31 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE Requirement, effluent releases via this pathway may continue for up to 28 days, pro-vided samples are continuously collected with auxiliary sampling equipment for periods on the order of seven (7) days and analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the end of sample collection.
242h
I Basis:
Specification 3.12.1.A. is provide <I to ensure that the dose at the unrestricted area boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CPR Part 20.
The annual dose limits are the dosea associated with the concentrations of 10 CFR Part 20, Appendix B,.
Table II.
These limits provide reasonable assurance that radioactive material discharged in gaseous effluents 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 II of 10 CFR Part 20 (10 CFR Part 20,, 106 (b) ).
The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the unre-stricted boundary to:E500 mrem / year to the total body or te13000 mrem / year to the skin.
These release rate limits also restrict, at a 1 times, the corresponding thyroid dose rate above backgound to an infant via the cow-milk-infant pathway to il500 mrem / year for the nearest cow to the plant.
For purposes of calculating doses resulting from airbarne releases the two stacks are considered a mixed mode release.
242i
Specification 3.12.1C. is provided to implement the requirements of Sections II.B, III.A.
and IV.A. of Appendix I, 10 CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable."
The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by cal-culational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated.
Section 3.12.lE is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50.
The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."
The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by the calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.
These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.
The release rate specifications for radioiodines, radioactive material in particulate form and radioiodines other than noble gases are dependent on the existing radio-nuclide pathways to man, in the unrestricted area.
The pathways which are examined in the develop-ment of these calculations are:
1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man.
The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases.
The alarm / trip setpoints for these instruments shall be calculated in accordance with the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
Restricting the quantity of radioactivity contained in each gas decay tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest unrestricted area boundary will not exceed 0.5 rem.
242j
9 LIMITINd CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.14 Plant Radiation Monitoring (non-e f fluent) 4.14 Plant Radiation Monitoring Applicability:
Applicability:
Applies to radiation measurement Applies to surveillance of radiation instrumentation.
measurement instrumentation.
Objective:
Objective:
To measure radiation levels in certain To establish surveillance frequencies critical areas of the plant.
for the associated radiation monitors.
Specification:
Specification:
250
o LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 1.
The radiation monitoring instrumentation 1.
Each radiation monitoring channels shall be operable except as instrumentation channel shall be specified in 3.14.2.
demonstrated OPERABLE by performance of the CHANNEL CHECK,, SOURCE CHECK, CIIANNEL 2.
With one or more radiation monitoring CALIBRATION and CHANNEL FUNCTIONAL TEST channels inoperable, take the action operations during the MODES and at the c5own in Table 3.14.1.
If the channel frequencies shown in Table 3.14.1.
is not operable for more than 30 days, it will be considered a degraded mode
- 2.. Not applicable for reporting purposes (6. 6. 2. a. 2 ).
3.
Not applicable 3.
In the event a limiting condition for operation and/or associated action re-quirements cannot be satisfied because of circumstances in excess of those addressed in the specification, no
~
changes are required to the operational conditions of the plant non does it prevent changes in the operational mode.
251
Table 3.14.1 Table 4.14.1 Radiation Monitoring Radiation Monitoring Surveillance Minimum Channel Channels channel Check Channel Functional Instrument Applicability Action Operable Check Source Calibration Test
- 1. Control Room ORT-PROlA 20 1
D W
R Q
ORT-PROlB 20 1
D W
R Q
- 2. Fuel Bldg.
ORT-AR03 20 1
D W
R Q
Table Notation:
Action 19
- With the number of channels OPERABLE less than required by the MINIMUM channels operable requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- At All Times 252/253
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.16 Environmental Radiological Monitoring 4.16 Environmental Radioactivity Monitoring Program Program Applicability Applicability Applies to Radiological Monitoring of Applies to Radiological Monitoring plant environs.
of plant evirons.
Objective Objective To establish a radiological monitoring To establish a radiological monitoring program adequate to measure changes in prog, ram adequate to measure changes the levels of environmental radioactiv-in the levels of anvironmental radioactiv-ity.
ity due to plcat' efflents.
Specification Specification 1.
The radiological environmental monitorin9 1.
The radiological environmental monitoring samples program shall be conducted as specified shall be collected pursuant to Table 3.16-1 from in Table 3.16-1.
the locations shown on Figure 3.16-1 2.
With the radiological environmental The results of analyses performed an the radio-monitoring program not being conducted logical environmental monitoring samples shall as specified in Table 3.16-1, prepare he summarized
, the Annual Radiological Environ-and submit to the Commission, in the mental Operating 3eport.
See sections 6.6.3.c Annual Radiological Operating Report, and 6.6.3.d.
a description of the reasons ~for not conducting the program as required and 2.
Not applicable.
the plans for preventing a recurrence.
Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous con-ditions, seasonal unavailability, con-tractor omission which is corrected as soon as discovered, malfunction of sam-pling equipment, or if a person who participates in the program by providing samples goes out of business.
275
o LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.16.2 Continued
[4.16.2 Continued If the equipment malfunctions, corrective actions shall be completed as soon as practical.
If a person supplying sample ;
goes out of business, a replacement will be found as soon as possible.
All deviations from the sampling schedule shall be described in the annual report.
3.
With the level of radioactivity in an l3.
Not Applicable.
environmental sampling medium at one or more of the locations specified in Figure 3.16-1 exceeding the limits of Table 3.16 i column 5,when averaged over any calena quarter, prepare and submit to the cor ission within 30 days from the e-d of L'e affected calendar quarter, a Special L. aort which includes an evaluat on of any release conditions, environmental factors or other aspects which caused the limits of Table 3.16-1, column 5, to be exceeded.
This report is not required i' the measured level of radioactivity was not the result of plant effluents; how-ever, in such an event, the condition shall be, reported and described in the Annual Radiological Environmental Operating Report.
4.
With milk samples unavailable from any o l4.
Not Applicable, the sample locations required by Figure 3.16-1, prepare and submit to the Com-mission within 30 days, pursuant to Spec ification 6.6.3.c.h.,
a Special Report which identifies the cause of the unavai ability of samples and identifies loca-tions for obtaining replacement samples.
The locations from which samples were unavailable may then be deleted from the program provided the locations from which the replacement samples were obtained are added to the environmental monitoring 276 program as replacement locations.
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.16.5.
A census of animals producing milk for 4.16.5.
The land use census shall be conducted human consumption shall be conducted at least once per 12 months between the annually during the grazing season to dates of (June 1 and October 1) by a determine their location and number with door-to-door survey,. road survey, respect to the site.
The census shall be aerial survey, or by consulting local conducted under the following conditions:
agriculture authorities.
A.
Within a 2-mile radius from the plant The results of the land use census shall site, enumeration by a door-to-door be included in the Annual Radiological or equivalent counting technique.
Environmental Operating Report.
B.
Within a 5-mile radius, enumeration by using referenced information from county agricultural agents or other relia'cle sources.
6.
With a land use census identifying a loca-6.
Not Applicable; tion (s) which yield (s) a calculated dose or dose commitment greater than the values currently being calculated in Specificati on 3.11.2.3 prepare and submit to the Com-mission within 30 days, pursaant to Spec-ification 6.9.2, a.Special Report which identifies the new location (s).
The new location (s) shall be added to the radiologic'al environmental monitoring program within 30 days, if poss~ible, The sampling location having the lowest cal-culated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.
277
LIMITI?:G CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.16.7.
Radiological Analyses shall be performed 4.16.7.
The results of analyses performed as part on samples representative of those in of the above required crosscheck program shall Table 3.16-1 that are supplied as part of be included in the Annual Radiological Environ-the Environmental Protection Agency's mental Operating Report.
(EPA) Environmental Radioactivity Labora-tory Intercomparisons Studies (Crosscheck)
Program.
At least one sample of each typc provided by the EPA shall be analyzed annually.
8.
With analyses not being performed as re-8.
Not Applicable quired above, report the corrective actior s taken to prevent a recurrence to the Com-mission in the Annual Radiological Enviror mental Operating Report.
9.
In the event a limiting condition for oper -
9.
Not Applicable ation and/or associated. Action requirement s cannot be satisfied because of circumstanc es in excess of those addressed in the speci-fications, no changes ase required to the operational conditions of the plant nor does it prevent changea in the operational mode.
278
LRATIONb 0F FIXED ENVIRONMENTAL RADIOLOGICAL MONITORING STATIONS I-0nsite Station I g/
- accin, re 3.16-1 2-Onsite Station 2 3-Onsite Station 3 h
4-Zion (W)
.:+
5-Zion (SW)
'9 g,no,n, N(
6-Zion (WNW)
. i3 Il Jl 7-WinIhrop Harbor Cily Garage
\\
^
8-Kenosha Road Farm
~~
,,y $th i n" $i s 7
9-Waukegan j
3 h
fg f o-North Chicago Il-Sheridan Lanes Bowling Alley
.f
)
.iwoukegon 12-Flood Farm
- l2 c"Q, u/
cn f
g a y
13-Pleasant Prairie
,f
.3'N W.
n j{
l L:he Bluf f L.S a * ?, v.n e
't t s:L_
Lohe fores
\\4 A
~
SCALE 8
?_
IS to vitts
-w
" A----
278a
TABLE 3.16-1 ZION STANDARD RADIOLOGICAL MONITORING PROGRAM Non-Routine A
Sample Media Collection Site Type of Analysis Frequency Reporting Levels 3
e 1.
Air Monitoring (a) Onsite ar d near field 1.
Filter - gross beta
- 1. Weekly CS-134 10, Cs-137 20 pCi/m 3
(1) Onsite. Station #1 2.
Charcoal - I-131 2.
Bi-Weekly
- 0.7 pC1/m (2) Onsite Station #2 (3) Onsite Station #3 3.
Sampling Train -
3.
Weekly Test and Maintenance (b) Far Field (1) Zion (W) 1.
Filter Exchange 1.
Weekly Sameas1(a)
(2) Zion (SW)
(3) Zion (WNW) 2.
Charcoal Exchange 2.
Bi-Weekly
- when analyses (4) Winthrop Harbor City Garage 3.
Sampling Train -
3.
Weekly are made (5) Kenosha Road Farm Test and Maintenance (6) Waukegan (7)
N. Chicago (8) Sheridan Lanes Bowling Alley (9) Flood Farm (10) Pleasant Prairie DC 2.
TLD Same as 1 Gamma Radiation Quarterly 4
4 3.
Fisti Lake Michigan near Gamma isotopic Semi-annual Mn-54 3x10, Fe-591x10 Zion Station C0-58 3x10, Co-60 lx10 4
3 Zn-65 2x10 Cs-134 lx10 3
Cs-137 2x10 pCi/Kg wet weight 279 f
TABLE 3.16-1 ZION STANDARD RADIOLOGICAL MONITORING PROGRAM (CONT'D)
Non-Routine A
Sample Media Collection Site Type of Analysis Frequency Reporting Levels 4.
Milk (a) 2 Dairy Farms I-131 1.
Weekly -
I-131 3 pCi/l Grazing Season -
Cs-134 60 pCi/l May to Oct Cs-137 70 pCi/l 2.
Monthly -
Nov to Apr Ba-La-140 300 pC1/1 5.
Public Water Supply (a) Kenosha Water Works 1.
Gamma Isotopic 1.
Monthly Analysis Footnote **
(b) Lake County Water Works of weekly Composites (c) Waukegan Water Works 2.
Tritium 2.
Quarterly (d)
N. Chicago Water Works Composite (e) Naval Training Center Water Works (f) Lake Forest Water Works 6.
Cooling Water (a)
Inlet 1.
Gross Beta 1.
Weekly Sample (b) Discharge 2.
Tritium 2.
Quarterly Composite 7.
Sediment Lake Michigan (a)
Illinois Beach St.
Gamma Isotopic Annual Park Note:
- r1-Weekly shall mean that the frequency is once every other week.
AAverage concentration over calendar quarter 4
3 2
2 2
2 2
oA gamma isotopic analysis shall be performed whenever the gross beta concentration in a sample exceeds by five times (5x) the average concentration of the proceeding calendar quarter for the sample location.
280
TABLE 4.16-1 PRACTICALLOWERLIMITSOFDETECTION(LLD)
FOR STANDARD ENVIRCNMENTAL RADIOLOGICAL MONITORING PROGRAM LLD Sample Media Analysis (4.66r)
Units 3
Airborne " Particulate" Gross Betat 0.01 pCi/m Gamma Isotopic 0.01 pC1/m3 Sr-89,90 0.01 pCi/m 3
Airborne I-131 Iodine-131 0.10 pCi/m Liquids Sr-89 10 pCi/1 Sr-90 2
pCi/1 I-131 5*
pCi/1 Cs-134 10 pCi/1 Cs-137 10+
pCi/1 Tritium 0.2 pCi/ml Gross Betat 5
pCi/1 Gamma Isotopic
< 20 pCi/1/nuclide Vegetation Gross Betat 2
pCi/g wet I-131 0.03 pCi/g wet Sr-89,90 1
pCi/g wet Gamma Isotopic 0.2 pCi/g wet Soil, Sediment Gross Betat 2
pCi/g dry Sr-89,90 1
pC1/g dry Gama Isotopic 0.2 pCi/g dry Animal Tissue Sr-89,90 0.1 pCi/g wet I-131 - Thyroid 0.1 pCi/g wet Cs-134, 137 0.1 pCi/g wet Gross Betat 1.0 pCi/g wet
- 0.5 pCi/1 on milk samples collected during the pasture season.
tReferenced to Cs-137
+5.0 pCi/1 on milk samples.
280a
TABLE 4.16-1 (Continued)
TABLE NOTATION The LLD is the smallest concentration of radioactive a
material in a sample that will be detected with 95%
probability with only 5% probability of falsely con-cluding its presence.
For a particular measurement system (tehich may include radiochemical separation):
LLD =
exp ( feat) 2.22 Y
t
.E V
+
where LLD is the lower limit of detection as defined above (as pCi per unit mass or volume) sb is the square root of the background count or of the count of a blank sample as appropriate (as counts).
E is the counting efficiency (as counts per transformation)
V is the sample size (in units of mass or volume) 2.22 is the number of transformation per minute per picocurie Y is the fractional radiochemical yield (when applicable)
A is the radioactive decay constant for the particular radionuclide At is the elapsed time between sample collection and analysis t is the duration of the count 280b
TABLE 4.16-1 (Continued)
TABLE NOTATION a.
(continued)
The valae of s used in the calculation of the LLD u
for a detection system shall be based on the actual observed background count or of the count of the blank samples (as appropriate) rather than on an unverified theoretically predicted value.
In cal-culating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the contributions of other radionuclides present in the sample, b-The LLD for environmental measurement is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
280c
Basis:
3.16/4.16 The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation.
This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measure-able concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The initially specified monitoring program will be effective for at least the first two years of commercial operation.
Following this period, program changes may be initiated based on operational experience.
The detection capabilities required by Table 4.16-1 are state of-the-art for routine environmental measurements in industrial laboratories.
The specified lower limits of detection for I-131 in water, milk and other food products correspond to approximately one-quarter of the Appendix I to 10 CFR Part 50 design objective dose-equivalent of 15 mrem / year for atmospheric releases and 10 mrem / year for liquid releases to the most sensitive organ and individual.
They are based on the assumptions given in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," October 1977, except the change for an infant consuming 330 liters / year of drinking water instead of 510 liters / year.
This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census.
This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CER Part 50.
The requirement for participation in the EPA crosscheck program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in en'rironmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
2800
a.
Offsite Review and Investigative (2) Proposed changes to procedures, equip-ment or systems which involve an unreviewed Function (Continued) safety question as defined in Section 50.59 comprehensive interdisciplinary review 10 CFR.
coverage under this function, (iv) independ-ently review and approve the findings and (3) Proposed tests or experiments which involve recommendations developed by personnel an unreviewed safety question as defined performing the review and investigative in Section 50.59 10 CFR.
function, (v) approve and report in a timely manner all findings of noncompliance with (4) Proposed changes in Technical Specifications NRC requirements and provide recomendations or NRC operating licenses.
to the Station Superintendent Division Manager Nuclear Stations, Manager of (5) Noncompliance with NRC requirements, or of Quality Assurance, and the Vice President internal procedures or instructions having of Construction Production, Licensing nuclear safety significance.
and Environmental Affairs. During periods when the Supervisor of the Offsite Review (6) Significant operating abnonnalities or deviations and Investigative Function is unavailable, from nonnal and expected performance of plant he shall designate this responsibility to equipment that affect nuclear safety as referred an established alternate who satisfies the to it by the Onsite Review and Investigative function.
formal training and experience requirements for the supervisor of the Offsite Review and Investigative Function.
(7) Reportable Occurrences requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notifi-cation to the Commission.
The responsibilities of the personnel performing this function are stated below.
(8) All recognized indications of an unanticipated The Offsite Review and Investigative deficiency in some aspect of design or operation of safety related structures, systems or components.
Function shall review:
(1) The safety evaluations for 1) changes (9) Review and report findings and recomendations to procedures, equipment or systems regarding all changes to the Generating Stations as described in the safety analysis Emergency Plan prior to implementation of such report and 2) tests or experiments change.
completed under the provision of 10 CFR Section 50.59 to verify that (10) Review and report findings and racomendations such actions did not constitute an regarding all items referred by the Technical unreviewed safety questions. Proposed Staff Supervisor, Station Superintendent, changes to the Quality Assurance Program Division Manager - Nuclear Stations and Manager description shall be reviewed and approved of Quality Assurance, by the Manager of Quality Assurance.
(11) Changes to the Offsite Dose Calculation Manual (ODCM).
301
- b. Audit Function (6) Audit of the Facility Security Plan and implementing procedures.
The Audit Function shall be the responsibility of the Managar of Quality Assurance independent (7) Audit onsite and offsite reviews.
of the Production Department.
Such responsibility is delegated to the Director of Quality Assurance (8) Audit the facility Fire Protection Program for Operating and to the Staff Assistant to the and implementing procedures at least once Manager of Quality Assurance for maintenance per M m es, quality assurance activities.
(9) Audit the radiological environmental Either shall approve the audit agenda and check-monitoring program at least once per lists, the findings and the report of each audit.
g Audits shall be performed in accordance with the Compnay Quality Assurance Program and Procedures. l(10) An independent fire protection and loss Audits shall be performed to assure that safety-prevention program inspection and audit related functions are covered within a period of shall be performed at least once per 12 two years or as designated below.
months utilizing either qualified offsite li nse p rsonnel or an outside fire (1) Audit of the Conformance of facility opera-protection firm.
tion to provisions contained within the Technical Specifications and applicable l(11) An inspection and audit of the fire license conditions at least once per year.
protection and loss prevention program shall be performed by a qualified outside (2) Audit of the adherence to procedures, fire consultant at least once per 36 months.
training and qualification of the station staff at least once per year.
l(12) Report all findings of noncompliance with NRC requirements and recommendations and (3) Audit of the results of actions taken t results of each audit to the Station correct deficiencies occurring in facility Superintendent, the Division Manager -
equipment, structures, systems or methods Nuclear Stations, Manager of Quality of operation that affect nuclear safety at Assurance, the General Superintendent of least once per six months.
Production Systems Analysis, and to the Vice President of Construction, Production, (4) Audit of the performance of activities Licensing and Environmental Affairs.
required by the Quality Assurance Program to meet the Criteria of Appendix "B",
(13) The Offsite Dose Calculation Manual at 10 CFR 50.
least once per 24 months.
(5) Audit of the Facility Emergency Plan and implementing procedures.
302
- c. Authority The Manager of Ouality Assurance reports to the Executive Vice-President and the Super-visor of the Offsite Review and Investigative Function reports to the General Superinten-dent of Production Systems Analysis.
Either the !!anager of Quality Assurance or the Super-visor of the Of fsite Review and Investigative Function has the authority to order unit shut-down or request any other action which he deems necessary to avoid unsafe plant conditions.
- d. Records (1) Reviews, audits and recommendations shall be documented and distributed as covered in 6.1.6.1.a and 6.1.6.1.b.
3 03
d.
Records (Continued) f.
Personnel (2) Copies of documentation, reports, and (1) The persons, including consultants, per-correspondence shall be kept on file forming the review and investigative function, at the station.
in addition to the Supervisor of the Offsite Review and Investigative Function, shall e.
Procedures have expertise in one or more of the following disciplines as appropriate for the subject Written administrative procedures shall be or subjects being reviewed and investigated.
prepared and maintained for the off-site reviews and investigative functions described in Specifications 6.1.G,1.a.
(a) nuclear power plant technology These procedures shall cover the following:
(b) reactor operations (c
utility operations (1) Content and method of submission of (d
power plant design presentations to the Supervisor of the (e
reactor engineering Offsite Review and Investigative Function.
(f) radiological safety (g) reactor safety analysis (2) Use of comittees and consultants.
(h) instrumentation and control (1) metallurgy (3) Review and approval.
(j) any other appropriate disciplines required by unique characteristics (4) Detailed listing of items to be reviewed, of the facility.
(5) Method of (a) appointing personnel, (b) performing reviews, investigations, (2)
Individuals performing the Review and (c) reporting findinos and recommendations Investigative Function shall possess a of reviews and investigations, (d) approving minimum formal training and experience reports, and (e) distributing reports.
as listed below for each discipline.
(6) Determining satisfactory completion of (a) Nuclear Power Plant Technology action required baseo on approved findings and recomendations reported by personnel Engineering graduate or egoiva-performing the review and investigative lent with 5 years experience function.
in the nuclear power field design and/or operation.
304
(h) Instrumentation and Control (b) Reactor Operations Engineering graduate or equivalent Engineering graduate or equivalent with at least 5 years of experience with 5 years experience in nuclear in instrumentation and control power plant operations.
design and/or operation.
(c) Utility Operations (i) Meta llurgy Engineering graduate or equivalent Engineering graduate or equivalent with at least 5 years of experience with at least 5 years of experience in utility operation and/or engineering.
in the metallurgical field.
(d) Power Plant Design (3) The Supervisor of t.
Offsite Review and Investigative Function shall have Engineering graduate or equivalent experience and training which satisfy with at least 5 years of experience ANSI N18.1 - 1971 requirements for in power plant design and/or operation.
plant managers.
(e) Reactor Engineering The Onsite Review and Investigative Function 2.
shall be supervised by the Station Superintendent.
Engineering graduate or equivalent.
In addition, at least 5 years of Onsite Review and Investigative Function experience in nuclear plant engineering, a.
operation, and/or graduate work in The Station Superintendent shall: (i) provide nuclear engineering or equivalent direction for the Review and Investigative in reactor physics is required.
Function and appoint the Technical Staff Supervisor, or other comparably qualified (f) Radiological Safety individual as a senior participant to provide appropriate direction; (ii) approve partici-Engineering graduate or equivalent Pants for this function; (iii) assure that a with at least 5 years of experience complement of more than one participant who in radiation control and safety.
collectively possess background and qualifi-cations in the subject matter under review (g) Safety Analysis are selected to provide: comprehensive inter-disciplinary review coverage under this Engineering graduate or equivalent function; (iv) independently review and approve with at least 5 years of experience the findings and reconinendations developed in nuclear engineering.
305
- a. Onsite Review and Investigative.
(5) Investigation of all noncompliance with NRC Function (Continued) requirements and shall prepare and forward a report covering evaluation and recommenda-by personnel performing the Review und tions to prevent recurrence to the Division Investigative Function; (v) report all Manager-Nuclear Stations and to the findings of noncompliance with NRC re-Supervisor of the Offsite Review and quirements, and provide recommendations Investigative Function.
to the Division Manager-Nuclear Stations (6) Review of facility operations to detect and the Supervisor of the Offsite Review and Investigative Function; and (vi) submit Potential safety hazards.
to the Offsite Review and Investigative (7) Performance of special reviewr and in-Function for concurrence in a timely manner, vestigations and reports thereon as re-those items described in Specification quested by the Supervisor of the Offsite 6.1.G.l.a which have been approved by the Review and Investigative Function.
Onsite Review and Investigative Function.
(8) Review of the Station Security Plan and The responsibilities of the personnel shall submit recommended changes to the performing this function are stated Division Manager-Nuclear Stations.
below:
(9) Review of the Emergency Plan and station (1) Review of:
- 1) procedures required by implementing procedures and shall submit Specification 6.2 and changes thereto, recomaended changes to the Division
- 2) any other proposed procedures or M nager-riuclear Scations.
changes thereto as determined by the (10) Review of reportable occurrences and actions Plant Superintendent to affect nuclear taken to prevent recurrence.
safety.
(11) Review of changes to the Offsite Dose (2) Review of all proposed tests and Calculation Manual (ODCM).
experir.ents that affect nuclear safety.
,(12) Review of every unplanned release of radioactive material to the environment (3) Review of all proposed changes to the Technical Specifications.
including evaluation of the event, corrective action to prevent recurrences, disp sition of the corrective action, and (4) Review of all proposed changes or documentation of the event.
modifications to plant systems or
- b. Authority equipment that affect nuclear safety.
The Technical Staff Supervisor is responsible to the Station Superintendent and shall n.ake recommendations in a timely manner in all areas of review, investigation, 306 and quality control phases of plant maintenance, opera-tion and administrative procedures relating to facility
b.
Authority (Continued)
(1) Content and method of submission and presentation to the Station Superintendent, Division Manager-Nuclear Stations and the operations and shall have the authnrity to Supervisor of the Offsite Review and In-request the action necessary to ensure compliance with rules, regulations, and procedures when in vestigative Function.
his opinion such action is necessary. The Station Superlatendent shall follow such (2) Use of comittees.
recommendations or select a course of action that is more conservative regarding safe (3) Review and approval.
operation of the facility. All such disagree-ments shall be reported imediately to the (4) Detailed listing of items to be reviewed.
Division Manager-Nuclear Stations and the Supervisor of the Offsite Review and ICstiga-(b) Procedures for administration of the quality control activities.
tive Function.
(6) Assignment of responsibilities.
c.
Records e.
Personnel (1) Reports, reviews, investigations, and recomendations shall be documented (1) The personnel performing the Onsite Review with copies to the Division Manager-and Investigative Function, in addition to Nuclear Stations, the Supervisor of the Station Superintendent, shall consists the Offsite Review and Investigative of persons having expertise in:
function, the Station Superintendent and the Manager of Quality Assurance.
(a) nuclear power plant technology (b) reactor operations (2) Copies of all records and documentation (c) reactor engineering shall be upt on file at the station.
(d) radiological safety and chemistry (e) instrumentation and control d.
Procedures _
(f) mechanical and electric systems.
~
Written administrative procedures shall be (2) Personnel performing the Onsite, Review prepared and maintained for conduct of the and Investigative Function shall meet Onsite Review and Investigative Function.
minimum acceptable levels as described These procedures shall include the following:
in ANSI N18.1 1971, Sections 4.2 and 4.4.
307
6.2 Plant Operating Procedures A.
Detailed written procedures including
- 8. Tests and experiments.
applicable checkoff lists covering items listed below shall
'.>o prepared, approved,
- 9. Procedure to ensure safe shutdown of and adhered to:
the plant.
- 1. Normal startup, operation, and shutdowr.
10 Station Security Plan and implementing of the reactor and other systems and procedures.
components involving nuclear safety of the facility.
- 11. Fire Protection Program implementation.
- 2. Refueling operations.
- 12. Offsite Dose Calculation Manual implementation.
- 3. Actions to be taken to correct specific and foreseen potential malfunctions of B. Radiation control procedures shall be systems or components including responses maintained, made available to all station to alarms, suspected primary system leaks, personnel and adhered to.
These procedures and abnormal reactivity changes, shall show permissible radiation exposure and shall be consistent with the require-
- 4. Emergency conditions involving potential ments of 10 CFR 20.
This radiation or actual release of radioactivity -
protection program shall be organized to
" Generating Stations Emergency Plan" and meet the requirements of 10 CFR 20.
station emergency and abnormal procedures.
- 5. Instrumentation operation which could have an effect on the safety of the facility.
6.
Preventive and corrective maintenance operations which could have an effect on the safety of the facility.
7.
Surveillance and testing requirements.
308
B.
- 1. Procedures for items identified in Speci-C. Temporary changes to procedures 6.2.A fication 6.2.A and any changes to such and 6.2.B above may be made provided:
procedures shall be reviewed and approved by the operating Engineer and the Tech-
- 1. The intent of the original orocedure nical Staff Supervisor in the areas of is not altered.
operation, fuel handling, or instrument maintenance, and by Maintenance Engineer
- 2. The change is approved by two and Technical Staff Supervisor in the members of the plant management areas of plant maintenance and plant in-staff, at least one of whom holds spection.
Procedures for items identified a Senior Reactor Operator's License in Specification 6.2.B and any changes to on the unit affected.
such procedures shall be reviewed and approved by the Technical Staff Supervi-
- 3. The change is documented, reviewed sor and the Radiological Chemical Super-by the Onsite Review and Investiga-visor.
At least one person approving tive Function and approved by the each of the above procedures shall hold Station Superintendent within 14 a valid senior operator's license.
In days of implementation.
addition, these procedures and changes thereto must have authorization by the D.
Drills of the emergency procedures Station Superintendent before being described in Specification 6.2.A.4 shall implemented.
be conducted quarterly.
These drills will be planned so that during the
- 2. Work and instruction type procedures course of the year, communication which implement approved maintenance or links are tested and outside agencies modification procedures shall be approved are contacted.
and authorized by the Maintenance Engineer where the written authority has been pro-vided by the Station Superintendent.
The
" Maintenance / Modification Procedure" utilized for safety related work shall be so approved only if procedures referenced in the " Maintenance / Modification Procedure" have been approved as required by 6.2.A.
Procedures which do not fall within the requirements of 6.2.A or 6.2.B may be approved by the Department Heads.
309
6.3 Action to'be Taken in the Event of an Reportable 2.
Record of principal maintenance Occurrence in Plant Operation activities, including inspection and repair, regarding principal Any reportable occurrence shall be promptly items of equipment pertaining to reported to the Division Manager-Nuclear nuclear safety.
Stations or his designated alternate.
The incident shall be promptly reviewed 3.
Records and reports of reportable pursuant to Specification 6.1.G,2.a(5) and safety limit occurrences.
and a separate report for each reportable occurrence shall be prepared in accordance 4.
Records and periodic checks, with the requirements of Specification inspection and/or calibrations 6.6.B.
performed to verify the Surveillance Requirements (See 6.4 Action to be Taken in the Event of a Safety Limit Section 4 of these Specifications) is Exceeded are being met. All equipment failing to meet surveillance If a safety limit is exceeded, the reactor requirements and the corrective shall be shut down imediately and reactor action taken shall be recorded.
operation shall not be resuned until authorized by the NRC. The conditions of shutdown shall 5.
Records of changes made to the be promptly reported to the Division Manager-equipment or reviews of tests and Nuclear Stations or his designated alternate.
experiments to comply with The incident shall be reviewed pursuant 10 CFR 50.59.
to Specification 6.1.G.l.a and 6.1.G.2.a and a separate report for each occurrence 6.
Records of radioactive shipments.
shall be prepared in accordance with 7.
Records of physic tests and other Specification 6.6.B.
tests pertaining to nuclear safety.
6.5 Plant Operating Records 8.
Records of changes to operating A.
Record and/or logs relative to the following procedures, items shall be kept in a manner convenient for review and shall be retained for at 9.
Shift Engineers Logs.
least five years.
- 10. By-product material inventory records 1.
Records of normal plant operation, and source leak test results.
including power levels and periods of operation at each power level.
310
B.
Records and/or logs relative to the following
- 10. Records of transient or operational items shall be recorded in a manner convenient cycling for those components that for review and shall be retained for the life have been designed to operate safely of the plant.
for a limited number of transient or operational cycles.
1.
Substitution or replacement of principal items of equipment pertaining to nuclear
- 11. Records of individual staff members sa fety.
indicating qualifications, experience, training and retraining.
2 Changes made to the plant as it is described in the Safety Analysis Report.
12.
Inservice inspections of the reactor coolant system.
3.
Records of new and spent fuel inventory and assembly histories.
- 13. ' Minutes of meetings and results of reviews performed by the off-site and
- 4. (Deleted) on-site review functions.
5 Updated, corrected, and as-built drawings of the plant, 6.
Records of plant radiation and contamination surveys.
7.
Records of off-site environmental monitoring surveys.
8.
Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant in accordance with 10 CFR 20.
9.
Records of radioactivity in liquid and gaseous wastes released to the envi ronment.
311
6.6 Reporting Requirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforce-ment unless otherwise noted.
1.
Routine Reports a.
Startup Report.
A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of ar.
operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifica-tions that may have significantly altered t.'e nuclear, thermal, or hydraulic performance of the plant.
The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or charac-teristics obtained during the test program and a comparison of these values with design pre-dictions and specifications.
Any corrective actions that were required to obtain satisfac-tory operation shall also be described.
Any additional specific details required in license conditions based en other commitments shall be included in this report.
Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.
If the Startup Report does not cover all three events (i.e.,
initial criticality, completion of startup test pro-gram, and resumption or commencement of com-mercial power operation), supplementary reports shall be submitted at Icast every three months until all three events have been completed.
b.
Annual Occupational Exposure Data. A tabula-tion covering the previous calendar year should be submitted prior to March 1 of each year on the number of atation, utility and other per-sonnel (including contractors) receivir.g expos-ures g.ewcer than 100 mrem / year and their 312
6.6 Reporting Reauirements (Cont'd) associated man rem exposure according to work and job functions, 1 e.g.,
reactor operations and surveillance, inservice inspectioq, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.
The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.
Small ;xposures totalling less than 20% of the individual total dose need not be accounted for.
In the aggregate, 80% c#
the total whole body dose received from external sources shall be assigned to specific major work functions.
THIS SPACE INTENTIONALLY LEFT BLA g 1.
This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
313
6.6 Reporting Recuirements (Cont ' d) c.
Monthly Operatinc Recort.
Routine reports using the " Operating Data", " Unit Shutdowns and Lower Reductions" and " Average Daily Unit Power Level" report formats,shall be submit-ted on a monthly basis to the Director, Office of Management, Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555 wi th a copy to the appropriate Regional Offf.cc, to arrive no later than the
N of each month following the calendar month 15th Ni/
covered by the report.
Included with the above three report formats will be " Narrative Summary of Operatina Experience" that describes the operation of the facility, including major safety-related maintenance, for the monthly report per-iod.
In addition, any changes to the Offsite Dose Calculation Manual shall be submitted with the monthly Operatina Peport within 90 days of the effective date of the chance.
2.
Reportable Occurrences Reportable occurrences, including corrective actions and measures to prevent reoccurrence, shall be re-ported to the NRC.
Supplemental reports may be required to fully describa final resolution of an occurrence.
In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
t I
314 I,
6.6 Reportine Requirenents (cont'd) a.
Pro: t h'o t i fi ca ticn !!ith t?t i t ten Fo 1 ~. meu n.
The types of cvents listed helcu shal] Le report <.d as expediticunly as possibic, but ui:hin 24 hourc by tc.cphone and confi.:cd by telegraph, cailgr.mm, or facciuile t ransrc.icsion to (.bc Director of the appropriate Fegional Office, or his decignate no later than thu firct 'corkir.g day follouing the event, with a uritten fo))ocup report within tuo uceks.
The uritten follouup report shall include, as a minimun, a completed copy of a licensec event report f o r:n.
Infornation provided on the licencee event report fo m shall be supplemented, as needed, by cdditional narrative natcrial to provide con-pletc explanation of the circunstances surroundint; the event.
(1)
Failure of the reactor protection cycten or other systems subject to limiting safety systen settings to initicrc the rcquired protective function by the time a moni. cored parameter reaches the setpoint snecified as the liciting safety nysten setting in the tc;hnical specifications or failure to complete the required protective functipn.
Note:
Instrument drift discovered as a result of testing need not be tcpcrted undc this item but t.my he reportab2 c under ite=n 2.a(5),, 2.a(6), or 2.b(1) h e.l ou.
A M s...
(2)
Operation of the unit or affected systems when any parameter or operativa culject to a limiting condition is less conservative than the Icast conservative aspect of the limiting condition for operation cLtablished in the technical specifications.
Note:
If cpecified action is taken when a system in found to be operating between the nost concorvctive and the least conservative aspects of a limiting conditien for operation listed.in Qte technical specifications, the limiting conditien for operation is not censidered to have been violated and need not be reported under this item, but it may be reportable under iter., 2.b(2) belou.
(3)
Abnormal degradation discovered in fuel claddinp, reactor coolant, pressure boundary, or primary containment.
Note:
Leahcce of valve packing or gaskets uithin the licits for identified Icahana set forth in technical, spec-ifications need not be reportcJ under this item.
315
Appendix A 6.6 Reporting Requirements (Cont'd)
(4) Reactivity anomalies involving disagreement with the prt iicted 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 conserva-tive than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if sub-critical, an unplanned reactivity in-sertion of more than 0.5% Ak/k; or occurrence of any unplanned criticality.
(5) Failure of malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.
(6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfi1Lment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
Note:
For items 2.a(5) and 2.a(6) reduced redundancy that does not result in a loss of system function need not be reported under this section but may be reportable under items 2.b(2) and 2.b(3) below.
(7) 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 tech-nical specifications.
(8) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
(9) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical ppecifications bases; or discovery during plant life on conditions net specifically considered in the safety analysis report or technical specifica-tions that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.
Note:
This item is intended to provide for reporting of potential generic problems.
@0) Radioactive material contained in liquid or gaseous holdup tanks in excess of that permitted by the limiting conditions of operation established in the Technical Specification.
(31) Occurrence of an event that causes a release of radioactive material from the site to unrestricted areas in excess of the limits in 10 CFR Part 20.
316
6.6 Reportine Requirements (cont'd) f le b.
Thirty Day Written Reports.
The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shal) include, as a minimum, a completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
(1)
Reactor protection system or engineered safety feature instrument settings which are found to be icss conserv-ative than those established by the technical specifica-tions but which do not prevent the fulfillment of the functions 1 requirements of affected systems.
(2)
Conditions leading to operatfor. in a degraded mode permitted by a 'liciting condition for operation or f,
plant shutdown required by a limiting condition Cor j
operation.
p Note:
Routine surveillance' testing, instrument c alib ra tior.,
or preventativ6 caintenance which require system configurations as ccscribed in itens 2.b(1) and 2.b(2) need not be reported except where test results thensclves reveal a degraded mode as described above.
.(3)
Observed inadequacies in the implementation of admin ~
istrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systcas.
(4)
Abnormal degradation of systems other than those specified in 'iten 2.a(3) above designed to contain radioactive caterial resulting from the fission process.
Note:
Scaled cources or calibration. sources are not included
'under this iten.
Leakage of valve packing or gaskets uithin the limits for identified leakage set forth in technical specifications need not be reported under this' item.
317 t
6.6 Reporting Requirements (Cont'd)
(5) An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid efflu-ents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents.
The re-port of an unplanned offsite release of radio-active material shall include the following information:
- 1. A description of the event and equipment
. involved.
- 2. Cause(s) for the unplanned release.
- 3. Actions taken to prevent recurrence.
- 4. Consequences of the unplanned release.
318
Appendix A 6.0 ADMINISTRATIVE CONTROLS 6.6.3 Unique Reporting Requirements a.
Semiannual Effluent Release Report (c) Particulate Release Within 60 days after January 1 and July 1 of (1) Total gross radio-each year a report shall be submitted covering activity (g,5) released the radioactive content of effluents released (in curies) excluding to unrestricted areas during the previous six background radioactivity.
months operation.
The data shall be in the format of RG 1.21, Rev. 1 (June 1974) and shall (2) Total gross alpha radio-be summarized on a monthly basis and shall activity released (in include as a minimum:
curies) excluding back-ground radioactivity.
- 1. Gaseous Effluents (3) Total gross radio-(a) Gross Radioactivity Releases activity released (in curies) of nuclides (1) Total gross radioactivity (in curies) with half-lives greater primarily noble and activation gases than eight days.
released.
(4) Percene of technical (2) Maximum gross radioactivity release specification limit for rate during any one-hour period.
particulate radioactivity with half-lives greater (3) Total gross radioactivity (in curies) than eight days.
by nuclide released, based on repre-sentative isotopic analyses performed.
(2) Liquid Eff]uents (4) Percent of technical specification (a) Total gross radioactivity limit.
@,1 ) released (in curies) dxcluding tritium and (b) Iodine Releases average concentration re-leased to the unrestricted (1) Total iodine radioactivity (in curies) area.
by nuclide released, based on repre-sentative isotopic analyses performed.
(2) Percent of technical specification limit for I-131 released 319
Appendix A ADMINISTRATIVE CONTROLS (b) Total tritium and total alpha (h) Percent of technical radioactivity released (in specification limit.
curies) and average concentration released to (3) Solid Radioactive Waste the unrestricted area.
(a) The total amount of solid waste packaged (in cubic (c) Total dissolved noble feet).
gas radioactivity released (in curies) and average concentration released (b) The total estimated radio-activity (in curies) to the unrestricted area.
involved.
(d) Total volume (in liters) of liquid waste released.
(c) The dates of shipment and disposition (if shipped (e) Total volume (in liters) offsite).
of dilution water used b.
Environmental Radiological Monitoring prior to release from the restricted area.
(1) Standard Radiological Monitoring (f) The maximum concentration Program of gross radioactivity (s, y) released to the (a) Non-Routine Report unrestricted area (aver-aged over the period of (i)
If a confirmed measured radionuclide con-release).
centration in an environ-(g) Total gross radioactivity mental sampling medium (in curies) by nuclide averaged over any calendar released, based on repre-quarter sampling period exceeds the reporting sentative isotopic analyses performed.
l level given in Table 3.16-1 I
and if the radioactivity is attributable to plant operation, a written report shall be submitted to the Director of the NRC Regional Office, with a copy to the Director, 320
Office of Nuclear Reactor Regulation, (i) Results of environmental sampling within 30 days from the end of the summarized quarter.
When more than one of the following the format of Regulatory radionuclides in Table 3.16-1 are de-Guide 4.8, Table 1 (December 1975);
tected in the medium, the reporting (Individual sample results will be level shall have been exceeded if:
retained at the station) ;
Ci (ii) An assessment of the monitoring results y
RL.i ~
and radiation dose via the principal pathways of exposure resulting from where C is the concentration of the plant emissions of radioactivity; in-th i
radionuclide in the medium and cluding maximum noble gas gamma and RL is the reporting level of radio-beta air doses in the unrestricted nuclide i.
area (dose calculations shall be performed in accordance with the ODCM).
(ii)
If radionuclides other than those in l
Table 3.16-1 are detected and are due (iii) results of the consus to determine the to plant effluents, a reporting level locations of animals producing milk is exceeded if the potential annual dose for human consumption.
of an individual is equal to or greater than the design objective doses of (iv) A summary of the metu rological 10 CFR 50, Appendix I.
conditions concurrent with the re-lease of gaseous effluents during (iii)
This report shall include an evaluation each quarter as outlined in of any release conditions, environmental Regulatory Guide 1,21 (Revision 1) factors, or other aspects necessary to dated June, 1974, following the explain the anomalous effect.
format of Appendix B thereof.
(b) Annual Operating Report An annual report containing the data taken in the standard radiological monitoring program l
(Table 3.16-1) shall be submitted by April 30 of the next year.
The content of the report shall include:
321
6.6.3.a Special Reports Reports on the following areas shall be as indicated:
AREA SUBMITTAL DATE a.
Primary Coolant Leakage Initial report within one year of initial criticality; 5 years after commercial service.
b.
In-Service Inspection Development First in-service inspection period after initial full power operation.
c.
In-Service Inspection Evaluation The requirements of Section 11 ASME Boiler Code (IS-620).
d.
Loose parts monitoring program On same date as in-service inspection report.
- e. Containment Building Structural Within 90 days following completion of each Testing Program test.
- f. Lake Michigan continuing Lake Within 60 days following the end of each 6 Study Projects No.
1, 6, 9, 10, month routine operating report period.
11, 12***
Each report shall contain tabular and graphical information, analysis and interpretation of information obtained pursuant to each project.
- g. Waukegan Memorial Airport Expansion The expansion plans and status of such plans Plans for the Waukegan Memorial Airport will be reported in the Annual Report including FAA #5010 report.
lh.RadioactiveEffluentReleases Per Sections 3.11 and 3.12 within 30 days of occurrence.
- Zion Nuclear Power Station Environmental Report, Section 2.3.6, Table 9.
SPECIAL REPORTS 322 Table 6.6.2
6.7 OFFSITE DOSE CALCULATION MANUAL (ODCM)
FUNCTION
- 6. 7.1 The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these Technical Specifications.
The ODCM shall contain a list of equipment contained in the Radioactive Waste Treatment System.
6. 7. 2 Any changes to the ODCM shall be made by either of the following methods:
A.
Licensee initiated changes:
1.
Shall-be submitted to the Commission by inclusion in the Monthly Operating Report pursuant to Specification 6.B.lc within 90 days of the date the change (s) was made effective and shall contain:
a.
sufficiently detailed information to support the change.
Information submitted should consist of a package of those pages of the ODCM to be changed, together with appropriate analyses or evaluations justifying the change (s) ;
b.
a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.
documentation of the fact that the change has been reviewed and found acceptable by both the onsite and offsite review functions.
2.
Shall become effective upon a date specified and agreed to by both the onsite and offsite review functions following their review and acceptance of the change (s).
B.
Commission initiated changes:
1.
Shall be determined by the onsite review function to be applicable to the facility after consideration of facility design.
323
2.
The licensee shall provide che Comission with written notification of their determination of applicability including any necessary revisions to reflect facility design.
3.
Shall be reviewed by the offsite review function.
324
i l
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EnECUTIVE VICE PRESID(NT I
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Zion Station Organization FIGURE 6.1-2
9 O
O
. Unit in Operating Mode (other than Position None 1 or 2 1&2 cold shutdown)
~
Shitt Engineer or 1
1 2
Shift Foreman 8 Nuclear Station Operator 1
2 3
Equipment Operator or 2
3 4
Equipment. Attendant Radiation Protection Man 1
1 1
TOTAL 5
7 10 MINIMUM" 5
6 9
- The minimum number refera only for the case of shift shortage, caused.by a sudden sickness or home emergency.
Notes:
1.
SRO shall be present on site at all times when there is fuel in the reactor.
2.
A itcensed man shall be in the control room at all times whenever fuel is in either reactor.
3.
Two licensed men shall be in the control room during reactor startups,' shutdowns, oporation, and other periods such as, planned control rod inanipulations.
4.
For the pericd of Unit I and Unit 2 Start up Test Program, two licensed men per unit shall be in the control room during any operation of the reactor or plant which cz.u cause chances in reactivity which by trie startup test brogram.have not been veri fied previ.>t s' -*
ZION SIIIFT MANNING CIIART Figure 6.1.3 327
h
\\
Door #
Location 1
Exterior from vestibule rocm 101 - Service Bldg.
5 Dock from storeroom #108 - Service Uldg.
10 Machine Shop #116 16 Exterior from Turbine B15g. Track door - Unit #1 20 Exterior from Service"ldg. East itall 69 Exterior from Turbine Bldg. West Wall Unit #1 70 Exterior from Turbine Bldg. East Wall between units 71 Exterior from Turbine Bldg. East Wall Unit #2 72 Exterior from Turbine Bldg. West ball Unit #2 73 Exterior fron. Turbine Bldg. West Wall Unit #2 277 Exterior to car shed - rouing steel door.
278 Exterior from car shed west wall 316 Exterior from cribhouse south wall 318 Exterior from cribhouse north wall 341 Exterior from isolation safety valve enclosure 1E north side Unit #1 342 Exterior from isolation safety valve enclosure 1E south side Unit #1 343 Exterior from isolation safety valve' enclosure 1W south side Unit #1 344 Exterior from isolation safety valve enclosure 1W j
north side Unit #1 Boundary Doors for Flood Conditions TABLE 6.3.1 I
328 l
O
.O Door #
Location 345 Exterior from isolation safety valve enclosure 2E south side 346 Exterior from isolation safety valve enclosure 2E north side 347 Exterior from isolation safety valve enclosure 2W north side 1
348 Exterior frou isolation safety vdive encisoure 2W I
sourth side i
l o
5 i
J i
l 1
I Doundary Doors for Flood Conditions I
TABLE 6.3.1 (Cont.)
l 1
329 1