ML20127P771
| ML20127P771 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 06/25/1985 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20127P767 | List: |
| References | |
| 0275K, 275K, NUDOCS 8507020513 | |
| Download: ML20127P771 (107) | |
Text
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ATTACHMENT 1 Proposed Zion Appendix A Technical Specification Changes to Sections:
3.11, 12, 14, 16 and 20 4.11, 12, 14, 16 and 20 6.1, 2, 3, 4, 5, 6, 7, and 8 Pages Changed Pages Added Pages Deleted iv xi v
vil ix x
'Section 3/4.11 222 223a 223 223b 224 227a 225 227b 225a 228a 226 228b 226a 228c
'227 229a 228 229 Section 3/4.12 230 232a 231 236a 232 236b 233 238a 234 241a 235 241b 236 242a 237 238 239 240 241 242 Section 3/4.14
-250 251a 251 251b 252 251c 253 252a 254 253a 253b 253c 253d NER'2889888885,,
0275K
f
' Pages Changed Pages Added Pages Deleted Section 3/4.16 275 275a 276 279a 277 279b 278 279c 279 280b 280 280c 280a Section 3/4.20 295a 295b Section 6.1 300
-301 302a 302 328a 303 330 304 331 305 332 306 333 307 334 Section 6.2 308 309 Sections 6.3, 310 4, 5' 311 Section 6.6 312 313 314 315 316 317 318 -
319 320 321 322 323 324 Section 6.7 325
.Section 6.8 326 327 328 329 0275K
SURVtlLLANCE LIMillNG COND_lil0N FOR OPERATION REQUIREMENT.
PAGE_
3.13 Refueling Opetations 4.13 243 3.13.1 Core Reactivity 4.13.1 243 3.13.2 Protection from Damaged Spent Fuel 4.13.2 244 3.13.3 Containment Status 4.13.3 245 3.13.4 Radiation Monitoring 4.13.4 246 3.13.5 Refueling Equipment Checkout 4.13.5 246 3.13.6 Refueling Equipment Operability 4.13.6 246 3.13.7 Spent Fuel Plt Cooling Systems 4.13.7 246 3.13.8 Fuel Inspection Program 4.13.8 247 3.13.9 Residual lleat Removal System Operation 247c 3.13.10 Water Level - Reactor Vessel 247d 3.13.11 Water level - Storage Pool 4.13.9 247a Bases 240 3.14 Plant Radiation Monitoring 4.14 250 Bases 254 3.15 Auxillary Electrical Power System 4.15 255 Bases 270 3.16 Radiological Environmental Monitoring 4.16 275 Bases 279 3.17.1 Ventilation 4.17.1 281 3.17.2 Aircraft Fire Detection 4.17.2 203 Bases 287 3.18 Steam Generator Activity 4.1R 289 Bases 290 3.19 Failed Fuel Monitoring 4.19 292 Bases 293 3.20 Radioactive Solids 4.20 295a Bases 3.21 Fire Protection 4.21 2950 Bases 2950 3.22 Shock Suppressors (Snubbers) 4.22 295W Dases 295Z 3.23 Special Test Exceptions 4.23 295AB Dases 295AC 1AOLE OF CONTENTS (Continued)
Iv 0531t/0532t 0098A
9 5.0 Design Features 296 5.1 Site 296 5.2 Reactor Coolant System 296 5.3 Reactor Core 296 4
296 5.4 Containment System 1
298 5.5 Fuel Storage 5.6 Seismic Design 299 1
i l
6.0 ADMINISTRATIVE CONTROLS 300 6.1 Organization, Review, Investigation and Audit....................
300 1
6.2 Plant Operating Procedures 308 6.3 Actions to be Taken in the event of a Reportable Event in Plant Operation 310 6.4 Action to be Taken in the Event A Safety Limit is Exceeded 310 i
]
6.5 Plant Operating Records...............................
310 6.6 Reporting Requirements 312
.i 6.7 Offsite Dose Calculation Manual (ODCM).
325 I
326 6.8 Flooding Protection i
Table of Contents (Continued) i i
V 06430 06430
Figure Page i
3.3.2-1 Reactor Coolant System Heatup Limitations 84 l
3.3.2-2 Reactor Coolant System Cooldown limitations 85 l
3.3.2-3 Effect of Fluence and Copper Content on Shift of 86 l
RTNDT for Reactor Vessel Steels Exposed to 550 degrees F t
Temperature I
3.3.2-4 Fluence at 1/4T and 3/4T as a Function of Full Power 87 Service Years 3.3.6-1 Dose Equivalent I-131 RC Limit versus Percent of Rated Thermal Power 124c t
3.4-1 High Steam Line Flow Setpoint 131a j
3.11-1 Restricted Area Boundry 226 6.1.1 Minimum Shift Crew Composition 327 i
I List of Figures (Continued) l vii 0531t/0532t 0098A
O Table Page 3.16-1 Zion Standard Radiological Environmental Monitoring Program 279 t
3.16-2 Reporting levels for Radioactivity Concentrations in Environmental 219c Samples 4.1-1 Reactor Protection System Testing and Calibration Requirements 35 4.3.B-1 Minimum Number of Steam Generators to be inspected During Inservice 741 Inspection i
4.3.D-2 Steam Generator Tube inspection 74j 4.3.6-1 Primary Coolant Specific Activity Sample and Analysis Program 124b 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 i
153 4.6-1 Containment Spray System Components Steam Generator Safety Valves, Set Pressures, Ortfice Sizes 160 4.7-1 and Steam Flows 4.7-2 Auxiliary Feedwater Puen System 161 l
4.7-3 Aux 111ary Feedwater Supply System 161a l
J 4.8-1 Centrifugal Charging Pump System 105 4.8-2 Safety injection Pump System 106 4.B-3 Residual Heat Removal Pump System 187 4.0-4 Accumulator Tanks 100 I
LIST OF TABLES 0531t/0532t 1x 0090A
Table Page 4.0-5 Component Cooling Pump System Ing i
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 200 4.11-1 Radioactive L1guld Effluent Sampling and Analysis 227 Surveillance '
t 4.11-2 Radioactive Liquid Effluent Monitoring Instrumentation 220b Surveillance 1
4.12-1 Radioactive Gaseous Effluent Sampling and Analysis Program 230 l
4.12-2 Radioactive Gaseous Effluent Monitoring Instrumentation 240 l
Surveillance j
4.14-1 Plant Radiation Monitoring Instrumentation Surveillance 253 4.15-1 Diesel Generator test. Schedule 268 4.15.la Additional D/G Reliability Action Reporting Requirements 268a
~
4.15-2 4160-Volt Engineered Safeguard Bus Main, Reserve, and 269 Standby Feeds 4.16-1 Maximum Valves for the Lower Limits of Detection (LLD) 200 2
List of Tables (Continued)
X 0531t/0532t
i lable Page 4.17-1 Charcoal Filters 284 4.17-2 HEPA Filters 285 4.19-1 Failed Fuel Monitoring Instruments 295 4.21-1 Fire Protection Instruments 295p 4.21-2 Fire Suppression Water System 295r 4.21-3 Sprinkler Systems 295s 4.21-4 CO2 Systems 295t 4.21-5 Fire Hose Stations 295u 6.6-2 Special Reports 323 6.8.1 Boundary Doors for Flooding Protection 328 List of Tables (Continued)
Xi 0531t/0532t 0098A
LIM 111NG CONDlil0N IDR OPERATION SURVEllLANCE REQUIREMENT i
3.11 Radioactive liquids 4.11 Radioactive liquids 1.
L1_ quid Effluent Concentration 1.
Liquid' Effluent Concentration A.
The concentration of radioactive lhe radioactivity content of each batch of i
material released from the site radioactive 11guld waste shall be determined (see Ilgure 3.11-1) shall be prior to release by sampling and analysis in limited to the concentrations accordance with Table 4.11-1.
The results specified in 10 CIR Part 20, of pre-release analyses shall be useil with Appendix B, lable 11, Column 2 for s
the calculational methods in the 00CM to radionuclides other than dissolved assure that the concentration at the point or entralned noble gases. For of. release is maintained within the limits dissolved or entrained noble of Specification 3.11.1. A gases, the limit is shown in table 3.11-1.
Post-release analyses of samples composited from batch releases shall be performed in 4
l accordance with Table 4.11-1.
The results l
B.
During the release of radioactive of the previous post-release analyses shall j
11guld wastes a minimum dilution be used with the calculational methods In water flow rate of 44,000 gpm 1s the 00CM to assure that the concentrations required.
at the point of release were maintained within the 11mits of Specification 3.11.1.A.
APPLICABill1Y: At All limes The radioactivity concentration of 11gulds discharged from continuous release points shall be determined by collcction and i
analysis of samples in accordance with lable 4.11-1.
The results of the analyses shall be used with the calculational methods in the 00CM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.A.
At least two service water pumps or a circulating water pump sha11 be operational on the discharge path.
0662t/0663t 222
LIMlllNG CON 0lil0N IDR OPERATION SURVEILLANCE REQUIREMENI AC110N: a)
With the concentration of radioactive material released from the site to unrestricted areas exceeding the limits specified in 3.11.1.A., immediately decrease the release rate of radioactive waterials and/or increase the dilution flow rate to restore the concentration to within the above limits.
b)
The provisions of specification 3.0.3 and 3.0.4 are not applicable.
t i
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6 i
I 0662t/0663t 223
+
llM111NG CONDillDN IDR OPERATION SURVEILLANCE REQUIREMEN1 3.11 Radioactive liquids (continued) 4.11 Radioactive Liquids (Continued) 2.
Dose 2.
Dose A.
The dose or dose commitment to a A.
Dose Calculations Cumulative dose member of the public above contributions from liquid effluents background from radioactive shall be determined by calculation at i
materials in 11guld effluents least once per month and a cumulative released from the site to summation of these total body and any unrestricted areas (see Figure organ doses shall be maintained for 3.11-1) shall be limited:
each calendar quarter.
1)
During any calendar quarter to t
less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ, and 2)
During any calendar year to less than or equal to 6 mrem to the total body and to less than or equal to 20 mrem to any organ. -
l APPLICAnilliV:
At all times I
i i
0662t/0663t 223a 9
LIMillNG CONDlil0N FOR OPERATION SURVEllLANCE REQUIREMENT 3.11.2 (Continued) 4.11.2 (Continued)
ACTION:
a)
With the calculated dose from the release of radioactive materials in liquid effluents exceeding twice the limits specified in 3.11.2.A.1 limit the subsequent releases such that the dose or dose commitment to a r.iember of the public f rom all uranium f uel cycle sources is limited to 5 25 mrem to the total body or any organ (except th}rold, which is limited to 5 75 mrem) over 12 consecutive months. Demonstrate that radiation exposures to all members of the pubile from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are less than the 40 CfR Part 190 and 40 CfR part 141 Standard, otherwise obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 141 or 190 Standard. The radiation exposure analysis shall use methods prescribed in the ODCM.
b)
The provisions of specifications 3.0.3 and 3.0.4 are not app 11 cable.
9 0662t/0663t 223b 1
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LIM 111NG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11 Radioactive liquids (Continued) 4.11 Radioactive liquids (Continued) 4.
Llquid Radwaste Treatment System 4.
Liguld Radwaste Treatment System A.
The 11guld radwaste treatment A.1. Doses ue to 11guld releases from system shall be OPERABLE.* The th site to unrestricted areas appropriate portions of the system shall be projected at least once shaII be used to reduce the per month in accordance with the radioactive materials in liquid 00CM.
wastes prior to their discharge when the projected dose due to
~
liquid effluent releases from the site to unrestricted areas-(see Figure 3.11-1) when averaged over 31 days would exceed 0.13 mrem to the total body or 0.42 mrem to any organ.
- 1he liquid radwaste system shall be considered OPERABLE if liquid waste can be held-up and/or discharged within applicable limits.
APPLICABIL11Y: At all times AC110N:
a)
With the liquid radwaste treatment system inoperable for more than 30 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, return the system to OPERABLE status and place the appropriate portions of the system in use.
b)
The provisions of specifications 3.0.3 and 3.0.4 are not applicable.
0662t/0653t 225
9 LIMITING CONDITION FOR OPFRATION SURVEILLANCE REQUIREMENT 3.11 Radioactive liquids (Continued) 4.11 Radioactive liquids (Continued) 1 l
S.
-Uutdoor Liquid Tanks 5.
-Outdoor Liquid Tanks 1he quantity of radioactive material A.
1he quantity of radioactive contained in the following tanks shall material excluding tritium and i
be limited to less than or equal to 10 noble gases contained in the l
j curies, excluding dissolved or listed tanks shall be determined entrained gases and tritium:
to be within the limit by i
analyzing a representative sample a.
Outside temporary radioactive of the tank (s) contents at least l
liquid storage tank (s) (when once per week when radioactive
?
applicable) materials are being added to the tank (s). These concentrations will be determined as prescribed APPLICABILITY: At all times in the 00CM.
AC110N:
a.
With the quantity of radioactive material in the above listed
! l tank (s) exceeding the above limit, inmediately suspend all additions of radioactive material to the tank (s) and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank (s) contents to within the i
limit.
i b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
0662t/0663t 225a i
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NUCL10E MPE(yc_1/::1)*
Kr-85m 2 A 10-4 Kr-85 5 A 10-4 Kr-81 4 A 10-5 Kr-8B 9 K 10-5 Ar-41 7 A 10-5 Xe-131m 7 A 10-4 Xe-133m 5 A 10-4 j
6 K 10-4 Ke-133 2
Re-135m 2 A 10-4 5
Ke-135 2 A 10-4
- Computed from Equation 20 of ICRP Publication 2(1959), adjusted for infinite cloud submersion in water, and R - 0.01 rem / week, density - 1.0 g/cc and Pw/Pt - 1.0.
l i
i MAXIMUM PERMISSIBLE CONCENTRAll0N OF DISSOLVED OR ENTRAINED N00LE GASES RELEASED FROM Tile Si1E j
10 UNRESIRICIED AREAS IN LIQUID EFFLUEMIS l
Table 3.11-1 226a 0662t/0663t
Minimum (a,e)
Liquid Sampling Analysis Type of Lower Limit of Release Type Frequency Frequency Activity Analysis Detection (LLD)(uci/ml)
A.
Prior lo Each Prior To Each Principal Ganuna SE-7 Release (c)
Dissolved and entrained 1E-5 Lake Discharge One Batch /M (c)
M Gases (Gansna emitters)(d)
Tank P
M Tritium IE-5 Each Batch (c)
Composite (b)
Gross Alpha 1E-7 P
Q Sr-89, Sr-90 SE-8 Each Batch (c)
Composite (b)
Principal Gamma Emitters (c)
SE-7 3
Continuous W
I-131 1E-6 During Releases Dissolved and entrained gases 1E-5 (d)
(Gamma Emitters)
I Turbine Building M
Tritium 1E-5 I b"*P Continuous (d)
Composite (b)
(7)
Gross Alpha 1E-7 Q (b)
Sr-89, Sr-90 SE-8 Continuous (d)
Composite Fe-55 1E-6 rincipal ga m em m er R-7 Waste Neutralizing Prior to each Prior to each Tank release release 1-131 1E-6 P
M Tritium
~
1E-5 Each Batch (c)
Composite (b)
Gross Alpha 1E-7 Radioactive Liquid Effluent Sampling & Analysis Surveillance Table 4.11-1 0662t/0663t 227
TABLE NOTATIDN a.
The LLD is the smallest cencentralien cf radicactiv2 material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemicJi separation):
4.66 sh LLD =
A E V 2.22 Y exp ( -A At)
Where:
LLD is the lower limit of detection as defined above in picocuries (pC1) per unit j
mass or volume, i
sb is the square root of the background co~nting rate or of the counting rate u
~
of a blank sample as appropriate (as counts per minute),
A is the number of gamma - rays emitted per disintregration for gamma-ray radio nuclide analysis ( A - 1.0) for gross alpha and tritium measurement.
l E is the counting efficiency (counts per sannnd t
i V is the sainple size (in units of mass or volume),
4 2.22 is the number of disintergrations per minute per picocurie, Y is the fractional radiochemical yield when applicable (otherwise Y = 1.0) l l
A is the radioactive decay constant for the particular radionuclide, and At is the~ elapsed time between midpoint of sample collection and time of l
l counting.
(for plant effluents, not environmental sample)
The value of sb used in the calculation of the LLD for a detection system shall be based on the 4
actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.
In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y, and a t shall be used in the calculation. The background count rate is calculated from the background counts that are determined to be within i ene FWHM (Full-Width-at-Half-Maximum) energy band about the energv of the gamma ray peak used for the quantitative analysis for that radionuclide.
4 I
i Radioactive Liquid Effluent Sampling & Analysis Surveillance i
Table 4.11-1 Continued j
0662t/0663t 227a 1
4
i TABLE NOTAil0N (Continued) a.
for certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in imich greater concentrations. Under these circumstances, it will be more appropriate to calculate the concentrations of such radionuc11 des using observed ratios with those radionuclides which are measurable.
b.
A COMP 0SliE SAMPLE 1s one in which the quantity of 11guld sampled is proportional to the quantity of l
liquid waste discharged and in which the method of samp)ing employed results in a specimen which is l
representative of the 11gulds released.
1)
To be representative of the quantities and concentrations of radioactive materials'in liquid i
effluents, all samples taken for the composite shall be throughly mixed in order for the composite sample to be representative of the effluent release.
]
2) 1he weekly and monthly proportional Composite samples are not required provided that (1) the analysis required for each of these composite samples has been run on each batch discharged, and (2) a monthly record of radionuclides discharged (isotope and quantity) 1s maintained.
c.
A BATCil RELEASE is the discharge of 11guld wastes of a discrete volume. prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.
d.
A CONTINUOUS RELEASE is the discharge of 11guld wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous release, e.
The principal gamma emitters for which the LLD specification applies exclusively are the following i
radionuclides:
Mn-54, Fe-59, 00-58, 00-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Cc-144.
1his list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLO level for that nuc11de. 1he "less than" values shall not be used in the required dose calculations.
f)
If the fire sump composite sampler is inoperable grab samples will be taken from the turbine building j
ftre sump once per shift.
Radioactive L1 quid Effluent Sampling & Analysis Surveillance Table 4.11-1 Continued 0662t/0663t 227b I
l l
Minimum Channels Applicable Instrument Operable Action #
Modes 1.
Gross Activity Monitors Providina Automatic'
. c Termination of Release.
,c-
~J A.
. Lake Discharge Tank 4
1.
OR-PR04 See Action 1 1
All 2.
OR-PR05 See Action 1 1
All B.
Turbine Bldg.
~
~
1.
OR-PR25 1'
2 All 2.
Continuous Composite Sampler A.
lurbine Building F.
. Sump 1
2 All 3.
Flow Rate,ttonitors A.
Lake Discharge Tank 1.
OF-WD63 1
3 All 2.
OF-WD67 1
3 All i
~x
^;-
~.
Radioactive Li,qujd Effluent Monitoring Instrumentation TABLE 3.11-2 0662t/0663t 228 s
TABLE NOTATION 2
ACTION 1 --
With one of the LDT monitors inoperable, all LDT releases shall be made through the OPERABLE monitored pathway.
If both monitors are inoperable, effluent releases from the tank may continue, for up to 14 l
days provided that prior to initiating the release:
1 1,
At least two independent samples of the tank's contents are analyzed, in accordance with 1
Specification 4.11.1 and 2.
At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge flow path valving; Otherwise, suspend release of radioactive effluents via this pathway.
1 ACTION 2 -
With the number of channels OPERABLE less than the minimum number required, effluent releases via this pathway may continue, provided that at least once per shif t grab samples are analyzed for gross radioactivity (beta /ganne or isotopic) at a lower limit of detection (LLD) of at least 10-7 uti/ml.
AC110N 3 -
With the number of channels OPERABLE less than the minimum number required, effluent releases via this pathway may continue, for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may be used to estimate flow.
I I,
Radioactive Liquid Effluent Monitoring Instrumentation TABLE 3.11-2 Continued 0662t/0663t 228a l
Channel Channel Source Channel Functional Instrument Check Check Calibration (1)
Test (2) 1.
Gross Activity Monitors Providing Automatic Termination Of Release A.
Lake Discharge Tank 1.
OR-PR04 P
P R
Q 2.
OR-PROS P
P R
Q B.
Turbine Bldg.
1.
OR-PR25 D3 M
R Q
2.
Continuous Composite Sampler A.
Turbine Building Fire Sump D
N/A N/A N/A 3.
Flow Rate Monitors A.
Lake Discharge Tank 1.
OF-WD63 03 N/A R*
N/A 2.
OF-WD67 D3 N/A R*
N/A 1
Radioactive Liquid Effluent Monitoring Instrumentation Surveillance TABLE 4.11-2 0662t/0663t 228b
O TABLE NOTATION (1) CHANNEL CAllBRATION shall include performance of a CHANNEL FUNCT10NAL TEST and a source check.
(2) 1he CHANNEL fUNC110NAL TEST shall also demonstrate that any automatic isolation of this pathway occurs and that control room alarm annunciation occurs if any of the following conditions exist. (if the capability is I
installed):
a)
Instrument indicate levels above the alarm setpoints.
b)
Circuit failure.
Instrume't indicates a downscale failure.
c) n d)
Instrument controls not set in OPERATE mode.
(3) CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.
Does not include flow sensor.
Radioactive liquid Effluent Monitoring Instrumentation Surveillance TABLE 4.11-2 Continued 0662t/0663t 228c
E Bases:
3.11.1 CONCENTRAil0N j
4.11.1 1 hts specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CIR 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 result in exposures within (1) the Section ll.A design objectives of Appendix 1, 10 CfR 50, to an Individual and (2) the limits of 10 CfR i
20.106(e) to the population. The concentration limit for dissolved or entrained noble gases, HPC in air i
(submersion) was converted to an equivalent concentration in water using the methods descr W in International Commission on Radiological Protection (ICRP) Pub 11 cation 2.
3.11.2 DOSE
[
4.11.2 This specification is provided to implement the requirements of Sections II.A, lit.A and IV.A. of l
Appendix 1, 10 CfR Part 50. The limiting Condition for Operation implements the guides set forth in i
Section II.A of Appendix 1.
The AC110N statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive material in 11guld effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CfR 141. 1he i
dose calculations in the ODCM 1mplement the requirements in Section III.A of Appendix I that conformance l
with the guides of Appendix I be shown by calculational procedures based on models and data, such that j
the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the DDCM for calculating the doses due to the actual release rates of radioactive materials in 11guld effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of An'nual Doses to Man from Routine Releases of Reactor Effluents i
for the Purpose of Evaluating Compliance with 10 CfR Part 50, Appendix 1," Revision 1, October 1971 and Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix 1," Apr11 1977.
This specification applies to the release of 11guld effluents from the site. for shared radwaste treatment systems, the 11guld effluents from the shared system are proportioned among the units sharing that system.
0662t/0663t 229 0288A
o Bases:
i 3.11.3 RAD 10AcilVE LIQUID EFFLUENT INSTRUMENTAT10N 4.11.3 The radioactive liquid effluent instrument is provided to monitor and control, as applicabic, the releases of radioactive materials in liquid effluents during actual or potential releases of 11guld effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CfR l
Part 20.
The OPERABill1Y and use of this instrumentation is consistent with the requirements of General
)
Design Criterla 60, 63 and 64 of Appendix A to 10 CfR Part 50.
1 3.11.3 LlyUID WASTE 1REA1 MENT, 4.11.4 The OPERABill1Y of the liquid radwaste treatment system ensures that this system w111 he available for use whenever 11guld effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents w111 be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CIR Part 50.36a, General Design Criterton 60 of Appendix A to 10 CIR Part 50 and the design objective given in Section 11.D of Appendix I to 10 CIR Part 50.
The specified limits governing the use of appropriate portions of the liquid i
radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section li. A of Appendix 1,10 CIR Part 50, for 11guld ef fluents.
1
.i 3.11.5 001000R LIQUID TANKS i
1 4.11.5 Restricting the quantity of radioactive material contained in those outdoor tanks specified, that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have 4
tank overflows and surrounding area drains connected to the liquid radwaste treatment system, provides I
assurance that in the event of an uncontrolled release of the tanks contents, the resulting j
concentrations would be less than the limits of 10 CfR Part 20, Appendix 8. Table 11 Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.
i 1
0662t/0663t 229a 0288A
i LIM 111NG CON 0lil0N IDR OPERATION SURVEILLANCE REQUIREMENT 3.12 Gaseous Effluents 4.12 Gaseous Effluents 1.
Dose Rate 1.
Dose Rate A.
The dose rate due to radioactive materials A.
1.
The dose rate due to radioactive l
released in gaseous effluents from the materials in gaseous effluents shall site (see figure 3.11-1) shall be limited be determined to be within the l
to the following:
prescribed 11mits in accordance with the methods and procedures of the l
1.
for noble gases: Less than or equal DDCM.
to 500 mrem /yr to the total body and i
less than or equal to 3000 mrem /yr to 2.
The dose rate due to radioactive i
the skin, and materials, other than nobic gases.
in gaseous effluents shall be 2.
For lodine-131, lodine-133 and all determined to be within the radionuclides in particulate form with prescribed limits in accordance with i
half lives greater than 8 days: Less the methods and procedures of the
~
than or equal to 1500 mrem /yr.
00CM by obtaining representative samples and performing analyses in APPLICABILITY: At all times.
accordance with the sampling and analysis program, specified in lable 4.12.1.
1 ACTION:
With a release exceeding the above limits, immediately reduce the release rate to within the above limits.
I 0664t/0665t 230 l
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12 Gaseous Effluents (Continued) 4.12.
Gaseous Effluents (Continued) 2.
Dose 2.
Dose A.
Noble Gas - The air dose due to noble A.
Cumulative dose contributions for the gases released in gaseous effluents current calendar quarter and current from the site (see Figure 3.11-1) calendar year shall be determined in shall be limited to the following:
accordance with the 00CM at least once every 31 days for noble gas and 1.
During any calendar quarter: Less radioiodines, radioactive materials in than or equal to 5 mrad for gamma particulate form and radionuclides (other radiation and less than or equal than noble gas) with half-lives greater to 10 mrad for beta radiation and, than eight days.
t 2.
During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
B.
Radioiodine - Particulate - Other Than Noble Gas The dose to a member of the public from radioiodines and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released from the site (see Figure 3.11-1) shall be limited to the following:
1.
During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, 2.
During any calendar year:
Less than or equal to 15 mrem to any organ.
0664t/0665t 231 i
l
LIMlilNG CON 0lT10N FOR OPERATION SURVEILLANCE REQUIREMENT 3.12.2 Gaseous Effluents (Continued) 4.12.2 Gaseous Effluents (Continued)
APPLICABILITY: At all times AC110N:
a.
With the calculated air dose from gaseous effluents exceeding the above limits, define the corrective action (s) to be taken i
to ensure that future releases are in compliance with 3.12.2.
t b.
With the calculated air dose from radioactive noble gases in gaseous effluents exceeding twice the limits of Specification 3.12.2.A.1 1.
Limit subsequent releases such that the dose or dose i
commitment to a member of the i
public from all urantum fuel j
cycle sources is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to 75 mrem) over 12 consecutive months.
1 2.
Prepare an analysis which demonstrates that radiation exposures to all members of the public from all uranium fuel cycle sources (including i
all effluent pathways and j
direct radiation) are less I
than the 40 CfR Part 190 l
Standard.
l 0664t/0665L 232 l
l
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12.2 Gaseous Effluents (Continued) 4.12.2 Gaseous Effluents (Continued) c.
With the calculated dose from the i
release of Iodine-131 Iodine-133, 4
tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents exceeding twice the limits of Specification 3.12.2.B.1 1.
Limit subsequent releases such i
that the dose or dose commitment to a member of the public from all
?
uranium to less than or equal to 25 mrem to the total body or organ (except the thyroid which is limited to less than or equal to 75 mrem) over 12 consecutive months.
j 2.
Prepare an analysis which demonstrates that radiation exposures to all members of the public from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are i
less than the 40 CFR Part 190
)l Standard. Otherwise, request a i
variance from the Commission to I
permit releases which exceed the 1
40 CFR Part 190 Standard. The radiation exposure analysis shall i
use the methods prescribed in the l
00CM.
i d.
The provisions of Specifications 3.0.3 i
and 3.0.4 are not applicable.
0664t/0665t 232a i
LIM 111NG CDNDlil0N IDR OPERATION SURVEILLANCE REQUIREMENT 3.12 Gaseous Effluents (Continued) 4.12 Easeous Effluents ~(Continued) i 3.
Gaseous Effluent Instrumentation 3.
Gaseous Effluent instrumentation A.
The radioactive gaseous effluent A.
1.
1he setpoints shall be determined in i
monitoring instrumentation channels accordance with procedures as shown in Table 3.12-1 shall be OPERABLE described in the ODCM.
with their alarm / trip set points set in accordance with the methods prescribed 2.
Each radioactive gaseous effluent in the 00CM to ensure that the limits monitoring instrumentation channel i
of 3.12.1 are met.
shall be demonstrated OPERABLE by performance of a CilANNEL CllECK, APPLICABilliY: At all times, except as indicated on SOURCE CHECK, CllANNEL CAllBRA110N g
i lable 3.12-1 and CHANNEL IUNCTIONAL 1EST at the frequencies shown in Table 4.12-2.
Atil0N:
l
- a. With a radioactive gaseous effluent monitoring j
instrumentation channel alarm / trip setpoint j
less conservative than required by the above Specification, immediately suspend the release i
of rad 101ctive gaseous effluents monitored by the affected channel or declare the channel inoperable.
- b. With one or more radioactive gaseous effluent' monitoring instrumentation channels inoperable, take the ACTION shown in Table 3.12-1.
i
- c. The provisions of Spec 1I1' cations 3.0.3 and 3.0.4 are not app 11 cable.
i 2
0664t/0665t 233 0293A i
}
i e
LIMillNG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT I
3.12.
Gaseous Effluents (Continued) 4.12.
Gaseous Effluents (Continued) 4.
Gas Decay Tanks 4.
Gas Decay Tanks A.
1he quantity of radioactivity contained A.
The quantity of radioactive material in each gas decay tank shall be limited contained in each gas decay tank shall to 22,000 curies (considered as Xe-133).
be determined to be within the above limit at least weekly when radioactive 4
APPLICABILITY: At all times materials are being added to the tank.
ACTION:
a.
With the quantity of radioactive material in any gas decay tank exceeding the above limit,'
immediately suspend addition of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.
1 b.
The provision of Specifications 3.0.3 and 3.0.4 are not applicable.
i l
0664t/0665t 234 l
l i
LIMillNG CDNDlilDN FOR OPERATION SURVEILLANCE REQUIREMENT t
i 3.12 Gaseous Effluents (Continued) 4.12 Gaseous Effluents (Continued)
- 5. Explosive Gas Mixture 5.
Explosive Gas Mixture A.
The concentrallon of hydrogen or oxygen in the A.
The concentrations of hydrogen or oxygen j
waste gas system
- shall be limited to less in the waste gas system shall be than or equal to 3% by volume.
determined to be within limits, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l APPLICABill1Y: At all times ACTION:
J a.
With the concentration of hydrogen and oxygen in the waste gas system each greater than 3%
by voluwe but either hydrogen or oxygen less than or equal to 4% by volume, initiate action within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore the concentration of hydrogen or oxygen to within the limit.
With the concentration of hydrogen and oxygen b.
in the waste gas system each greater than 4%
by volume, initiate action within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to reduce the concentration of hydrogen or oxygen j
to less than or equal to 3%.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- The waste gas system consists of the following: Gas Decay Tank when on f111 or cover gas and the Hold-up Tanks.
I e
i 0664t/0665t 235 i
l
Minimum Channels Applicable Instrument Operable Action Modes Notes 1.
Gas Decay Tank A.
Gas Activity Monitor 1.
UR-PR10A low range gas 1
5 All 2.
OR-PR100 High range gas 1
5 All B.
Particulate / lodine Monitor 1.
OR-PRIOC 1
5 All C.
Flow Rate Monitor 1.
Of-WG03 1
9 All 2.
Air Ejector Off-Gas A.
Gas Activity Monitor 1.
1R-0015 Gas 1
6 1,2,3,4,7 4
2.
2R-0015 Gas 1
6 1,2,3,4,7 B.
Particulate / Iodine Monitor
?
1.
1R-PR26 1
6 1,2,3,4,7 2.
2R-PR26 1
6 1,2,3,4,7 C.
Flow Rate Monitor 1.
1F-0G10 1
12 1,2,3,4,7 2.
2F-0G10 1
12 1,2,3,4,7 3.
Containment Purge or Vent A.
Gas Activity Monitor 1.
1R-PR09A Gas 1
62, 7' All
' during venting 2.
2R-PR09A Gas 1
62, 7' All
' during purging 3.
1R-PR40E (Channel 5) 1 62, 78 All 4.
2R-PR40E (Channel 5) 1 62, 72 All B.
Iodine Monitor 1.
IR-PR09B Iodine 1
62, 78 All 2.
62, 78 All 3.
1R-PR40C (Channel 3) 1 62, 78 All 4.
2R-PR40C (Channel 3) 1 62, 7' All C.
Particulate Monitor l
1.
1R-PR09C Particulate 1
61, 7' All 2.
2R-PR09C Particulate 1
62, 78 All l
3.
1R-PR40A (Channel 1) 1 62, 78 All 4.
2R PR40A (Channel 1) 1 62, 7' All Radioactive Gaseous Effluent Monitor Instrumentation j
{
1able 3.12-1 0664t/0665t 236
Minimum Channels
~ Applicable Instrument Operable Action Modes 4.
Auxiliary Building Ventilation and Miscellaneous Ventilation Stack A.
Gas Activity Monitor 1.
OR-0014 or 1
6 All 2.
IR-PR25 and 2R-PR25 1
6 All 3.
OR-PR18B Gas 1
6 All 4.
1R-PR49E (Channel 5) 1 6
All 5.
2R-PR49E (Channel 5) 1 6
All B.
Iodine Monitor l
1.
1R-PR49C (Channel 3) 1 8
All
}
2.
2R-PR49C (Channel 3) 1 8
All C.
Particulate Monitor l
1.
OR-PR18A Particulate 1
6 All 1.
1R-PR49A (Channel 1) 1 8
All t
2.
2R-PR49A (Channel 1) 1 8
All i
D.
Flow Rate Monitor 1.
ILP-084 1
9 All 2.
2LP-084 1
9 All l-5.
Service Building Ventilation A.
Gas Activity Monitor 1
8 All 1.
OR-PR22 1
8 All B.
Particulate / lodine Monitor 1.
OR-PR36 1
8 All i
1 6.
Steam Generator Atmospheric Relief and Safety Valves 1.
1R-PR58 1
10 1,2,3,7 2.
2R-PR58 1
10 1,2,3,7 3.
1R-PR59 1
10 1,2,3,7 4.
2R-PR59 1
10 1,2,3,7 5.
1R-PR60 1
10 1,2,3,7 6.
2R-PR60 1
10 1,2,3,7 7.
IR-PR61 1
10 1,2,3,7 8.
2R-PR61 1
10 1,2,3,7 Radioactive Gaseous Effluent Monitor Instrumentation (Continued) t Table 3.12-1 (Continued) 0664t/0665t 236a 4
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Table Notation i
ACTION 5 - With the number of channels OPERABLE less than the minimum number required, the contents of the tank may l
be released to the environment provided that prior to initiating the release:
At least two independent samples of the tank's content are analyzed, and 1.
2.
At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge flow path valving; otherwise, suspend release of radioactive effluents via this pathway.
i AC110N 6 - With the number of channels OPERABLE less than the minimum number required, effluent releases via this
}
pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these l
samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
AC110N 7 - With the number of channels OPERABLE less than the minimum number required, and no redundant monitor OPERABLE in this flow path, immediately suspend PURGING of radioactive effluents via this pathway.
j AC110N 8 - With the number of channels OPERABLE less than the minimum number required, effluent releases via this l
pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling
]
equipment as required in Table 4.12.1.
)
ACTION 9 - With the number of OPERABLE channels less than the minimum number required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per shift while release is in
{
progress.
ACTION 10 - With the number of channels OPERABLE less than the minimum number required, restore the inoperable monitor to OPERABLE status within 30 days or establish an alternate means of monitoring the parameter.
AC110N 11 - With the number of OPERABLE channels less than the minimum number required, suspend vent and purge j
operations and close each vent and purge valve providing direct access from the containment atmosphere to j
the outside atmosphere or suspend the movement of nuclear fuel and reactor components in the vicinity of j
the reactor, refueling cavity, and transfer canal (containment side).
ACTION 12 - With the number of OPERABLE channels less than the minimum number required, effluent releases via this l
pathway may continue provided the effluent flow is being accounted for in the total plant effluent.
i l
Radioactive Gaseous Effluent Monitor Instrumentation (Continued)
TABLE NOTATION Table 3.12-1 0664t/0665t 237 l
)
(.
Minimum Gaseous Sampling Analysis Type of Lower Limit of (e)
Release Type Frequency Frequency
' Activity Analysis Detection (LLD) (pC1/cc)
A.
Gas Decay Tank Grab Sample Prior to Each Noble Gases 1E-4 Prior to Each Release (C)
Principal Gamum Emitters Release (d)
Continuous' Sample After Each Particulate IE-11 During Each Release (c)
Principal Gamma Emitters Release (d)
Tritium 1E-6 I-131 (Charcoal Sample) 1E-12 I-133 (Charcoal Sample) 1E-10 3
Sr-89 Particulate 1E-11 Composite Quarterly (c)
Gross Alpha lE-11 B.
Containment Vent Prior to Each Prior to Each Principal Gaseous Gamma 1E-4 and Purge Release (a)
Release (c)
Emitters (d)
Particulate Gamma Emitters 1E-11 (d)
Tritium 1E-6 I-131 (Charcoal) 1E-12 1-133 (Charcoal) 1E-10 Sr-89 Particulate IE-11 Composite Quarterly (c)
Sr-90 Particulate IE-11 l
Gross Alpha 1E-11 Radioactive Gaseous Effluent Sampling and Anais is Program s
lable 4.12-1 0664t/0665t 238
Minimum Gaseous Sampling Analysis Type of Lower Limit of (e)
Release Type Frequency Frequency Activity Analysis Detection (LLD) (uti/cc)
C.
Continuous Grab (b)
Monthly Principal Gaseous and IE-4 Release Points Gamma Emitters
- 1. Air Ejector for Continuous (b)
Monthly Tritium 1E-6 Both (2) Units 1-131 (Charcoal Sample) 1E-12
- 2. Au dg ent.
Continuous (b)
. Weekly 1-133 (Charcoal Sample) 1E-10
.l Units Continuous (b)
Weekly (c)
Principal Particulate IE-11
- 3. Misc. Vent 11tn.
Gamma Emitters Stack
- 4. Serv. Bldg. Vent Composite Quarterly (c)
Sr-89 Particulate 1E-11 Sr-90 Particulate 1E-11 Bo
(
lt l
- 6. Turbine Bldg.
Gross Alpha 1E-11 1
Radioactive Gaseous Effluent Sampling and Analysis Program (Continued)
Table 4.12-1 0664t/0665t 238a
Table Notation a.
Should a shutdown, startup or power change greater than 50% occur which could alter the mixture of radionuclides after sampling, another analysis shall be performed prior to release.
b.
The ratto of the sample flow rate to the sampled stream flow rate shall be known for the time period in Specification 4.12.
c.
The particulate filter (s) from this/these release points shall be saved for a quarterly composite analysts for Sr-89 and Sr-90.
d.
The principal gamma emitters for which the LLD specification appIles exclusively are the following radionuclides: Kr-87 Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions, and Mn-54, re-59, i
Co-60, Zn-65, Co-58, Mo-99, Cs-134. Cs-137, Ce-141, and Ce-144 for particulate emissions. Other peaks which are measureable and identifiable by gamma ray spectrometry,'together with the above nuclides, shall be also identified and reported when an actual analysis is performed on a sample. Nuclides which are i
below the LLD for the analyses shall not be reported as being at the LLO level for that nuclide.
i e.
The LLD is defined in Notation a. of Table 4.11-1 page 227a.
I i
Radioactive Gaseous Effluent Sampling and Analysis Program (Continued) l l
Table Notation Table 4.12-1 0664t/0665t 239 0293A
Chann21 Chann21 Source Chann21 Functicnal Instrument Check
. Check Calibration (1)
Test (2) 1.
Gas Decay Tank A.
Gas Activity Monitor 1.
OR-PR10A low range gas P
P R
Q 2.
OR-PR10B High range gas P
P R
Q B.
Particulate / Iodine Monitor 1.
OR-PR10C P
P R
Q C.
Flow Rate Monitor 1.
OF-WG03 P
N/A N/A Q (5) 2.
Air Ejector Off-Gas A.
Gas Activity Monitor 1.
1R-0015 Gas D
t M
R Q
2.
2R-0015 Gas D
M R
Q B.
Particulate / Iodine Monitor 1.
1R-PR26 O
M R
Q 2.
2R-PR26 D
M R
Q C.
Flow Rate Monitor 1.
1F-0G10 0
N/A R
N/A 2.
2F-0G10 D
N/A R
N/A 3.
Containment Purge or Vent A.
Gas Activity Monitor 1.
1R-PR09A D
M R
Q 2.
2R-PR09A D
M R
Q 3.
1R-PR40E (Channel 5)
D M
R Q
4.
2R-PR40E (Channel 5)
D M
R Q
B.
Iodine Monitor 1.
1R-PR09B D
M R
Q 2.
2R-PR098 D
M R
Q 3.
1R-PR40C (Channel 3)
D M
R Q
4.
2R-PR40C (Channel 3)
D M
R Q
C.
Particulate Monitor 1.
IR-PR09C D
M R
Q 2.
2R-PR09C D
M R
Q 3.
1R-PR40A (Channel 1)
D M
R Q
4.
2R-PR40A (Channel 1)
D M
R Q
Radioactive Gaseous Effluent Monitor Instrumentation Surveillance Table 4.12-2 9
0664t/0665t 240
Channel.
Channel Source Channel Functicnal Instrument Check Check Calibration (1)
Test (2) 4.
Auxiliary Building Ventilation and Miscellaneous Ventilation Stack A.
Gas Activity Monitor 1.
OR-0014 Gas or D
M R
Q 2.
M R
Q 3.
OR-PR18B D
M R
Q 4.
IR-PR49E (Channel 5)
D M
R Q
5.
2R-PR49E (Channel 5)
D M
R Q
B.
lodine Monitor 1.
1R-PR49C (Channel 3)
D M
R Q
2.
2R-PR49C (Channel 3)
D M
R Q
C.
Particulate Monitor 1.
OR-PR18A D
M R
Q 2.
1R-PR49A (Channel 1)
D M
R Q
3.
2R-PR49A (Channel 1)
D M
R Q
D.
Flow Rate Monitor 1.
1LP-084 0
N/A R
Q 2.
2LP-084 D
N/A R
Q 5.
Service Building ventilation A.
Gas Activity Monitor 1.
OR-PR22 D
M R
Q B.
Particulate / Iodine Monitor 1.
OR-PR36 N/A N/A N/A N/A 6.
Steam Generator, Atmospheric Relief and Safety Valves 1.
1R-PR58 D
M R
Q 2.
2R-PR58 0
M R
Q 3.
1R-PR59 D
M R
Q 4.
2R-PR59 D
M R
Q 5.
IR-PR60 D
M R
Q 6.
2R-PR60 D
M R
Q 7.
1R-PR61 D
M R
Q 8.
2R-PR61 D
M R
Q Radioactive Gaseous Effluent Monitor Instrumentation Surveillance (Continued)
Table 4.12-2 0664t/0665t 241
Channel Channel Source Channel functional Jnstrumeg Check Check talibration (1)
_lest (2) 1.
Accident Monitoring A.
Containment 1.
IR-PR40G (Channel 1)
N/A N/A R
Q 2.
2R-PR40G (Channel 1)
N/A N/A R
Q 3.
IR-PR401 (Channel 9)
N/A N/A R
Q 4.
2R-PR401 (Channel 9)
N/A N/A R
Q B.
Miscellaneous Vent Stack 1.
IR-PR49G (Channel 1)
N/A N/A R
Q 2.
2R-PR49G (Channel 7)
N/A s
N/A R
Q 3.
1R-PR491 (Channel 9)
N/A N/A R
Q 4.
2R-PR491 (Channel 9)
N/A N/A R
Q C.
Fuel llandling Area 1.
IR-AR04A D
M(3) g g((4) i 2.
1R-AR04B D
M(3)
R Q(4) 3.
2R-AR04A D
M(3) g 4) g(4) 4.
2R-AR04B D
M(3)
R Q
I Radioactive Gaseous Effluent Monitor Instrumentation Surve111ance (Continued)
Tabic 4.12-2 (Continued) 0664t/0665t 241a
i Table Notation (1) CHANNEL CALIBRATION shall include performance of a CHANNEL FUNCTIDNAL TEST.
(2) The CHANNEL FUNC110NAL TEST shall also demonstrate that any automatic isolation occurs; and that Control Room alarm annunciation occurs if any of the following conditions exist (if the capability is installed):
a)
Instrument indicates measured levels above the alarm setpoint.
b) Circuit failure.
I c)
Instrument indicates a downscale failure.
d)
Instrument controls not set in " operate" mode.
(3) Daily when purging the containment during fuel handling operations.
(4) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to commencing refueling operations l
(5) Operability test only.
1 i
1 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance (Continued)
TABLE NOTATION Table 4.12-2 0664t/0665t 241b
1 I
]
Basts:
3.12 Specification 3.12.1.A is provided to ensure guides set forth in Section IV.A of Appendix I that the dose at the unrestricted area to assure that the releases of radioactive j
4.12 boundary from gaseous effluents from all units material in gaseous effluents will be kept "As i
on the site will be within the annual dose Low As Is Rea'sonably Achievable". 'The j
limits of 10 CFR Part 20.
The annual dose Surveillance Requirements implement the f
j limits are the doses associated with the requirements in Section III.A of Appendix I that concentrations of 10 CFR Part 20, Appendix B, conformance with the guides of Appendix I is to
{
Table II Column 1.
These limits provide to be shown by calculation procedures based on reasonable assurance that radioactive material models and data'such that the actual exposure of discharged in gaseous effluents will not result an individual through the appropriate pathways in the exposure of an individual in an 1s unlikely to be substantially underestimated.
i unrestricted area, to annual average l
concentrations exceeding the limits specified in Section 3.12.2.B/4.12.7 implements the j
Appendix B. Table II of 10 CFR Part 20 '(10 CFR requirements of Sections II.C. III.A and IV.A of Part 20, 106 (b)(1)). The specified release Appendix 1, 10 CFR Part 50.
The Limiting rate limits restrict, at all times, the Conditions for Operation are the guides set corresponding gama and beta dose rates above forth in Section II.C of Appendix 1.
The ACTION l
j backgraound to an individual at or beyond the statements provide the required operating i
unrestricted boundary to 500 mrem / year to the flexibility and at the same time implement the j
total body or to less than or equal to 3000 guides set forth in Section IV.A of Appendix I j
j mrem / year to the skin. These release rate to assure that the releases of radioactive r
l 11mits also restrict, at all times, the materials in gaseous effluents will be kept "as corresponding thyroid dose rate above background low as is reasonably achievable". The ODCM to a child via the inhalation pathway to less calculation methods specified in the than or equal to 1500 mres/ year. For purposes surveillance requirements implement the of calculating dose resulting from airborne requirements in Section III.A of Appendix I that i
releases the two stacks are considered a mixed conform with the guides of Appendix I is to be mode release.
shown by calculational procedures based on models and data such that the actual exposure of l
)
Specification 3.12.2.A implements the an individual through appropriate pathways is j
requirements of Sections II.8, III.A and IV.A of unlikely to be substantially underestimated.
J Appendix 1, 10 CFR Part 50.
The Limiting These equations also provide for determining the Condition for Operation implements the guides actual doses based upon the historical average set forth in Section 11.8 of Appendix 1.
The atmospheric conditions. The release rate j
ACTION statements provide the required operating specifications for radioiodines, radioactive I
flexibility and at the same time implement the material in particulate form and radiolodines j
5 l
0664t/0665t 242 i
4
Basts: (Continued) other than noble gases are dependent on the existing radionucIlde pathways to man, in the unrestricted area. The pathways which are i
examined in the development of these calculations are: 1) Individual inhalation of alrborne radionuclides, 2) deposition or radlonuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals grare with consumption of the milk and meat by man.
s The radioactive gaseous effluent instrumentation is provided to monitor, record and control, as applicable, the releases of radioactive i
materials in gaseous effluents during actual or 4
potential releases. The alarm / trip setpoints i
for these instruments shall be calculated in
]
accordance,with the 00CM to ensure that the j
alarm / trip will occur prior to exceeding the 1
limits of 10 CfR Part 20.
J Restricting the quantity of radioactivity j
contained in each gas decay tank provides i
assurance that in the event of an uncontrolled
]
release of the tanks contents, the resulting 3
total body exposure to an individual at the nearest unrestricted area boundary will not i
exceed 0.5 rem.
J l
i r
i i
l 0664t/0665t 242a r
i
~r
LIMillNG CDNDITIDN FDR OPERAllDN SURVEILLANCE REQUIREMENT 3.14 Plant Radiation Monitoring 4.14 Plant Radiation Monitoring 1.
Radiation Monitoring Instrumentation 1.
Radiation Monitoring Instrumentation A.
The radiation sonttoring A.
Each radiation monitoring x.
instrumentations shown in table 3.14 1 instrumentation channel shall be shall be OPERABLE.,
demonstrated OPERABLE by the n
' performance of the CilANNEL CllECK, CHANNEL CAllBRATIDN and CHANNEL FUNCTIONAL TEST operations for the s
APPLICABilliVIAs indicated on table 3.14.1 app 11 cable MODES at the frequencies shown in Table 4.14-1.
ACilDNt a.
With one or more radiation monitoring channels inoperable, take the ACTIDN a
shown in Table 3.14-1.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
y w
G S
070lt/Olont 250
s Mlaimum
/
Channels Applicable Instrument DRerable Action #
Modes l
1.
Area Monitors i
A.
Fuel Storage Pool Area i
1.
OR-0005 1
24 All 2.
OR-AR03 1
21 During Fuel llandling
]
Operations or Crane Operation in or near SFP.
3.
OR-AR13 1
21 During Fuel llandling Operations B.
Containment Purge Isolation 1.
1R-AR04A 1
22 6
When purging during 2.
IR-AR04B 1
22 6
fuel handling 3.
2R-AR04A 1
22 6
operations 4.
2R-AR04B 1
22 6
C.
Containment Area (Illgh Range) 1.
IR-AR02 1
30 1,2,3,4,7 I
2.
2R-AR02 1
30 1,2,3,4,7 3.
IR-AR03 1
30 1,2,3,4,7 4.
2R-AR03 1
30 1,2,3,4,7 0.
Control Room 1.
OR-0001 1
24 All E.
Portable Area Monitor 1
24 All l
RADIA110N MONiiORING INSTRUMENTATION TABLE 3.14-1 0107t/0708t 251 I
1
Minimum Channels Appilcable Instrument operable Action #
Modes F.
Auxillary Building Area 1.
OR-AR04 1
24 All 2.
OR-AR08 -
1 24 All 3.
OR-AR09 1
24 All 4.
DR ARIO 1
24 All 5.
OR-AR11 '.
1 24 All 6.
OR-0006 1
24 All 2.
Process Monitors t
A.
Containenent
- 1., Reacter Leak Detection e
a.
IR-PR12A 1
28 1,2,3,7
" - ^
b.
1R-PR128 1
28 1,2,3,7 c.
2R-PR12A 1
28 1,2,3,7 d.
2RIPR128 1
28 1,2,3,7}
s 2.
Ventilation q-a.
IR-PR40A (Channel 1) 1 25 1,2,3,7 b.
1R-PR40C (Channel'3) 1 25 1,2,3,7 c.
IR-PR40E (Channel 5) 1
~'t 25 1,2,3,7 d.
2R-PR40A (Channel 1) 1 25 1,2,3,7 e.
2R-PR40C (Channel 3) 1 25 1,2,3,7
^
f.
2R-PR40E (Channel 5) 1 25 1,2,3,7 B.
Component Cooling 26 v:
All 1.
-OR-PR07 1
2.
IR-0017 1
26 All RADIAil0N MONITORING INSTRUMENTATION 1ABLE 3.14-1 (Continued) 0107t/0iont 251a
1 Mlatmum Channels Applicable Ins trument_
Deerable Action #
Modes C.
Pipe Chase i
1.
1R-PR07A 1
20 1,2,3,4,7 2.
IR-PROTB 1
20 1,2,3,4,7 3.
2R-PR07A 1
20 1,2,3,4,7 4.
2R-PR07B 1
20 1,2,3,4,7 D.
Failed Fuel 1.
30 1,2,7 2.
30 1,2,7 s
E.
UR-PR06 1
27 All 2.
OR-PR08 1
27 1,2,3,4,7 3.
OR-PR09 1
27 1,2,3,4,7 4.
IR-PR08 1
27 1,2,3,4,7 5.
2R-PR08 1
27 1,2,3,4,7 F.
Steam Generator Blowdown 1.
1R-019 1
26 1,2,3,4,7 2.
2R-019 1
26 1,2,3,4,7 G.
Gas Monitors l
1.
1R-PR15 1
26 1,2,3,4,7 2.
2R-PR15 1
26 1,2,3,4,7 3.
OR-PR02 1
26 1,2,3,4,7 RADIAT10M MON 110 RING INSTRUMENTAilDN 1ABLE 3.14-1 (Continued) 070ft/Olost 251b
Mlaimum Channels Applicable Instrument Operable Ac_ tion #
Nodes it.
Control Room 1.
DR-PR29A (Channel 1) 1 23 1,2,3,4,7 2.
UR-PR29C (Channel 3) 1 23 1,2,3,4,7 3.
OR-PR29E (Channel 5) 1 23 1,2,3,4,7 4.
OR-PR29G (Channel 7) 1 23 1,2,3,4,7 1.
OR-PR30A (Channel 1) 1
\\
23 1,2,3,4,7 2.
OR-PR30C (Channel 3) 1 23 1,2,3,4,7 3.
OR-PR30E (Channel 5) 1 23 1,2,3,4,7 4.
OR-PR30G (Channel 7) 1 23 1,2,3,4,7 i
I i
RADIATION MON 110 RING INSTRUMENTATION TABLE 3.14-1 (Continued)
(
010Tt/0iO8t 251c l
l
Action 20:
With the number of channels DPERABLE lets than the minimum number required, verify that the i
pipe tunnel and fuel building exhaust ventilation systems are diverted through the charcoal filters.
J l
Action 21:
With the number of channels OPERABLE less than the minimum number required, stop all movement of fuel within the spent fuel pool and crane operation with loads over the spent fuel pool.
r a
l Action 22:
With the number of channels OPERABLE less than the minimum number required, stop all movement j
of nuclear fuel and reactor components in the vicinity of the reactor, refueling cavity, and transfer canal (containment side) ori suspend vent and purge operations and close each vent and purge valve providing direct access from the containment atmosphere to the outside atmosphere.
Action 23:
With the number of OPERABLt channels less than the minimum number required, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> i
initiate and maintain the ventilation system in the recirculation mode of operation.
l Action 24:
With the number of OPERABLE channels less than the minimum number required, 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 />.
Action 25:
With the number of OPERABLE channels less than the minimum number required, restore the inoperable channel to OPERABLE status within 14 days or conduct a station review to determine a plan of action to restore the channel to operability.
Action 26:
With the number of OPERABLE channels less than the minimum number required, perform a grab sample analysis at least once per shift. If the inoperable channel is not returned to OPERABLE status within 30 days, conduct a Station Review to determine a plan of action to '
restore the channel to operability.
TABLE NOTAllDN 1ABLE 3.14-1 (Continued) 0107t/0iO81 252 i
I
ro s
i h
e e
t t
h h
s n
t t
r e
e a
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v h
i sL e h
no e
t viBr t
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t)Rt w
e2 h
r ncEs t7 t
o eiP e y
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upOr l
i n f
s l o p
ti o
e ft oo m
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f ott o
nt s
es c
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o idn w
a t
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drni d
d,
n eort e
e) 7 o
r uc r
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i(
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a ea o
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rgt r
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n besp b
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(
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4
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l s o
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1
(
2 ofi a
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Erit E
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n 0
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Lg a
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1 B
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.t Ae S
A A.
Af l
t A
1 Rh y R
Ro b
i.
1 Et aa E3 E
a ws 0
3 P
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P) r u
N Od t
O3 Oe e
wt E
erc l
p ea E
L sd eu s4 sb o
it L
B l i pd l
l i n
vs B
A ev n
es es i
e A
1 noeo nt na RE 1
nrcc.
nh ne e
L apn y
ae af h
nB*
h ost hm h
t oA ce yi ce cf iR ut al
- r i
e tE f nsdi fi f(
r aP oia b
ou o
o tO t e0a q
d t
S rnl 3r re ro s
o eo e
er eh e
at bct np b
bt r
m aio me me tl uy h
uc um r
ce nadt o nn n
e un meit a
e h
dn e
nw el et t
na h yi l
hl ha i
oh t amse ti t n E
Cc wrun e
r:
hh et n h v he ttt aa t r ttd i aeth iu il n
)
)
W pd s c Ws Waa 1
2 7
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2 2
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A A
0 0
f 0
/
t 7
0 l
O
Channel.
Channel-Source Channel Functional Instrument Check Check Calibration (1)
Test (2) 1.
Area Monitor _s A.
Fuel Storage Pool Area 1)
OR-0005 D
M3 R
2)
OR-AR03 D
M3 R
Q 3)
OR-AR13 D
M3 R
Q B.
Containment Purge isolation 1) 1R-AR04A D
s M4 R
QS 2) 1R-AR048 0
M4 R
QS 4
3) 2R-AR04A D
M4 R
QS 4) 2R-AR04B D
M4 R
QS C.
ContainmentArea(HiginRange) 1)
1R-AR02 D
M R
Q 2) 2R-AR02 D
M R
Q 3) 1R-AR03 D
M R
Q 4) 2R-AR03 D
M R
Q D.
Control Room 1)
OR-0001 0
M R
Q E.
Portable Area Monitor D
M R
Q l
PLANT RA01A110N MON 110 RING INS 1RUMENTATION SURVEILLANCE REQUIREMENTS l
TABLE 4.14-1 0707t/0iO0t 253 i
I
^
Channel Channel Source Channel functional instrument Check Check Calibration (1)
Test
(_2_),
F.
Auxiliary Bu11 ding Area 1)
OR-AR04 D
M R
Q 2)
OR-AR08 D
M R
Q 3)
OR-AR09 D
M R
Q 4)
OR-ARIO D
M R
Q 5)
OR-AR11 D
M R
Q 6)
OR-0006 0
M R
Q 2.
Process Monitors A.
Containment 1.
Reactor leak Detection a) 1R-PR12A D
N/A R
Q b) 1R-PR128 D
N/A R
Q c) 2R-PR12A D
N/A R
Q d) 2R-PR12B D
N/A R
Q 2.
Vent 11ation a) 1R-PR40A (Channel 1)
D M
R Q
b) 1R-PR40C (Channel 3)
D M
R Q
c) 1R-PR40E (Channel 5)
D M
R Q
d) 2R-PR40A (Channel 1)
D M
R Q
e) 2R-PR400 (Channel 3)
D M
R Q
f) 2R-PR40E (Channel 5)
D M
R Q
D.
Component Cooling 1.
OR-PR07 D
M R
Q 2.
IR-0017 D
M R
Q Plant Radiation Monitoring Instrumentation Surveillance Requirements (Continued) 1ABLE 4.14-1 (Continued)
Ol07t/0708t 253a
Channal thannel Source Channel Functional Instrument Check Check Calibration (1)
Test (2).
C.
Pipe Chase 1.
IR-PR07A D
M R
Q 2.
IR-PR07B D
M R
Q 3.
2R-PR07A D
M R
Q 4.
2R-PR07B D
M R
Q D.
Failed Fuel 1.
IR-PR18 D
N/A R
Q OR 2.
IR-PR27 D
N/A R
Q 3.
2R-PRIB D
N/A R
Q or 4.
2R-PR27 D
N/A R
Q E.
OR-PR06 D
M R
Q 2.
OR-PR08 D
M R
Q 3.
OR-PR09 D
M R
Q 4.
IR-PRO 8 D
M R
Q 5.
2R-PR08 D
M R
Q F.
Steam Generator D10wdown 1.)
1R-019 0
M R
Q 2.)
2R-019 D
M R
Q Plant Radiation Honitoring Instrumentation Surveillance Requirement (Continued)
Table 4.14-1 (Continued)
Dio7t/0f08t 253b
Chinn21 Channel Source Channel Functional Instrument Check Check _
Calibration (1)
Test L21 G.
Ces Monitors 1.
IR-PRIS D
M R
Q 2.
2R-PR15 0
M R
Q 3.
OR-PR02 D
M R
Q H.
Control Room 1.
OR-PR29A (Channel 1) 9 M
R Q
2.
OR-PR29C (Channel 3)
D M
R Q
3.
OR-PR29E (Channel 5)
O M
R Q
4.
DR-PR29G (Channel 1)
N/A N/A R
Q l.
OR-PR30A (Channel 1)
D M
R Q
2.
OR-PR300 (Channel 3)
D M
R Q
3.
OR-PR30E (Channel 5)
D M
R Q
4.
DR-PR30G (Channel 1)
N/A N/A R
Q l
9 Plant Radiation Monitoring Instrumentation Surveillance Requirements (Continued)
Table 4.14-1 (Continued)
Ol07t/Oloot 253c
NoiE:
1.) CHANNEL CAllDRATION shall include performance of a CHANNEL FUNCilDNAL 1EST.
2.)
1he CllANNEL IUNC110NAL TEST shall also demonstrate that, any automatic isolation occurs and that local and remote annunciation (if installed) occurs, if any of the following conditions exist:
a)
Instrument indicates measured levels greater than the alarm setpoint.
b)
Circuit failure c)
Instrument indicates a downstale failure-d)
Instrument controls not set in the " operate" mode.
3.)
Daily during fuel handling operations or load handling operations in or near the spent fuel 4
pool.
I 4.)
Dally when purging the containment during fuel handling operations.
d 5.) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to commencing refueling operations.
t I
l Plant Radiation Monitoring Instrumentation Surve111ance Requirements (Continued)
TABLE NOTATION i
Table 4 14-1 (Continued) 0707t/0iO0t 253d I
BASES:
3.14 and 4.14 1he OPERABit.lTY of the plant radiation monitoring thannels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. (1)
(1) ISAR Section 11.3 e
i l
G 0707t/0700t 254
LIM 111NG COND1110N FOR OPERATION SURVEILLANCE REQUIREMENT 3.16 RADIOLOGICAL ENVIRONMENTAL MONITORING 4.16 ENVIRONMENTAL RADI0 ACTIVITY MONITORING 1.
MONITORING PROGRAM 1.
MONITORING PROGRAM The radiological environmental monitoring The radiological environmental program shall be conducted as specified in monitoring samples shall be collected Table 3.16-1.
from the locations specified in the ODCM and analyzed pursuant to Table 3.16-1 and the detection capabilities APPLICABILITY: At all times required by Table 4.16.1.
The results of analyses performed on ACTION:
the radiological environmental With th radiological e ronmen'tal monitoring samples shall be summarized monitor $ngprogramnotSengconductedas in the Annual Radiological a.
specified in Table 3.16-1, prepare and Environmental Operating Report. See submit to the Commission, in the Annual Technical Specification 6.6.3.B.l.b.
Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, contractor omission which is corrected as soon as discovered, malfunction of sampling equipment, or if a person who participates in the program by providing samples goes out of business, i
I 0693t/0694t 275
O LIMillNG CONDlil0N FOR OPERATION SURVElllANCE REQUIREMENT 3.16.1 Monitoring Program (Continued) 4.16 If the equipment malfunctions, corrective actions shall be cosapleted as soon as practical.
If a person supplying samples i
goes out of business, a replacement w111 be found as soon as possible. All deviations from the sampling schedule shall be described in the Annual Radiological Operating Report.
b.
With the level of radioactivity in an 1
i environmental sampling medtwa at one or were of the locations specified in the DDCM exceeding the limits of Table 3.16-2, when averaged over any calendar quarter, prepare and submit to the Commission t
within 30 days from the end of the affected calendar quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 3.16-2, to be exceeded. This report is not required if the measured j
+
level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
l l
l l
OG93t/0694t 275a i
+
l
i LIMlilMG CDNDlilDN fDR DPERAIlDN SURVEILLANCE REQUIREMENT 3.16.1 Monitoring Program (Continued) 4.16 c.
With milk samples unavailable from any of the sample locations required by Figure 3.16-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.
The locations from which samples were unavailable may 1
then be deleted from the program.
Identify the cause of the unava11ab111ty of samples and Identify the new location (s) for obtaining replacement samples in the Annual Radiological Environmental Operating report and also include in the report a revised figure (s) and table for the DDCM reflecting the new i
location (s).
d.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
i e
0693t/0694t 276
LIMlilNG CON 0lilDN IDR OPERATIDM SURVEILLANCE REQUIREMENT 3.16.
RADIOLOGICAL ENVIRDNMENTAL MONITORING 4.16.
RADIOLOGICAL ENVIRONMENTAL MONITORING 2.
LAND USE CENSUS 2.
LAND USE CENSUS A land use census shall be conducted to The land use census shall be conducted identify the location of the nearest at'least once per 12 months between the residences and of animals producing slik dates of June 1 and October 1, by a for human consumption in each of the door-to-door survey, road survey, following meteorological sectors, A, J K,
aerial survey, or by consulting local L, M, N, P, Q and R within a distance of 5 agriculture authorttles.
miles.
The results of the land use census shall be included in the Annual Radiological Environmental Operating APPLICABillTY: At all times Report.
ACTION:
a.
With a land use census identifying a i
location which yields an 00CM calculated dose or dose commitment greater than the values currently being calculated in Specification 3.12.2.B.1, this new location shall be added to the Radiological Environmental Monitoring Program within 30 days. The sampling l
location excluding the Control Station Location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this l
monitoring program after October 31 of the year in which this land use census was
]
conducted, b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
0693t/0694t 277
LIMilING CONDill0N FOR OPERATION SURVElllANCE REQUIREMENT 3.16.
RADIOLOGICAL ENVIRONMENTAL 4.16.
RADIOLOGICAL ENVIRONMENTAL MONITORING (Continued)
MONITORING (Continued) 3.
Interlaboratory Comparison Program 3.
Interlaboratory Comparison Program Analyses shall be performed on radioactive A summary of the results obtained as materials supplied as part of an part of the above required Interlaboratory Comparison Program which Interlaboratory Comparison Program and has been approved by the Commission.
In accordance with the 00CM shall be included in the Annual Radiological Environmental Operating Report.
APPLICABilliY: At all times.
ACTION:
a.
With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
b.
The provisions of specifications 3.0.3 and 3.0.4 are not applicable.
+
I i
l 0693t/0694t 278
r-i Sample Media
.Collectirn Sit?
Type af Analysis Frequency 1.
Airbor_ne (A) Unsite and near field a) F11ter-gross beta 2 a) Weekly (1) Unsite Station #1 b) Charcoal-1-131 b) 01-Weeklyl (2) Onsite Station #2 (3) Onsite Station #3 c) Sampling Train -
c) Weekly Test and Maintenance (B) Offsite - Far Field 10 locations a) F11ter Exchange a) Weekly b) Charcoal Exchange b) 01-Weeklyl s c) Sampling Train -
c) Weekly Test and Maintenance i
2.
Direct Radiation (TLD)
Forty locations (minimum of Gamma Radiation Dose Quarterly i
two ILD's per packet)
Zion Standard Radiological Environmental Monitoring Program lable 3.16-1 279 0693t/0694t I
Exposure Pathway and/or Sample Media Collection Site Type of Analysis
[requency 3.
Waterborne A.
Public Water Supply 6 Locations a)
Ganne Isotopic a)
Monthly Analysis of weekly composites b)
Tritium b)
Quarterly Composite B.
Cooling Water (1) Inlet a)
Gross Beta a)
Weekly Sample (2) Discharge b)
Tritium b)
Quarterly Composite C.
Sediment Lake Michigan Shoreline 1 Location Gamma Isotopic Semi-Annually 4.
Ingestion A.
Milk 2 Dairy Farms 1131 and gamma isotopic Semi monthly -
l May to Oct, Monthly at all other times.
i B.
Fish Lake Michigan Near Gamma Isotopic on edible Semi Annually Zion Station portions Zion Standard Radiological Environmental Monitoring Program (Continued)
Table 3.16-1 0693t/0694t 279a s
t I
Exposure Pathv'y and/or Sample Media Collection Site Type of Analysis Frequency 4.
Ingestion (continued)
C.
Food Products Samples of three different kinds Gaune isotopic and 1-131 Monthly when of broad leaf vegetation grown
- analysis, available, nearest each of two different off-site locations of highest predicted annual average ground-level D/Q if milk sampling is not performed.
One sample of each of the sim'111ar Gamma isotopic and 1-131 Monthly when broad lead vegetation grown 15 to analysis.
available.
30 km distant in.the least prevalent wind direction if milk sampling is not performed.
Footnotes:
Bi-Weekly shall mean at the f requency of once every other week.
1.
2.
A gamma isotopic analysis shall be performed whenever the gross beta concentration in a sample exceeds by five times (5x) the average concentration of the preceeding calendar quarter for the sample location.
Zion Standard Radiological Environmental Monitoring Program Tabic 3.16-1 (Continued) 0693t/0694t 279b l
i
i REPORTING LEVELS Water AirborneParticujate Fish Milk Food Products Analysts (pCl/1) or Gases (pC1/m )
(pC1/Kg, Wet)
(pC1/1)
(pC1/Kg, wet) 4(
H-3 2 x 10 54Mn 1 x 103 3 x 104 59 e 4 x 102 1 x 104 F
50C0 1 x 103 3 x 104 60C0 3 x 102 1 x 104 65 n 3 x 102 2 'x 104 2
II 95 r-Nb 4 x 102 Z
131 1 2
0.9 3
1 x 102 134Cs 30 10 1 x 103 60 1 x 103 137 s 50 20 2 x 103 70 2 x 103 0
140Ba-La 2 x 102 3 x 102 (a) for drinking water samples. This is 40 CFR Part 141 value.
(b) Total for parent and daughter Reporting Levels for Radioactivity Concentrations in Environmental Samples Table 3.16-2 0693t/0694t 279c
)y r
d.
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I a.
The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95%
probability with only 5% probability of falsely concluding.its presence.
For a particular measurement system (which may include radiochemical separation):
LLD =
4.66 sh l
AEV 2.22
- Y
- esp (-AAt) where LLO is the lower limit of detection as defined above (as pCi per unit mass or volume) i sb is the square root of the background count or of the count of a blank sample as appropriate (as counts per minute)
A is the number of gamma-rays emitted per disintegration for gamma-ray radio-nuclide analysis j
(A = 1.0 for gross alpha and tritium measurements)
E is the counting efficiency (counts per ganma)
V is the sample size (in units of mass or volume) i 2.22 is the number of transformations per minute per picocurie Y is the f ractional radiochemical yield when applicable (otherwise Y = 1.0)
A is the radioactive decay constant for the particular radionuclide, and 4
at is the elapsed time between sample collection and analysis.
i l Table Notation Table 4.16-1 (Continued) 0693t/0694t 280a
a.
(continued)
The value of sb used in the calculation of the LLD for a d>tection system shall be based on the actual observed background count or of the count of the blank samples (as appropriate) rather than on an unverifled theoretically predicted value. lypical values of E, Y, Y, At, shall be used in the calculation.
for gamma-ray radionuclide analyses the background counts are determined from the total counts in the thannels which are within plus or minus one IWHM (full Width at Half Maximum) of the gamme-ray photopeak energy normally used for the quantative analysis for that radlonuclide. Typical values of the IWilM shall be used in the calculation, b.
1he 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 posterior 1 (after the fact) limit for a particular measurement.
c.
LLD for drinking water.
i lable Notation (Continued) lable 4.16-1 (continued) 0693t/0694L 200b
Basts:
3.16/4.16 1he radiological monitoring program required by Table 3.16-1 provides for measurements of radiation and of radioactive materials in those exposure pathways and for those radionuc11 des, 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 measureable j
concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modellr.g of the environmental exposure pathways. Changes to the inttlally specified monitoring program may be initiated based on operational expertence.
l The detection capab111tles 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, allk and other food products correspond to approximately one-quarter of the Appendix ! to 10 CfR Part 50 design t
objective dose-equivalent of 15 mrem / year for atmospheric releases and 10 mres/ year for 11guld 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 11ters/ year.
The Land Use 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.8.3 of Appendix I to 10 CfR Part 50.
The requirement for participation in the Interlaboratory Comparison (crosscheck) Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
0693t/0694t 290c
LIMITING CON 01110N FOR OPERATION SURVE1LLANCE REQUIREMENT 3.20 Radioattive Solids 4.20 Radioactive Solids 1.
Solid Radioactive Waste 1.
Solid Radioactive Waste A.
1hc solid radwaste system shall be A.
The solid radwaste system shall be used, as applicable in accordance verified OPERABLE at least once per with a PROCESS CONTROL PROGRAM, for quarter by:
the 50L101FICAT10N and packaging of wet radioactive wastes to meet
- 1) Verifying the solid radwaste system shipping and burial ground has been operated at least once per requirements, quarter in accordance with the PROCESS CONTROL PROGRAM, or APPLICABillTY: At all times
- 2) Verification of the existence of a ACTION:
a) With the provisions of the valid contract for SOLIDiflCATION to PROCESS CON 1ROL PROGRAM not be performed by a contractor in satisfied, suspend shipments of accordance with a PROCESS CONTROL defectively processed or PROGRAM.
packaged solid radioactive wastes from the site.
B.
The PROCESS CONTROL PROGRAM shall l
specify:
b)
The provisions of Specifications 3.0.3 and 3.0.4 are not 1)
The method and frequency of analysis applicable.
to verify SOLIDIFICATION of l
radioactive waste.
- 2) Actions to be taken if SOLIDIFICATION is not verified.
l 0664t/0665t 295a
t B.ases:
Solid Radioactive Waste 3.20 1he OPERA 61LliY of the solid radwaste system ensures that the system will be available 4.20 for use whenever solid radwastes require processing and packaging prior to being shipped offsite. Ihls specification implements the requirements of 10 CTR Part 50.36a and General Design Criterton 60 of Appendix A to 10 CFR Part 50.
1he process parameters included in establishing the PROCESS CON 1ROL PROGRAM may s
include, but are not limited to waste type, waste pit, waste /liquld/ solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.
I 1
0 0664t/0665t 295b 0293A
4.
Qunlificaticns of the Statien management and 6.0 ADMINISTRATIVE CONTROLS operating staff shall meet minimum acceptable levels as described in ANSI N18.1
" Selection and Training of Nuclear Power 6.1.
Organization, Review, Investigation, and Audit Plant Personnel," dated March 8, 1971 with 1.
The Station Manager shall have overall the exception the Radiation-Chemistry full-time responsibility for safe operations Supervisor or Lead Health Physicist, who of the facility.
During periods when the shall meet or exceed the qualifications of Station Manager is unavailable, he shall Radiation Protection Manager of Regulatory designate this responsibility to an Guide 1.8 September,1975. The Shift established alternate who satisfies the ANSI Control Room Engineer shall have a N18.1 cxperience requirements for plant bachelor's degree or equivalent in a manager.
scientific or engineering discipline with specific training in plant design, and 2.
The organization chart of the corporate response and analysis of the plant for management which relates to the operation of transient, and accidents. The individual this station and the normal functional filling the position of Services organization chart for operation of the Superintendent shall meet the minimum station is shown in Figures 1-6 of the acceptable level for " Technical Manager" as Commonwealth Edison Quality Assurance Manual.
described in~4.2.4 of ANSI N18.1, 1971.
3.
1,he Shift manning for the station shall be as 5.
Retraining and replacement training of shown in Figure 6.1-1.
The Assistant Stat *.on personnel shall be in accordance Superintendent Operating, Operating Engineer, with ANSI N18.1, " Selection and Training of Shift Engineers, and Shift Foreman shall have Nuclear Power Plant Personnel," dated March a Senior Reactor Operators License.
The Fuel 8, 1971. A training program for the Fire Handling Foreman shall have a Senior Reactor Brigade shall be maintained under the Operator or Limited Senior Reactor Operators direction of the Station Fire Marshall and License.
The Division Vice President, and shall meet or exceed the requirements of General Manager Nuclear Stations on the Section 27 of the NFPA Code - 1975 except corporate level has responsibility for the that Fire Brigade training will be conducted Fire Protection Program. An Operating quarterly.
Engineer at the station will be responsible for implementation of the Fire Protection 6.
Retraining shall be conducted at intervals Program. A Fire Brigrade of at least 5 not exceeding two years.
members snall be maintained onsite at all times. The Fire Brigade shall not include the 7.
The Review and Investigative Function and minimum shift crew necessary for safe shutdown the Audit Function of activities affecting of the plant (4 members) or any personnel quality during facility operations shall be required for other essential functions during constituted and have the responsibilities a fire emergency.
and authorities outlined below:
300 0615t/0616t 0266A
6.1.7. (Continued)
Offsite Review and Investigative Function this responsibility to an established alternate who satisfies the formal training A.
1he Supervisor of the Offsite Review and and experience requirements for the Investigative Function shall be appointed by Supervisor of the Offsite Review and i
the Manager of Nuclear Safety. The Audit Investigative Function.
l Function shall be the responsibility of the 1
Manager of Quality Assurance and shall be The responsibilities of the personnel independent of operations.
performing this function are stated below.
The Offsite Review and Investigative 1.
lhe Supervisor of the Offsite Review and Function shall review:
Investigative Function shall:
(1) provide directions for the review and (a) The safety evaluations for changes to investigative function and appoint a procedures, equipment or systems as senior participant to provide appropriate described in the safety analysis report direction, (2) select each participant and for tests or experiments completed for this function, (3) select a under the provision of 10 CFR Section complement of more than one participant 50.59 to verify that such actions did who collectively possess background and not constitute an unreviewed safety 2
qualifications in the subject matter question.
Proposed changes to the under review to provide comprehensive Quality Assurance Program description interdisciplinary review coverage under shall be reviewed and approved by the t
this function, (4) independently review Manager of Quality Assurance.
l and approve the findings and recommendations developed by personnel (b)
Proposed changes to procedures, performing the review and investigative equipment or systems which involve an j
function, (5) approve and report in a unreviewed safety question as defined timely manner all findings of in 10CFR Section 50.59.
j noncompilance with NRC requirements and j
provide recommendations to the Station (c)
Proposed tests or experiments which Manager, Division Vice-President and involve an unreviewed safety question General Manager Nuclear Stations, Manager as defined in 10CFR Section 50.59.
of Quality Assurance, and the Executive j
Vice-President of Construction, (d)
Proposed changes in Technical Production, Engineering, Licensing and Specifications or NRC facility Environmental Affairs.
During periods operating licenses.
j when the Supervisor of the Offsite Review and Investigative Function is (e) Noncompliance with NRC requirements, or unavailable, he shall designate of internal procedures or instructions 4
having nuclear safety significance.
0615t/0616t 301 0266A W
O 6.1.7.A.
6.1.7.A.1 (Continued)
~
2.
Audit Function
~
(f)
Significant operating abnormalities or deviations from normal and expected The Audit Function shall be the performance of plant equipment that responsibility of the Manager of Quality affect nuclear safety as referred to it Assurance independent of the Product 1on by the Onsite Review and Investigative Department. Such responsibility is Function.
delegated to the Director of Quality Assurance Operations and to the General (g)
Reportable Events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Supervisor of Quality Assurance notification to the Commission (Maintenance).
(h)
All recognized indications of an Either of'the above, or designated Corporate unanticipated deficiency in some aspect Staff or Supervision approved by the Manager of design or operation of safety of Quality Assurance, shall approve the related structures, systems or audit agenda and checklists, the findings components, and the report of each audit. Audits shall be performed in accordance with the Company (i)
Review and report findings and Quality Assurance Program and Procedures.
recommendations regarding all changes Audits shall be performed to assure that to the Generating Stations Emergency safety-related functions are covered within
~
Plan prior to implementation of such a period of two years or as designated below:
change.
(a) Audit of the conformance of facility (j)
Review and report findings and operation to provisions contained recommendations regarding all items referred by the: Technical Staff
_ within the Technical Specifications and applicable license conditions at least Supervisor, Station Manager, Division once per year.
Vice President and General Manager -
Nuclear Stations or Manager of Quality (b)
Audit of the adherence to procedures, Assurance.
training and qualification of~the (k)
Changes to Offsite Dose Calculation
~
Station staff at least once pcl. year.
Manual (0DCM)
(c) Audit of the results of actions taken to correct deficiencies occurring in (1)
Changes to the PROCESS CONTROL PROGRAM.
facility equipment, structures, systems or methods of operation that affect nuclear safety at least once per six months.
0615t/0616t 302
6.1.7.A.2 (Continued)
(m)
Report all findings of noncompliance with NRC requirements and (d)
Audit of the performance of activities recommendations and results of each required by the Quality Assurance audit'to the Station Manager the Program to meet the Criteria of 10 CFR Division Vice-President and General 50 Appendix "B" Manager Nuclear Stations, Manager of Quality Assurance, Vice-President (e)
Audit of the Generating Stations (Nuclear Operations), and Manager of Emergency Plan and Implementing Nuclear Safety.
Procedures.
(n) Audit the Offsite Dose Calculation (f)
Audit of the Facility Security Plan and Manual at least once per 24 months.
implementing procedures.
6.1.7.A.3.
Authority (g)
Audit Onsite and Offsite Reviews.
The Manager of Quality Assurance (h)
Audit the Radiological Environmental reports to the Vice-Chairman and the Monitoring Program at least once per 12 Supervisor of the Offsite Review and months.
Investigative Function reports to the Manager of Nuclear Safety who reports (1)
Audit the Facility Fire Protection to the Chairman and President. Either Program and implementing procedures at the Manager of Quality Assurance or the least once per 24 months.
Supervisor of the Offsite Review and Investigative Function has the (j)
An independent fire protection and loss authority to order unit shutdown or prevention program inspection and audit request any other action which he deems shall be performed at least once per 12 necessary to avoid unsafe plant months utilizing either qualified conditions.
offsite licensed personnel or an outside fire protection firm.
4.
Records (k)
An inspection and audit of the fire (a) Reviews, audits and recommendations protection and loss prevention program shall be documented and distributed as shall be performed by a qualified covered in 6.1.7.A.1 and 6.1.7.A.2 outside fire consultant at least once per 36 months.
(b) Copies of documentation, reports, and correspondence shall be kept on file at (1)
The PROCESS CONTROL PROGRAM and the station.
implementing procedures at least once per 24 months.
0615t/0616t 303 0266A
O 6.1.7.A. (Continued) 6.1.7.A.6 (Continued) 5.
Procedures Written administrative procedures shall be prepared (2)
Reactor operations and maintained for the Offsite Review and (3)
Utility operations Investigative functions described in specifications (4)
Power plant design 6.1.7.A.l. These procedures shall cover the (5)
Reactor engineering following (6)
Radiological safety (7)
Reactor safety analysis (a) Content and method of submission of (8)
Instrumentation and control presentations to the Supervisor of the Offsite (9)
Metallurgy Review and Investigative Function.
(10)
Any other appropriate disciplines required by unique characteristics of (b) Use of committees and consultants.
the facility.
(c)
Review and approval.
(b)
Individuals performing the Offsite Review and Investigative Function shall possess a minimum (d)
Detailed listing of items to be reviewed.
formal training and experience as listed below for each discipline.
(e) Method of (1) appointing personnel, (2) performing reviews and investigations, (3)
(1)
Nuclear Power Plant Technology reporting findings and recommendations of reviews and investigations, (4) approving Engineering graduate or equivalent with reports, and (5) distributing reports.
5 years experience in nuclear power field design and/or operation.
(f)
Determining satisfactory completion of action required based on approved findings and (2)
Reactor Operations recommendations reported by personnel performing the review and investigative Engineering graduate or equivalent with function.
at least 5 years experience in nuclear power plant operations.
6.
Personnel (3)
Utility Operations (a)
The persons, including consultants, performing the review and investigative function, in Engineering graduate or equivalent with addition to the Supervisor of the Offsite at least 5 years of experience in Review and Investigative Function, shall have utility operation and/or engineering.
I expertise in one or more of the following disciplines as appropriate for the subject or l
subjects being reviewed and investigated.
l (1)
Nuclear power plant technology 0615t/0616t 304 0266A
i 6.1.7.A.6 (Continusd) 6.1.7.A.6.(b) (Continued)
(c) The Supervisor of the Offsite Review and (4)
Power Plant Design Investigative Function shall have experience and training which satisfy Engineering graduate or equivalent with at ANSI N18.1 - 1971 requirements for plant least 5 years of experience in power plant managers.
design and/or operation.
6.1.7.B Onsite Review and Investigative Function (5)
Reactor Engineering Engineering graduate or equivalent with at 1.
The Onsite Review and Investigative Function least 5 years of experience in nuclear plant shall be supervised by the Station Manager.
engineering, operation, and/or graduate work in nuclear engineering or equivalent in The Station Manager shall: (a) Provide reactor physics.
direction for the Onsite Review and Investigative Function and appoint the (6)
Radiological Safety Technical Staff Supervisor, or other comparably qualified individual as a senior Engineering graduate or equivalent with at participant to provide appropriate least 5 years of experience in Radiation direction; (b) Approve participants for this Control and Safety function; (c) Assure that a complement of more than one participant who collectively (7) Safety Analysis possess background and qualifications in the subject matter under review are selected to Engineering graduate or equivalent with at provide comprehensive interdisciplinary least 5 years experience in nuclear review coverage under this function; (d) engineering.
Independently review and approve the findings and recommendations developed by (8)
Instrumentation and Control personnel performing the Review and Investigative Function; (e) Report all Engineering graduate or equivalent with at findings of noncompliance with NRC least 5 years of experience in instrumentation requirements, and provide recommendations to and control design and/or operation.
the Division Vice-President and General manager-Nuclear Stations and the Supervisor (9)
Metallurgy of the Offsite Review and Investigative Function; and (f) submit to the Offsite Engineering graduate or equivalent with at Review and Investigative function for least 5 years of experience in the concurrence in a timely manner, those items metallurgical field.
described in Specification 6.1.7.A.1 which have been approved by the Onsite Review and Investigative Function.
0615t/0616t 305 0266A
6.1.1.0. (Continued)
(h)
Review of the Station Security Plan and shall submit recommended changes to the Division Vice-President and General Manager-Nuclear 2.
The responsibilities of the personnel performing Stations.
this function are stated below:
(1) Review of the Emergency Plan and station (a)
Review of procedures required by Specification Implementing Procedures and shall submit 6.2 and changes thereto, and any other recommended changes to the Division proposed procedures or changes thereto as Vice-President and General Manager-Nuclear determined by the Station Manager to affect Stations.
nuclear safety.
(j)
Review of Reportable Events and actions taken (b)
Review of all proposed tests and experiments to prevent recurrence.
that affect nuclear safety.
(k)
Review of changes to the Offsite Dose (c) Review of all proposed changes to the Calculation Manual (ODCM).
Technical Specifications.
(1)
Review of changes to the PROCESS CONTROL (d) Review of all proposed changes or PROGRAM (PCP).
modifications to plant systems or equipment that affect nuclear safety.
(e)
Investigation of all noncompliance with NRC requirements and shall prepare and forward a report covering evaluation and recommendations to prevent recurrence to the Division Vice President and General Manager-Nuclear Stations and to the Supervisor of the Offsite Review and Investigative Function.
(f)
Review of facility operations to detect potential safety hazards.
(g)
Performance of special reviews and investigations and reports thereon as requested by the Supervisor of the Offsite Review and Investigative Function.
0615t/0616t 306 0266A
6.1.7.B (Continued)
(a) Content and method of submission and presentation to the Station Manager, 3.
Authority Division Vice President and General Manager-Nuclear Stations and the The Assistant Superintendent - Technical Services Supervisor of the Offsite Review and is responsible to the Station Manager and shall Investigative function.
make recommendations in a timely manner in all areas of review, investigation, and quality control (b) Use of committees.
phases of plant maintenance, operation and administrative procedures relating to facility (c)
Review and approval.
operations and shall have the authority to request the action necessary to ensure compliance with (d)
Detailed listing of items to be reviewed.
rules, regulations, and procedures when in his opinion such action is necessary.
The Station (e)
Procedures for administration of the Manager shall follow such recommendations or select quality control activities.
a course of action that is more conservative regarding safe operation of the facility. All such (f) Assignment of responsibilities.
disagreements shall be reported immediately to the Division Vice President and General Manager-Nuclear 6.
Personnel Station and the Supervisor of the Offsite Review and Investigative Function.
(a) The personnel performing the Onsite Review and Investigative Function, in addition to the 4.
Records Station Manager, shall consist of persons having expertise in:
(a) Reports, reviews, investigations, and (1) Nuclear power plant technology recommendations shall be documented with (2)
Reactor operations copies to the Division Vice President and (3)
Reactor engineering General Manager-Nuclear Stations, the (4)
Radiological safety Supervisor of the Offsite Review and (5)
Instrumentation and control Investigative Function, the Station Manager (6)
Chemistry and Radiochemistry and the Manager of Quality Assurance.
(7) Mechanical and electric systems.
(b)
Copies of all records and documentation shall (b)
Personnel performing the Onsite Review and be kept on file at the station.
Investigative Function shall meet minimum acceptable levels as described in ANSI N18.1 5.
Procedures 1971, Sections 4.2 and 4.4.
Written administrative procedures shall be prepared and maintained for conduct of the Onsite Review and Investigative Function.
These procedures shall include the following:
0615t/0616t 307 0266A
~
- - - =
J.
Statien Security Plan and implementing 6.2.
Plant Operating Procedures prcctdures.
l.
Written procedures including applicable K.
Fire Protection Program implementation.
checkoff lists covering items listed below shall be prepared, implemented, and maintained:
L.
Post Accident Sampling Program A.
Normal startup, operation, and shutdown M.
Working hours of the Shift Engineer, of the reactor and other systems and Shift Control Room Engineer, Shift components involving nuclear safety of Foreman, and Nuclear Station Operator the facility.
shall be controlled such that the heavy use of overtime is not routinely required.
B.
Refueling operations.
2.
Radiation control procedures shall be C.
Actions to be taken to correct specific prepared, implemented and maintained.
These and foreseen potential malfunctions of procedures shall specify permissible radiation systems or components including exposure limits and shall be consistent with responses to alarms, suspected primary the requirements of 10 CFR 20.
The radiation system leaks, and abnormal reactivity protection program shall meet the requirements changes.
of 10 CFR 20.
D.
Emergency conditions involving 3.
Procedures for items identified in potential or actual release of Specification 6.2.1 and any changes to such radioactivity
" Generating Stations procedures shall be reviewed and approved by Emergency Plan" and station emergency the Operating Engineer and the Technical Staff and abnormal procedures.
Supervisor in the areas of operation and fuel handling, and by the Assistant Superintendent E.
Instrumentation operation which could Maintenance and Technical Staff Supervisor in have an effect on the safety of the the areas of plant maintenance, instrument facility.
maintenance, and plant inspection.
Procedures for items identified in Specification 6.2.2 F.
Preventive and corrective maintenance and any changes to such procedures shall be operations which could have an effect reviewed and approved by the Technical Staff on the safety of the facility.
Supervisor and the Radiation Protection Manager (RPM). At least one person approving i
G.
Surveillance and testing requirements.
each of the above procedures shall hold a valid senior operator's license.
In addition, 3
H.
Tests and experiments.
these procedures and changes thereto must have i
the authorization of the cognizant I.
Procedures to ensure safe shutdown of Superintendent before being implemented.
the plant.
l 308 0615t/0bl6t 0266A
6.2.3 (Continued)
Work and instruction type procedures which implement 4
approved maintenance or modification procedures shall be approved and authorized by the Production 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 6.2.1.
Procedures which do not fall within the requirements of 6.2.1 or 6.2.2 may be approved by the Department Heads.
4.
Temporary changes to procedures 6.2.1 and 6.2.2 above may be made provided:
A.
The intent of the original procedure is not altered.
B.
1he change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
C.
The change is documented, reviewed by the Onsite Review and Investigative function and approved by the cognizant Superintendent within 14 days of implementation.
5.
Drills of the emergency procedures described in Specification 6.2.1.D shall be conducted at the frequency specified in the Generating Station Emergency Plan. These drills will be planned so that during the course of the year, communication Tinks are tested and outside agencies are contacted.
0615t/0616t 309 0266A
B.
Record of principal maintenance activities, 6.3 Action to be Taken in the Event of a Reportable including inspection, repair and replacements, Event in Plant Operation:
regarding principal items of equipment pertaining to nuclear safety.
Any reportable Event shall be promptly reported to the Division Vice President and General C.
Records and reports of Reportable Events and Manager-Nuclear Stations or his designated Safety Limit occurrences.
alternate. The incident shall be promptly reviewed pursuant to Specification 6.1.7.B.l.(e) and a D
Records and periodic checks, inspection and/or separate report for each reportable event shall be calibrations perforned to verify the prepared in accordance with the requirements of Surveillance Requirements (Section 4 of these Section 50.73 of 10 CFR Part 50.
Specifications) are being met. All equipment failing to meet surveillance requirements and 6.4 Action to be Taken in the Event of a Safety Limit the corrective action taken shall be recorded.
is Exceeded:
E.
Records of changes made to procedures required If a safety limit is exceeded, the reactor shall be by Specifications 6.2.1 and 6.2.2.
placed in HOT SHUTODWN within I hour and reacter operation shall not be resumed until authorized by F.
Records of radioactive shipments.
the NRC. The conditions of shutdown shall be promptly reported to the Division Vice President G.
Records of physics tests and other tests and General ManagerlNuclear. Stations or his pertaining to nuclear safety.
designated alternate. The incident shall be reviewed pursuant to Specification 6.1.7.A.1 and H.
Records of sealed source and fission detector 6.1.7.B.1 and a separate report for each occurrence leak tests and results.
shall be prepared in accordance with Section 50.36 of 10 CFR Part 50.
I.
Shift Engineers Logs.
6.5 Plant Operating Records J.
By-product material inventory records.
1.
Records relative to the following items shall be retained for at least five years:
A.
Records of normal plant operation, including power levels and periods of operation at each power level.
0615t/0616t 310 0266A
1.
Records of transient or operational cycling 6.5. (Continued) for those components that have been designed to operate safely for a limited number of 2.
Records relative to the following. items shall transient or operational cycles, be retained for the duration of the Operating J.
Records of training and qualification for License.
current members of the station staff.
A.
Substitution or replacement of principal items of equipment pertaining to nuclear K.
Inservice inspections performed pursuant to safety.
the Technical Specifications.
B.
Records and drawing changes reflecting L.
Minutes of meetings and results of reviews plant design modifications made to performed by the Offsite and Onsite Review systems and equipment described in the Functions.
Safety Analysis Report.
M Records of secondary water sampling and water C.
Records of new and irradiated fuel quality.
inventory, fuel transfers and assembly burnup histories.
N.
Records for Environmental Qualification of components required by the environmental D.
Updated, corrected, and as-built drawings qualification program.
of the plant.
O.
Records of the service lives of all snubbers E.
Records of plant radiation and covered by Specification 3.22 including the contamination surveys.
date at which the service life commences and associated installation and maintenance F.
Records of offsite environmental records.
monitoring surveys.
P.
Records of reviews performed for changes made G.
Records of radiation exposure for all to procedures or equipment or reviews of tests plant personnel, including all and experiments pursuant to 10 CFR 50.59.
contractors and visitors to the plant in accordance with 10 CFR 20.
Q.
Records of Quality Assurance activities required by the Q.A. Manual.
H.
Records of radioactivity in liquid and gaseous wastes released to the R.
Records of reactor tests and experiments, environment.
0615t/0616t 311 026bA
O 6.6 REPORTING REQUIREMENTS Startup reports shall be submitted within (a) 90 days following completion of the In addition to the applicable reporting startup test program, requirements of Title 10, Code of Federal (b) 90 days following resumption of Regulations, the following identified reports shall commercial power operation, or be submitted to the Director, of the Regional (c) 9 months following initial criticality, Office of Inspection and Enforcement unless whichever is earliest.
Otherwise noted.
If the Startup Report does not cover all three events (i.e., initial criticality, completion 1.
Routine Reports of startup test program, and resumption of i
A.
Startup Report. A summary report of commercial power operation), supplementary i
plant startup and power escalation reports shall be submitted at least every testing shall be submitted following three months until all three events have been 1
(1) receipt of an operating license, completed.
(2) amendment to the license involving a planned increase in power level, B.
Annual Occupational Exposure Report. A (3) installation of fuel that has a tabulation covering the previous calendar year different design or has been should be submitted prior to March 1 of each manufactured by a different fuel year on the number of station, utility and supplier, and other personnel (including contractors)
(4) modifications that may have receiving exposures greater than 100 mrem / year significantly altered the nuclear, and their associated man rem exposure thermal, or hydraulic performance of according to work and job functions, 1 e.g.,
the plant.
reactor operations and surveillance, inservice The report shall address each of the tests inspection, routine maintenance, special identified in the FSAR and shall in general maintenance (describe maintenance), waste include a description of the measured values processing, and refueling. The dose of the operating conditions or characteristics assignments to various duty functions may be obtained during the test program and a estimates based on pocket dosimeter, TLD, or comparison of these values with design film badge measurements. Small exposures predictions and specifications. Any totaling less than 20% of the individual total corrective actions that were required to dose need not be accounted for.
In the obtain satisfactory operation shall also be aggregate, at least 80% of the total whole described. Any additional specific details body dose received froa external sources shall required in license conditions based on other be assigned to specific major work functions.
commitments shall be included in this report.
I 1
This tabulation supplements the requirements l
of 20.407 of 10CFR Part 20.
I 0615t/0616t 312 026bA
6.6.1. (Crntinu;d)
(5) A summary description of the radiological C.
Annual Radiological Environmental Operating Report environmental monitoring program.
An annual report containing the data taken in the (6) A map of all sampling locations keyed to a standard radiological monitoring program (Table table giving approximate distances and 3.16-1) shall be submitted by April 30 of the directions from one reactor.
following year. The content of the report shall include:
(7)
The results of the Interlaboratory Comparison Program required by Specification 3.16.3 (1)
Results of radiological environmental sampling, summarized and tabulated, following (8) This report shall also include an annual the format of Regulatory Guide 4.8, Table 1 summary of hourly meteorological data (December 1975); individual sample results collected over the previous year. This will be retained at the station; in the event annual summary may be either in the form of that some results are not available for an hour-by-hour listing of wind speed, wind inclusion with the report, the report shall direction, atmospheric stability, and be submitted noting and explaining the reason precipitation (if measured) on magnetic tape, for the missing results. The missing data or in the form of joint frequency shall be submitted as soon as possible in a distributions of wind speed, wind direction, supplementary report.
and atmospheric stability. This same report shall include an assessment of the radiation (2)
An assessment of the monitoring results and doses due to the radioactive liquid and radiation dose via the principal pathways of gaseous effluents released from the station exposure resulting from plant emissions of during the previous calendar year. This same radioactivity; including maximum noble gas report shall also include an assessment of gamma and beta air doses in the unrestricted the radiation doses from radioactive liquid area (dose calculations shall be performed in and gaseous effluents to individuals due to accordance with the ODCM).
their_ activities inside the site boundary (Figure 3.11-1) during the report period.
(3)
Results of the census to determine the All assumptions used in making these locations of animals producing milk for human assessments (i.e., specific activity, consumption.
exposure time and location) shall be included in these reports.
The meteorological (4)
A summary of the meteorological conditions conditions concurrent with the time of concurrent with the release of gaseous release of radioactive materials in gaseous effluents during each quarter as outlined in effluents (as determined by sampling Regulatory Guide 1.21 (Revision 1) dated frequency and measurement) shall be used for June, 1974, following the format of Appendix determining the gaseous pathway doses.
The B thereof.
assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (0DCM).
0611t/0618t 313 026bA
6.6.1.C (Crntinued)
This report shall also include an assessment D.
Semiannual Radioactive Release Report:
of radiation doses to the likely most exposed real individual from reactor Within 60 days aafter January 1 and July 1 of releases and other nearby uranium fuel cycle each year a report shall be submitted sources (including doses from primary covering the radioactive content of effluents effluent pathways and direct radiation) for released to unrestricted areas during the the previous 12 consecutive months to show previous six months operation. The data conformance with 40 CFR 190, Environmental shall be in the format of Regulatory Guide Radiation Protection Standards for Nuclear 1.21 Rev. 1 (June 1974) and shall be Power Operation. Acceptable methods for summarized on a quarterly basis and shall calculating the dose contribution from include as a minimum:
liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.
1.
Gaseous Effluents (a)
Gross Radioactivity Releases (1) Total gross radioactivity (in curies) primarily noble and activation gases released.
(2) Maximum gross radioactivity release rate during any one-hour period.
(3) Total gross radioactivity (in curies) by nuclide released, based on representative isotopic analyses performed.
(4)
Percent of technical specification limit.
(b)
Iodine Releases (1)
Total iodine radioactivity (in curies) by nuclide released, based on representative isotopic analyses performed.
(2)
Percent of technical Specification for I-131 released.
0617t/0618t 314 0766A
6.6.1.D.1. (Centinued)
(c)
Particulate Release (f) The maximum concentration of gross radioactivity (Beta-Sanna) released to the (1) Total gross radioactivity (Beta-Ganna) unrestricted area (averaged over the period released (in curies) excluding of release).
background radioactivity.
(g) Total gross radioactivity (in curies) by (2) Total gross Alpha radioactivity nuclide released, based on representative released (in curies) excluding isotopic analyses performed.
background radioactivity.
(h)
Percent of Technical Specification limit.
(3) Total gross radioactivity released (in curies) of nuci) des with half-lives greate~r~tha6 Olgn oays.
3.
Solid Radioactive Waste:
(4)
Percent of Technical Specification The following information for each type of solid limit for particulate radioactivity waste shipped offsite during the report period:
with half-lives greater than eight days.
(a.) Container volume 2.
Liquid Effluents (b.) Total curie quanity (specify whether (a)
Total gross radioactivity (Beta, Gamma) determined by measurement or estimate).
released (in curies) exclusing tritium
)
and average concentration released to (c.) Principal radionuclides (specify whether the unrestricted area.
determined by measurement or estimate)
(b) Total tritum and total Alpha (d.) Type of waste (e.g., spent resin, compacted l
radioactivity released (in curies) and dry waste evaporator bottoms),
average concentration released to the unrestricted area.
(e.) Type of container (e.g. LSA, Type A, Type B, l
Large Quantity), and (c) Total dissolved noble gas radioactivty released (in curies) and average (f.) Solidification agent (e.g., cement, urea concentration released to the unrestricted formaldehyde).
area.
(g)
Dates of shipment and disposition (if (d)
Total volume (in liters) of liquid waste shipped offsite) released.
4.
The radioactive effluent release reports shall (e) Total volume (in liters) of dilution water include unplanned releases from the site to used prior to release from the restricted unrestricted areas of radioactive materials in area.
gaseous and liquid effluents on a quarterly basis.
0617 t/0618t 315
6.6.1.0. (Continu:d) 5.
The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.
E.
Monthly Operating Report Routine reports using the " Operating Data", " Unit Shutdowns and Load Reductions" and " Average Daily Unit Power Level" report formats, shall be submitted on a monthly basis to the 4
Director, Office of Management, Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 with a copy to the appropriate NRC Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.
Included with the report will be a
" Narrative Summary of Operating Experience" that describes the operation of the facility, including major safety related maintenance, for the monthly report period.
Any changes to the Offsite Dose Calculation Manual shall be submitted with the monthly Operating Report within 90 days of the effective date of the change.
i I
\\
i I
0617t/0618t 316
3 6.6.2.
Reportable Events 6.6.2.A.2 Non-Emergency Reports:
A.
~Immediate Notification:
l.
Emergency Reports:
a.
One-Hour Reports - If not reported as a declaration of an Emergency a.
The licensee shall notify the NRC Class under paragraph 6.6.2.A.1 of Operations Center via the Emergency this section, the licensee shall Notification System (ENS) of:
notify the NRC as soon as practical 1)
The declaration of any of the and in all cases within one hour of Emergency Classes specified in the occurrence of any of the the Generating Station following.
Emergency Plan (GSEP).
2)
Those Non-Emergency events 1)
The initiation of any nuclear specified in paragraph plant shutdown required by the 6.6.2.A.2 below.
plant's Technical Specifications.
l b.
If the Emergency Notification System is inoperative, the required 2)
Any deviation from the plant's notifications shall be made via Technical Specifications j
commercial telephone service, other authorized pursuant to dedicated telephone system, or any 10CFR50.54(x).
other method which will ensure that a report is made as soon as 3)
Any event or condition during practical to the NRC Operations operation that results in the Center.
condition of the nuclear I
powerplant, including its c.
The licensee shall notify the NRC principal safety barriers, immediately after notification of being seriously degraded, or the appropriate state or local results in the nuclear power agencies and not later than one hour plant being:
after the time the licensee declares one of the Emergency Classes.
a)
In an unanalyzed condition that significantly d.
When making a report under paragraph compromises plant safety, 6.6.2.A.l.c above, the licensee shall identify:
b)
In a condition that is 1) the Emergency Classification outside the design basis declared: or of the plant, or 2)
Either paragraph (2)(a),
"One-Hour Report," or paragraph c)
In a condition not covered (2)(b), "Four-Hour Report." as by the plant's operating the paragraph of this section and emergency procedures.
requiring notification of the Non-Emergency Event.
0615t/0616t 317
6.6.2.A.2.a (Centinu;d) 4)
Any natural phenomenon or other b.
Four-Hour Reports - If not reported external condition that poses an under paragraphs 6.6.2.A.1 or actual threat to the safety of 6.6.2.A.2.a of this section, the the nuclear power plant or licensee shall notify the NRC as significantly hampers site soon as practical and in all cases, personnel in the performance of within four hours of the occurrence duties necessary for the safe of any of the following:
operation of the plant.
j 1)
Any event found while the reactor is 5)
Any event that results or should shutdown, that had it been found have resulted in Emergency Core while the reactor was in operation, i
Cooling System (ECCS) discharge would have resulted in the nuclear into the reactor coolant system power plant, including its principal as a result of a valid signal.
safety barriers, being seriously degraded or being in an unanalyzed 6)
Any event that results in a major condition that significantly loss of emergency assessment compromises plant safety.
capability, offsite response i
capability, or communications 2)
Any event or condition that results capability (e.g. significant in manual or automatic actuation of i
portion of control room an Engineered Safety Feature (ESF),
indication, Emergency including the Reactor Protection Notification System, or offsite System (RPS). However, actuation of i
notification system.)
an ESF, including the RPS, that results from and is part of the 7)
Any event that poses an actual preplanned sequence during testing or threat to the safety of the reactor operation need not be nuclear power plant or reported.
significantly hampers site I
personnel in the performance of 3)
Any event or condition that alone duties necessary for the safe could have prevented the fulfillment j
operation of the nuclear power of the safety function of structures plant including fires, toxic gas or systems that are need to:
releases, or radioactive releases.
a)
Shutdown the reactor and i
maintain it in a safe l
shutdown condition.
i b) Remove residual heat l
c) Control the release of radioactive material, or d) Mitigate the consequences i
of an accident.
0615t/0616t 318 1
I 6.6.2.A.2.b (Continued) 6.6.2.A. 3.
Followup Notification:
i 4)
.a)
-Any airborne radioactive release that exceeds 2 times i
the-applicable concentrations With respect to the telephone i
of the limits specified in notifications made under paragraphs Appendix B. Table II of 10CFR20 6.6.2.A.1 and 6.6.2.A.2 in addition to in unrestricted areas, when making the required initial notification, averaged over a time period of each licensee, shall during the course of one hour.
the event:
i b)
Any liquid effluent release a.
humediately Report 7
that exceeds 2 times the limiting combined Maximum 1)
Any further degradation in the level i
Permissible Concentration (MPC) of safety of the plant or other (see Note 1 of Appendix 8 to worsening plant conditions, 10CFR 20) at the point of entry including those that require the into the receiving water (i.e.
declaration of any of the Emergency l
unrestricted area) for all Classes, if such a declaration has l
radionuclides except tritium not been previously made, or i
and dissolved noble gases, when averaged over a time period of 2)
Any change from one Emergency Class j
one hour.
(Immediate to another, or notifications made under-this l
paragraph also satisfy the 3)
A termination of the Emergency Class requirements of paragraphs (a)(2) and (b)(2) of 10CFR 4)
The results of ensuring evaluations 20.403.)
or assessments of plant conditions.
5)
Any event requiring the transport of 5)
The effectiveness of response or a radioactively contaminated person protective measures taken to an offsite medical facility for treatment.
6)
Information related to plant j
behavior that is not understood.
6)
Any event or situation, related to
}
the health and safety of the public b.
Maintain an open, continuous j
or onsite personnel or protection of communication channel with the NRC j
the environment, for which a news Operations Center upon request by the NRC.
4 release is planned or notification I
to other government agencies has
}
been or will be made. Such an event j
may include an onsite fatality or j
inadvertent release of radioact vely
(
j contaminated materials.
[
0615t/0616t 319
6.6.2 (Continued)
(2) The following shall be reported:
B.
Thirty Day Written Reports 1.
Reportable Events, including (I) A)
The completion of any plant corrective actions and measures to shutdown required by the prevent reoccurrence, shall be Technical Specifications; or reported to the NRC pursuant to Title 10. Code of Federal B)
Any operation or condition Regulations Part 50 Section 50.73.
prohibited by the Technical Supplemental reports may be required Specifications; or to fully describe the final resolution of an occurrence.
In C)
Any deviation from the case of corrected or supplemental Technical Specifications reports, a Licensee Event Report authorized pursuant to 10CFR shall be completed and reference 50.54(x) shall be made to the original report number.
(II)
Any event or condition that resulted in the condition of the plant, (a) Reportable Events including its principal safety barriers, being seriously degraded, (1) A Licensee Event Report or that resulted in the plant being:
(LER) shali Lc inhmitted for any event of the typc A)
In an unanalyzed condition that described below within 30 significantly compromised plant days after the discovery safety; of the event. Unless B) in a condition that was outside otherwise specified in the oesign basis of the plant; this section, the licensee or shall report an event C)
In a condition not covered by regardless of the plant the plant's operating and MODE or power level, and emergency procedures.
regardless of the significance of the structure, system, or (III) Any natural phenomenon or other component that initiated external condition that posed an the event.
actual threat to the safety of the plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the plant.
0615t/0616t 320 0266A
6.6.2.B.l.a (Continued)
(IV)
Any event or condition that (VII) '
Any event where a single cause or resulted in manual or automatic condition caused at least one actuation of any Engineered Safety independent train or channel to Feature (ESF), including the become inoperable in multiple Reactor Protection System (RPS).
systems or two independent trains However actuation of an ESF, or channels to become inoperable including the RPS, that resulted in a single system designed to:
from and was part of the preplanned sequence during testing A)
Shutdown the reactor and or reactor operation need not be maintain it in a safe reported.
shutdown condition; B)
Remove residual heat; (V)
Any event or condition that alone C)
Control the release of could have prevented the radioactive material; or fulfillment of the safety function D)
Mitigate the consequences of of structures or systems that are an accident.
needed to:
(VIII)
A)
Any airborne radioactivity A)
Shutdown the reactor and release that exceeded 2 times maintain it in a safe the applicable concentrations shutdown condition; of the limits specified in B)
Remove residual heat; Appendix B Table II 10CFR C)
Control the release of Part 20 in unrestricted radioactive material; or areas, when averaged over a D)
Mitigate the consequences of time period one hour.
an accident.
B)
Any liquid effluent release that exceeded 2 times the (VI)
Events covered in paragraph (V) limiting combined Maximum above may include one or more Permissible Concentration procedural errors, equipment (MPC) (see Note 1 of Appendix failures, and/or discovery of B to 10CFR Part 20) at the design, analysis, fabrication, point of entry into the construction, and/or procedural receiving water (i.e.
inadequacies. However, individual unrestricted area) for all component failures need not be radionuclides except tritium reported pursuant to this and dissolved noble gases, paragraph if redundant equipment when averaged over a time in the same system was operable period of one hour.
and available to perform the required safety function.
0617t/0618t 321 0266A
6.6.2.8.1.a (Continued)
(IX)
Reports submitted to the Commission in
~
accordance with paragraph (VIII) above
]{~_
Ci
>l also meet the effluent release RL.i reporting requirements of 10CFR Part 20 paragraph 20.405(a)(5) where C is the concentration of the ith radionuclides in the medium and RL is the (X)
Any event that posed an actual threat reporting level of radionuclide i.
to the safety of the plant or significantly hampered site personnel (2)
If radionuclides other than those in Table in the performance of duties necessary 3.16-1 are detected and are due to plant for the safe operation of the nuclear effluents, a reporting level is exceed if power plant including fires, toxic gas the potential annual dose of an individual releases, or radioactive releases.
is equal to or greater than the design objective doses of 10 CFR 50, Appendix 1.
6.6.3 Unicue Reporting Reauirements (3) This report shall include an evaluation of any release conditions, environmental A.
Non-Routine Reports:
factors, or other aspects necessary to explain the anomalous effect.
(1) Environmental Radiological Monitoring Program
~
If a confirned measured radionuclide concentration in an environmental sampling medium averaged over any calendar quarter sampling period exceeds the reporting level given in Table 3.16-1 and if the radioactivity is attributable to plant operation, a w-itten report shall be submitted to the Director of the NRC Regional Office of Inspection and Enforcement with a copy to the Director, Office of Nuclear Reactor Regulation within 30 days from the end of the quarter. When more than I
one of the radionuclides in Table i
3.15-1 are detected in the medium, the reporting level shall have been exceeded if:
i 0611t/0618t 322 026bA
6.6.3.B Special Reports Reports on the following areas shall be submitted as indicated:
Topic SUBMITTAL DATE a.
In-Service Inspection Evaluation In accordance with the requirements of Section 11 ASME Boiler Code (15-620) b.
Loose parts monitoring program On same date as in-service inspection report.
c.
Containment Building Structural Testing Within 90 days following completion of each test.
Report d.
Changes to the ODCM Within 90 days of effective date as part of the Monthly Operating Report.
i e.
Waukegan Memorial Airport Expansion The expansion plans and status of such plans for i
Plans.
the Waukegan Memorial Airport will be reported 1
yearly in the Annual Report, including FAA form
- 5010 report.
l f.
Overpressure Protection System Operation Within 30 days of operation.
g.
Primary Coolant Specific Activity Within 30 days of exceeding 1.0 microcurie per gram.
DOSE EQUIVALENT I-131 for greater than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any consecutive 6-month period, h.
Pressurizer PORV or Safety Valve Within 30 days as an LER.
Failure SPECIAL REPORTS Table 6.6-2 0617t/0618t 323 0766A
6.6.3.B Special Reports (Continued)
TOPIC SURMITTAL DATE i.
Pressurizer PORV or Safety Valve Document the event in the annual report challenges j.
Service Water leaks inside containment Within 30 days of exceeding T. S. 4.8.7.C.1.
k.
Steam generator tube inspection and/or Per surveillance requirements 4.3.1.B.5 plugging.
1.
Emergency core cooling system actuation Within 30 days as LER - include nozzle usage
& injection when RCS temp > 350*F factor per T.S.
3.3.2.F.3 m.
Fire detector inoperability Within 30 days per T.S. 3.21.1.C n.
Fire pump system inoperability Within 30 days per T.S. 3.21.2.B o.
Fire suppression system inoperability Within 30 days per T.S. 3.21.2.C or 3.21.2.0 p.
Sprinkler system inoperability Within 30 days per T.S. 3.21.3.C q.
Low pressure CO2 system inoperability Within 30 days per T.S. 3.21.4.C l
l SPECIAL REPORTS Table 6.6-2 (Continued) 0617t/0618t 324 0266A i
6.7 0FFSITE DOSE CA!.CULATION MANUAL (ODCM)
~
b.
A determination that the change will not reduce the 1.
The ODCM shall describe the methodology and accuracy or reliability of parameters to be used in the calculation of dose calculations or setpoint offsite doses due to radioactive gaseous and determinations; and liquid effluents and in the calculation of gaseous and liquid effluent monitoring c.
Documentation of the fact instrumentation alarm / trip setpoints consistent that the change has been with the applicable LCO's contained in these reviewed and found acceptable Technical Specifications.
by both the Onsite and Offsite Review functions.
2.
Any changes to the ODCM shall be made by either of the following methods:
B.
Commission initiated changes:
A.
Licensee initiated changes:
1.
Shall be determined by the Onsite Review function to be 1.
Shall be submitted to the applicable to the facility Commission by inclusion in the after consideration of Monthly Operating Report pursuant facility design, to Specification 6.6.lc within 90 days of the date the change (s) was 2.
The licensee shall provide made effective and shall contain:
the Commission with written notification of their a.
Sufficiently detailed determination of information to support the applicability including any change.
Information necessary revisions to submitted should consist of a reflect facility design, package of those pages of the ODCM to be changed, together 3.
Shall be reviewed by the with appropriate analyses or Offsite Review function.
evaluations justifying the change (s);
0611t/0618t 325 0266A
6.8 Floodina Protection 1.
In the event of the possibility of flooding.
all doors listed in table 6.8-1 shall be verified closed.
P k
0617t/0618t 326 0266A
Units in Modes 1. 2. 3. 4 or 7 Unit 1 Unit 1 No or and Position Unit Unit 2 Unit 2 Shift Engineer 1
1 2
or Shift Foreman Shift Control Room Engineer None 1
1 Nuclear Station Operator 2
3 3
Equipment Operator 2
3 4
or Equipment Attendant Radiation Protection Person 1
1 1
TOTAL 6
9
'll MINIMUM
- 6 8
10
- The minimum rumber refers only for the case of shift shortage, caused by a sudden sickness or home emergency.
NOTES:
1.
Senior Reactor Operator (SRO) shall be present on-site at all times when there is fuel in the reactor.
2.
An NRC licensed individual shall be in the control room at all times whenever fuel is in either reactor.
3.
Two NRC licensed individuals shall be in the control room during reactor startups, shutdowns, operation, and other periods such as planned control rod manipulations.
MINIMUM SHIFT CREW COMPOSITION Figure 6.1-1 0617t/0618t 327 0266A
DOOR #
Door Zone LOCATION 1
1 Exterior from Service Building vestibule.
5 Loading from Service Building storeroom.
16 58 Extertor from' Turbine Bldg. Track door - Unit #1 20 4
Exterior from Service Bldg. East Wall 69 SA Exterior from Turbine Bldg. West Wall Unit #1 70 3
Exterior from Turbine Bldg. East Wall between units 71 8
Exterior from Turbine Bldg. East Wall Unit #2 72 9B Exterior from Turbine Bldg. West Wall Unit #2 73 9A Exterior from Turbine Bldg. West Wall Unit #2 277 17B Exterior to car shed - rolling steel door.
1 278 12A Exterior from car shed west wall
]
31b 10 Exterior from cribhouse south wall 318 11 Exterior from cribhouse north wall 341 6
Exterior to Radwaste Annex - rolling steel door i
342 19B Exterior from isolation safety valve enclosure 1E south side Unit #1 i
343 190 Exterior from isolation safety valve enclosure 1W south side Unit #1 344 19E Exterior from isolation safety valve enclosure IW north side Unit #1 1
Boundary Doors for Flooding Protection 7ABLE 6.8-1 0611t/0618t 328 0266A
i h
9-DOOR #
Door Zone LOCATION 345 14A Exterior from isolation safety valve enclosure 2E south side.
346 14B Exterior from isolation safety valve enclosure 2E north side.
347 14D Exterior from isolation safety valve enclosure 2W north side.
348 14C Exterior from isolation safety valve enclosure 2W south side.
14C Exterior from containment escape hatch enclosure Unit 1 19C Exterior from containment escape hatch enclosure Unit 2 Boundary Doors for Flooding Protection TABLE 6.8-1 (Continued) 0611t/0bl8t 329 0266A
yl
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ATTACHW_NT 2 Evaluation of Significant Hazards Consideration Proposed Changes to Zion Appendix A Technical Specification Sections:
3.11, 12, 14, 16 and 20 4.11, 12, 14, 16 and 20 6.1, 2, 3, 4, 5, 6, 7, and 8 Description of Amendment Request An amendment to Facility Operating License Nos_. DPR-39 and DPR-48 is proposed to modify eighteen separate sections of the Technical Specifications. These changes are being submitted as part of an effort to convert the Zion Technical Specifications to the Standardized Technical Specifications.
A description of the significant variations between the proposed Technical Specifications and the Standardized Technical Specifications (STS) is provided below:
A.
Sections 3.11 and 4.11 - Liquid Effluents (1) The allowable dose specified in 3.11.2 is twice the STS per Reactor ~
Value since Zion uses a per site value and has 2 reactors on site.
The STS Bases B 3/4 11.1 indicate the STS Value is per reactor.
(2) The Gaseous Radwaste Treatment specification of STS 3.11.2.4 is omitted since Zion Gaseous Radwaste is processed via the gas decay tanks, containment and auxiliary building ventilation systems all of which are adequately monitored and Technical Specifications exists which require monitor operability or alternative action.
(3) Specification 3.11.3 " Liquid Effluent Monito$1ng Instrumentation" has been included in this section, instead of in a separate instrumentation section to facilitate continuity.
(4) Specification 3.11.4 " Liquid Radwaste Treatment System".
Additionally defines operable as applied to the liquid radwaste treatment system.
B.
Section 3.12 and 4.12 - Gaseous Effluents (1) Specification 3.12.1.A.2 has been reworded slightly from STS 3.11.2.1.b to agree with other corporate submittals.
(2) Specification 3.12.2 combines STS specifications 3.11.2.2 and 3.11.2.3 into a single specification 3.12.2.
The requirement for a special report has been omitted from the Action statement since the condition will be reportable under the LER program per 10 CFR 50.36(c)(2), and 10 CFR 50.73 if the action statement is exceeded.
Duplicate reporting is an unnecessary administrative burden.
e (3) Specification 3.12.3 " Gaseous Effluent Instrumentation" has been included in this section to facilitate continuity.
(4) Specification 3.12.5 " Explosive Gas Mixture" has been reworded to reflect the present grab sampling method of testing. The concentration of hydrogen or oxygen may be above the specified limit as long as one of the two is below the limit without adverse effects. This proposed change is in agreement with Westinghouse Technical Bulletin NSD-TB-80-3.
C.
Sections 3.14 and 4.14 - Plant Radiation Monitoring (1) Alarm / trip setpoints are included in the ODCM instead of the Tech Specs, in order to facilitate changes due to operational ~
requirements.
D.
Sections 3.16 and 4.16 - Radiological Environmental Monitoring (1) Specification 3.16.1 " Monitoring Program" Action statement "a" includes alternate actions if samples are unobtainable due to various reasons. A clarifying statement is included in the table notation of table 4.16-1 regarding LLD measurement.
(2) Specification 3.16.2 " Land Use Census" lists only those sectors which are on land.
E.
Section 3.2.0 and 4.2.0 - Solid Radioactive Waste (1) Specification 3.20.1.A includes an additional reference to the 10 CFR 61 requirements. The requirement for a special report has been omitted because without an operable system it would be impossible to meet the LCO and no solid radioactive waste could be shipped of f-site.
F. -Sections 6.1, 2, 3, 4, 5, 6, 7, and 8 - Administrative Controls
-Sections 6.0 continues to incorporate the elements of STS. The changes to 6.0 involve the following aspects:
(1) The entire section has been reformatted to mirror the STS.
(2) The management structure and titles reflect a recent reorganization.
(3) Incorporation of the reporting requirements of 10 CFR 50.72 and 50.73.
(4) Includes recent changes to 10 CFR 50.54.
(5) Includes new requirements to review the changes to the Process Control Program (PCP) and the Off-Site Dose Calculation Mode (00CM).
o
. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission's examples of actions involving no significant hazards considerations include:
(i) A purely administrative change to technical specifications:
for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature; (ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications:
for example, a more stringent surveillance requirement; (vii) A change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.
It is clear that the changes described in A through E above all involve additional limitations or restrictions. For example, the dose limitations of sections 3/4.11 and 3/4.12, the larger number of monitors
/
addressed in Sections 3/4.14, the increased detail of Sections 3/4.16, and the implementation of the PCP in Sections 3/4.20 are all additional limitations or restrictions. Thus, these changes fit example (ii) above.
(
The changes to Sections 6.1 through 6.8 also fit the examples provided. For example: the changes identified above as F.1 and F.2 are both purely administrative changes and thus conform to example (1).
The changes identified above as F.3 and F.4 are designed to make Zion's license conform with 10 CFR 50.72, 50.73, and 50.54. Since facility operation is clearly within the applicable regulations, these amendments fit example (vil).
The change identified above as F.5 imposes the new requirements to review the ODCM and PCP. Thus, this change fits example (ii).
Therefore, since the application for amendment involves proposed changes that are similar to examples to which no significant hazards consideration exists, Commonwealth Edison has made a proposed determination that the application involves no significant hazards consideration.
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