ML031270586
| ML031270586 | |
| Person / Time | |
|---|---|
| Site: | Cooper  (DPR-046) |
| Issue date: | 05/01/2003 |
| From: | Coyle M Nebraska Public Power District (NPPD) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NLS2003051 | |
| Download: ML031270586 (118) | |
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APPENDIX D ODAM SPECIFICATIONS
INFORMATION ONLY OFFSITE DOSE ASSESSMENT MANUAL APPENDIX D TABLE OF CONTENTS D 1.0 Use and Application............................................
D 1.0-1 D 2.0 Definitions....................................................
D 2.0-1 D 3.0 ODAM Specifications Applicability...................................
D 3.0-1 D 3.1 LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration................................
D 3.1-1 D 3.1.2 Liquid Waste Concentration..................................
D 3.1-5 D 3.1.3 Liquid Effluents Dose.....................................
D 3.1-6 D 3.1.4 Outside Temporary Storage of Radioactive Liquid................
D 3.1-9 D 3.2 GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Concentration.............................
D 3.2-1 D 3.2.2 Noble Gas Dose.............
D 3.2-3 D 3.2.3 Iodine and Particulates................................
D 3.2-5 D 3.2.4 Offgas Treatment System................................
D 3.2-1 0 D 3.2.5 Exhaust Ventilation Treatment System...........
............ D 3.2-12 D 3.2.6 Hydrogen Concentration................................
D 3.2-14 D 3.2.7 Primary Containment Venting and Purging.........
........... D 3.2-15 D 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring.......
D 3.3-1 D 3.3.2 Gaseous Effluent Monitoring........................
D 3.3-8 D 3.4 LIQUID/GASEOUS DOSE D 3.4.1 Liquid/Gaseous Effluents Dose..............................
D 3.4-1 D 3.5 SOLID RADIOACTIVE WASTE D 3.5.1 Solid Radioactive Waste.................
D 3.5-1 D 4.0 MONITORING PROGRAM D 4.1 Monitoring Program Compliance.............................
D 4.1-1 D 4.2 Monitoring Program Concentration............................
D 4.2-1 D 4.3 Monitoring Program Dose..
D 4.3-1 D 5.0 MISCELLANOUS PROGRAMS/REPORTS D 5.1 Interlaboratory Comparison Program..........................
D 5.1-1 D 5.2 Annual Radiological Environmental Report......................
D 5.2-1 D 5.3 Radioactive Effluent Release Report...........................
D 5.3-1 D 5.4 Special Reports..............
D 5.4-1 D 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid).....
D 5.5-1 CNS ODAM iii 1 0/10101
INFORMATION ONLY Use and Application D 1.0 D 1.0 USE AND APPLICATION The Offsite Dose Assessment Manual (ODAM) Specifications are contained in Section 3.0 of this appendix. They contain operational requirements, Surveillance Requirements. and reporting requirements. Additionally, the Required Actions and associated Completion Times for degraded Conditions are specified.
The format is consistent with the Technical Specifications (Appendix A to the CNS Operating License).
The definitions contained in Technical Specifications Section 1.1.
"Definitions." apply to the ODAM Specifications. Defined terms are shown in all capital letters, consistent with the Technical Specifications.
The rules of usage for the ODAM Specifications are the same as those for the Technical Specifications.
These rules are found in Technical Specifications Sections 1.2. "Logical Connectors:" 1.3.
Completion Times;" and 1.4;
'Frequency."
i eRMATION ONLYFigure D2.a-1 D 2.0 DEFINITIONS Figure D2.a-1 (Page 1 of 1)
Site and Exclusion Area Boundaries CNS ODAM D 2.0-1 1 2/20101
INFORMATION ONLY ODAM Specifications Applicability D 3.0 D 3.0 ODAM Specifications Applicability DLCO 3.0.1 DLCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in DLCO 3.0.2.
DLCO 3.0.2 Upon discovery of a failure to meet a DLCO, the Required Actions of the associated Conditions shall be met, except as provided in DLCO 3.0.5.
If the DLCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.
DLCO 3.0.3 When a DLCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, action shall be initiated to:
- a.
Restore compliance with the DLCO or associated ACTIONS, and
- b.
Enter the circumstances into the Corrective Action Program.
--------- NOTE------------------------ ------
DLCO 3.0.3.b shall be completed if DLCO 3.0.3 is entered.
Exceptions to this Specification are stated in the individual Specifications.
DLCO 3.0.4 Not Applicable to ODAM Specifications.
DLCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to DLCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
INFORMATION Oh&L ODAM Specifications Applicability D 3.0 D 3.0 ODAM Specifications Applicability DSR 3.0.1 DSRs shall be met during the MODES or other specified conditions in the Applicability for individual DLCOs, unless otherwise stated in the DSR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the DLCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the DLCO except as provided in DSR 3.0.3.
Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
DSR 3.0.2 The specified Frequency for each DSR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
DSR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the DLCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance.
If the Surveillance is not performed within the delay period, the DLCO must immediately be declared not met, and the applicable Condition(s) must be entered.
When the Surveillance is performed within the delay period and the Surveillance is not met, the DLCO must immediately be declared not met, and the applicable Condition(s) must be entered.
INFORMATION ONL Liquid Effluents Concentration
.1.
D 3.1.1 D 3.1 LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration DLCO 3.1.1 The concentration of radioactive material in water beyond the Site and Exclusion Area Boundary (Figure D2.a-1) due to radioactive liquid effluent shall not exceed:
- a. The concentration specified in 10 CFR Part 20.1302 for radionuclides other than dissolved or entrained noble gases: and
- b.
2 x 10-' pCi/ml total activity concentration for dissolved or entrained noble gases.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A.1 Initiate action to Immediately radioactive material restore concentration beyond the Site and to within limits.
Exclusion Area Boundary due to radioactive liquid effluent exceeds limits.
INFORMATION ONLY Liquid Effluents Concentration
- 1.
D 3.1.1 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.1.1 Perform radioactive liquid waste sampling In accordance and activity analysis.
with Table D3.1.1-1 DSR 3.1.1.2 The analytical results shall be used with In accordance methods in the ODAM to verify that the with ODAM average concentration beyond the Site and Section 2.4 Exclusion Area Boundary does not exceed DLCO 3.1.1 when SR-89. SR-90, and Fe-55 concentrations are averaged over no more than 3 months and other radionuclide concentrations are averaged over no more than 31 days.
t4F-ORMATION ONLY Liquid Effluents Concentration D 3.1.1 Table D3.1.1-1 (Page 1 of 2)
Radioactive Liquid Waste Sampling and Analysis LIQUID SAMPLE LOWER RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE LIMIT OF DETECTION TYPE TYPE FREQUENCY FREQUENCY ANALYSIS (LLD)(h)
Batch Waste Grab Each batch (a)
Each batch (a)
Principal Gamma 5 X 107 pCi/ml (i)
Release Tanks (c) sample Eninitters (X)(k) 1-131 1 X 104 pCi/mli Grab sample One batch/
31 days (b)
Dissolved and I X lo,, pCi/mI 31 days (a)
Entrained Gases (gamma emitters)
Proportional Each batch (a) 31 days (b)
H-3 I X 10' pCi/mi Composite of grab samples (f)
Gross Alpha I X 10' pCi'mI 92 days (b)
Sr-89 5 X 104 pCi/mi Sr-90 5 X 10' pCi/ml Fe-55 I X 104 pCi/mI Plant Service Grab Sample 7 days 7 days (b)
Principal Gamma 5 X 10' pCi/mi (X)
Water Effluent (d)
Emitters (X)(k)
Plant Continuous Proportional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7 days (b)
Principal Gamma 5 X 10'I Ci/m" (l)
Discharge (e)
Composite of Emitters (X) (k)
Grab Samples (g) 1-131 1 X 10' pCi/ml Grab Sample 31 days 31 days (b)
Dissolved and I X 10, PCowmI Entrained Gases (gamma emitters)
Proportional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 31 days (b)
H-3 I X 10' pCi/ml Composite of Grab Samples (g)
Gross Alpha I X I17 pCi/ml 92 days (b)
Sr-89 5 X 10 ' pCi/ml Sr-90 5 X 10' pCilml Fe-55 I X 10 pCi/ml (a)
Complete prior to each release.
(b)
Analysis may be performed after release.
(c)
A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed.
(d)
A grab sample of plant service water effluent shall be analyzed at least once each week in accordance with Table D3.1.1-1, Plant Service Water Effluent. In the event the radioactivity concentration in a sample exceeds 3 X 10.6 IJCi/ml, or in the event the plant service water effluent monitor indicates the presence of an activity concentration greater than 3 X le0 pCi/ml, sampling and analysis according to Table D3.1.1-1, Plant Continuous Discharge, shall commence and shall be performed as long as the condition persists.
(e)
A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g. from a volume of system that has an input flow during the continuous release.
HNITORMIAIION O~NLY Table D3.1.1-1 (Page 2 of 2)
Radioactive Liquid Waste Sampling and Analysis (I)
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released (g) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, daily grab samples shall be collected in proportion to the rate of flow of the effluent stream. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release (h) The LLD is the smallest concentration of the radioactive material in a sample that will be detected with 95% probability (5%
probability of falsely concluding that a blank observation represents a 'real' signal).
For a particular measurement system (which may include radiochemical separation):
LLD-(4 66)(Sb)
(E)(V)(2.22) (Y) e Where LLD is the 'a priori' lower limit of detection as described above (as picocurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2.22 is the number of transformations per minute per picocurie.
Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples)
The value of s b used in the calculation of the LLD for a detection system shall be based on the 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 at shall be used in the calculation.
(i)
For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (I e., 5 X 1 0-'I1 where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the value specified In 10 CFR 20. Appendix B, Table 2, Column 2.
0 The principal gamma emmitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be 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 analysis should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.
(k)
If an isotopic analysis is unavailable, batch releases may be made for up to 14 days provided the gross betalgamma concentration to the unrestricted area Is s 1 x 1 iO pclml and the sample is analyzed when the instrumentation is once again available.
H-WTORMATION ONLY Liquid Waste Concentration D 3.1.2 D 3.1 LIQUID EFFLUENTS D 3.1.2 Liquid Waste Concentration DLCO 3.1.2 The concentration of radioactive materials in liquid wastes from pre-release analysis shall be < 0.01 pCiml, excluding tritium and noble gases.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Concentration of A.1 Appropriate parts of the liquid Prior to liquid waste radioactive materials in radwaste treatment system discharge liquid wastes from pre-shall be used to reduce the release analysis concentration.
> 0.01 pCi/ml, excluding tritium and noble gases.
B.
Required Action and B.1 Prepare and submit a Special 31 days following the associated Completion Report to the NRC pursuant to end of the quarter in Time not met.
Specification D 5.4 that which the limit was identifies equipment or exceeded AND subsystems not OPERABLE and the reason for the Radioactive liquid waste inoperability, action(s) taken to being discharged without restore the inoperable treatment in excess of equipment to OPERABLE 0.01 pCilml, excluding status and a summary tritium and noble gases.
description of the action(s) taken to prevent a recurrence.
- 7LiquMidff QJs 0LY D 3.1.3 D 3.1 LIQUID EFFLUENTS D 3.1.3 Liquid Effluents Dose DLCO 3.1.3 The dose to a Member of the Public due to radioactive material in liquid effluents beyond the Site and Exclusion Area Boundary (Figure D2.a-1) shall be limited to:
- a.
s 1.5 mrem to the total body or < 5.0 mrem to any body organ during any calender quarter; and
- b.
s 3.0 mrem to the total body or < 10.0 mrem to any body organ during any calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Calculated dose due to A.1 Prepare and submit a Special 31 days following the radioactive material in Report, in lieu of any other end of the quarter in liquid effluents beyond report, pursuant to Specification which the limit was the Site and Exclusion D 5.4 to the NRC which exceeded Area Boundary identifies the cause(s) for exceeds the limit.
exceeding the limit(s) and defines the corrective actions to be taken.
I NF 0 R W"AT i 0N t N Pi Liquid Effluents Dose D 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
Calculated dose due to B.1 Prepare and submit a Special 31 days radioactive material in Report, in lieu of any other liquid effluents beyond report, pursuant to Specification the Site and Exclusion D 5.4 to the NRC which: 1 )
Area Boundary defines actions to be taken to exceeds two times the reduce releases and prevent limit.
recurrence, and 2) results of an exposure analysis including effluent pathways and direct radiation to determine whether the dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station during the calender year through the period covered by the calculation was -5 75 mrem to the thyroid and -< 25 mrem to the total body and all other body organs.
INFORMAl IBM GLYDose D 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.3.1 Perform an assessment of compliance with 31 days DLCO 3.1.3.
DSR 3.1.3.2 Project a prospect of compliance with In any quarter in DLCO 3.1.3 for radioactive liquid releases without which Radioactive radwaste system in operation.
liquid releases are made and the radwaste system is not operated CNS ODAM D 3.1-8 10/10/01 I
I VI Outside Temporar y QNtg8cMIJaQRt LitUr-D 3.1.4 D 3.1 LIQUID EFFLUENTS D 3.1.4 Outside Temporary Storage of Radioactive Liquid DLCO 3.1.4 Radioactive liquid contained in unprotected outdoor temporary liquid storage tanks shall conform to the requirements of Technical Specification (TS) 5.5.8.b.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Level of radioactivity A.1 Suspend addition of Immediately exceeds the limits of TS radioactive material.
5.5.8.b.
AND A.2 Begin measures to reduce Immediately content to within the limits of TS 5.5.8.b.
AND A.3 Describe the events leading to In accordance with the condition in the the Radioactive Radioactive Effluent Release Effluent Release Report.
Report frequency I
I I
Outside Temporary JtNo 0a AN IQci.LY D 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.4.1 Sample and analyze radioactive liquid located in 7 days during unprotected outdoor temporary liquid storage tanks addition of for level of radioactivity.
radioactive liquid to the tanks CNS ODAM D 3.1 -1 0 1 0110/01 I
toRMATION ONLY Gaseous Effluents Concentration D 3.2.1 D 3.2 GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Concentration DLCO 3.2.1 The dose rate beyond the Site and Exclusion Area Boundary (Figure D2.a-1) due to radioactive gaseous effluents shall be limited to the following:
- a.
For noble gases, s 500 mrem per year to the total body and
- 3000 mrem per year to the skin; and
- b.
For H-3, 1-131, 1-133, and radioactive material in particulate form with half lives 2 8 days, - 1500 mrem per year to any organ when;
- 1.
The dose rate due to H-3, Sr-89, Sr-90, and alpha emitting radionuclides is averaged over, 3 months and;
- 2.
The dose rate due to other radionuclides is averaged over < 31 days.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Dose rates beyond the A.1 Decrease release rate to Immediately Site and Exclusion Area comply with the limits.
Boundary due to radioactive gaseous effluents exceeds limits.
INFORMATION Outt1%
Gaseous Effluents Concentration
<.D 3.2.1 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.1.1 Perform an assessment of compliance for 31 days DLCO 3.2.1(b).
I Eb9 e6KffAQ9N ONLD D 3.2.2 D 3.2 GASEOUS EFFLUENTS D 3.2.2 Noble Gases Dose DLCO 3.2.2 The air dose beyond the Site and Exclusion Area Boundary (Figure D2.a-1) due to noble gases released in gaseous effluents shall be limited to the following:
- a.
For gamma radiation, < 5 mrad during any calender quarter and s 10 mrad during any calender year; and
- b.
For beta radiation, - 10 mrad during any calender quarter and
- 20 mrad during any calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Calculated air dose due A.1 Prepare and submit a Special 31 days following the to radioactive noble Report pursuant to end of the quarter in gases beyond the Site Specification D 5.4 to the NRC which the limit was and Exclusion Area in lieu of any other report exceeded Boundary exceeds the which identifies the cause(s) limit.
and defines the corrective actions taken.
INF qOej9ip4 ONLY e2 ose D 3.2.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated air dose B.1 Prepare and submit a 31 days due to radioactive Special Report, in noble gases beyond the lieu of any other Site and Exclusion report, pursuant to Area Boundary exceeds Specification D 5.4 two times the limit.
to the NRC which 1) defines actions to be taken to reduce releases and prevent recurrence and 2) results of an exposure analysis including effluent pathways and direct radiation to determine whether the dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station during the calender year through the period covered by the calculation was < 75 mrem to the thyroid and < 25 mrem to the total body or any other body organ.
SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Perform an assessment of compliance for 31 days DLCO 3.2.2.
t hi'RMATION ONLY Iodine and Particulates D 3 2.3 D 3.2 GASEOUS EFFLUENTS D 3.2.3 Iodine and Particulates DLCO 3.2.3 The dose to a Member of the Public due to 1-131, 1-133 and radioactive material in particulate form having a half-life > 8 days in gaseous effluents beyond the Site and Exclusion Area Boundary (Figure D2.a-1) shall be limited to:
- a.
- 7.5 mrem to any organ during any calender quarter; and
- b.
- 15 mrem to any organ during any calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Calculated dose due to I-A.1 Prepare and submit a Special 31 days following the 131, 1-133 and Report, in lieu of any other end of the quarter in radioactive material in report, pursuant to which the limit was particulate form having a Specification D 5.4 to the NRC exceeded.
half-life > 8 days beyond which identifies the cause(s) the Site and Exclusion for exceeding the limit(s) and Area Boundary exceeds describes the corrective action the limit.
taken.
INFORMATION ONLY Iodine-and Particulates D 3.2.3 CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose due to B.1 Prepare and submit a 31 days I-131. I-133 and Special Report, in radioactive material lieu of any other in particulate form report, pursuant to having a half-life Specification D 5.4
> 8 days beyond the to the NRC which 1)
Site and Exclusion defines actions to be Area Boundary exceeds taken to reduce two times the limit.
releases and prevent recurrence and 2) results of an exposure analysis including effluent pathways and direct radiation to determine whether the dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station was < 75 mrem to the thyroid and < 25 mrem to the total body or any other body organ.
SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.3.1 Perform radioactive gaseous waste sampling In accordance and activity analysis on effluents other with than noble gases.
Table D3.2.3-1 DSR 3.2.3.2 Perform a dose assessment to determine 31 days compliance with DLCO 3.2.3.
INFORMATION ONLY Iodine and Particulates D 3.2.3 Table D3.2.3-1 (Page 1 of 3)
Radioactive Gaseous Waste Sampling and Analysis GASEOUS SAMPLE LOWER LIMIT RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE OF DETECTION(LLD)
TYPE TYPE FREQUENCY FREQUENCY ANALYSIS (h)
- 1. Elevated Release Point (ERP)
Grab Sample Grab Sample Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor
- 2. Reactor Building Vent Grab Sample 31 days (c) 31 days (c)
Principal Gamma Emitters (f) 92 days (a) 92 days 11-3 Continuous (b) 7 days (d) 1-131 1-133 Continuous (b) 7 days (d)
Principal Gamma Emitters (1)
(1-131, Others)
Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma) 31 days (c) 31 days (c)
Principal Gamma Emitters (f) 92 days (a) 92 days H-3 Continuous (b) 7 days (d) 1-131 1-133 Continuous (b) 7 days (d)
Principal Gamma Emitters (D (1-131, Others)
Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma) 31 days (c) 31 days (c)
Principal Gamma Emitters (M 92 days (a) 92 days H-3 Grab Sample Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor I X 104 pCi/ml (i)
I X 10-6 pCi/ml I X 10'12 pCi/ml I X 10-10 PCI/ml I X 10" PCuinm (i)
I X 101, pCi/nw I X 10 11 PIC,/nl Ix10 IfpCi/mI I X 10 o PCi/ml I X 106 6pCi/ml I X 10f pCi/al (i)
I X 10 pCi/mI I X o12 PCi/mI I X 10o"IJCi/mI I X 10o" PCi/mI (i)
I X 10711 PUCi/m I X 10" P[Cilm I X10"1 pCi/mI Ix1 XpO6ICi/nil I X 10' pCi/mi (i)
I X 1o-6 Paw I X10 pCi/mI
- 3. Augmented Radwaste Building Vent Grab Sample Grab Sample Charcoal Sample Continuous (b) 7 days (d) 1-131 (continued)
4,G~~
srjid ;
aggt7t
'dodjin e ;c 3n~
yat, PQJNLY D 3.2.3 Table D3.2.3-1 (Page 2 of 3)
Radioactive Gaseous Waste Sampling and Analysis GASEOUS SAMPLE LOWER RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE LIMIT OF TYPE TYPE FREQUENCY FREQUENCY ANALYSIS DETECTION (LLDX1h)
- 3. (continued)
Charcoal Sample Particulate Sample Continuous (b)
Continuous (b) 7 days (d) 7 days (d) 1-133 Principal Gamma Emitters (f)
(1-131, Others)
I X 10o' PC,/mI I X 10.11 pCilmI (i)
Composite Particulate Sample (e)
Noble Gas Monitor
- 4. Turbine Building Vent (Gaseous)
Grab Sample Grab Sample Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma) 31 days (c) 31 days (c)
Pnncipal Gamma Emitters (f) 92 days (a) 92 days H-3 Continuous (b) 7 days (d) 1-131 1-133 Continuous (b) 7 days (d)
Principal Gamma Emitters (f)
(1-1 31, Others)
Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma)
Continuous (b) 7 days (d) 1-131 I X 10"11,UCi/ml I X lo'1 PJC,/MI I X 10-11 PCi/ml I X 0l pCi/mI I X 104 pCI/mI (i) lXI0 - PC,/mI 1 X 10.12 pCi/mI I x 10' 0 PCiml I X 10-, pCI/ml (1)
I X 10-11 pCi/mI I X 1 0lo pCilmI I X ID-II PICIlml I X 106 pCI/mil I X 10-12 pCi/mI I X 10.
PCilml I X lo-,, PCimI (i)
I X lo-II pCi/rn]
I X 10-11 pCi/ml I X 10.11 PCi/ml 5
Multi Purpose Facility (MPF)
Building Vent (Gaseous)
Charcoal Sample 1-133 Particulate Sample Composite Particulate Sample (e)
Continuous (b) 7 days (d)
Principal Gamma Emitters (f)
(1-13 1, Others)
Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha (a) A H-3 grab sample will also be taken when the reactor vessel head is removed. This sample will be taken at the ERP or Reactor Building Vent whichever will be representative dependent upon the head removal vacuum procedure.
(b) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications D 3 2 1, D 3.2 2 and D 3.2 3.
(continued)
INO M'0Q T rP LjAa kitef D 3.2.3 Table D3.2.3-1 (Page 3 of 3)
Radioactive Gaseous Waste Sampling and Analysis (c) Analyses shall also be performed following an increase as indicated by the gaseous release monitor of greater than 50% in the steady state release, after factoring out increases due to power changes or other operational occurrences, which could alter the mixture of radionuclides (d) Analysis shall also be performed following an increase as indicated by the gaseous release monitor of greater than 50% in the steady state release, after factoring out increases due to power changes or other operational occurrences, which could alter the mixture of radionuclides.
When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less are analyzed, the corresponding LLD's may be increased by a factor of 10.
(e) A quarterly composite particulate sample shall include a portion of each week's particulate samples collected during the quarter.
(f) The principal gamma emitters for'which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for the gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions This 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 should not be reported as being present at the LLD level for that nuclide When unusual circumstances cause LLD's higher than required for more than 31 days, the reasons shall be documented in the Radioactive Effluent Release Report (g) The noble gas continuous monitor shall be calibrated using laboratory analysis of the grab samples from Table D3 23-1 or using reference sources (h) The LLD is the smallest concentration of radioactive material in sample that will be detected with 95% probability (5% probability of falsely concluding that a blank observation represents a "real' signal )
For a particular measurement system (which may include radiochemical separation)
(4 66)(s b)
LLD =
(E)(V)(2 22) (Y)e Where LLD is the "a priori" lower limit of detection as described above (as picocurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2 22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),
X is the radioactive decay constant for the particular radionuclide, and t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).
The value of sb used in the calculation of the LLD for a detection system shall be based on the 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 t shall be used in the calculation (i) For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD Under these circumstances, the LLD may be increased inversely proportional to the magnitude of the gamma yield (i e., I x I 0'V, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the values specified in 10 CFR 20, Appendix B, Table 2, Column 1.
X i
-J.a i 2 t A, i I
rh I Offgas Treatment System D 3.2.4 D 3.2 GASEOUS EFFLUENTS D 3.2.4 Offgas Treatment System DLCO 3.2.4 APPLICABILITY:
Gaseous releases discharged through the Offgas Treatment System shall have at least one train of charcoal adsorbers in service.
Main condenser air ejector in service, except during startup or shutdown with reactor < 10% rated power or when system cannot function due to low offgas flow.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Gaseous releases A.1 Restore release of gaseous 7 days discharged without either discharge via charcoal train of charcoal adsorbers.
adsorbers in service.
B.
Required Action and B.1 Prepare and submit a Special 31 days following the associated Completion Report pursuant to end of the quarter in Time not met.
Specification D 5.4 to the NRC which the release which identifies the inoperable occurred equipment and describes the corrective action taken.
I CNS ODAM D 3.2-1 0 1 0/10/01
Offgas teae SYStm0 Y
D 3.2.4 CZIIMIFTI I AWTF PRFniTPFMFNTS I
SURVEILLANCE FREQUENCY I
DSR 3.2.4.1 Verify operation of the Offgas Treatment System charcoal adsorbers by using the gaseous effluent monitoring program in D 3.3.2. Gaseous Effluent Monitoring.
In accordance with the DSR frequencies of D 3.3.2.
DSR 3.2.4.2 Project the prospect of compliance with DLCO 3.2.5. Condition B.
Every 31 days when radioactive material in gaseous effluent is released without treatment.
as.,
MA:N UNSRLYt Exhaust Ventilation Treatment Systems (EVTS)
D 3.2.5 D 3.2 GASEOUS EFFLUENTS D 3.2.5 Exhaust Ventilation Treatment Systems (EVTS)
DLCO 3.2.5 APPLICABILITY:
The Exhaust Ventilation Treatment Systems (EVTS) shall be operated to treat radioactive materials in effluent air.
When radioactive material in gaseous effluent is being released via the associated pathway.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Radioactive material in A.1 Ensure DSR 3.2.5.1 is met.
31 days gaseous effluent released without treatment.
B. Air is discharged without B.1 Prepare and submit a Special 31 days following the treatment for > 31 days.
Report pursuant to end of the quarter in Specification D 5.4 to the which the release AND NRC which identifies the occurred inoperable equipment and The projected dose to a describes the corrective Member of the Public due action taken.
to activity in air effluent via that pathway exceeds 0.3 mrem to any body organ.
I s
) ;,-s I i~
ONE Exhaust Ventilation-Treatment System D 3.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.5.1 Project the prospect of compliance with Every 31 days DLCO 3.2.5.
when radioactive material in gaseous effluent is released without treatment.
Hydrogen Citsncgelfr A
xhN
, r D 3.2.6 D 3.2 GASEOUS EFFLUENTS D 3.2.6 Hydrogen Concentration DLCO 3.2.6 APPLICABILITY:
The concentration of hydrogen in the augmented offgas treatment system downstream of the recombiners shall be limited to < 2% by volume.
During augmented offgas treatment system operation.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Concentration of A.1 Restore concentration to within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hydrogen exceeds limits.
limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.6.1 Verify Hydrogen concentration in the augmented offgas 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> treatment system downstream of the recombiners is within limits.
Primary Containment Ventig A
ia D 3 2.7 D 3.2 GASEOUS RELEASES D 3.2.7 Primary Containment Venting and Purging DLCO 3.2.7 Venting and purging of the primary containment shall be through the Standby Gas Treatment System.
NOTES--------------------
- 1.
This specification does not apply to Normal Ventilation.
- 2.
This specification does not apply during startup while performing primary containment inerting in accordance with Technical Specification 3.6.3.1, Primary Containment Oxygen Concentration, following a shutdown of > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirement not met.
A.1 Suspend all venting and Immediately purging of the primary containment.
INFORMA1iON ONLY Primary Containment Venting and Purging D 3.2.7 SURVETILANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.7.1 The primary containment shall be determined to be aligned for venting or purging through the Standby Gas Treatment System.
Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to venting or purging of the primary containment AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter during venting or purging of the primary containment.
LQiX0ATIPr ONLY D 3.3.1 D 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring DLCO 3.3.1 The liquid effluent radiation monitoring instrumentation channels shown on Table D3.3.1-1 shall be OPERABLE with:
- a.
The minimum OPERABLE channel(s)in service.
- b.
The alarm and trip setpoints set to ensure that the limits of DLCO 3.1.1 are not exceeded.
APPLICABILITY:
According to Table D3.3.1-1.
ACTIONS
- ---------------------- NOTE------
I Separate condition entry is allowed for each channel.
I CONDITION REQUIRED ACTION COMPLETION TIME A.
Liquid effluent radiation A.1 Suspend liquid effluent Immediately monitoring instrumentation radiation release monitored channel alarm and trip by the inoperable channel.
setpoint less conservative than required.
OR A.2 Declare channel inoperable.
Immediately CNS ODAM D 3.3-1 1 0/10/01
INFR 0, Oj0 Y Liquid Efflu3.3.1 INFORI~D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One or more channels B.1 Enter the Condition Immediately inoperable.
referenced in Table D3.3.1-1 for the channel.
AND B.2.1 Restore inoperable 31 days channel(s) to OPERABLE status.
OR B.2.2 In lieu of any other In accordance with report, explain in the the Radioactive Radioactive Effluent Effluent Release Release Report why the Report frequency.
instrument was not repaired in a timely manner.
INFORMATION ONLY Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by Required C.1 Analyze a minimum of 2 Prior to initiating a Action B.1 and referenced independent samples in release in Table D3.3.1-1.
accordance with Table D3.1.1-1.
AND C.2
NOTE------
Determination Action and Verification Action will be performed by two separate technically qualified members of the Facility Staff.
Determine and Prior to initiating a independently verify the release release rate calculations and discharge valving.
D. As required by Required D.1 Collect and analyze a grab 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Action B.1 and referenced sample for gross beta in Table D3.3.1-1.
radioactivity or gross gamma AND radioactivity (as applicable) at a lower limit of detection Once per
< 10 pCi/ml.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter E.
As required by Required E.1 Estimate flow rate during 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action B.1 and referenced actual release.
in Table D3.3.1-1.
AND Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter CNS ODAM D 3.3-3 1 0/10/01 I
INFORMATION ONLY Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F.
Required Action and F.1 Suspend liquid effluent Immediately associated Completion releases monitored by the Time for Condition C or E inoperable channel(s).
not met.
t '
4t r-f-.
A 11 i ()N (t':
E ; I'l Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.1.1 Perform CHANNEL CHECK.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.1.2 Perform CHANNEL CHECK for each channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on any demonstrate OPERABILITY by verifying indication of day on which flow during periods of release.
continuous, periodic, or batch releases are made DSR 3.3.1.3 Perform SOURCE CHECK.
Completed prior to each release DSR 3.3.1.4 Perform SOURCE CHECK.
31 days DSR 3.3.1.5 Perform CHANNEL CALIBRATION 18 months DSR 3.3.1.6 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate automatic isolation of the pathway for instrument indication levels measured above the alarm/trip setpoint and circuit failure; and control room alarm annunciation for instrument indication levels measured above the alarm/trip setpoint, circuit failure and instrument indicating downscale failure.
C;NS ODAM D 3.3-5 10/10/01 I
INFORMATION ONLY Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.1.7 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation for instrument indication levels measured above the alarm/trip setpoint, circuit failure, instrument indicating downscale failure, and instrument controls not set in operate mode.
DSR 3.3.1.8 Perform CHANNEL FUNCTIONAL TEST.
184 days DSR 3.3.1.9 Perform LOGIC SYSTEM FUNCTIONAL TEST 184 days CNS ODAM D 3.3-6 1 0/10101 I
INFORMATION ONLY Liquid Effluent Monitoring D 3.3.1 Table D3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION B.1 REQUIREMENTS
- 1. Gross Beta or Gamma Radioactivity Monilors Providing Automatic Isolation a Liquid Radwaste (a) 1 (b)
C DSR 3.3.1.1 Effluent Line DSR 3.3.1 3 DSR 3.3.1.5 DSR 3.31 6 DSR 3 3.1.9
- 2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm but not Providing Automatic Isolation a Service Water System (a) 1 D
DSR 3.3 1.1 Effluent Line DSR 3 3 1.4 DSR 3 31.5 DSR 33 1.7
- 3. Flow Rate Measurement Devices (a) 1 E
DSR 3 3.1.2
- a. Liquid Radwaste DSR 33 1.5 Effluent Line DSR 3 3 1.8 (a) During releases via this pathway.
(b) Set to alarm and automatically close the waste discharge valve prior to exceeding the limits of DLCO 3.1.1.
I
'MMAT§I 0 N ONLY Gaseous Effluent Monitoring D 3.3.2 D 3.3 INSTRUMENTATION D 3.3.2 Gaseous Effluent Monitoring DLCO 3.3.2 The gaseous effluent radiation monitoring instrumentation channel(s) shown in Table D3.3.2-1 shall be OPERABLE with:
- a.
The minimum OPERABLE channel(s) in service.
- b.
The alarm and trip setpoints set to ensure that the limits of DLCO 3.2.1 are not exceeded.
APPLICABILITY:
According to Table D3.3.2-1.
ACTIONS NOTE Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
Gaseous effluent A.1 Suspend gaseous effluent Immediately radiation monitoring radiation release monitored instrumentation by inoperable channel.
channel alarm and trip setpoint less OR conservative than required.
A.2 Declare channel inoperable.
Immediately CNS ODAM D 3.3-8 1 0110/01
Gaseous MTtQg1 ONLY D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
One or more channels B.1 Enter the Condition Immediately inoperable.
referenced in Table D3.3.2-1 for the channel.
AND B.2.1 Restore inoperable 31 days channel(s) to OPERABLE status.
OR B.2.2 In lieu of any other report, In accordance with explain in the Radioactive the Radioactive Effluent Release Report Effluent Release why the instrument was not Report frequency repaired in a timely manner.
C.
As required by C.1 Ensure the offgas delay Immediately Required Action B.1 system is not bypassed.
and referenced in Table D3.3.2-1.
AND C.2 Ensure the Elevated Immediately Release Point Monitoring noble gas activity monitor is OPERABLE.
AND C.3 Restore inoperable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> channels to OPERABLE status.
D.
Required Action and D.1 Be in MODE 2.
12 hburs associated Completion Time for Condition C not met.
t.uSMU tw N hI V en o
Gaseous Ef uent onitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.
As required by E.1 Estimate flowrate.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action B.1 and referenced in AND Table D3.3.2-1.
Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter F.
As required by F.1 Take grab samples.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action B.1 and referenced in AND Table D3.3.2-1.
Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter AND F.2 Analyze for gross activity.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of sampling completion CNS ODAM D 3.3-10 1 0110/01I
it-MT"
`A0O ONLY Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION I
REQUIRED ACTION I COMPLETION TIME G.
As required by Required Action B.1 and referenced in Table D3.3.2-1.
G.1.1 Verify one Function 2.a monitor OPERABLE.
AND G.1.2 Monitor recombiner exhaust temperature.
OR G.2.1.1 Collect gas sample.
AND G.2.1.2 Analyze gas sample.
AND G.2.2.1 Verify one Function 2.a monitor OPERABLE.
OR G.2.2.2 Monitor recombiner exhaust temperature.
Immediately Immediately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from time of sampling completion Immediately Immediately H.
Required Action and H.1 Discontinue operation of Immediately associated Completion the augmented offgas Time for Condition G treatment system.
not met.
5 " ' i,, 1
/
i 1'
2 3 be ~ tI r, k! U N LY Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I.
As required by Required Action B.1 and referenced in Table D3.3.2-1.
1.1
NOTE ------
When the primary monitoring system is inoperable and the backup system is in service, sampling may be discontinued for up to 30 minutes only for changing particulate filters and iodine cartridges.
Continuously collect samples with auxiliary sampling equipment as required in Table D3.2.3-1.
OR 1.2.1 If auxiliary sampling equipment cannot be established within the specified completion time, enter the problem into the Corrective Action Program to evaluate particulate and iodine effluent releases.
AND 1.2.2 Report this event in the Radioactive Effluent Release Report.
4 Hours Immediately In accordance with the Radioactive Effluent Release Report Frequency J.
Required Action and J.1 Discontinue effluent Immediately associated Completion releases via this pathway.
Time for Condition E, F, or I not met.
INFORMATION ONLY Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME K.
Function 1.a trip K.1 Close the offgas isolation Immediately capability not valve.
maintained.
AND AND K.2 Initiate reactor shutdown.
Immediately Radiation level exceeds 1.0 ci/sec AND (prior to 30 min. delay line) for > 15 K.3 Be in MODE 4.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> consecutive minutes.
it'4@T~g5ZMoni0torin Gaseous f uluen on itonng~
D 3.3.2 SURVEILLANCE REQUIREMENTS
NOTES --------------------------------------
- 1.
Refer to Table D3.3.2-1 to determine which DSRs apply for each instrument.
- 2.
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function is maintained.
SURVEILLANCE FREQUENCY DSR 3.3.2.1 Perform CHANNEL CHECK.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.2.2 Perform CHANNEL CHECK.
7 days DSR 3.3.2.3 Perform SOURCE CHECK.
31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST.
31 days DSR 3.3.2.5 Perform SOURCE CHECK.
92 days DSR 3.3.2.6 Perform CHANNEL CALIBRATION. The CHANNEL 92 days CALIBRATION shall include the use of a standard gas sample containing a percentage of hydrogen to verify accuracy of the monitoring channel in its operating range.
DSR 3.3.2.7 Perform CHANNEL FUNCTIONAL TEST.
92 days DSR 3.3.2.8 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint, circuit failure, instrument indicates a downscale failure, or instrument controls not set in operate mode.
f Is i V
V
\\<l5 h
IWrs f~K'
~2 I
Gaseous Effluent Monitoring D 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.2.9 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint or circuit failure.
DSR 3.3.2.10 Perform CHANNEL CALIBRATION. For 18 months Function 1.a, the time delay setting for closure of the steam jet air ejector isolation valves shall be
- 15 minutes and trip settings shall correspond to Technical Specification 3.7.5.
DSR 3.3.2.11 Perform CHANNEL FUNCTIONAL TEST. The 18 months CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint, circuit failure, instrument indicates a downscale failure, or instrument controls not set in operate mode.
DSR 3.3.2.12 Perform LOGIC SYSTEM FUNCTIONAL TEST.
18 months CNS ODAM D 3.3-15 03/20/03
Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 1 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION B 1 REQUIREMENTS
- a. Noble Gas Activity (a) 1(e)
C DSR 3 3 2.1 Monitor DSR 3 3 2 3 DSR 3328 DSR 3 3 2 1 0 DSR 3 3 2.11 DSR 3 3 2.12 b Effluent System Flow (b) 1 E
DSR 3 32.1 Rate Measuring Device DSR 3 3 2.7 DSR 3 3 2.10
- 2. Augmented Offgas Treatment System Explosive Gas Monitoring System a Hydrogen Monitor (c) 2 G
DSR 33 2 1 (2% monitor)
DSR 3324 DSR 3 32 6 3
Reactor Building Ventilation Monitoring System
- a. Noble Gas Activity (b) (f) 1 F
DSR 3 3 2 1 Monitor DSR 3 3 2.3 DSR 3 3 2 9 DSR 3 3 2 1 0 b Iodine Sampler (b)(I) 1 I
DSR 33 2.2 Cartridge C Particulate Sampler (b) (1) 1 I
DSR 3.3 2 2 Filter d Effluent System Flow (b) 1 E
DSR 3 3 2 1 Rate Measunng Device DSR 3 3 2 7 DSR 3 3.2.10 e Sampler Flow Rate (b) 1 E
DSR 3 3 2.1 Measuring Device DSR 3 3 2.7 DSR 3 3 2.10
- f. Isolation Monitor (d)
(d)
(d)
DSR 3.3 2.5 DSR 3 3 2 11 (continued)
(a) During operation of the steam jet air ejector (b) During releases via this pathway (c) During augmented offgas treatment system operation (d) See Technical Specification 3.3 6.2 (e) Second channel must either be OPERABLE or be in the tripped condition (f) A channel may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for low flow alarm check without entering Conditions or Required Actions CNS ODAM D 3.3-16 03120/03
Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 2 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITION OPERABLE ACTION B.1 REQUIREMENTS
- 4. Elevated Release Point Monitoring System
- a. Noble Gas Activity Monitor
- b. Iodine Sampler Cartridge C. Particulate Sampler Fillter
- d. Effluent System Flow Rate Measuring Device
- e. Sampler Flow Rate Measuring Device (b) (f) 1 (b) (f) 1 (b) (f) 1 (b)
(b) 1 1
F E
E F
- 5. Radwaste Building Ventilation Monitoring System a Noble Gas Activity (b) (f)
Montor 1
DSR 3.3 2.1 DSR 332.3 DSR 3.3 2.9 DSR 3.3 2.10 DSR 3 3 2.2 DSR3 322 DSR 3.3 2.1 DSR 3 3 2.7 DSR 3 3 2.10 DSR 3 3 2 1 DSR 3327 DSR 3 3 2 10 DSR 3 3 2.1 DSR 3.3 2 3 DSR 3 3 2.9 DSR 3.3 2 10 DSR 3 3.2 2 DSR 3 3.2 2 DSR 3 3 2.1 DSR 3 3.2.7 DSR 3 3 2.10 DSR 3 3 2.1 DSR 3.3.2.7 DSR 3.3 2.10 DSR3321 DSR 3.3.2 3 DSR 3.3.2 9 DSR 332 10 DSR 332.2 I
I I
- b. Iodine Sampler (b)
Cartridge
- c. Particulate Sampler (b)
Filter
- d. Effluent System Flow (b)
Rate Measuring Device (f) 1 I
I (9
1 I
1 E
E
- e. Sampler Flow Rate Measuring Device
- 6. Turbine Building Ventilation Monitoring System
- a. Noble Gas Activity Monitor
- b. Iodine Sampler Cartridge (b)
(b) (f) 1 1
F I
I (b) (f) 1 I
(continued)
(b) During releases via this pathway (f) A channel may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for low flow alarm check without entering Conditions or Required Actions I
Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 3 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNFLS FROM SRi I
LL IANCE INSTRUMENT CONDITIK
- 6. (continued)
- c. Particulate Sampler (b)
Filter
- d. Effluent System Flow (b)
Rate Measunng Device
- e. Sampler Flow Rate (b)
Measuring Device
- 7. Multi Purpose Facility (MPF)
Building Ventilation Monitoring System a Iodine Sampler (b) (f)
Cartridge b Particulate Sampler (b) (f)
Filter
- c. Effluent System Flow (b)
Rate Measuring Device
- d. Sampler Flow Rate (b)
Measuring Device ON OPERABLE ACTION B.1 REQUIREMENTS 1
E 1
E DSR3322 DSR 3 3.2 1 DSR 3 3.2.7 DSR 3 3.2.10 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 33 2.2 DSR 3 3.2.2 DSR 3.3.2.1 DSR 3 3.2 7 DSR3.3210 DSR 3.3 2.1 DSR 3.327 DSR 332 10 I
1 1
1 E
I I
1 E
(b) During releases via this pathway (f) A channel may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for low flow alarm check without entering Conditions or Required Actions.
Liquid/Gaseous uents Dose D 3.4.1 D 3.4 LIQUIDIGASEOUS DOSE D 3.4.1 Liquid/Gaseous Effluents Dose DLCO 3.4.1 The dose or dose commitment to an actual Member of the Public due to radiation and radioactive releases from Cooper Station shall be limited to
< 75 mrem to the thyroid and, 25 mrem to the total body or any other body organ during a calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Estimated dose or dose A.1 Verify the condition resulting in Immediately commitment due to doses exceeding these limits is radiation and radioactive corrected.
releases exceeds limits.
B.
Required Action and B.1 ------
NOTE---
associated Completion This is the Special Report Time not met.
required by D 3.1.2, D 3.1.3, or D 3.2.3 supplemented with the following.
Submit a Special Report 31 days pursuant to Specification D 5.4, including information specified in 40 CFR Part 190.11(b). This submission shall be deemed a timely request for variance in accord with provisions of 40 CFR Part 190. The variance is granted until NRC staff action on the item is complete.
Liquid/Gaseous Mfluehls Dose,
D 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.4.1.1 Perform a cumulative dose calculation due to 12 months radioactive material in gaseous and liquid effluents to determine compliance with DLCO 3.4.1.
Solid adoac ive as e D 3.5.1 D 3.5 SOLID RADIOACTIVE WASTE D 3.5.1 Solid Radioactive Waste DLCO 3.5.1 The appropriate equipment of the solid radwaste system shall be OPERABLE to process radioactive waste containing liquid and liquid destined for disposal subject to 10 CFR Part 61 to a form that meets applicable requirements of 10 CFR Part 61.56 before the waste is shipped from the site.
APPLICABILITY:
During solid radwaste processing.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Container of waste does A.1 Suspend delivery to a carrier Immediately not comply with 10 CFR for transport.
Part 61.56.
ir N;3 -0 Rd' M v 1 0 N ONLY Solid Radioactive Waste D 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.1.1 Sample and analyze the dewatered radioactive waste Prior to for pH.
solidification of every 10th batch of dewatered waste.
DSR 3.5.1.2 Inspect solidified or dewatered radioactive waste to Prior to capping insure that there is no free standing liquid on top of each drum or High the solid waste.
Integrity Container (HIC)
DSR 3.5.1.3 Record the following information for radioactive solid In accordance waste shipped offsite during the report period for the with Radioactive Effluent Release Report per 10 CFR 50.36a the Reporting Requirements in Technical Specification 5.6.3; a)
Container burial volume, b)
Total curie quantity (determined by measurement or estimate),
c)
Principal gamma radionuclides (determined by measurement or estimate),
d)
Type of waste, and e)
Solidification agent.
, N <0 W
0 i gl R' ~ON ONL"V Monitoring Program Compliance D 4.1 D 4.0 MONITORING PROGRAM D 4.1 Monitoring Program Compliance DLCO 4.1 The radiological environmental monitoring program shall be conducted as specified in Table D4.1-1, using analytical techniques such that the detection capabilities in Table D4.1-2 are achieved.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Radiological A.1 Prepare and submit to the May 15th following environmental monitoring NRC in the Annual the end of the year program not conducted Radiological Environmental as specified in Report the reasons for not Table D4.1-1.
conducting the program in accordance with Table D4.1-1 and the plans for preventing recurrence.
B. Environmental sampling B.1 Report in the Annual May 15th following medium is not available Radiological Environmental the end of the year from a sampling location Report the cause and location as specified in where replacement samples Table D4.1-1.
were obtained.
~'I p0 r4 ' PAR IF'tN 0N CiU Monitoring Program Compliance I
D 4.1 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 4.1.1 Perform radiological environmental sampling In accordance and analysis.
with Table D4.1-1 DSR 4.1.2 Conduct a land use census to identify the 12 months location of the nearest garden that is greater than 500 square feet in area and that yields edible leafy vegetables, the location of the nearest milk animal, and the location of the nearest resident in each of the 16 meteorological sectors within 3 miles of the Station.
DSR 4.1.3 Summarize results of radiological May 15th environmental analysis in the Annual following the Radiological Environmental Report.
end of the year DSR 4.1.4 Submit results of the land use census in May 15th the Annual Radiological Environmental following the Report.
end of the year CNS ODAM D 4.1-2 REVISION 0
'11
- 4.
St IR ai
-vi
.1 4}
I N
0, e
Monitoring Program Compliance D 4.1 Table D4.1-1 (Page 1 of 2)
Radiological Environmental Monitoring Program EXPOSURE PATHWAY NUMBER OF SAMPLING AND TYPE AND FREQUENCY ANDJOR SAMPLE SAMPLE STATIONS COLLECTION FREQUENCY OF ANALYSIS
- 1. Airborne
- a. Radioiodine and Particulate
- 2. Direct Radiation At least 5 locations At least 32 locations Continuous operation of sampler with sample collection as required by dust loading but at least once per 7 days.
Thermoluminescent Dosimeters (TLD)(b) exchange and read-out at least once per 92 days.
Radioiodine canister: Analyze at least once per 7 days for 1-131.
Particulate sample: Analyze for gross beta radioactivity X 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following filter change. Perform gamma isotopic(') analysis on each sample in which gross beta activity is >10 times the yearly mean of control samples.
Perform gamma isotopic()
analysis on composite (by location) sample at least once per 92 days Gammna dose: At least once per 92 days.
- 3. Waterborne
- a. River Water b.Ground Water
- c. Sediment from Shoreline At least 2 locations At least 2 locations At least I location Collect a one (I) gallon grab sample at least once per 31 days.
Collect a one (1) gallon grab sample at least once per 92 days.
Two (2) times a year, once in the spring and once in the fall.
Gamma isotopic>') analysis of each sample. Composite grab sample for tritium analysis at least once per 92 days.
Gamma isotopic<') and tritium analysis of each sample.
Gamma isotopic<') analysis of each sample.
- 4. Ingestion
- a. Milk (nearest producer)
At least I location At least once per 15 days during Peak Pasture PeriodSc); at least once per 31 days at other times.
At least once per 92 days.
Gamma isotopict ') and 1-131 analysis of each sample.
- b. Milk (other producers)
At least 2 locations Gamma isotopic(') and 1-131 analysis of each sample.
Gamma isotopic(') analysis on edible portions.
- c. Fish At least 2 locations Two times per year (once in the summer and once in the fall).
Attempt to include the following:
I. Bottom feeding species
- 2. Middle-Top feeding species (continued)
Monitoring Program Compliance D 4.1 Table D4.1-1 (Page 2 of 2)
Radiological Environmental Monitoring Program EXPOSURE PATHWAY NUMBER OF SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE STATIONS COLLECTION FREQUENCY OF ANALYSIS
- 4. (continued) d Food Products Samples of three different Monthly when available Gamma isotopic(') and 1-131 kinds of broad leaf analysis vegetation grown nearest each of two different offsite locations of highest predicted annual average ground-level D/Q if milk sampling is not performed One sample of each of the Monthly when available Gamma isotopic() and 1-131 similar broad leaf analysis vegetation grown 15-30 km distant in the least prevalent wind direction if milk sampling is not performed.
(a) Gamma isotopic analysis refers to high resolution gamma spectrum analysis as follows the sample is scanned for gamma-ray activity If no activity is found for a selected nuclide, the detection sensitivity for that nuclide will be calculated using the counting time, detector efficiency, gamma energy, geometry, and detector background appropriate to the particular sample in question The following nineteen (19) nuclides shall be analyzed routmely I
BaLa-140 Be-7 Ce-141 Ce-144 Co-58 Co-60 Cs-134 Cs-137 Fe-59 1-131 K-40 Nb-95 Mn-54 Ra-226 Ru-103 Ru-106 Th-228 Zn-65 Zr-95 Any radionuclide detected, i e, having a measured concentration greater than the LLD, whether or not it is one of the 19 nuclides listed above, shall be regarded as present in the sample.
(b) Thermoluminescent Dosimeters (ILD) is a single pbosphore. Two or more phosphores in one package are considered to be two or more dosimeters (c) Peak Pasture Period is June I through September 30 of each year.
Monitoring Program Compliance D 4.1 Table D4.1-2 (Page 1 of 2)
Detection Capabilities for Environmental Sample Analysis LOWER LIMIT OF DETECTION"') (LLD) ("
AIRBORNE PARTICULATE FOOD WATER OR GAS FISH MILK PRODUCTS SEDIMENT ANALYSIS (pCI/I)
(puimn)
(pCi/kg, wet)
(pCI/i)
(pCi/kg, wet)
(pCi/kg. dry) gross beta 4
1 x 10-2 H-3 2000 Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 1-131 1 ()
7 x 10-1 60 Cs-134 15 5 x 10 130 15 60 150 Cs-137 18 6x 10 2
150 18 80 180 Ba-140 60 60 La-140 15 15 (a) This 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 CNS ODAM D 4.1-5 8126199
Monitoring Program Compliance D 4.1 Table D4.1-2 (Page 2 of 2)
Detection Capabilities for Environmental Sample Analysis (b)
The LLD is the 'a priori" smallest concentration of radioactive material in a sample that will be detected with 95% probability (5% probability of falsely concluding that a blank observation represents a 'real" signal).
For a particular measurement system (which may include radiochemical separation):
(4.66)(s b)
=
(E)(V)(2 22) (Y) (e
)
Where:
LLD is the "a priori" lower limit of detection as described above (as picocurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute).
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2 22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable)
A is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between sample collection (or midpoint of the sample collection period) and time of counting.
The value of sb used in the calculation of the LLD for a detection system shall be based on the 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 (e g.,
potassium-40 in milk samples)
Analysis shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Report.
(c)
LLD for drinking water.
Monitoring Program Concentration D 4.2 D 4.0 MONITORING PROGRAM D 4.2 Monitoring Program Concentration DLCO 4.2 Radioactivity concentrations in sampled medium from the radiological environmental monitoring program shall not exceed values specified in Table D5.4-1 when averaged over a calender quarter.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Radioactivity A.1 Prepare and submit to the 31 days following the concentrations of NRC a Special Report in end of the quarter sampled medium from accordance with Specification the radiological D 5.4 which includes an environmental monitoring evaluation of any release program exceeds values conditions, environmental specified in Table D5.4-1, factors or other conditions averaged over a calender which caused the value(s) to quarter which is be exceeded.
attributable to release(s) from the Station.
B. Radioactivity B.1 Report and explain in the May 15th following concentrations of Annual Radiological the end of the year sampled medium from Environmental Report the the radiological results of the sample(s).
environmental monitoring program exceeds values specified in Table D5.4-1, averaged over a calender quarter which is not attributable to release(s) from the Station.
l R§ Si l
t_
i j
o I
A Monitoring Program Dose D4.3 D 4.0 MONITORING PROGRAM D 4.3 Monitoring Program Dose DLCO 4.3 The calculated personal dose associated with sampled exposure pathway(s) shall not exceed 120% of the calculated dose at the maximum dose location associated with like pathways at a location where sampling is conducted as specified in Table D4.1-1.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Location(s) identified at A.1
NOTE-------
which the calculated dose Only applicable if associated with the samples are reasonably exposure pathway(s) attainable at the new exceeds 120% of the location.
calculated dose at the maximum dose location associated with like Add new sampling 92 days pathways at a location location(s) identified where sampling is having maximum conducted as specified in exposure potential to the Table D4.1-1.
radiological environmental monitoring program and Table D4.1-1.
AND A.2 Describe change made to May 15th following Table D4.1-1 in the the end of the year Annual Radiological Environmental Report.
InterlaboratKg1KAWR\\
ONLY D 5.1 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.1 Interlaboratory Comparison Program DLCO 5.1 Analyses shall be performed on radioactive materials supplied as part of the Interlaboratory Comparison Program.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Analyses not performed.
A.1 Report to the NRC in the May 15th Annual Radiological following the end of Environmental Report the the year corrective actions taken to prevent recurrence.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 5.1.1 Submit a brief summary of the results obtained from May 15th the Interlaboratory Comparison Program in the following the end Annual Radiological Environmental Report pursuant of the year to Technical Specification 5.6.2.
INFORMATION ONL-.
Annual Radiological Environmental Report D 5.2 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.2 Annual Radiological Environmental Report The Annual Radiological Environmental Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Assessment Manual (ODAM), and in 10 CFR50, Appendix I,Section IV.B.2, IV.B.3, and IV.C.
The Annual Radiological Environmental Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODAM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in Regulatory Guide 4.8, December 1975. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
Details of the report shall include the following:
- a.
A summary of doses to a Member of the Public beyond the Site and Exclusion Area Boundary due to Cooper Nuclear Station aqueous and airborne radioactive effluents, calculated in accordance with methods compatible with the ODAM.
- b.
A summary of the results of the land use census required in DSR 4.1.2.
- c.
Summarized and tabulated results in the format of Table D5.2-1 of analysis of samples required by the radiological environmental monitoring program, and taken during the report period.
- d.
A summary description of the radiological environmental monitoring program including any changes; a map of all sampling locations keyed to a table giving distances and directions from the reactor; and, the results of participation in the Interlaboratory Comparison Program, required by D 5.1.
- e.
Summaries, interpretations, and analysis of trends of the results of the radiological environmental monitoring program for the reporting period.
TABLE jJ5.2-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM
SUMMARY
Name of Facility Cooper Nuclear Station Docket No.
50-298 Location of Facility Nemaha. Nebraska (County, State)
Reporting Period Type &
Lower Limit All Indicator Control Medium of Pathways Total No.
of Locations Location with Highest Annual Mean Locations No. of Sampled of Analyses Detection(l)
Meant](2)
Name Mean[](2)
Mean[](2) Reportable (Unit of Measurement)
Performed (LLD)
Ranae (2)
Distance & Direction Range (2) Range (2) Occurrences ZM Table Notes:
(1)
Nominal Lower Limit of Detection (LLD).
(2)
Mean and Range based upon detectable measurements only.
Fraction of detectable measurements at specified locatin indicated in brackets [.
I
Annual Radioactive Effluent Release Report D 5.3 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.3 Radioactive Effluent Release Report The Radioactive Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I,Section IV.B.1.
The Radioactive Effluent Release Report shall be submitted to the NRC by May 1 of each year and shall include the following:
- a.
A summary by calendar quarter of the quantities of radioactive liquid and gaseous effluents released from the Station, reported in the format recommended in Regulatory Guide 1.21, Appendix B, Tables 1 and 2.
- b.
A summary of radioactive solid waste shipped from CNS, including information provided in DSR 3.5.1.3.
- c.
A summary of meteorological data collected during the year.
- d.
A list and brief description of each unplanned release of gaseous or liquid radioactive effluent that causes a limit in DLCO 3.1.1, DLCO 3.1.3, DLCO 3.2.1, DLCO 3.2.2 or DLCO 3.2.3 to be exceeded.
- e.
Calculated offsite dose to humans resulting from the release of effluents and their subsequent dispersion on the atmosphere reported in accordance with Regulatory Guide 1.21.
- f.
A summary of changes made to the CNS Process Control Program. The summary shall contain sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; a determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and documentation of the fact that the change has been reviewed and found acceptable by the SORC.
Annual Radioactive Effluent Release Report D 5 3 D 5.3 Radioactive Effluent Release Report (continued)
- 9.
Changes made to the CNS Offsite Dose Assessment Manual shall be submitted to the NRC in the form of a complete, legible copy of the entire ODAM as part of, or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODAM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and l
year) the change was implemented.
INFORMATION ONLY Special Reports D5.4 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.4 Special Reports Special reports shall be submitted to the Director, Nuclear Reactor Regulation, USNRC, Washington, D.C. 20555 and to the NRC Regional Administrator within the time period specified for each report.
Special reports (in lieu of Licensee Event Reports) may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Offsite Dose Assessment Manual.
A special report is required if measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table D5.4-1 when averaged over any calendar quarter sampling period. When more than one of the radionuclides in Table D5.4-1 are detected in the sampling medium, this report shall be submitted if:
Concentration (1) + Concentration (2) +
Limit Level (1) Limit Level (2)
When radionuclides other than those in Table D5.4-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of DLCO 3.1.3 and 3.2.3.
This report is not required if the measured 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 Report.
2--
4 1 N '
IAI *4.
0 js/i t9 2k~g i Special Reports D 5.4 Table D5.4-1 Reporting Levels for Radioactivity Concentrations in Environmental Samples Reporting Levels AIRBORNE PARTICULATE OR FISH FOOD PRODUCTS ANALYSIS WATER (pCi/i)
GASES (pCi/mr)
(pCi/Kg, Wet)
MILK (pCt/I)
(pCi/Kg, Wet)
H-3 2E + 4(a) 3E + 4(c)
Mn-54 IE + 3 3E + 4 Fe-59 4E + 2 IE + 4 Co-58 IE+3 3E+4 Co-60 3E+2 IE+4 Zn-65 3E+2 2E+4 Zr-Nb-95 4E + 2(b) 1-131 2
0.9 3
IE+2 Cs-134 30 10 IE + 3 60 IE+3 Cs-137 50 20 2E + 3 70 2E + 3 Ba-La-140 2E + 2(b) 3E + 2(b)
(a)
(b)
(c)
For drinking water samples. This is the 40 CFR 141 value.
Concentration of parent or daughter.
For samples of water not used as a source of drinking water.
INFORMATION ONLYfv Major Changes to Radioactive Waste Treatment Systems D 5.5 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid)
The radioactive waste treatment systems (liquid, gaseous. and solid) are those systems described in the facility Safety Analysis Report and amendments thereto, which are used to maintain that control over radioactive materials in gaseous and liquid effluents and in solid waste packaged for offsite shipment required to meet the DLCO's set forth in Specifications D 3.1.1, D 3.1.2, D 3.1.3, D 3.1.4, D 3.2.1, D 3.2.2. D 3.2.3. D 3.2.4. D 3.2.5, D 3.2.6, D 3.2.7, D 3.3.2. D 3.4.1.
and D 3.5.1. The NRC is notified of major changes to these systems under the provisions of 10 CFR Part 50.59 and Part 50.71 (USAR revisions).
INFORMATION ONLY OFFS1TE DOSE ASSESSMENT MANUAL APPENDIX D TABLE OF CONTENTS - BASES B 3.0 ODAM SPECIFICATION APPLICABILITY.........
.................. B 3.0-1 B 3.1 LIQUID EFFLUENTS B 3.1.1 Liquid Effluents Concentration..............................
B 3.1-1 B 3.1.2 Liquid Waste Concentration....................
B 3.1-2 B 3.1.3 Liquid Effluents Dose.....................................
B 3.1-5 B 3.1.4 Outside Temporary Storage of Radioactive Liquid......
......... B 3.1-9 B 3.2 GASEOUS EFFLUENTS B 3.2.1 Gaseous Effluents Concentration............................
B 3.2-1 B 3.2.2 Noble Gas Dose...........
B 3.2-2 B 3.2.3 Iodine and Particulates................................
B 3.2-3 B 3.2.4 Offgas Treatment System...............................
B 3.2-4 B 3.2.5 Exhaust Ventilation Treatment System............
........... B 3.2-5 B 3 2.6 Hydrogen Concentration...............................
B 3.2-6 B 3 2.7 Primary Containment Venting and Purging.........
........... B 3.2-7 B 3.3 INSTRUMENTATION B 3.3.1 Liquid Effluent Monitoring..................................
B 3.3-1 B 3.3.2 Gaseous Effluent Monitoring.......
B 3.3-2 B 3.4 LIQUID/GASEOUS EFFLUENTS DOSE B 3.4.1 Liquid/Gaseous Effluents Dose.............................
B 3.4-1 B 3.5 SOLID RADIOACTIVE WASTE B 3.5.1 Solid Radioactive Waste.................
B 3.5-1 B 4.0 MONITORING PROGRAM B 4.1 Monitoring Program Compliance.........................
B 4.1-1 B 4.2 Monitoring Program Concentration.........................
B 4.2-1 B 4.3 Monitoring Program Dose............
B 4.3-1 B 5.0 MISCELLANEOUS PROGRAMSIREPORTS B 5.1 Interlaboratory Comparison Program.........................
B 5.1-1 B 5.2 Annual Radiological Environmental Report.........
........... B 5.2-1 B 5.3 Radiological Effluent Release Report.........................
B 5.3-1 B 5.4 Special Reports...........
B 5.4-1 B 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid)
INFORM, i1 N'
ODAM Specification Applicability B 3.0 B 3.0 ODAM Specification Applicability BASES DLCOs 3.0.1, 3.0.2, and 3.0.5, and DSRs 3 0.1, 3.0.2, and 3.0.3 reflect parallel requirements in the Technical Specifications. Refer the Technical Specification Bases for appropriate discussion.
ODAM Specification DLCO 3.0.3, in lieu of imposing a plant shutdown as paralleled in Technical Specification 3.0.3, requires: (a) an Action to initiate efforts to restore compliance with the l
ODAM or associated Actions; and (b) an Action that requires entering the circumstances into the Corrective Action Program (CAP). These actions ensure that the appropriate actions continue l
to be focused on and that the circumstances concerning failure to comply with the ODAM 1
Actions would be reviewed. This review will be conducted in accordance with the procedural I
guidance for CAP Notifications.
There are no ODAM 3.0 Specifications that parallel Technical Specification LCO 3.0.4 or SR 3.0.4. Restrictions in entering MODES or other specified conditions in the Applicability have historically not been applied to ODAM Specifications. There are also no ODAM 3.0 Specifications that parallel Technical Specification LCO 3.0.6 and LCO 3.0.7, which allow for exceptions and revisions of other Technical Specifications. They are not applicable to the ODAM since it is not permitted to allow the ODAM to revise a Technical Specification.
(Note that there currently are no identified ODAM DLCOs that support Technical Specification systems; however, this discussion is presented to address the philosophy that would be applied.) An allowance similar to Technical Specification LCO 3.0.6 does not apply to the I
ODAM. When a Technical Specification supported system LCO is discovered to be not met solely due to a ODAM support system DLCO not met, appropriate Technical Specification ACTIONS are required to be entered immediately. This applies even in instances where the ODAM contains a delay prior to declaring a Technical Specification supported system inoperable. In this case, certain ODAM inoperabilities may not directly impact the OPERABILITY of the Technical Specification supported system and delayed declaration of inoperability of the supported system is acceptable. In other cases, discovered support system inoperabilities that directly result in supported system inability to perform the safety function, should result in immediate declaration of inoperability of the supported system.
Technical Specification LCO 3.0.7 has no parallel in the ODAM since it provides for explicit changes to specified Technical Specifications by the Section 3.10 Specifications. However, in the event that LCO 3.0.7 provides for changes to the Technical Specification MODE definitions by the Section 3.10 Specifications, the revised MODE definitions apply to all plant references, including ODAM references.
INFORMATI N ONLY Liquid Effluents Concentration B 3.1.1 B 3.1 LIQUID EFFLUENTS B 3.1.1 Liquid Effluents Concentration BASES This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20.1302. 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 IV.A guides on technical specifications in Appendix I, 10 CFR Part 50, for an individual and (2) the limits of 10 CFR Part 20.1301 and 20.1302(b)(2)(i) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
Since Service Water is not a normal or expected source of significant radioactive release, routine sampling and monitoring for radioactivity is precautionary. An activity concentration of 3 x 10-6 pCi/ml in Service Water effluent is diluted in the discharge canal to about 1.5% of the 10 CFR 20 Appendix B Table 2 Column 2 concentration with only one circulating water pump operating. During normal Station operation the dilution would be even greater. By monitoring Service Water effluent continuously for radioactivity and by confirmatory sampling weekly, reasonable assurance that its activity concentration can be kept to a small fraction of the 10 CFR Part 20.1302 limit and within the Specification D 3.1.3 limit is provided.
By monitoring Service Water continuously and liquid radwaste continuously during discharge with the monitor set to alarm or trip before the limit specified in 10 CFR 20.1302 is exceeded, reasonable assurance of compliance with Specification D 3.1.1 is provided. Verification that radioactivity in liquid effluent averaged only a small fraction of the concentration limit is provided by calculations demonstrating compliance with Specification D 3.1.3.
Compliance with 10 CFR Part 20.1302(b)(2)(i) implies that the concentration limit represented by 10 CFR Part 20, Appendix B, Table 2 will be met within a suitable and reasonable averaging time for assessing compliance. That averaging time is dependent upon the resolving time of the measurements or estimates which are used to evaluate compliance. Assessment of compliance is done by sampling and analysis according to DSR 3.1.1.2. by estimating or measuring the maximum release flow and the minimum dilution flow coincident during the period of release represented by the sample, and by computing the concentration as a fraction of the limit beyond the site and exclusion area boundary periodically on the basis of these data.
Reporting by Special Reports and other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Liquid Waste Concentration B 3.1.2 B 3.1 LIQUID EFFLUENTS B 3.1.2 Liquid Waste Concentration BASES Specification D 3.1.2 implements the requirements of 10 CFR Part 50.36a(a)(1) that operating procedures be established and followed and that equipment be maintained and used to keep releases to the environment as low as is reasonably achievable. The OPERABILITY of the liquid radwaste treatment system ensures that the appropriate portions will be available for use whenever liquid effluents require treatment prior to release to the environment. The specification that the portions of the system which were used to establish compliance with the design objectives in 10 CFR Part 50.
Appendix I.Section II be used when specified provides reasonable assurance that releases of radioactive material in liquid effluent will be kept as low as is reasonably achievable. The activity concentration. 0.01 pCi/ml, below which liquid radwaste treatment would not be cost beneficial, and therefore not required, is demonstrated below:
The quantity of radioactive material in liquid effluent released annually from Cooper Station has been calculated to be' total iodines 3.65 curies total others (less H3) 0.7 total 4.35 curies The population dose commitment resulting from the radioactive material in liquid effluent released annually has been calculated to be thyroid 1.95 manrem total body 0.56 total 2.5 manrem Therefore, population doses are about 0.5 manrem per curie of iodine released and about 0.8 manrem per curie of other radionuclides (less H3) released in liquids. It would be conservative to assume one manrem committed per curie released in liquid effluent.
The volume of liquid waste processed and intended for discharge is estimated to be:
Low Purity Waste 5700 gal/day 1.8 x 106 gal/yr Chem Waste + Demin Regenerant Waste 4000 gal/day 1.2 x 106 gal/yr (continued)
.AftiNRNVATION ONU Liquid Waste Concentration B 3.1.2 BASES (continued)
The annual credit for to be:
costs to operate the radwaste processing equipment, neglecting capital recovery, are estimated according to Regulatory Guide 1.110 Dirty Waste Ionex
$ 88,000/yr Evaporator
$114,000/yr Unit volume operating costs are about:
Cost to ion exchanger = $ 88.000
= $0.05/gal 1.8E+6 gal Cost to evaporate
= $114.000 1.2E+6 gal
= $0.10/gal Assuming the cost-benefit balance is $1,000 expenditure per manrem reduction and assuming treatment removes all radioactivity from the liquid, then (1) the activity concentration in a batch below which treatment is not cost-beneficial is
$88,000 1 curie 106 pCi x
x x
manrem curie 1 manrem
$1,000 1.8E+6 gal x 3785 m1_
gal C = 0.013 pCi/ml (2) the activity concentration below which evaporation is not costbeneficial is
$114,000 1 curie 106 PCi 1.2E+6 gal x 3785 ml manrem curie ga 1 manrem
$1,000 C = 0.025 pCi/ml Therefore, to one si nificant digit, radwaste treatment of liquids containing less than 0.01 pCi/mi is not justified.
(continued)
Liquid Waste Concentration B 3.1.2 BASES (continued)
Demonstration of Compliance with 10 CFR 50 Appendix I. Revision 1 and Supplement 2, Nebraska Public Power District, Cooper Nuclear Station, January 9, 1978.
Liquid Effluents Dose B 3.1.3 B 3.1 LIQUID EFFLUENTS B 3.1.3 Liquid Effluents Dose BASES Note: The Bases discussion refers to "technical specifications' and quotes the Staff's use of "technical specifications." The statements and opinions pre-date Generic Letter 89-01, Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the relocation of procedural details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program.
Generic Letter 89-01 provides the guidance and justification for relocation of these "technical specifications" to the Offsite Dose Assessment (ODAM)
Manual and the Process Control Program (PCP).
Therefore, 'technical specifications' as used in this Bases refers to ODAM Specifications.
Specifications D 3.1.3. D 3.2.2 and D 3.2.3 implement the requirements of 10 CFR Part 50.36a and of 10 CFR Part 50. Appendix I,Section IV. These specifications state ODAM LIMITING CONDITID S FOR OPERATION (DLCO) to keep levels of radioactive materials in LWR effluents as low as is reasonably achievable. Compliance with these specifications will also keep average releases of radioactive material in effluents at small percentages of the limits specified in 10 CFR Part 20.1301. Surveillance Requirements provide for the measurement of releases and calculation of doses to verify compliance with the Specifications. Action statements in these Specifications implement the requirements of 10 CFR Part 50.36(c)(2) and 10 CFR Part 50, Appendix I,Section IV.A in the event an LCO is not met. Annual dose limitations stated in Specifications D 3.1.3, D 3.2.2 and D 3.2.3 are not strict limits as used elsewhere in the Technical Specifications (are not an immediate safety concern) but do obligate NPPD to take the applicable Required Action in Specifications D 3.1.3, D 3.2.2 and D 3.2.3.
(continued)
Liquid Effluents Dose B 3.1.3 BASES (continued) 10 CFR Part 50 contains two distinctly separate statements of requirements pertaining to effluents from nuclear power reactors. The first concerns a description of equipment to maintain control over radioactive materials in effluents, determination of design objectives, and means to be employed to keep radioactivity in effluents ALARA.
This requirement is stated in Part 50.
Section 34a and Appendix I,Section II.
Appendix I.Section III stipulates that conformance with the guidance on design objectives be demonstrated by calculations (since demonstration is expected to be prospective). The other is a requirement for developing limiting conditions for operation in technical specifications. It is stated in 10 CFR Part 50, Section 36a and Appendix I,Section IV. Both the intent of the Commission and the requirement are clearly stated in the Opinion of the Commission; I relevant paragraphs from that document follow:
Section 50.36a(b) of 10 CFR Part 50 provides that licensees shall be guided by certain considerations in establishing and implementing operating procedures specified in technical specifications which take into account the need for operating flexibility and at the same time ensure that the licensee will exert his best efforts to keep levels of radioactive materials in effluents as low as practicable.
The Appendix I that we adopt provides more specific guidance to licensees in this respect.
A. The Rule Section IV of Appendix I specifies action levels for the licensee. If, for any individual light water cooled nuclear power reactor, the quantity of radioactive material actually released in effluents to unrestricted areas during any calendar quarter is such as to cause radiation exposure, calculated on the same basis as the design objective exposure, which would exceed one-half the annual design objective exposure. the licensee shall make an investigation to identify the causes of these high release rates, define and initiate a program of action to correct the situation, and report these actions to the Commission within 30 days of the end of the calendar quarter.
The conclusion of the NRC Staff in the Appendix I Rulemaking Hearing 2 agrees with that of the Commission.
The Staff recommended, "...that the limiting conditions for operation described in Appendix I.Section IV be applicable upon publication to technical specifications included in any license authorizing operation of a light water cooled nuclear power reactor..."
(p. 73).
(continued)
Liquid Effluents Dose B 3.1.3 BASES (continued)
The action to be taken by a licensee in the event a limiting condition is exceeded, is stated in Appendix I,Section IV.A and in the Opinion of the Commission.
3 ODAM Specifications D 3.1.3, D 3.2.2, D 3.2.3 and Surveillances DSR 3.1.3.1. 3.1.3.2. 3.2.2.1, 3.2.3.1 and 3.2.3.2 for Cooper Station conform to this requirement.
Guidance for developing limiting conditions for operation for surveillance and monitoring is included in Appendix I,Section IV.B.
Although "it is expected that the annual releases of radioactive material in effluents from light water cooled nuclear power reactors can generally be maintained within the levels set forth as numerical guides for design objectives in Section II" (Appendix I.Section IV), no recommendation was made by either the Staff in its Concluding Statement I or by the Commission in its Opinion 5 that design objective values should appear as technical specification limits.
The Opinion of the Commission and the statement of Appendix I are clear.
Limiting conditions of operation (LCO) related to the quantity of radioactive material in effluents released to an unrestricted area stated in technical specifications shall conform to Appendix I.Section IV.A.
Licensee action in the event an LCO is exceeded should be in accord with Section IV.A.
Finally, surveillance and monitoring of effluents and the environment should conform to Section IV.B.
With the implementation of Specification D 3.1.3 and Surveillances DSR 3.1.3.1 and 3.1.3.2. there is reasonable assurance that Station operation will not cause a radionuclide concentration in public drinking water taken from the River that exceeds the standard for anthropogenic radioactivity in community drinking water.
1NRC Commissioners, "Opinion of the Commission," in the Appendix I Rulemaking hearing, Docket Rm 502, p. 101-102, April 30, 1975.
2NRC Staff, "Concluding Statement of the Regulatory Staff," in the Appendix I Rule-making Hearing, Docket RM 502, pp. 17, 69, 73, 115, February, 1974.
3NRC Commissioners, p. 101 4NRC Staff, op. cit.
5NRC Commissioners, op. cit.
(continued)
Liquid Effluents Dose B 3.1.3 BASES (continued) 6Generic Letter 89-01, Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the relocation of procedural details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program.
k
0 I -w i;-,
, ",, -
_ L, -
t Outside Temporary Storage of Radioactive Liquid B 3.1.4 B 3.1 LIQUID EFFLUENTS B 3.1.4 Outside Temporary Storage of Radioactive Liquid BASES Custom Technical Specifications Bases did not exist.
INFORMATION ONI Gaseous Effluents Concentration B 3.2.1 B 3.2 GASEOUS EFFLUENTS B 3.2.1 Gaseous Effluents Concentration BASES DLCO 3.2.1(a) is included to assure that a measure of control is provided over the concentration of radionuclides in air leaving the exclusion area.
Radioactive noble gases are monitored by instruments that provide a measure of release rate and cause automatic alarm when the noble gas concentration beyond the Site and Exclusion Area Boundary is expected to exceed the dose rate specified in DLCO 3.2.1(a).
With prompt action to reduce the radioactive noble gas concentration in effluent following alarm initiation. it can be maintained at a small fraction of the annual limit.
The specified release rate limits restrict the corresponding gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to
- 500 mrem/year to the total body or to
- 3000 mrem/year to the skin.
Radioiodines and radionuclides in particulate form are sampled with integrating samplers that permit assessment of the average release rate during each sample collection period.
By complying with DLCO 3.2.2 and 3.2.3 the average concentration beyond the Site and Exclusion Area Boundary will be maintained at a small fraction of the 10 CFR Part 20.1302(b)(2)(i) concentration limit.
Noble Gases Dose B 3.2.2 B 3.2 GASEOUS EFFLUENTS B 3.2.2 Noble Gases Dose BASES Assessments of dose required by Surveillances DSR 3.2.2.1 and DSR 3.2.3.2 to verify compliance with Appendix I,Section IV is based on measured radioactivity in gaseous effluent and on calculational methods stated in the ODAM. Pathways of exposure and location of individuals are selected such that the dose to a nearby resident is unlikely to be underestimated. Dose assessment methodology described in the ODAM for gaseous effluent will be consistent with the methodology in Regulatory Guides 1.109 and 1.111. Cumulative and projected assessments of dose made during a quarter are based on historical average, or reference (the same period of record used in the design objective Appendix I evaluation) atmospheric conditions. Assessments made for the Annual Radiological Environmental Report will be based on quarterly and annual averages of atmospheric conditions during the period of release.
The bases for Specification D 3.2.2 and Surveillance DSR 3.2 2.1 are also discussed in the bases for Specification D 3.1.3 and Surveillances DSR 3.1.3.1 and 3.1.3.2.
-INFORMATION ON.h Iodine and Particulates B 3.2.3 B 3.2 GASEOUS EFFLUENTS B 3.2.3 Iodine and Particulates BASES231an3232 ification D 3.2.3 and Sjurveil~lainc~esn DSR 3.
an S DSR This bases for Spec rSeification D 3.2.3 and SurveillesOR..31andc
.23.
are discussed in the bases for pcfcto
.. n u~i1ne S 3
(1S3.1 and 3 12 3 2 3
REVISION 0 ENS ODAMB 3.2-3.
Offgas Treatment System B 3.2.4 B 3.2 GASEOUS EFFLUENT B 3.2.4 Offgas Treatment System BASES The OPERABILITY of the gaseous radwaste treatment system ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50. and design objective Section IID of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of this system are specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I. 10 CFR Part 50, for gaseous effluents.
INFORMA 7UYI1(Thtt j
Exhaust Ventilation Treatment Systems B 3.2.5 B 3.2 GASEOUS EFFLUENTS B 3.2.5 Exhaust Ventilation Treatment Systems BASES An Exhaust Ventilation Treatment System (EVTS) is a system intended to remove radioiodine or radioactive material in particulate form from gaseous effluent by passing exhaust ventilation air through charcoal absorbers and/or HEPA filters before exhausting the air to the environment. An EVTS is not intended to affect noble gas in gaseous effluent.
Engineered Safety Feature (ESF) gaseous treatment systems are not considered to be EVTS.
The Standby Gas Treatment System is an ESF and not an EVTS.
EVTS are specifically identified in ODAM Figure 3-1.
The OPERABILITY of the exhaust ventilation treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a. General Design Criterion 60 of Appendix A to 10 CFR Part 50. and design objective Section IID of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of the system are specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50. for gaseous effluents.
IN0FORMATION U1-13,wo Hydrogen Concentration B 3.2.6 B 3.2 GASEOUS EFFLUENTS B 3.2.6 Hydrogen Concentration BASES This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen. While the Augmented Treatment System is in service the hydrogen and oxygen concentrations are prevented from reaching the flammability limits.
Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR 50.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
iqF (i RfM '
l i;*
Containment B 3 2.7 B 3.2 GASEOUS RELEASES B 3.2.7 Primary Containment Venting and Purging BASES This specification provides reasonable assurance that releases of Iodine from drywell purging during power operations, and during startup while performing primary containment inerting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown will not be excessively large, particularly due to Iodine spiking.
The exemptions to using the SBGT system are intended to minimize the time the SBGT system is on line while coolant temperature is greater than 200'F, hence to decrease the probability of damage to the SBGT filters that could occur from overpressurization due to a LOCA and the main purge and vent valves open.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
INFGRMA'fV fH Liquid Effluent Moni or n -
B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Liquid Effluent Monitoring BASES The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the release of radioactive material in liquid effluents.
The OPERABILITY and use of these instruments implements the requirements of 10 CFR Part 50. Appendix A. General Design Criteria 60, 63, and 64. The alarm and/or trip setpoints for these instruments are calculated in the manner described in the ODAM to assure that the alarm and/or trip will occur before the limit specified in 10 CFR Part 20.1302 is exceeded.
Control of the normal liquid discharge pathway is assured by station procedures governing locked discharge valves and valve line-up verification.
The liquid radwaste monitor assures that all liquid discharged to the discharge canal does not exceed the limits of Specification D 3.1.1.
Upon sensing a high discharge level, an isolation signal is generated which closes the radwaste discharge valve.
The set point is adjustable to compensate for variable isotopic discharges and dilution flow rates.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Gaseous Efhk'urnt'
' I r
- ~
d; B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Gaseous Effluent Monitoring BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The location of this instrumentation is indicated by a Figure in the ODAM, a simplified flow diagram showing gaseous effluent treatment and monitoring equipment. The alarm/trip setpoints for these instruments shall be calculated in accordance with methods in the ODAM, which have been reviewed by NRC, to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20. The process monitoring instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the augmented offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line. Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip. There is a fifteen minute delay accounted for by the 30-minute holdup time of the off-gas before it reaches the stack. Both instruments are required for trip but the instruments are so designed that any instrument failure gives a downscale trip. The trip setting of 1.0 ci/sec (prior to 30 min. delay) provides an improved capability to detect fuel pin cladding failures to allow prevention of serious degradation of fuel pin cladding integrity which might result from plant operation with a misoriented or misloaded fuel assembly. This limit is more restrictive than 0.39 ci/sec noble gas release rate at the air ejectors (after 30 min. delay) which was used as the source term for an accident analysis of the augmented off-gas system. Using the.39 ci/sec source term, the maximum off-site total body dose would be less than the.5 rem limit.
A footnote has been added to Table D3.3.2-1 to indicate that a channel of the Noble Gas Activity Monitor, the Iodine Sampler Cartridge, or the Particulate Sampler Filter, in the ventilation monitoring systems for the Reactor Building, Radwaste Building, Turbine Building, Elevated Release Point, and Multi Purpose Facility, may be removed from service for up to 30 minutes for changing particulate filters or iodine cartridges or for the low flow alarm check without entering the conditions and required actions. The channel will continue to be declared inoperable when removed from service, but the Conditions and Required Actions will not be entered due to the short duration that the channel is inoperable and the facts that the system must be removed from service to perform the required ODAM surveillance and is returned to an OPERABLE condition prior to expiration of the Completion Time for initiating the Required Action.
A note has been added to Required Action 1.1 stating that continuous monitoring may be discontinued for up to 30 minutes only for changing particulate filters and iodine cartridges when the backup monitoring system is in service. Replacement of these filters and cartridges is CNS ODAM B 3.3-2 03120/03
Gaseous Effluent Monito6rinrg B 3.3.2 a normal activity, and not a condition that is appropriate for entry into the Corrective Action Program or for including in the Radioactivity Effluent Release Report. This allowance to l
discontinue the sampling is acceptable based on (1) four hours are allowed to begin the l
continuous collection of samples with auxiliary equipment when the primary monitoring system l
is inoperable, (2) occasions when the primary monitoring system is inoperable and the backup l
monitoring system is in service are expected to be rare, and (3) the short period of time (30 minutes) that the monitor is allowed to be out of service.
In the event no flow rate measurement device is operable on a gaseous stream, alternative 24-hour estimates are adequate since the system design is constant flow and loss of flow is alarmed in the control room.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
9, Liquid/Gaseous Effluents Dose B 3.4.1 B 3.4 LIQUID/GASEOUS DOSE B 3.4.1 Liquid/Gaseous Effluents Dose BASES This specification is provided to meet the reporting requirements of 40 CFR Part 190. In the event an analysis is required to determine compliance with 40 CFR 190. the dose to a member of the public due to radiation direct from the station will be estimated with the aid of environmental TLD, PIC. or similar environmental radiation dosimetry. A contribution from another fuel cycle facility is not added since there is no licensed fuel cycle facility within 50 miles of Cooper Station.
I'llFO R MAT;¢e$
Ih Solid Radioactive Waste B 3.5.1 B 3.5 SOLID RADIOACTIVE WASTE B 3.5.1 Solid Radioactive Waste BASES The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require materials processing and packaging prior to being shipped offsite. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.
Monitoring Program Compliance B 4.1 B 4.0 MONITORING PROGRAM B 4.1 Monitoring Program Compliance BASES The radiological environmental monitoring program, including the land use census, is conducted to satisfy the requirements of 10 CFR Part 50, Appendix I.Section IV.B.2 and 3.
The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation.
This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measureable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The environmental monitoring program described in Table D4.1-1 is the minimum program which will be maintained. The Offsite Dose Assessment Manual(ODAM) describes in detail the actual monitoring program which is performed to ensure compliance with the specified minimum program.
The land use census is conducted annually to identify changes in use of the unrestricted area in order to recommend modifications in monitoring programs for evaluating individual doses from principal exposure pathways.
The need to adjust the program to current conditions and to assure that the integrity of the program is maintained are thereby provided. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/square meter.
V In* I I
i I
a 3,
Monitoring Program Concentration B 4.2 B 4.0 MONITORING PROGRAM B 4.2 Monitoring Program Concentration BASES Custom Technical Specifications Bases did not exist.
i i 5i Monitoring Program Dose B 4.3 B 4.0 MONITORING PROGRAM B 4.3 Monitoring Program Dose BASES Like pathways monitored (sampled) at a location. excluding the control station location(s), having the lowest associated calculated personnal dose may be deleted from Table D4.1-1 at the time the new pathway(s) and locations are added.
Interlaboratory Comparison Program B 5.1 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.1 Interlaboratory Comparison Program BASES The requirement for participation in a Interlaboratory Comparison 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 a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
11 I
4 h Annual Radiological Environmental B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.2 Annual Radiological Environmental Report BASES Report B 5.2 Custom Technical Specifications Bases did not exist.
T-e ! -- F
. -,- f,
. 1 6"
11 - I -.
I, "
f
-, r -,JJ,4 i " -" r F I [ I,(
1
.y ieldv -.
Annual Radiological Effluent Release Report B 5.3 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.3 Annual Radiological Effluent Release Report BASES Custom Technical Specifications Bases did not exist.
Special Reports B5.4 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.4 Special Reports BASES Custom Technical Specifications Bases did not exist.
INFORMOIATION ONLY Major Changes to Radioactive Waste Treatment Systems B 5.5 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid. Gaseous.
and Solid)
BASES Custom Technical Specifications Bases did not exist.
INFORMATION ONLY PROCESS CONTROL PROGRAM FOR COOPER NUCLEAR STATION
'K4rUR IAWAY TiN ONL' Cooper Nuclear Station Process Control Program List of Effective Pages Pace Date 1
8/7100 2
8/7/00 3
8/7/00 4
8W1M00 5
8/7/00 6
817100 7
8/7/00 8
8/7/00 9
8/7/00 10 8/7/00 11 8/7/00 12 817/00 13 8/7/00 14 8/7/00 15 8/7/00 CNS PCP Page 1 of 1 8f7/0
A.
Purpose:
> v The Cooper Nuclear Station (CNS) Process Control Program (PCP) establishes the processing conditions for assuring the SOLIDIFICATION, dewatering or stabilization of CNS radioactive waste streams produced from the CNS liquid radioactive waste treatment system and from activities producing radioactive waste requiring SOLIDIFICATION, dewatering or stabilization such as decontamination system resins, irradiated components and highly contaminated equipment.
The CNS PCP is comprised of the Dewatering Process Control Program (DPCP),
SOLIDIFICATION Process Control Program (SPCP) and vendor Process Control Programs. The DPCP utilizes the CNS dewatering system to process solid wet waste streams from the CNS liquid radioactive waste treatment system or from chemical decontamination resins. The SPCP utilizes the cement solidification system to process solid wet waste from the liquid radioactive waste treatment system. Vendor Process Control Programs utilize NRC approved PCP's, stabilization processes and High Integrity Containers to process various forms of solid and liquid radioactive wastes at CNS.
B.
Definitions:
- 1.
SOLIDIFICATION, SOLIDIFY, SOLIDIFIED - The conversion of radioactive wastes from liquid systems to a solid which is as uniformly distributed as reasonably achievable with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).
- 2.
Stability - Structural stability which ensures that the waste does not degrade and (a) promote slumping, collapse, or other failure of the cap or cover over a near-surface disposal facility and thereby lead to water infiltration, or (b) impart a substantial increase in surface area of the waste form that could lead to an increase in leach rate. Stability is also a factor in limiting exposure to an inadvertent intruder since it provides greater assurance that the waste form will be recognizable and nondispersable during its hazardous lifetime. Structural stability of a waste form can be provided by the waste form itself (as with activated stainless steel components),
by processing the waste to a stable form (e.g., stabilization), or by placing the waste in a container or structure that provides stability (e.g, high integrity container or engineered structure).
C.
Discussion:
The regulation, "Licensing Requirements for Land Disposal of Radioactive Waste,"
10 CFR Part 61, establishes a waste classification system based on the radionuclide concentrations in the waste. All Class B and C wastes are required to be stabilized. Class A wastes have lower concentrations and may be segregated without stabilization. All Class A liquid wastes, however, require solidification or absorption to meet the free liquid requirements. Class A wastes may also be stabilized and disposed of with stabilized Class B and C wastes.
D.
References:
- 1.
CNS Operating Procedure 2.5.4.1, Solid Wet Waste Packaging, Storage and Transfer System.
- 2.
CNS Operating Procedure 2.5.4.2, Solid Wet Waste Drum Filling.
- 3.
CNS Operating Procedure 2.5.4.3, Radwaste Drum Mixer Operation.
- 4.
CNS Operating Procedure 2.5.4.4, NuPac Dewatering System.
- 5.
CNS Operating Procedure 2.5.1.3, Radwaste Low Conductivity Liquid Waste Spent Resin Tank Fluid Transfer.
- 6.
CNS Operating Procedure 2.5.2.1, Radwaste High Conductivity Liquid Waste and Waste Sludge Tank Fluid Transfer.
- 7.
CNS Operating Procedure 2.5.1.2, Radwaste Low Conductivity Liquid Waste Condensate Phase Separator Fluid Transfer.
- 8.
CNS Operating Procedure 2.5.1.7, RWCU Phase Separator Tank Transfer
- 9.
CNS Radiological Protection Procedure 9.RW.1, Radioactive Shipments
- 10.
CNS Radiological Protection Procedure 9.RW.2, Condensate Waste Resins, Spent Resins, RWCU Resins and Waste Sludge Classification and Listing.
- 11.
CNS Radiological Protection Procedure 9.RW 7, Waste Stream Sampling
- 12.
Cooper Nuclear Station Offsite Dose Assessment Manual, D.3.5.1 - Solid Radioactive Waste.
- 13.
Cooper Nuclear Station Technical Specifications,5.6.3 - Radioactive Effluent Release Report.
- 14.
Cooper Nuclear Station Updated Safety Analysis Report, XI 11-13 0- Process Control Program.
- 15.
10CFR61 Licensing Requirements for Land Disposal of Radioactive Waste.
- 16.
Waste Form Technical Position, Revision 1, January 1991, U. S. NRC.
- 17.
Certificate of Compliance WN-EB-01, State of Washington Department of Health.
- 18.
Certificate of Compliance WN-EB-02, State of Washington Department of Health.
- 19.
Safety Evaluation Report TP-02-P, Revision 1, U. S. NRC.
- 20.
Cooper Nuclear Station Offsite Dose Assessment Manual, D.5.3. - Radioactive Effluent Release Report.
E.
Process Descriptions:
SOLIDIFICATION Process Control Program
- a.
Overview The Cooper Nuclear Station (CNS) SOLIDIFICATION Process Control Program (SPCP) is a program of sampling, processing, analysis and formulation determination by which SOLIDIFICATION of radioactive waste from liquid systems is assured to be consistent with the CNS Offsite Dose Assessment Manual. Compliance with the SPCP ensures that the resultant waste form characteristics are acceptable for burial at a licensed low level radioactive waste burial facility.
The SPCP is not intended to be a substitute for the CNS Operations Manual. The SPCP outlines the general methods of sampling, processing, analysis and waste formulation during the SOLIDIFICATION of solid wet radioactive waste. The CNS Operations Manual details the actual methods by which solid wet waste is SOLIDIFIED.
Solid wet waste streams produced from the liquid radioactive waste treatment system can be SOLIDIFIED utilizing the CNS SPCP. These wet waste streams are comprised of condensate resins, spent resins, reactor water cleanup resins and waste sludges.
When regulations require SOLIDIFICATION,thesewastestreams are SOLIDIFIED in accordancewith SPCP requirements. The waste processing is performed with equipment installed at CNS under the direct supervision of CNS personnel utilizing CNS approved procedures. ALARA considerations and radiological controls are incorporated into these procedures. After the waste has been processed in its burial container, the waste classification and type is determined so that the waste may be shipped to a licensed burial site in accordance with appropriate NRC, DOT, state and burial site regulations.
- b.
Operation - General:
The solid wet radioactive waste SOLIDIFICATION process is operated on a batch basis. A batch consists of all the resulting drums sequentially processed from the contents of a single source such as a phase separator.
The batch is completed when the waste source is empty or the batch process is terminated due to radioactive waste system operational constraints. Solid wet waste sources which are typically processed on a batch basisforSOLIDIFICATION includethe condensate phase separators, reactor water clean-up phase separators, waste sludge tank or spent resin tank.
Waste steams being SOLIDIFIED from the condensate or reactor water cleanup phase separators, waste sludge tank, or spent resin tank are routed through centrifuge units for the purpose of dewatering the waste slurry. The wastes enter a storage hopper after dewatering in the centrifuge units. The waste in the storage hopper ranges from a relatively dry granular consistency to a wet, putty-like consistency, depending on the waste material (e.g., sludge, resin). Department of Transportation (DOT) 17H specification 55 gallon drums containing dry Portland Type I cement are transferred under the hopper and filled with the dewatered waste and demineralized water. The drum then progresses to the drum mixing section where an in-drum mixes the cement and waste.
During the mixing process additional water is added to the mixture in quantities sufficient to ensure SOLIDIFICATION.
Due to the varying degrees of moisture content subsequent to the centrifuge dewatering process, a visual inspection of the initial drums being mixed is necessary to establish the exact amount of water required to SOLIDIFY the waste batch.
Once the correct volume of water for the batch is determined the mixing system programmer is adjusted to add the correct amount of mixing water automatically. The water, cement and waste are thoroughly mixed by the mixing system's programmed sequence. After the mixing sequence is 817/00 CNS PCP
complete the radiation levels are measured on each drum. The drum is then transferred to the drum storage line.
The filled drums are allowed to SOLIDIFY a minimum of twenty-four hours before being transferred from the drum storage line to the drum capping station. Here the drums are visually inspected for freestanding liquid and SOLIDIFICATION. Drums having novisible freestanding liquid and meeting SOLIDIFICATION acceptance criteria are capped, decontaminated and returned to the drum storage line in preparation for shipment to a licensed burial facility.
Drums not complying with freestanding liquid of SOLIDIFICATION requirements are retained for corrective disposition prior to capping and storage.
- c.
Operation - Specific:
1.8 +/- 0.3 cubic feet of dry Portland Type I cement is added to each DOT 17H specification 55 gallon drum. It has been demonstrated that this volume of cement will properly SOLIDIFY the waste streams processed by the drum mixing system.
The waste to be SOLIDIFIED is added in quantities sufficient to fill the drum along with approximately seven gallons of demineralized water. When the drum filling is complete the drum is transferred to the drum mixing station. A representative waste sample from one drum of each batch is taken and analyzed for pH. Experience has conclusively shown that the SOLIDIFICATION process is unaffected by pH as long as the pH of the waste remains within the range of 2 to 13.
The filled drum is mixed by an automatic, sequence controlled, in-drum mixer. During the mixing sequence approximately five gallons of additional demineralized water are added to the filled drum. When the water addition is complete and the final mixing steps are sequencing to completion, the mixer motor amps are checked to ensure sufficient water has been added.
Slight changes in the amount of water being added during the mixing sequence may be made to adjust the mixer motor amps to approximately 8/7/00 CNS PCP
seven amps. These adjustments are made on the first several drums being processed from the waste batch, and are then programmed into the drum mixing sequence for the remaining drums of the batch. Upon completion of the mixing process, and prior to the cement setting, one drum from the batch is sampled in order to determine its isotopic content and distribution.
The drum's radiation level is also measured. By comparing this drum's isotopic distribution, concentration, and radiation level with the radiation readings on each of the other drums in the batch the total concentration of the radionuclides present in each drum can be determined. This data comparison method may also be used to determine waste carry-over from previous batches into the batch currently being processed.
Following a minimum of twenty-four hours after mixing, each SOLIDIFIED drum is visually inspected for freestanding liquid and SOLIDIFICATION.
Every tenth drum of the batch is quality control checked for resistance to penetration to verify SOLIDIFICATION. The penetration test verifies that the SOLIDIFIED waste has a minimum compressive strength of 50 psi. Filled drums meeting the freestanding liquid and SOLIDIFICATION criteria are capped and transferred to the drum storage line for later shipment to a licensed burial site. The packaging, classification and shipping of wastes processed via the CNS SPCP are in accordance with the applicable sections of 10CFR61, 10CFR71 and 49CFR Parts 171 through 178.
- d.
Qualification:
Qualification of the SPCP process for stabilization must be conducted for each waste stream requiring stabilization (e g., Class B waste, Class C waste) prior to burial. Qualification shall be conducted in accordance with the stabilization testing recommendations of the "Waste Form Technical Position" Revision 1, January 1991, U. S. NRC. The SPCP stabilization qualification testing is typically performed by a CNS approved vendor.
- e.
Scaling Factors:
As permitted by 10CFR61.55, waste stream radionuclide scaling factors based on certain gamma emitting nuclides (e.g., Co-60, Cs-137) may be used to calculate the concentrations of difficult-to-measure radionuclides.
Scaling factor determinations are made using a detailed radionuclide analysis of waste stream samples conducted by a CNS approved vendor.
- f.
Reports:
The volume, curie content, principle nuclides, type of waste and solidification agent for the wet wastes SOLIDIFIED at Cooper Nuclear Station are documented in the Cooper Nuclear Station Radioactive Effluent Release Report. This information is listed in the format outlined in Revision 1 of Regulatory Guide 1.21.
F.
Dewatering:
- a.
General The CNS Dewatering PCP (DPCP) utilizes dewatering equipment and disposable waste containers to dewater solid wet waste streams produced from the liquid radioactive waste treatment system or from chemical decontamination resins.
Compliance with the DPCP ensures that the resultant waste meets the free standing liquid criteria set forth in 1 OCFR61. The DPCP is not intended to be a substitute for the CNS Operations Manual nor does the process meet the waste stability form requirements of 10CFR61. Stabilization/solidification of the dewatered waste is provided by High Integrity Containers (HICs) and/or Engineered Concrete Barriers.
Solidification of the dewatered waste can also be accomplished by dewatering the waste in metal liners.
Solid wet waste streams produced from the liquid radioactive waste treatment system and chemical decontamination resins can be dewatered utilizing the CNS CNS PCP 8/7/00 1
DPCP. The liquid radioactive waste treatment streams include condensate resins, spent resins, reactor water cleanup resins and waste sludges. When regulations require solidification or stabilization, these waste streams can be dewatered in HICs and burial liners per DPCP requirements. The waste processing is performed with the Nuclear Packaging, Inc. dewatering system equipment installed at CNS, under the direct supervision of CNS personnel utilizing CNS approved procedures. The design and arrangement of the CNS dewatering system components are based on maintaining the operator radiation exposure ALARA. ALARA considerations and radiological controls are also incorporated into the system operating procedures.
After the waste has been processed in its burial liner the waste classification and type are determined so that the waste may be shipped to a licensed burial site in accordance with appropriate NRC, DOT, state and burial site regulations.
- b.
Operation - General.
The solid wet radioactive waste dewatering process is operated on a batch basis.
A batch consists of all the resulting dewatered liners processed from the contents of a waste source such as a phase separator. The batch is complete when the waste source is empty or the batch process is terminated due to radioactive waste system operational constraints.
Solid wet waste sources which are typically dewatered on a batch basis include the spent resin tank, waste sludge tank, condensate phase separators, reactor water clean-up phase separators, decontamination resins, or a combination of these waste sources.
Prior to dewatering the solid wet waste, verification is made that the waste being dewatered is characterized in accordance with the NUPAC Services Division, Inc. Waste Characterization Form. Selection of the burial liner (steel liner or HIC) utilized in the dewatering process is based upon the classification and type of waste being processed.
The CNS dewatering system consists of a dewatering waste container, a dewatering pump, an off-gas vent unit, a container level indicator, a waste fill head, a water separator with water chiller unit, an air blower, a relative humidity instrument, a control panel, and interconnecting piping and valves.
{,~~~
-i -,,a, After wet radioactive waste is charged into a waste container, dewatering is achieved with continuous suction on the waste container by the dewatering pump.
Various types, numbers, and configurations of filters are used within the waste container to retain spent resin and filter precoat materials. Water removed from the waste container is returned to the liquid radwaste system.
The dewatering pump is operated for given time internals in accordance with the NuPac Services Division, Inc. Waste Characterization Form. The pumping time may range from eight to sixteen hours depending upon the waste stream and waste container. After most of the free water in the waste container has been removed, drying air is continuously recirculated in a loop from the air blower to the waste container through the water separator to remove any residual free water in the waste container.
The dewatering system is provided with temperature instrumentation which is interlocked to automatically shut down the dewatering process on high air temperature due to the potential for exothermic reactions during ion exchange resin dewatering. The waste container is considered dewatered when at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of pumping time has been completed and, the relative humidity in the recirculating drying air meets the acceptance criteria specified in the NUPAC Services Division, Inc. Waste Characterization Form. A relative humidity instrument and monitor are provided to remotely and continuously monitor the waste container outlet air. This instrument is used to establish the positive end point to the dewatering process and the point at which the free standing liquid criteria of 10CFR61 is met. Filled waste containers which do not meet the dewatering end point criteria are retained until a satisfactory corrective disposition is determined.
Stability or solidification of the waste is provided by the waste container and/or engineered barriers such as concrete vaults.
No airborne or liquid radwastes are released to the environment from the dewatering operation. The dewatered liquid radwastes are routed to the liquid radioactive waste treatment system and resin drying air is vented through a HEPA system. The dewatering system is designed to prevent the uncontrolled release of radioactive materials by monitoring liquid levels in the waste container by a level indicator. During the waste filling operation, the operator visually monitors the CNS PCP
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waste transfer process by observing the dewatering control panel video monitor and the local area radiation monitors. Potential inadvertent spills and overflows are contained in the Augmented Radioactive Waste Building and can be routed back to the liquid radioactive waste treatment system via floor drains.
- c.
Operation - Specific-The CNS Dewatering System fillhead is installed on a waste container compatible with the dewatering fillhead. The waste container typically has four filter levels. The filters provide the flow path from the water container, through the fillhead, to the dewatering pump. The fillhead is fitted with connections to transfer solid wet waste from the Waste Transfer System to the waste container. The fillhead also contains a remote video camera for internal monitoring of the waste container, dewatering pump suction connections, electrical service connections for the waste container level indication system, a service air supply to provide cooling, a spray ring header for flushing, a pressure switch to monitor container pressure, and connections for the waste container temperature indicating system.
Once the fillhead is installed and the bottom filter level dewatering pump suction valve is opened, solid wet waste from the Condensate Phase Separator, Spent Resin Tank, or Reactor Water Cleanup Phase Separator is transferred as a homogeneously mixed slurry to the waste container. During the waste transfer the fillhead video monitor is used to verify that the waste container is being filled.
When the waste slurry reaches the second filter level, the dewatering pump is started. The dewatering pump is cycled on and off to maintain the water level in the waste container slightly above the level of the waste. When the slurry level reaches the third level of filters the second level dewatering pump suction valve is opened and the first level dewatering suction valve is closed. At this point a sample of the waste slurry is drawn for gamma isotopic analysis. As the waste slurry level reaches the fourth filter level the third level dewatering pump suction valve is opened. When the slurry level reaches the fill line deflector plate the fourth level dewatering pump suction valve is opened. Waste slurry transfer is continued until CNS PCP 8n/7/0 1
the waste container is full. Demineralized water is then used to flush the residue in the waste transfer lines and fill head connections to the waste container. The dewatering pump is stopped, the second and third level dewatering suction valves closed, and the Waste Transfer System lineup secured. The dewatering cycle is then commenced by starting the dewatering pump.
When the fourth level dewatering pump suction line vacuum fluctuates rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the fourth level dewatering suction valve is closed and the third level suction valve is opened. As the third level dewatering pump suction line vacuum begins to fluctuate rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the third level dewatering suction valve is closed and the second level dewatering suction valve is opened. Similarly, as the second level dewatering pump suction line vacuum begins to fluctuate rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the second level dewatering suction valve is closed and the first level dewatering suction valve is opened When the first level dewatering pump suction line vacuum fluctuates rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the first level dewatering suction valve is closed and the dewatering pump secured.
After the temperature of the radioactive waste in the waste container is measured and recorded the drying cycle is commenced. The drying cycle utilizes a blower to blow warm dry air through the waste to evaporate the residual water remaining in the waste container. The warm moist air which exits from the waste container is pumped through an entrainment separator tank where refrigeration coils condense the water vapor and the entrained water is removed. The water collected in the entrainment separator tank is pumped to the radioactive waste floor drain system.
The dehumidified warm air is recirculated through the waste until the endpoint humidity and minimum recirculation dry time requirements of the NuPac Waste Characterization Form are met.
The drying cycle begins by closing the fillhead waste isolation valve, aligning the blower to the waste container and starting the dewatering system chiller. The dewatering pump and blower are started and the blower inlet pressure, blower CNS PCP 8t7/0 1
discharge pressure, dewatering pump suction pressure, blower inlet temperature, blower discharge temperature, waste container temperature parameters are then monitored and recorded each hour. After 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of blower operation the waste level in the waste container is checked Due to volume reduction inherent with the drying cycle, additional waste may have to be added to the waste container. To add waste, the drying cycle is secured and waste is added per the filling and dewatering cycles. Once additional waste is added the drying cycle is resumed. Several iterations of blower operation and waste addition may be required When 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of blower operation have been completed, and waste addition is not required, the Relative Humidity Monitoring System is placed in operation. This system monitors the relative humidity of the air stream from the waste container before the air enters the entrainment separator tank. The sample tubing is wrapped with heat tape and insulated to prevent condensation in the sample lines The air sample flows through a filter/separator and then into the sample chamber. Air flow is maintained and monitored with a flow meter. A General Eastern 100DP Dew Point Hygrometer measures the system air temperature and the system air dew point temperature.
From this data, the percent (%) relative humidity of the entrainment separator inlet air is determined.
The air temperature and dew point temperature of the air stream at the inlet of the entrainment separator tank is then monitored and recorded hourly in addition to the temperature and pressure parameters previously noted. The percent relative humidity of the air stream is calculated using the air temperature and dew point temperatures. The drying cycle is considered complete when the percent relative humidity of the air stream reaches the end point relative humidity determined by the NuPac Services Division, Inc. Waste Characterization Form and the NuPac Services Division, Inc. Humidity Endpoint Graph and, when at least eight hours of blower operation has been completed. Once the drying cycle is successfully completed the CNS Dewatering System is shutdown, the fillhead removed from the waste container and the container primary and secondary closure devices installed.
If the percent relative humidity endpoint can not be attained the waste container is retained for corrective disposition. Waste containers which have been successfully CNS PCP 8s7/00 1
dewatered are classified and shipped to a licensed burial site in accordance with the applicable sections of 10CFR61, 10CFR71, and 49CFR Parts 171 through 178.
- d.
Qualification:
Qualification of the DPCP processfordewateringwas conducted in accordancewith reference 19 Qualification of High Integrity Containers used for stabilization was conducted in accordance with references 17 and 18.
- e.
Scaling Factors:
As permitted by 10CFR61.55, waste stream radionuclide scaling factors based on certain gamma emitting nuclides (e.g., Co-60, Cs-137) may be used to calculate the concentrations of difficult-to-measure radionuclides Scaling factor determinations are made using a detailed radionuclide analysis of waste stream samples, by a CNS approved vendor.
- f.
Reports:
The volume, curie content, principle nuclides, and type of waste for the wet wastes dewatered at Cooper Nuclear Station are documented in the Cooper Nuclear Station Radioactive Effluent Release Report This information is listed in the format outlined in Revision 1 of Regulatory Guide 1.21.
G.
Vendor Process Control Programs, Stabilization Processes, and High Integrity Contain Qualifications:
Vendor supplied Process Control Programs, Stabilization Processes and High Integrity Containers which have NRC approved Topical Reports or, which are approved in Low-Level Radioactive Waste Disposal Facility licenses, can be utilized to process radioactive waste at CNS.
H.
District Initiated Changes I-i INFOR
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f !, F 11 11
11,
I
- 1.
Shall be submitted to the Commission by inclusion in the Radioactive Effluent Release Report for the period in which the change(s) was made effective and shall contain:
- a.
Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
- b.
A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and I
I I
I I
I I
- c.
Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
- 2.
Shall become effective upon review and acceptance by the SORC.