ML031140399
| ML031140399 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/24/2003 |
| From: | Cooper D Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML031140399 (149) | |
Text
ODCM Appendix A Revision 10 Issued Date 12/16/02 PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Procedure Sponsor
/
Date WWDoolittle
/
2/3/00 Technical Reviewer Date N/A
/
User Reviewer Date DECooper
/
3/27/00 Plant Manager - Operations Date
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page i TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table of Contents INTRODUCTION 1
DEFINITIONS.................................................................................................................1 A.
CHANNEL CALIBRATION
.1 B.
CHANNEL CHECK
.1 C.
CHANNEL FUNCTIONAL TEST
.2 D.
SOURCE CHECK
.2 E.
OFFSITE DOSE CALCULATION MANUAL
.2 F.
GASEOUS RADWASTE TREATMENT SYSTEM
.2 G.
MEMBERS OF THE PUBLIC
.2 H.
.3 I.
SITE BOUNDARY
.3 J.
UNRESTRICTED AREA
.3 K.
VENTILATION EXHAUST TREATMENT SYSTEM
.3 III.
PROCEDURAL AND SURVEILLANCE REQUIREMENTS AND BASES 4
A.
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.....................................................................................4
- 1.
Requirement
.4
- 2.
Action.4
- 3.
Surveillance Requirements
.4
- 4.
Bases
.5 B.
GASEOUS EFFLUENTS DOSE RATE.10
- 1.
Requirement.......................................................................... 10
- 2.
Action..................................................................................... 10
- 3.
Surveillance Requirements.10
- 4.
Bases.11
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page ii TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table of Contents C.
NOBLE GASES DOSE..................................
12
- 1.
Requirement.......................................................................... 12
- 2.
Action..................................................................................... 12
- 3.
Surveillance Requirements..................................
12
- 4.
Bases..................................
13 D.
1-131,1-133, TRITIUM, AND PARTICULATES.......................
........... 14
- 1.
Requirement.......................................................................... 14
- 2.
Action..................................................................................... 14
- 3.
Surveillance Requirements..................................
14
- 4.
Bases..................................
15 E.
GASEOUS WASTE TREATMENT SYSTEM............................................... 18
- 1.
Requirement.......................................................................... 18(
- 2.
Action..................................................................................... 18
- 3.
Surveillance Requirements..................................
18
- 4.
Bases..................................
19 F.
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 19
- 1.
Requirement........................................................................... 19
- 2.
Action..................................................................................... 19
- 3.
Surveillance Requirements...............................
20
- 4.
Bases...............................
20 G.
LIQUID EFFLUENTS CONCENTRATION 25
- 1.
Requirement.......................................................................... 25
- 2.
Action..................................................................................... 25
- 3.
Surveillance Requirements...............................
25
- 4.
Bases...
26
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page iii TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table of Contents H.
LIQUID EFFLUENT DOSE 29
- 1.
Requirement.........................................
29
- 2.
Action.........................................
29
- 3.
Surveillance Requirements.........................................
29
- 4.
Bases.........................................
30 I.
TOTAL DOSE
....................................... 31
- 1.
Requirement.........................................
31
- 2.
Action.........................................
31
- 3.
Surveillance Requirements.........................................
32
- 4.
Bases.........................................
33 J.
RADIOLOGICAL ENVIRONMENTAL MONITORING.
34
- 1.
Requirement.........................................
34
- 2.
Action.........................................
34
- 3.
Surveillance Requirements.........................................
35
- 4.
Bases.........................................
36 K.
SIRW OR TEMPORARY LIQUID STORAGE TANK.
46
- 1.
Requirement.........................................
46
- 2.
Action............................................4........................................46
- 3.
Surveillance Requirement.46
- 4.
Bases.47 L.
SURVEILLANCE REQUIREMENT TIME INTERVALS.
47
- 1.
Requirement.........................................
47
- 2.
Action.........................................
47
- 3.
Surveillance Requirements.........................................
48
- 4.
Bases...
48
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page iv TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table of Contents M.
SEALED SOURCE CONTAMINATION.
48
- 1.
Requirement...............................................
48
- 2.
Action...............................................
48
- 3.
Surveillance Requirements................................................
49
- 4.
Bases...............................................
50 IV.
REPORTING REQUIREMENTS
.................... 50 A.
RADIOLOGICAL EFFLUENT RELEASE REPORT
........................ 50
- 1.
Supplemental Information......................
51
- 2.
Gaseous Effluents.....................
51
- 3.
Liquid Effluents.....................
53
- 4.
Radiological Impact on Man.....................
54
- 5.
ODCM Changes.....................
55(
B.
RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 55 C.
NONROUTINE REPORTS 56 V.
MAJOR MODIFICATIONS TO RADIOACTIVE LIQUID AND GASEOUS WASTE TREATMENT SYSTEMS 58 A.
LICENSEE MODIFICATIONS 58 B.
DEFINITION OF MAJOR RADWASTE SYSTEM MODIFICATION..
58 TABLES A-1 Radioactive Gaseous Effluent Monitoring Instrumentation A-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements B-1 Radioactive Gaseous Waste Sampling and Analysis Program C-1 Radioactive Liquid Effluent Monitoring Instrumentation C-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements D-1 Radioactive Liquid Waste Sampling and Analysis Program E-1 Radiological Environmental Monitoring Program E-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples E-3 Detection Capabilities for Environmental Sample Analysis F-1 Environmental Radiological Monitoring Program Summary
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 1 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
INTRODUCTION The NRC, through 1 OCFR50.36a, requires implementation of Technical Specifications on effluents from nuclear power plants. NRC Generic Letter 89-01, dated January 31, 1989, allowed relocation of the existing procedural requirements from the Technical Specifications (implemented in Amendment 85, November 9, 1984). The relocated procedural requirements related to gaseous and liquid effluents, total dose, environmental monitoring program, and associated procedural reporting requirements follow below. Programmatic controls are retained in the Administrative Controls section of the Technical Specification to satisfy the regulatory requirements of 1 OCFR50.36a. The Technical Specifications programmatic controls include requirements for the establishment, implementation, maintenance, and changes to the Offsite Dose Calculation Manual (ODCM) as well as record retention and reporting requirements.
II.
DEFINITIONS A.
CHANNEL CALIBRATION
- a Channel Calibration shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The Channel Calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the Channel Function Test.
The Channel Calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
B.
CHANNEL CHECK
- a Channel Check shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 2 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
C.
CHANNEL FUNCTIONAL TEST
- a Channel Functional Test shall be:
- 1.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions.
- 2.
Bistable channels - the injection of a simulated signal into the sensor to verify operability including alarm and/or trip functions.
D.
SOURCE CHECK
- a source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
E.
OFFSITE DOSE CALCULATION MANUAL
- (per Plant Technical Specification) - the Offsite Dose Calculation Manual (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain; 1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by the Technical Specifications and, 2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by the Technical Specifications.
F.
GASEOUS RADWASTE TREATMENT SYSTEM
- any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
G.
MEMBERS OF THE PUBLIC
- all persons who are not occupationally associated with the Plant. This category does not include employees of the utility, its contractors or vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 3 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
H.
- shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10CFR Part 20, 10CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste.
I.
SITE BOUNDARY
- that line beyond which the land is neither owned nor otherwise controlled by the licensee.
J.
UNRESTRICTED AREA
- any area at or beyond the Site Boundary access which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or, any area within the Site Boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
K.
VENTILATION EXHAUST TREATMENT SYSTEM
- any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 4 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
III.
PROCEDURAL AND SURVEILLANCE REQUIREMENTS AND BASES A.
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
- 1.
Requirement The radioactive gaseous effluent monitoring instrumentation channels shown in Table A-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of requirement lIl.B.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
- 2.
Action
- a.
With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above requirement, without delay, suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable or change the setpoint so it is acceptably conservative.
- b.
With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, take the action shown in Table A-1. Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
- 3.
Surveillance Requirements Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated operable by performance of the Channel Check, Source Check, Channel Calibration and Channel Functional Test operations at the frequencies shown in Table A-2.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 5 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 4.
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 alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 1 OCFR Part 20.
The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to IOCFR Part 50.
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL ODCM Appendix A Revision 10 Page 6 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table A-1 Radioactive Gaseous Effluent Monitoring Instrumentation Minimum Instrument Operable Applicability Action Channels
- 1.
WASTE GAS HOLDUP SYSTEM
- a.
Noble Gas Activity Monitor (RIA 1113) Providing Alarm and Automatic Termination of Release (1)
At All Times 1
- 2.
CONDENSER EVACUATION SYSTEM (RIA 0631)
- a.
Noble Gas Activity Monitor (1)
Above 210F 3
Modes 1, 2, 3, 4 Above 210F
- b.
Evacuation Flow Indicator (FI-0631 or Fl-0632)
(1)
Modes 1, 2, 3, 4
- 3.
STACK GAS EFFLUENT SYSTEM
- a.
Noble Gas Activity Monitor (RIA 2326)*
(1)
At All Times 3
- b.
Iodine/Particulate/Sampler/Monitor (RIA 2325)
(1)
At All Times 3
- c.
Sampler Flow Rate Monitor (FE-2346)
(1)
At All Times 2
- d.
Hi Range Noble Gas (RIA 2327)*
(1)
Above 21, 2F 4
Modes 1, 2,3, 4
- 4.
STEAM GENERATOR BLOWDOWN VENT SYSTEM
- a.
Noble Gas Activity Monitor (RIA 2320)
(1)
Above 210F 3
Modes 1, 2, 3, 4
- 5.
MAIN STEAM SAFETY AND DUMP VALVE DISCHARGE LINE
- a.
Gross Gamma Activity Monitor*
1 per Main Above 3250F 4
(RIA 2323 and 2324)
Steam Line Modes 1, 2, 3 He 1e 1e
- 6.
ENGINEERED SAFEGUARDS PUMP ROOM VENTILATION HIGH RADIATION SYSTEM
- a.
Noble Gas Activity Monitor **
(RlA 1810 and 1811) 1 per Room Above 210F Modes 1, 2, 3, 4 5
Setpoints for these instruments are exempted from III.B.1 limits, but are governed by Emergency Implementing Procedures or Operating procedures.
Setpoints for these instruments are exempted from III.B.1 limits, but are governed by Technical Specification 3.3.10.3.
- Documentation of operability not required.
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
ODCM Appendix A Revision 10 Page 7 of 59 Table A-1 (Cont'd)
TABLE NOTATION - ACTION STATEMENTS ACTION 1 -
With the number of channels operable less than required by the Minimum Operable Channels requirements, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:
- a.
At least two independent samples of the tank's contents are analyzed, and
- b.
At least two technically qualified members of the Facility staff independently'verify the release rate calculations and discharge valve line up; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 2 -
ACTION 3 -
ACTION 4 -
With the number of channels operable less than required by the Minimum Operable Channels requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With the number of channels operable less than required by the Minimum Operable Channels requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With the number of operable channels less than required by the Minimum Operable Channels requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:
- a.
Either restore the inoperable channel(s) to operable status within 7 days of the event, or
- b.
Prepare and submit a Special Report to the NRC within 30 days following the event outlining the actions taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.
ACTION 5 -
If either channel fails low or is otherwise inoperable, the ventilation dampers associated with that channel shall be closed immediately and action shall be taken to have the affected channel repaired. The dampers associated with the channel shall not be opened until the affected channel has been declared operable.
(Reference Technical Specification LCO 3.3.10.)
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
ODCM Appendix A Revision 10 Page 8 of 59 Table A-2 Monitorinq Instrumentation Surveillance Requirements Radioactive Gaseous Effluent IntuetChannel Source Channel Channel Modes in Which IntuetCheck Check Calibration Functional Surveillance Test Required
- 1.
WASTE GAS HOLDUP SYSTEM
- a.
Noble Gas Activity Monitor-Providing Alarm and D(4)
P R(3)
Q(1)(2)
Automatic Termination of Release
- 2.
CONDENSER EVACUATION SYSTEM
- a.
Noble Gas Activity Monitor D
M R(3)
Q(2)
Above 2100 F
- b.
Evacuation Flow Indicator (FI-0632) or Modes 1, 2, 3, 4
- c.
Evacuation Flow Indicator (FI-0631)
- 3.
STACK GAS EFFLUENT SYSTEM
- a.
Noble Gas Activity Monitor D
M R(3)
Q(2)
- b.
Iodine Particulate Sampler/Monitor W
M**
R(3)**
NA
- c.
Sampler Flow Rate Monitor D
NA R
NA
- d.
Hi Range Noble Gas D
M R(3)
Q(2)
Above 210F Modes 1, 2, 3, 4
- 4.
STEAM GENERATOR BLOWDOWN VENT SYSTEM Above 2100 F
- a.
Noble Gas Activity Monitor D
M R(3)
Q(2)
Modes 1, 2, 3, 4
- 5.
MAIN STEAM SAFETY AND DUMP VALVE DISCHARGE LINE Above 3250 F
.0 M
R(3)
Q(2)
Modes 1, 2, 3
- a.
Gross Gamma Activity Monitor
________oe__2_
- 6.
ENGINEERED SAFEGUARDS PUMP ROOM VENTILATION HIGH RADIATION SYSTEM
- a.
Noble Gas Activity Monitor 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 31 days 18 months (3) 31 days(1)(2)
Above 210F (Technical Specification Table 4.17.3 Item 4)
Modes 1, 2, 3, 4
{ITS - SR 3.3.10 and SR 3.7.13.1)
At all times other than when the line is valved out and locked.
- Sampler not applicable
- This type of Flowmeter doesn't have any surveillance requirements.
ie-l' 1e I e C
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 9 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table A-2 (Cont'd)
TABLE NOTATION (1) The Channel Functional Test shall also demonstrate that automatic isolation of this pathway occurs if instrument indicates measured levels above the alarm/trip setpoint.
(2)
The Channel Functional Test shall also demonstrate that control room alarm annunciation occurs if either of the following conditions exists.
- a.
Instrument indicates measured levels above the alarm setpoint (not applicable for Item 3.d, Hi Range Noble Gas).
- b.
Circuit failure.
(3)
- a.
The Channel Calibration shall be performed using one or more of the reference standards traceable to the National Institute of Standards and Technology or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range.
- b.
For subsequent Channel Calibration, sources that have been related to the (1) calibration may be used.
(4)
Channel Check shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous or batch releases are made.
TABLE FREQUENCY NOTATION S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M
At least once per 31 days P
Prior to radioactive batch release Q
At least once per 92 days R
At least once per 18 months W
At least once per week
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 10 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
B.
GASEOUS EFFLUENTS DOSE RATE
- 1.
Requirement The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the Site Boundary (see Figure 2-1) shall be limited to the following:
- a.
For noble gases: Less than or equal to 500mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and
- b.
For lodine-1 31, for lodine-1 33, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
- 2.
Action With the dose rate(s) averaged over a period of one hour exceeding the above limits, without delay, restore the release rate to within the
(
above limit(s).
- 3.
Surveillance Requirements
- a.
The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of B.1.a in accordance with the methodology and parameters in the ODCM.
- b.
The dose rate due to lodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the limits of B.1.b in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table B-1.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 11 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 4.
Bases This is provided to ensure that the dose at any time at and beyond the Site Boundary from gaseous effluents from all units on the site will be within 10 times the annual dose limits of 1 OCFR Part 20 to Unrestricted Areas. The annual dose limits are the doses associated with the concentrations of 10 times 1 OCFR Part 20, Appendix B, Table 2, Column 1. These restrictions provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a Member of the Public in an Unrestricted Area, either within or outside the Site Boundary, to annual exposure greater than design objectives of 10CFR 50, Appendix I,Section II.B.1. For Members of the Public who may at times be within the Site Boundary, the occupancy of the Member of the Public will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the Site Boundary. Examples of calculations for such Members of the Public, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding dose rate above background to a Member of the Public at or beyond the Site Boundary to less than or equal to 500 mrems/yr to the total body.
The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300, Currie, L A, "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal Chem 40, 586-93 (1968), and Hartwell, JK, "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 12 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
C.
NOBLE GASES DOSE
- 1.
Requirement The air dose due to noble gases released in gaseous effluents to areas at and beyond the Site Boundary (see Figure 2-1) shall be limited to the following:
- a.
During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
- b.
During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
- 2.
Action With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC within 30 days a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- 3.
Surveillance Requirements Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 13 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 4.
Bases This requirement is provided to implement the requirements of Sections Il.B, Ill.A, and IV.A of Appendix I, 10CFR Part 50. The limiting Condition for Operation implements the guides set forth in Section Il.B of Appendix I. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to Unrestricted Areas will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a Member of the Public through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 1 OCFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.1 11, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.
The ODCM equations provided for determining the air doses at and beyond the Site Boundary are based upon the historical average atmospheric conditions.
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D.
1-131,1-133, TRITIUM, AND PARTICULATES
- 1.
Requirement The dose to a Member of the Public from lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the Site Boundary (see Figure 2-1) shall be limited to the following:
- a.
During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
- b.
During any calendar year: Less than or equal to 15 mrems to any organ.
- 2.
Action With the calculated dose from the release of lodine-131, lodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC within 30 days a Special Report that identifies the cause(s) for exceeding the limit and define(s) the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
- 3.
Surveillance Requirements Cumulative dose contributions for the current calendar quarter and current calendar year for lodine-1 31, lodine-1 33, tritium and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
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- 4.
Bases This requirement is provided to implement the requirements of Sections 1I.C, III.A, and IV.A of Appendix I, 10CFR Part 50. The requirements are the guides set forth in Section II.C of Appendix I.
The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to Unrestricted Areas will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section II.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a Member of the Public through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 1 OCFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.1 11, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water-Cooled Reactors," Revision 1, July 1977.
These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate requirements for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in areas at and beyond the Site Boundary. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.
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Table B-1 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Type Sampling Minimum Type of Lower Limit Frequency Analysis Activity Analysis Detection (LLD)a Frequency (tiCi/ml)
A.
Waste Gas Storage Tank P
Each Tank P
Grab Sample Each Tank Principal Gamma Emittersb 1 x10-4 B.
Containment PURGE P
Each PURGE P
Grab Sample Each PURGE Principal Gamma Emittersb 1 x 10-4 C.
Stack Gas Effluent Wd,e ContinuousC Charcoal 1-131, 1-133 1 x 10o12 Sample Wd,e Continuousc Particulate Principal Gamma Emittersb 1 x 10-Sample (1-131, Others)
Q ContinuousC Composite Sr-89, Sr-90 and 1 x 10-11 Particulate Gross Alpha Sample Noble Gas Noble Gases 1 E-06 ContinuousC Monitor Gross Beta or Gamma I
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Table B-1 (Cont'd)
TABLE NOTATION a
The LLD is defined, in Table E-3, note C.
The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous 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 considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report.
- Ten times the LLD because of low gamma yields.
C The ratio of the sample flow rate to the sample stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with requirements lll.B.1, 111.C.1, and Ill.D.1.
d Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing or after removal from sampler.
e With channels operable on iodine monitor RIA 2325 less than required per Il.A.1, sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, start-up or Thermal Power change exceeding 15 percent of Rated Thermal Power in one hour and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if, 1) analysis shows that the Dose Equivalent 1-131 concentration in the primary coolant has not increased more than a factor of 3, and 2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
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E.
GASEOUS WASTE TREATMENT SYSTEM
- 1.
Requirement When gaseous waste exceeds a Xe-1 33 concentration of 1 E-05 piCi/cc, the Waste Gas Decay Tank System shall be used to reduce radioactive gaseous effluents by holding gaseous waste collected by the system for a minimum of 15 days up to 60 days.
- 2.
Action
- a.
If a waste gas decay tank is required to be released with less than 60 days holdup time, the system waste gas tank contents shall be evaluated and the waste gas decay tank with the lowest Xe-1 33 content shall be released.
- b.
Gaseous waste may be discharged directly from the waste gas surge tank through a high-efficiency filter or from a waste gas decay tank with less than 15 days of holdup directly to the
(
stack for a period not to exceed 7 days if the holdup system equipment is not available and the release rates meet requirements IlI.B, C, and D.
- 3.
Surveillance Requirements Not Applicable.
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- 4.
Bases 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" by meeting the design objectives given in Section Il.D of Appendix I to 10CFR50.
It is expected that releases of radioactive materials in effluents shall be kept at small fractions of the limits specified in 20.1302 of 10CFR20.
At the same time the licensee is permitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power even under unusual operating conditions which may temporarily result in releases higher than such small fractions, but still within the limits specified in III.B, C, and D.
F.
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION
- 1.
Requirement The radioactive liquid effluent monitoring instrumentation channels shown in Table C-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of IlI.G are not exceeded. The alarm/trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Offsite Dose Calculation Manual (ODCM).
- 2.
Action
- a.
With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
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- b.
With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels Operable, take the Action shown in Table C-1. Exert best efforts to return the instruments to Operable status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
- 3.
Surveillance Requirements Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated Operable by performance of the Channel Check, Source Check, Channel Calibration and Channel Functional Test operations at the frequencies shown in Table C-2.
- 4.
Bases The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 1 OCFR Part 20. The Operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 1 OCUR Part 50.
t
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Table C-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Operable Applicability Action Channels
- 1.
GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE
- a.
Liquid Radwaste Effluent Line (RIA 1049)
(1)
For Effluent Releases 1
- b.
Steam Generator Blowdown Effluent Line (1)
For Effluent Releases 2
(RIA 0707)
- 2.
GROSS BETA OR GAMMA RADIOACTIVE MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE
- a.
Service Water System Effluent Line (RIA 0833)
(1)
For Effluent Releases 3
- b.
Turbine Building (Floor Drains) Sumps Effluent (1)
For Effluent Releases 3
Line (RIA 521 1)
- 3.
FLOW RATE MEASUREMENT DEVICES
- a.
Liquid Radwaste Effluent Line (FIC 1051 or (1)
For Effluent Releases 4
1050)
- 4.
CONTINUOUS COMPOSITE SAMPLERS (Alarm/Trip Setpoints are not applicable)
- a.
Turbine Building Sumps Effluent Line (1)
For Effluent Releases 3
- b.
Service Water System Effluent (1)
For Effluent Releases 3
- c.
Steam Generator Blowdown Effluent (1)
For Effluent Releases 3
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Table C-1 (Cont'd)
TABLE NOTATION ACTION 1 -
With the number of channels operable less than required by the Minimum Operable Channels requirement, effluent releases may continue provided that prior to initiating a release:
- a.
At least two independent samples are analyzed in accordance with requirements and
- b.
At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 2 -
With the number of channels operable less than required by the Minimum Operable Channels requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection as specified in Table D-1 for principle gamma emitters and 1-131 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(
NOTE:
The Steam Generator blowdown monitor is normally used in a clean up closed loop system instead of as an effluent monitor. The action statement only applies when the monitor is used as an effluent monitor.
ACTION 3 -
ACTION 4 -
With the number of channels operable less than required by the Minimum Operable Channels requirement, effluent releases via this pathway may continue provided that, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for radioactivity at a lower limit of detection as specified in Table D-1 for principle gamma emitters and 1-131.
With the number of channels operable less than required by the Minimum Operable Channels requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves or tank levels may be used to estimate flow.
I
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Table C-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Source Channel Channel1 Instrument Check Check Calibration Functional Chec Chek Caibraion Test
- 1.
GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE
- a.
Liquid Radwaste Effluent Line (RIA 1049)
P P
R(3)
Q(1)(2)
- b.
Steam Generator Blowdown Effluent Line (RIA 0707)
D M
R(3)
Q(1)(2)
- 2.
GROSS GAMMA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC.
TERMINATION OF RELEASE
- a.
Service Water System Effluent Line (RIA 0833)
D M
R(3)
Q(2)
- b.
Turbine Building (Floor Drains) Sumps Effluent Line (RIA 5211)
D M
R(3)
Q(2)
- 3.
FLOW RATE MEASUREMENT DEVICES (5)
- a.
Liquid Radwaste Effluent Line (FIC 1051 or 1050)
D(4)
NA R
NA
- 4.
TURBINE SUMP EFFLUENT COMPOSITER D(4)
NA NA NA
- 5.
SERVICE WATER SYSTEM EFFLUENT COMPOSITE SAMPLER D(4)
NA NA NA
- 6.
STEAM GENERATOR BLOWDOWN EFFLUENT COMPOSITER D(4)
NA NA NA
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Table C-2 (Cont'd)
TABLE NOTATION (1) The Channel Functional Test shall also demonstrate that automatic isolation of this pathway occurs if instrument indicates measured levels above the alarm/trip setpoint.
(2)
The Channel Functional Test shall also demonstrate that Control Room alarm annunciation occurs if either of the following conditions exists:
- a.
Instrument indicates measured levels above the alarm setpoint.
- b.
Circuit failure.
(3)
- a.
The Channel Calibration shall be performed using one or more of the reference standards traceable to the National Institute of Standards and Technology or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range.
- b.
For subsequent Channel Calibration, sources that have been related to the (a) calibration may be used.
(4)
Channel Check shall consist of verifying indication of flow during periods of releases. Channel Check shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous or batch releases are made.
(5)
Turbine Sump Discharge Flow Meter FQI-5210 was calibrated at factory and doesn't require recalibration.
TABLE FREQUENCY NOTATION D
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Q
At least once per 92 days M
At least once per 31 days R
At least once per 18 months P
Prior to radioactive batch release W
At least once per week N
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G.
LIQUID EFFLUENTS CONCENTRATION
- 1.
Requirement The concentration of radioactive material released in liquid effluents to Unrestricted Areas shall be limited to 10 times the concentrations specified in 1 OCFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microcuries/ml total activity.
- 2.
Action With the concentration of radioactive material released in liquid effluents to Unrestricted Areas exceeding the above limits, without delay, restore the concentration to within the above limits.
- 3.
Surveillance Requirements
- a.
Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table D-1.
- b.
The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of G.1 above.
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- 4.
Bases This requirement is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to Unrestricted Areas will be less than 10 times the concentration levels specified in 10CFR Part 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in Unrestricted Areas will result in exposures within the Section ll.A design objectives of Appendix I, 1 OCFR Part 50, to a Member of the Public. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and 10 times the effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300, Currie, LA. "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal Chem 40, 586-93 (1968), and Hartwell, JK, "Detection Limits for Radioanalytical Counting Techniques,' Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
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Table D-1 Radioactive Liquid Waste Sampling and AnalVsis Program Liquid Release Type Sampling Minimum Type of Lower Limit Frequency Analysis Activity Analysis Detection (LLD)a Frequency (ItCi/ml)
A. Batch Waste Release Tanksb P
P Principal Gamma Emittersc 5 x 107 Each Batch Each Batch 1-131 1 x 1 0-6 P
M Dissolved and Entrained Gases 1 x 10-5 One Batch/M (Gamma Emitters)
P M
H-3 1 x 10'5 Each Batch Composited Gross Alpha 1 x 1 0 q7 P
Q Sr-89, Sr-90 5 x 10 8 Each Batch Composited B. Continuous Releases' Continuous' W
Principal Gamma Emittersc 5 x 10-7 (Turbine Sump, Steam Generator Composite 1-131 1 X 10-6 Blowdown, and Service Water)
M M
Dissolved and Entrained Gases Grab Sample (Gamma Emitters) 1 x 10 5 Continuousf M
H-3 1 x l0o Composite' Gross Alpha 1 x 10-7 Continuoust' Sr-89, Sr-90 5 x 1 0-8 Composite' Frequency Notation P
Prior to batch release M
Calendar month Q
Calendar quarter W
Calendar week
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Table D-1 (Cont'd)
TABLE NOTATION a
The LLD is defined, in Table E-3, Note C.
b A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling.
C The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99*, Cs-134, Cs-137, Ce-141, and Ce-144*. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report.
- LLD - 5E-06 because of low gamma yields.
d 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 that is representative of the liquids released.
A continuous release is the discharge of liquid wastes of a nondiscrete volume; eg, from a volume of a system that has an input flow during the continuous release.
To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected in a series of aliquots of constant volume collected at regular time intervals and combined to form a single sample. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
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H.
LIQUID EFFLUENT DOSE
- 1.
Requirement The dose or dose commitment to a Member of the Public from radioactive materials in liquid effluents released from each reactor unit to Unrestricted Areas shall be limited:
- a.
During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and
- b.
During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.
- 2.
Action With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the NRC within 30 days a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. This Special Report shall also include the results of radiological analyses of the drinking water source.
- 3.
Surveillance Requirements Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once every 31 days.
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- 4.
Bases This requirement is provided to implement the requirements of Sections II.A, Il.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to Unrestricted Areas will be kept "as low as is reasonably achievable." Also, for freshwater sites with drinking water supplies that can be potentially affected by Plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a Member of the Public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 1 OCFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents From Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.
(,
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TOTAL DOSE
- 1.
Requirement The annual (calendar year) dose or dose commitment to any Member of the Public due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
- 2.
Action With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of I I.0.1.a, III.C.1.b, IlI.D.1.a, IlI.D.1.b, IlI.H.1.a, or IlI.H.1.b, calculations should be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of 111.1.1 have been exceeded. If such is the case, prepare and submit to the NRC within 30 days a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 1 OCFR Part 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a Member of the Public from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR Part 190. Submittal of the report is considered a timely request and a variance is granted until staff action on the request is complete.
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- 3.
Surveillance Requirements
- a.
Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with 11I.C.1, 1Il.D.1 and Ill.H.1 and in accordance with the methodology and parameters in the ODCM.
- b.
Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in Action 1.2 above.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 33 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 4.
Bases This requirement is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 1 OCFR Part 20 by 46 FR 18525. It also requires the preparation and submittal of a Special Report whenever the calculated doses from Plant generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except for thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a Member of the Public will exceed the dose limits of 40CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the reactor units and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a Member of the Public to within the 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the Member of the Public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any Member of the Public is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11 and 1 OCR Part 20.2203, is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 1 OCUR Part 20. An individual is not considered a Member of the Public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 34 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
J.
RADIOLOGICAL ENVIRONMENTAL MONITORING
- 1.
Requirement The radiological environmental monitoring program shall be conducted as specified in Table E-1.
- 2.
Action
- a.
With the radiological environmental monitoring program not being conducted as specified in Table E-1, prepare and submit to the NRC, in the Annual Radiological Environmental Operating Report a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b.
With the level of radioactivity as the result of Plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table E-2 when averaged over any calendar quarter, prepare and submit to the NRC within 30 days a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents. When more than one of the radionuclides in Table E-2 are detected in the sampling medium, this report shall be submitted if:
Concentration (1) + Concentration (2) +....> 1.0 Reporting Level (1)
Reporting Level (2)
When radionuclides other than those in Table E-2 are detected and are the result of Plant effluents, this report shall be submitted if the potential annual dose to a Member of the Public is equal to or greater than the calendar year limits of Ill.C.1, llI.D.1, and lll.H.1. 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 Operating Report.
Q,
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 35 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- c.
With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table E-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program.
Identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Annual Radiological Environmental Report.
- 3.
Surveillance Requirements
- a.
The radiological environmental monitoring samples shall be collected pursuant to Table E-1 and shall be analyzed pursuant to the requirements of Table E-1 and the detection capabilities required by Table E-3.
- b.
A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 9 overland meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden of greater than 50 m2 (500 ft2) producing broad leaf vegetation.
- c.
The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report and shall be included in a revision of the ODCM for use in the following calendar year.
- d.
Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the NRC.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 36 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- e.
A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.
- f.
The environmental air samplers shall be operationally checked monthly and airflow verified annually.
- 4.
Bases
- a.
Monitoring Program The radiological environmental monitoring program provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of Members of the Public resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 1 OCFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.
The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table E-3 are considered optimum for routine environmental measurements in industrial laboratories.
Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300, Currie, LA, "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal Chem 40, 586-92 (1968), and Hartwell, JK, "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-15 (June 1975).
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 37 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- b.
Land Use Census:
This requirement is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the radiological environmental monitoring program are made if required by results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10CFR Part 50. Restricting the census to gardens of greater than 50 m2 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 (16 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (ie, similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2.
- c.
Interlaboratory Comparison Program:
The requirement for participation in an approved 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 the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10CFR Part 50.
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
ODCM Appendix A Revision 10 Page 38 of 59 Table E-1 Radiological Environmental Monitoring Program Exposure Pathway Number of Representative Sampling and Type of Frequency and/or Sample Samples and Sample Locationsa Collection Frequency of Analysis
- 1. DIRECT RADIATIONb 21 routine monitoring stations either with two or Quarterly Gamma dose quarterly more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:
An inner ring of stations, one in each overland meteorological sector (9) in the general area of the Site Boundary.
An outer ring of stations, one in each overland meteorological sector (9) within the 12 km range from the site.
The balance of the stations (3) to be placed to l_
serve as control stations.
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
ODCM Appendix A Revision 10 Page 39 of 59 Exposure Pathway Number of Representative Sampling and Type of Frequency and/or Sample Samples and Sample Locatlonsa Collection Frequency of Analysis
- 2. AIRBORNE Radioiodine and Samples from 5 locations.
Continuous sample Radioiodine Canister:
Particulates operation with sample 1-131 analysis weekly for 3 samples from within 6 km of the Site Boundary collection weekly or more each filter change.
in different sectors (2.4 km-SSW, 5.6 km-ESE, frequently if required by and 1.6 km-N).
dust loading.
Particulate Sampler:
Gross beta radioactivity 1 sample from the vicinity of a community having analysis following filter the highest calculated annual average ground changed. Gamma isotopic level D/Q (Covert-5.6 km-SE).
analysise ify ross beta
>1.0 pCi/m.
1 sample from a control location in the least prevalent wind direction' (Grand Rapids 89 km-NNE);
- 3. WATERBORNE
- a.
Lake (surface)
Plant lake water inlet.
Composite sample over Gross beta (>10 pCi/I 1-month periodf.
requires gamma) and tritium monthly.
- b.
Well (drinking)
Samples from Plant, State Park, and Covert Monthly - grab sample.
Gross beta (>10 pCi/l Township Park wells.
requires gamma) and tritium monthly.
- c.
Lake (drinking) 1 sample of South Haven drinking water supply.
Composite sample over Gross beta (>10 pCi/I 1-month periodf.
requires gamma) and tritium monthly.
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL ODCM Appendix A Revision 10 Page 40 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LE1TER 89-01 (TAC NO 75060)
Exposure Pathway Number of Representative Sampling and Type of Frequency and/or Sample Samples and Sample Locationsa Collection Frequency of Analysis
- d.
Sediment from 1 sample from between north boundary and Van Semiannually Gamma isotopic analysise shoreline Buren State Park beach, approximately 1/2 mile semiannually.
north of the Plant discharge.
- 4.
INGESTION
- a.
Milk Samples from milking animals in 3 locations Monthly Gamma isotopic' and 1-131 between 5-13 km distance.
analysis monthly.
1 sample from milking animals at a control location, 15-30 km distance.
NOTE: Samples of 3 different kinds of broad leaf Sample in season or Gamma isotopic analysis' vegetation grown nearest each of two different semiannually if they are on edible portions.
offsite locations of highest predicted annual not seasonal.
average ground level D/Q if milk sample is not performed. (SE or SSE sectors near site).
1 sample of each of the similar broad leaf At time of harvests Gamma isotopice and 1-131 vegetation grown 15-30 km distance in the least analysis.
prevalent wind direction if milk sampling is not performed. (SSW or S sectors).
- b.
Fish Sample 2 species of commercially and/or Sample in season or Gamma isotopice and 1-131 recreationally important species in vicinity of Plant semiannually if they are analysis.
discharge area. 1 sample of same species in not seasonal.
areas not influence by Plant discharge.
- c.
Food Products 1 sample each of two principal fruit crops At time of harvest Gamma Isotopic' and 1-131 (blueberries and apples).
analysis.
I
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 41 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table E-1 (Cont'd)
Table Notation Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program.
b One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors or phosphor readout zones in a packet are considered as two or more dosimeters.
C The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.
d Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
A composite sample is one in which the quantity (aliquot) of liquid samples is proportional to the quantity of liquid discharged and in which the method of sampling employed results in a specimen that is representative of the liquid released (continuous composites or daily grab composites which meet this criteria are acceptable).
9 If harvest occurs more than once a year, sampling shall be performed during each discrete harvest.
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
ODCM Appendix A Revision 10 Page 42 of 59 Table E-2 Renortina Levels for Radioactivitv Concentrations in Environmental Samnles Reporting Levels Analysis Water Airborne Particulates Fish Milk Food Products (pCi/I) or Gases (pCi/rn3)
(pCi/kg, Wet)
(pCi/I)
(pCi/kg, Wet)
H-3 20,000*
Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 1-131 2
0.9 3
100 Cs-134 30 10 1,000 60 1,000 Cs-1 37 50 20 2,000 70 2,000 Ba-La-140 200 300 For drinking water samples. This is 40CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/I may be used.
1r~
PALISADES NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
ODCM Appendix A Revision 10 Page 43 of 59 Table E-3 Detection Capabilities for Environmental Sample Analysisa Lower Limit of Detection (LLD)bc Water Airborne Particulates Fish Milk Food Products Sediment Analysis (pCi/I) or Gases (pCi/m3)
(pCi/kg, Wet)
(pCi/I)
(pCi/kg, Wet)
(pCi/kg, Dry)
Gross Beta 4
0.01 H-3 2,000*
Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 1-131 1d0.07 1
60 Cs-134 15 0.05 130 15 80 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 If no drinking water pathway exists, a value of 3,000 pCi/I may be used.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 44 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table E-3 (Cont'd)
TABLE NOTATION a
This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.
b Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13.
C The LLD is defined as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 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, b
E V 2.22 Y
Exp (-XAt)
Where:
LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume.
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute.
E is the counting efficiency, as counts per disintegration.
V is the sample size in units of mass or volume.
2.22 is the number of disintegrations per minute per picocurie.
Y is the fractional radiochemical yield, when applicable.
X is the radioactive decay constant for the particular radionuclide.
At for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting.
Typical values of E, V, Y, and At should be used in the calculation.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 45 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table E-3 (Cont'd)
Table Notation It should be recognized that the LLD is defined as an "a priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.
d LLD for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 46 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
K.
SIRW OR TEMPORARY LIQUID STORAGE TANK
- 1.
Requirement The concentration of radioactive material contained in the SIRW tank or any unprotected outside temporary tank* shall be limited such that the mixture radionuclides do not exceed 1,000 times the effluent concentration (EC) as listed in 10CFR Part 20, Appendix B, Table 2, Column 2.
Ca + Cb... + Ci = <1000 ECa ECb EC,
- 2.
Action With the quantity of radioactive material in any of the above listed tanks exceeding the above concentration, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this
(
condition in the next Radiological Effluent Release Report.
- 3.
Surveillance Requirement The concentration of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.
or A calculational methodology performed prior to the material being transferred may be used to show compliance with the requirement of this section if a representative sample cannot be obtained at least once per seven days. A representative sample of the radioactive material to be added to the SIRW or Temporary Liquid Storage Tank shall be analyzed and a calculation performed to show compliance with the 1000 EC limit.
Q_
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 47 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 4.
Bases This requirement will provide reasonable assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 1 OCFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an Unrestricted Area.
(The dilution between Palisades and the South Haven drinking water supply has been established as 1000.)
- Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
NOTE:
The limit for the SIRW Tank may be exceeded for operational flexibility if the conditions of this section are met.
L.
SURVEILLANCE REQUIREMENT TIME INTERVALS
- 1.
Requirement Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.
- 2.
Action Failure to perform a Surveillance Requirement within the allowed surveillance interval shall constitute noncompliance with the operability requirements. The time limits of the action requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The action requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowed outage time limits of the action requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Surveillance Requirements do not have to be performed on inoperable equipment.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 48 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 3.
Surveillance Requirements The applicable surveillance interval frequencies are specified in Tables A-2 and C-2. The applicable sampling and/or analysis frequencies are specified in Tables A-1, B-1, C-1, D-1, and E-1.
Extendable surveillance requirements are limited to channel checks, source checks, channel calibrations, channel functional checks, sampling frequencies and/or analysis frequencies.
- 4.
Bases The maximum allowable extension for a surveillance interval is consistent with the surveillance requirements specified in the Technical Specifications, Section 4.0. Until relocated in the ODCM, all of the effluent surveillances were subject to these same requirements.
M.
SEALED SOURCE CONTAMINATION
- 1.
Requirement
(
Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcuries of removable contamination.
- 2.
Action
- a.
With a sealed source having removable contamination in excess of 0.005 microcuries, immediately withdraw the sealed source from use and either:
(1)
Decontaminate and repair the sealed source, or (2)
Dispose of the sealed source in accordance with applicable regulations.
- b.
A report shall be prepared and submitted to the Commission on an annual basis if sealed source leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 49 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 3.
Surveillance Requirements
- a.
Each category of sealed sources as described in the requirement with a half-life greater than 30 days (excluding Hydrogen-3), and in any other form than gas, shall be tested for leakage and/or contamination at intervals not to exceed 6 months.
- b.
The test shall be performed by the licensee or by other persons specifically authorized by the Commission or an Agreement State. The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.
- c.
The test sample shall be taken from the sealed source or, in the case of permanently mounted sources, from the surfaces of the mounting device on which contamination would be expected to accumulate.
- d.
The periodic leak test does not apply to sealed sources that are stored and not being used. These sources shall be tested prior to use or transfer to another licensee, unless tested within the previous 6 months. Sealed sources which are continuously enclosed within a shielded mechanism (ie, sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
- e.
Sealed sources transferred without a certificate indicating the last test date shall be tested prior to being placed in use.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 50 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 4.
Bases The requirement, actions, and surveillance requirements are the same as contained in the Technical Specification 6.21 prior to relocation to the ODCM and will provide assurance that sealed sources are tested to demonstrate that source integrity is being maintained.
IV.
REPORTING REQUIREMENTS A.
RADIOLOGICAL EFFLUENT RELEASE REPORT The Radioactive Effluent Release Report shall be submitted in accordance with 1 OCFR 50.36a by March 31 of each year. 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 (1) consistent with the objectives outlined in the ODCM and Process Control Program and (2) in conformance with 10CFR 50.36a and Section IV.B.1 of Appendix I to 10CFR 50.
The report shall include an estimate of the uncertainty associated with the measurement of radioactive effluents. This error term is included to provide an estimate of the uncertainty and is not to be considered the absolute error associated with the measurements or to be used in determining compliance with these requirements.
These estimates will be based on a statistical analysis of a series of sample results (weighed appropriately for counting statistics) taken once a year from a minimum of one typical gaseous waste tank and from a minimum of one typical liquid waste tank. For noble gases released to the atmosphere from other than the waste gas system the error term will be estimated (and weight-averaged with the waste gas tank error) based on a statistical analysis of a series of sample results taken once a year (or the stack gas monitor counting statistics taken over one release per year) from each source contributing more than 10% of the total annual release.
The error term for iodine and particulates released to the atmosphere will be based on the counting statistics for one stack gas sample taken during the year.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 51 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
The report shall include an estimate of the lower level of detection (in iiCi/ml) if the unidentified portion of the release exceeds 10% of the total annual releases. This estimate of the lower level of detection will be made for those gamma emitting isotopes listed in Appendix B of Regulatory Guide 1.21 (June 1974) and will be provided based on a typical background gamma spectrum.
The report shall provide the following specific terms:
- 1.
Supplemental Information
- a.
Batch Releases:
The report should provide information relating to batch releases of liquid and gaseous effluents which are discharged to the environment. This information should include the number of releases, total time period for batch releases, and the maximum, mean, and minimum time period of release.
- b.
Abnormal Releases The number of abnormal releases of radioactive material to the environment should be reported. The total curies of radioactive materials released as a result of abnormal releases should be included.
- 2.
Gaseous Effluents
- a.
Gases (1)
Total curies of fission and activation gases releases.
(2)
Average release rates (1iCi/s) of fission and activation gases for the quarterly periods covered by the report.
(3)
Percent of limit for releases of fission and activation gases.
(4)
Quarterly sums of total curies for each of the radionuclides determined to be released, based on analyses of fission and activation gases.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 52 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- b.
lodines (1)
Total curies of each of the isotopes, lodine-131, lodine-133 and lodine-1 35 determined to be released.
(2)
Average release rate (jiCi/s) of lodine-131/133.
(3)
Percent of limit for lodine-1 31/133.
- c.
Particulates (1)
Total curies of radioactive material in particulate form with half-lives greater than 8 days determined to be released.
(2)
Average release rate (glCis) of radioactive material in particulate form with half-lives greater than 8 days.
(3)
Percent of limit for radioactive material in particulate form with half-lives greater than 8 days.
(4)
Total curies for each of the radionuclides in particulate form determined to be released based on analyses performed.
(5)
Total curies of gross alpha radioactivity determined to be released.
- d.
Tritium (1)
Total curies of tritium determined to be released in gaseous effluents.
(2)
Average release rate (gCis) of tritium.
(3)
Percent of applicable limits for tritium.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 53 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 3.
Liquid Effluents
- a.
Mixed Fission and Activation Products (1)
Total curies of radioactive material determined to be released in liquid effluents (not including tritium, dissolved and/or entrained gases, and alpha-emitting material).
(2)
Average concentrations ([tCVml) of mixed fission and activation products released to unrestricted areas, averaged over the quarterly periods covered by the report.
(3)
Percent of applicable limit of average concentrations released to unrestricted areas.
(4)
Quarterly sums of total curies for each of the radionuclides determined to be released in liquid effluents based on analyses performed.
- b.
Tritium (1)
Total curies of tritium determined to be released in liquid effluents.
(2)
Average concentrations (uCiml) of tritium released in liquid effluents to unrestricted areas, averaged over the quarterly periods covered by the report.
(3)
Percent of applicable limit of average concentrations released to unrestricted areas.
C.
Dissolved and/or Entrained Gases (1)
Total curies of gaseous radioactive material determined to be released in liquid effluents.
(2)
Average concentrations (giCi/ml) of dissolved and/or entrained gaseous radioactive material released to unrestricted areas, averaged over the quarterly periods covered by the report.
I PALISADES NUCLEAR PLANT 01CM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 54 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
(3)
Percent of applicable limit of average concentrations released to unrestricted areas.
(4)
Total curies for each of the radionuclides determined to be released as dissolved and/or entrained gases in liquid effluents.
- d.
Alpha Radioactivity Total curies of gross alpha-emitting material determined to be released in liquid effluents.
- e.
Volumes (1)
Total measured volume (liters), prior to dilution, of liquid effluent released.
(2)
Total determined volume, in liters, of dilution water used during the period of the report.
- 4.
Radiological Impact on Man The Radioactive Effluent Release Report shall include potential doses to individuals and populations calculated using measured effluent and averaged meteorological data in accordance with the methodologies in the ODCM.
- a.
Total body and significant organ doses (greater than 1 millirem to individuals in unrestricted areas from receiving water-related exposure pathways.
- b.
The maximum offsite air doses (greater than 1 millirad) due to beta and gamma radiation at locations near ground level from gaseous effluents.
- c.
Organ doses (greater than 1 millirem) to individuals in unrestricted areas from radioactive iodine and radioactive material in particulate form from the major pathways of exposure.
Q.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 55 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- d.
Total body doses-(greater than 1 manrem) to the population and average doses (greater than 1 millirem) to individuals in the population from receiving water-related pathways to a distance of 50 miles from the site.
- e.
Total body doses (greater than 1 manrem) to the population and average doses (greater than 1 millirem) to individuals in the population from gaseous effluents to a distance of 50 miles from the site.
- 5.
ODCM Changes The Radiological Effluent Release Report shall include any changes made during the reporting period to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to IlI.J.3.c.
B.
RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT The Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis 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; (1) the ODCM, and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix 1 to 1 OCFR50.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretation and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the Plant operation on the environment. The reports shall also include the results of land use census pursuant to IlI.J.3.c.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 56 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
The Annual Radiological Environmental Operating Reports shall include summarized and tabulated results in the format of Table F-1 of all radiological environmental samples taken during the report period. In the event that some 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 as soon as possible in a supplementary report.
The reports shall also include the following; a summary description of the radiological environmental monitoring program, including sampling methods for each sample type, a map of all sampling locations keyed to a table giving distances and directions from the reactor and the results of land use census required by IlI.J.3.c and results of the Interlaboratory Comparison Program required by III.J.3.e.
C.
NONROUTINE REPORTS A report shall be submitted to the NRC in the event that; 1) the Radiological Environmental Monitoring Programs are not substantially conducted as described in Section lll.J, or 2) an unusual or important event occurs from Plant operation that causes a significant environmental impact or affects a potential environmental impact. Reports shall be submitted within 30 days.
lK
PALISAi NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL ODL Appendix A Revision 10 Page 57 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
Table F-1 Environmental Radiological Monitoring Program Summary Name of Facility Docket No Location of Facility (Countv. State)
Reporting Period Medium or Pathway Type/Total Lower Limit All Indicator b
Control Locations Number of SmldINumber of of*
Locations Name Mean (_
MenfbEPRAL (Unit of Measure)
Analyses Detection Mean (geb Distance & Direction FRangeM n
OCCURRENCES (Unit__of__Measure)
_J Performed (LLD)
Range RangeCURENE Air Particulates Gross B 416 0.003 0.08 Middletown 0.10 (5/52) 0.08 (8/104)-
1 (pCi/rn3)
(200/312) 5 miles 3400 (0.08-2.0)
(0 05-1.40) y-Spec 32 (0.05-2.0)
Cs-137 0 003 0.05 (4/24)
Smithville 0.08 (2/4)
<LLD 4
(0.03-0.13) 2.5 miles 1600 (0 03-0.13)
Ba-140 0.003 0.03 (2/24)
Podunk 0.05 (2/4) 0.02 (1/8) 1 (0.01-0.08) 4 miles 2700 (0.01-0.08)
Sr-89 40 0.002
<LLD
<LLD 0
Sr-90 40 0.0003
<LLD
<LLD 0
Fish y-Spec 8 pCi/kg (dry weight)
Cs-137 80
<LLD
<LLD 90 (1/4) 0 Cs-1 34 80
<LLD
<LLD
<LLD 0
Co-60 80 120 (3/4)
River Mile 35 See
<LLD 0
(90-200)
Podunk River Column 4 Nominal Lower Limit of Detection (LLD) as defined in table notation c of Table E-3.
b Mean and range based upon detectable measurements only. Fraction of detectable measurements at specific locations is indicated in parentheses (f) d NOTE: The example data are provided for illustrative purposes only.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 58 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
V.
MAJOR MODIFICATIONS TO RADIOACTIVE LIQUID AND GASEOUS WASTE TREATMENT SYSTEMS A.
LICENSEE MODIFICATIONS Licensee initiated major modifications to the radioactive liquid and gaseous waste systems.
- 1.
Shall be reported to the NRC pursuant to 1 OCFR 50.59. The discussion of each modification shall contain:
- a.
A summary of the evaluation that led to the determination that the modification could be made in accordance with 1 OCFR Part 50.59.
- b.
A description of the equipment, components and processes involved, and the interfaces with other Plant systems.
- c.
Documentation of the fact that the modification was reviewed K
and found acceptable by the PRC.
- 2.
Shall become effective upon review and acceptance by the Plant General Manager.
B.
DEFINITION OF MAJOR RADWASTE SYSTEM MODIFICATION 1.
Purpose:
The purpose of this definition is to assure that this requirement will be satisfied under clearly identifiable circumstances, and with the objective that current radwaste system capabilities are not jeopardized.
PALISADES NUCLEAR PLANT ODCM OFFSITE DOSE CALCULATION MANUAL Appendix A Revision 10 Page 59 of 59 TITLE: RELOCATED TECHNICAL SPECIFICATIONS PER NRC GENERIC LETTER 89-01 (TAC NO 75060)
- 2.
Definition:
A major radwaste system modification is a modification which would remove (either by bypassing for greater than 7 days or physical removal) or replace with less efficient equipment, any components of the radwaste system:
- a.
Letdown filters or demineralizers.
- b.
Vacuum degassifier (not applicable when the reactor is in cold shutdown and depressurized).
- c.
Miscellaneous or clean waste evaporators.
- d.
The present waste gas compressor/decay tank system.
- e.
Fuel Pool filters/demineralizers.
- f.
Radwaste polishing demineralizers.
- g.
Radwaste Solidification system.
Improvements or additions to improve efficiency will not be considered major modifications unless a complete substitution of equipment or systems is made with equipment of unrelated design. Examples would be; 1) replacement of mechanical degassifier with steam, jet degassifier, 2) replacement of waste gas system with cryogenic system, 3) replacement of asphalt solidification with cement system, and 4) change from deep bead resins to Powdex, etc.
ODCM Appendix B DOCKET 50-255 - LICENSE DPR PALISADES PLANT REQUEST TO RETAIN SOIL IN ACCORDANCE WITH 10CFR 20.302 Revision 0 August 1, 1991 Approved KM Haas 8/20/91 Manager - Radiological Services Date TP Neal 8/1/91 RMC Administrator Date MRN ML Grogan Tech Review PF Bruce 91-082d 8/14/91 Date 8/14/91 Date Applicability Reviewed:
TPNeal 5/5/93 Gerald B. Slade Plant General Manager Applicability Reviewed:
11/30/93 Date S/(LpiL
ODCM Appendix B Page 1 Request to Retain Soil in Accordance with 10CFR 20.302 Consumers Power Company correspondence dated November 12, 1987 and January 25, 1988 requested authorization to dispose of contaminated soil in place as specified by 10CFR 20.302.
The area known as the South Radwaste Area has been contaminated by numerous cooling tower overflows and contamination was redistributed by heavy rain showers. Although the majority of the radioactive material has been packed for waste shipment, a large volume of very low activity radioactive material remains.
This volume of material would be very expensive to ship as waste.
The specific area contaminated is noted as AreaB on the attached survey grid map in reference 1. The entire area is fenced and is about 12,000 sq ft of soil exposed with the remainder buildings and asphalt. The inhalation pathway is for breathing suspended soil from this area.
The radworker could receive 8.03E-04 mRem/50-year maximum organ (liver) dose and the infant could receive 3.16E-05 mRem/50-year maximum organ (liver) dose, both of which are insignificant. Direct dose to a radworker is less than 2E-03 mRem/hr. Occupancy in this area should not average more than 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s/week or 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />s/year, which would result in a dose of <1 mRem/year.
The radwaste activities which caused the contamination of the soil were completely relocated to a new east-radwaste area.
The South Building has been deconned and is being used for non-radwaste activities. Some fixed contamination is present in floor cracks and vaults.
This has been documented for plant decommissioning. No further contamination will be added to the south area from the South Radwaste Building. In spite of this commitment, revocation of Michigan shipping privileges in November 1990 require the use of this area to store packaged low level radioactive waste (LLV).
Use of this building is addressed in CPCo's letter to NRC Document Control Desk, April 24, 1991 which is entered as reference 6.
This LLW, in the form of dry active waste (DAW) will be packaged in metal boxes and labeled, ready for future shipment to burial sites. The DAW metal shipping boxes will be stored off the floor to prevent water damage. The metal shipping boxes are strong, tight containers designed to prevent any leakage of radioactive material during transportation. Incidental water contact will not result in the spread of contamination.
Radioactive waste will not be processed in the South Radwaste Building and the building will be maintained as a normally clean (radiologically) area.
Rev 0 8-1-91
ODCM Appendix B Page 2 References (1)
CPCo's letters, T.C. Bordine to NRC Document Control Desk, November 12, 1987 and January 25, 1988.
(2)
Memorandum from L.J. Cunningham, DREP to T.R. Quay, T.V. Wambach, "Request for Additional Information (RAI)",
Harch 15,
- 1988, April 7,
- 1989, and January 12, 1990.
(3)
CPCo's supplement to Reference (1),
J.L. Kuemin to NRC Document Control Desk, June 27, 1988.
(4)
CPCo's supplement to References (1, 2), G.B. Slade to NRC Document Control Desk, August 31, 1990.
(5)
CPCo's letter, T.P. Neal to B. Holian, October 23, 1990. (Typo of 10/13/90 in original reference).
(6)
CPCo's letter, G.B.
Slade to NRC Document Control Desk, April 24, 1991.
"Use of South Storage Building as an Interim Radioactive Waste Storage Building".
(7)
NRC Letter, Brian Holian to G.B. Slade, CPCo, June 7, 1991, "Approval and Conditions to Retain Soil in Place".
Rev 0 8-1-91
ODCM - APPENDIX B REFERENCE 1
,l:
consumers Power POWERING MIEHIGAN¶S PROGRESS General Offices:
1945 West Parnall Road, Jackson, MI 49201 * (517) 788-0550 November 12, 1987 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 -
LICENSE DPR PALISADES PLANT -
REQUEST TO RETAIN SOIL IN ACCORDANCE WITH 10CFR20.302 N
The Code of Federal Regulations, Title 10, section 20.302 allows for approval of proposed procedures to dispose of licensed material in a manner not other-10 wise authorized in the regulations.
Flooding of the South Radwaste Building has caused contamination of 4,173 cubic feet of soil with 2,992.6 VCi of Cs-137 and 79.3 pCi of Co-60. The area is approximately 30 meters from Lake Michigan.
Site hydrology (Attachment 2, FSAR 2.2) indicates most of the activity will migrate to Lake Michigan in a few years.
In July, 1986 a two-fold evaluation began to identify and map the extent of the ground contamination in the flood plain. The initial findings and evaluation were provided to NRC and the Michigan Department of Public Health by internal letter dated September 26, 1986, to LHueter, NRC, Region III.
Lin Consumers Power Company requests authorization to dispose of this soil inplace as the costs of disposal at a burial ground is estimated at $270,000 while radiological consequences to the general public and site employees is very low.
The activities in the contaminated soil were input as a single radio-active liquid release to Lake Michigan into the NRC LADTAP Code.
The output indicated an estimated wholebody dose to the general public (50 mile radius population 1.05E06) of 1.69E-02 manRem or 1.6E-05 millirem per person. The maximum estimated wholebody dose to an individual would be 5.13E-03 millirem and maximum organ dose (teenage liver) would be 8.67E-03 millirem.
The maximum whole body dose rate was assumed to be at 18 inches from contaminated soil. The maximum whole body dose rate calculated using the Microshield Code was 1.02E-02 mR/hr. Occupancy of this area is controlled by the Radiological Safety Department and secured by a locked fence.
Average yearly occupancy is approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per week per individual for 4 to 5 individuals. A radiation worker should not exceed an additional wholebody dose of 4.08 millirem/year.
Flooding of the South Radwaste Building as a result of the cooling tower overflows is being addressed in two stages. For the short term the cooling tower bypass valve is now electrically isolated during cooling tower opera-tion.
Most previous flooding has been due to instrument failures that cause the valve to open during normal operation.
In addition the South Radwaste Building has been decontaminated to eliminate or minimize contamination that TPN-HP01-NLO1 Rev 0 8-1-9 1
Nuclear Regulatory Commission Palisades Plant 10CFR20.302 Request November 12, 1987 2
could be transported to the environment. A long term solution to remove radwaste activities from this area is being included in the Five-Year Plan.
The activity released to the environment from a flooding release prior to 1986 was estimated and added to the liquid section of the Semi-Annual Radiological Effluent Release Report dated February 28, 1986.
Following approval of this application, it is proposed to account for current activity as an abnormal liquid release included in the semi-annual effluent report. A background, evaluation and survey results discussion follows in Attachment A. Attachment i is the Microshield Code output and Attachment 2 is FSAR section 2.2 (including referenced tables and figures) on site hydrology.
Pursuant to 10CFRI70.12(c) a check in the amount of $150 is attached.
Thomas C Bordine (Signed)
Thomas C Bordine Administrator, Nuclear Licensing CC Administrator, Region III, NRC NRC Resident Inspector -
Palisades Attachment TPN-HPOI-NLO1 Rev 0 8-1-9 1
ATTACHMENT A Consumers Power Company Palisades Plant Docket 50-255 EVALUATION AND SURVEY RESULTS November 12, 1987 10 Pages Rev 0 8-1-91 TPN-HP01-NLOI
=
Background===
In 1986, a soil survey was conducted south of the Turbine Building which included the South Radwaste area.
The survey was conducted due to the South Radwaste Building being in the main flowpath of 'A' Cooling Tower, which has overflowed on three separate occasions in 8 years. The survey found that radioactive material was deposited in the soil due to the flooding of contami-nation and radioactive material areas inside the South Radwaste Building.
Other areas sampled that were not in the flood plain were; liquid radwaste storage tanks, T-90, T-91, storm drains, the beach and the sand dunes.
The a~n survey included a survey grid, surface sample results and core sample results.
All contaminated areas found in Area A (Figure 1) were packaged as radwaste.
In addition, the highest activity areas adjoining the South Radwaste Building were also packaged. A total of 16-98 cubic foot boxes were packaged containing over 85 percent of the estimated activity.
LOl Evaluation In August of 1987, the survey was conducted again to prepare this report and to verify the location of the ground contamination and if any contamination migrated further into the ground since the 1986 survey.
The survey was a two phase evaluation with the first phase being a mapped area consisting of 25' x 25' squares south of the Turbine Building. Once mapped out, surface samples were taken in this area.
The intent of this phase was to accurately map the location and determine the activity in pCi/gram of all ground surface contami-nation. Each surface sample consisted of approximately 20 grams of soil taken Rev 0 MI0987-0066A-NLO2 8-1-91
from the top 1/2" of ground and placed in a petri dish for analysis on the Multi-Channel Analyzer (MCA).
Over 275 samples were collected and analyzed with two surface samples being taken in each sector.
All samples were counted on MCA, Intrinsic Detector #1. Figure 2 shows the sector where activity was detected and their highest levels in VCi/gram.
Phase II was initiated after completion of the "surface" sample analysis.
This consisted of taking core samples in 6" increments where activity was detected.
Core samples were taken until two consecutive core samples reflected no activ-ity.
Core samples were also taken below the activity levels found in the 1986 soil survey until two consecutive core samples revealed no activity.
Figure 3 indicates the depth level where activity was no longer detectable.
For example, 6 inches is indicated in H-10 on Figure 3. This indicates that activity was only detected on the surface. H-9 and 1-10 indicate 18" which means activity was detected only at 12".
Table 1 shows the results summary in pCi/gram of the highest activity at all sample locations.
The sector numbers respond to grid coordinates shown on Figures 1, 2 and 3.
In addition to the sample sectors shown on Figure 1, 25 samples were collected at various locations on site.
These include surface and core samples around T-90, and T-91 on the Northwest side of the Turbine Building (location not shown on figures).
Surface samples were taken under the asphalt around the South Radwaste Building.
These are indicated by a hexagon on Figure 1 in F-ll, 1-12 and K-10 sectors. Core samples taken under the South Radwaste Building are indicated by circles on Figure 1. Of the areas sampled above activity was Rev 0 M10987-0066A-NLO2 8-1-91
found only under the East side of the South Radwaste Building in sector I-9 (Table 1).
In the 1986 soil survey other areas were sampled that were not in the flood plain of the South Radwaste Building. Those included Feedwater Purity Build-ing, North Storage Building, beach areas North and South of Plant, North and Northeast sand dunes and various storm drains.
In all of these areas no activity was detected.
Therefore, they were not sampled in the 1987 soil survey as they were not in the flood plain.
Since the 1986 soil survey, asphalt has been placed over various locations in the protected area. Asphalt was placed around all storm drains and approxi-mately 50% of the South end of the Turbine Building.
Before asphalt was laid down, about 3-6" of the top soil was removed and taken offsite. The soil before leaving site was sampled and counted with no activity detected.
Results To quantify activity and determine impact, the areas of ground contamination were separated into two areas.
Area A which contains all the sectors (A-L, 1-8) North of the "black top" to the Turbine Building. Area B contains all sectors (C-L, 9-14) South of "black top" in the vicinity of the South Radwaste Building.
In Area A no activity was detected, therefore it was not used in determining activity or impact.
Rev 0 8-1-91 M10987-0066A-NLO2
In Area B activity was detected in almost all sectors to the East of the South Radwaste Building (Figure 2).
Activities ranged from 2.07E-6 UCi/gram (E-11) to 3.75E-5 uCi/gram (H-11).
Cs-137 was the primary radionuclide present in all samples with two other samples containing Co-60 1.12E-5 pCi/gram at 6" and 5.80E-6 UCi/gram at 12" (1-9 East Figure 2).
The greatest depth where activity was detected was in sector H-l1 at 18" and when compared to the 1986 soil survey the activity has migrated down into the soil 6" inches further.
Activity was detected at the surface in sectors E-11, E-13, J-12 and L-9 and at 6" in L-9.
This was a result of moving the sand deposited on the asphalt during the flood to these sectors and the movement of soil during the grading and dumping during the asphalting of the South Radwaste area.
Activity in pCi was calculated for each sector (Table 2) by the following formula:
sector ft2 x depth of activity ft x *48144 grams/ft3 x activity (pci/gram) -
pCi.
- Average liter of soil weighed 1700 grams x 28.32 L/ft3 -
48144 grams/ft3.
The first level at which no activity was detected was used to determine depth of activity.
In a few sectors, activity was only detected on the first 1/2" of soil, but for determining cubic feet and activity a depth of 6" was used.
For example, activity for H-10 was calculated as follows:
625 ft2 x.5 ft. depth x 48144 grams/ft3 x 2.6E-6 pCi/gram activity of surface sample equals 39.12 pCi.
Total volume in cubic feet and total activity in pCi were calculated for each sector of Area B. For sectors with activity, the highest activity detected per sector was used in the pCi calculation.
Total contaminated area in Area B is Rev 0 MI0987-0066A-NL02 8-1-91
4173 ft3, total activity is 3071.9 uCi.
Sector H-11 contains 73.5% of the total activity which comprises 14.5% of the total contaminated area of Area B.
To quantify the dose to the population projections, 2992.6 uCi of Cs-137 and 79.3 uCi of Co-60 was entered into the LADTAP computer program.
Assuming that the total 3071.9 pCi was eventually released to Lake Michigan thru the water table, and the uptake pathways which included fish, drinking, swimming, boating and shoreline the 50 mile population estimated at 1.05E6 would receive a total body dose of 1.69E-2 manRem, or 1.61E-5 millirem per person. The maximum wholebody dose to an individual would be 5.13E-3 millirem and maximum organ dose (teenage liver) would be 8.67E-03 millirem.
Direct dose to an individual working in the affected areas was calculated using the MICROSHIELD code. The activities from sectors H-11 and I-9 were used for a dose 18 inches above the surface.
The dose rates from H-11 and I-9 are in 8.75E-06 R/hr and 1.02E-05 R/hr respectively (Attachment 1).
Therefore, a 50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> occupancy in one week could result in a maximum exposure of 0.51 millirem.
Normal occupancy of this area is on an as needed bases and averages less than 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s/week per individual in contact with contaminated soil.
Rev 0 MI0987-0066A-NLO2 8-l-9
Table 1 Soil qamnle Core Results (uCi/gram)
Sector E-1 E-13 H-9 H-10 H-l1 1-9 1-10 J-9 J-12 L-9 T-90 C)
T-91
- ve9 #1 0O
- 1-9 #2
- I-IO #3 1N
- ]-11 #4
-**K-10 0-.;:-;1-12
':.':"*I -
1 1 Surface 2.07E-6 4.39E-6 4.19E-6 2.60E-6 3.75E-5 1.24E-5
<tMDA 5.39E-6 6.39E-6
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA CHDA
<MDA
<MDA
<MDA 6"
12"1
<MDA
<MDA
<MDA
<MDA
<MDA 4.79E-6
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA 5.39E-6
<MDA C<DA
<MDA
<MDA 6.77E-6
<MDA
<MDA
<MDA
<MDA
<MDA 1.40E-5* 5.80E-61
<MDA
<MDA I Ro 2t4 30"11 36" 42"
.flLJA
<MDA
<MDA
<MDA
<MDA 8.45E-6
<MDA
<MDA
<MDA
<MDA
<MDA
<HDA
<HDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA Activity is all Co-60
- Includes 1.12E-5 pCi/gram of Co-60.
All other activities listed were identified u1) as Cs-137.
Core samples under foundation of the South Radwaste Building.
tN.
Surface samples under asphalt in South Radwaste area.
I (D
H ti1I098b-0062A-111O01
Table 2 Activity Calculations per Sector Area B Sector #
E-11 E-13 H-9 H-10 H-11 1-9 I-1l J-9 J-12 L-9 I-9 east Sq.ft.
X Depth =
ft' X
g/ft' X uCi/g
=
Total uCi 375 375 625 625 625 527 275 450 200 150 98 4325 0.5 0.5 1.5 0.5 2.0 0.5 1.5 0.5 0.5 1.0 1.5 187.5 187.5 937.5 312.5 1250. 0*
263.5 412.5 225 100 150 147 4173 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 2.07E-6 4.39E-6
- 4. 79E-6 2.60E-6
- 3. 75E-5 1.24E-5 5.39E-6 5.39E-6 6.39E-6
- 6. 77E-6 18.7 39.6 216.2 39.1 2256.8*
157.3 107.0 58.4 30.8 48.9 99.1 3071.9 48144 1.40E-5
- 1250
- 2256.8 =
73.5% of total activity Rev 0 8-1-91 MI0986-0062A-HP01
FIGURE 1 SURVEY GRID LEGEND E
ASMALT AREAS Lull25'X 25' SECTOS
/
lz AREA AREA A
/
N.
Itt
-1I (D
II H
AUGUST 1987 FIGURE 2 A.
I II I4 -
e.
C.A VI is I <
a t-IUR FACE
. I I
/
p i_
JS
/Z1/
If;7 I
1a
'A 7
_ I s r
f 7 t3 mk'~
Ri
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Rev 0 8-1-91 Attachment I Page 8 of 8
ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 FSAR SECTION 2.2 -
SITE HYDROLOGY November 12, 1987 10 Pages Rev 0 8-1-91 TPN-HP01-NLO1
2.2 HYDROLOGY The Palisades Plant site is surrounded on the north, east and south sides by sand dunes.
The west side of the site is the Lake Michigan shoreline.
As a result of this local topography, the site drainage is independent of the Brandywine Creek drainage basin which drains the hinterland.
All sur-face water runoff drains directly to the lake and the percolating runoff also discharges to the lake (see Reference 3).
There are no data available to verify the amount of surface runoff from the site; however, the flow from the Brandywine Creek drainage basin should be useful for the purpose of comparison.
Data obtained to establish base flow figures for Van Buren County streams indicate that the Brandywine Creek drainage basin is about 17 square miles (see Reference 4).
The average annual rainfall for the area is 34 inches.
During the period September 1962 to October 1963, the base flow measure-ments varied from a minimum of 0.90 cubic feet per second (ft3/s) to a maximum of 11.4 ft3/s. This resulted in a mean annual 7-day minimum flow of 1.6 ft3/s or 0.094 ft3/s/sq mi (cubic feet per second per square mile).
The period of stream measurements was representative of drought conditions.
The deposits of Brandywine Creek drainage basin are of low permeability which results in a nearly total runoff to Lake Michigan. This runoff prob-ably occurs soon after precipitation.
Minor groundwater storage in the old beach and reworked older sandy lake deposits observed on the surface to the east of the site area probably maintain Brandywine Creek during periods of low rainfall.
2.2.1 GROUNDWATER Almost all the water used in Van Buren County is obtained from wells.
Exceptions are the City of South Haven that obtains its municipal supply from Lake Michigan and some irrigation supplies that are obtained from streams, lakes and local ditches (see Reference 4).
The glacial drift is the only known source of fresh groundwater in the county.
All the glacial deposits are capable of yielding some water to wells, but the sand and gravel outwash deposits yield the largest quantities (see Reference 4).
The area of sand dunes along Lake Michigan is not generally favorable for obtaining large supplies of groundwater.
Probably most of the dune sand is above the water table and most wells must be drilled into the underlying.
lake deposits (see Reference 4).
- 1. General Groundwater levels were established by the 1966 Geology and Groundwater Investigation conducted by Bechtel Company for Consumers Power Company (see Reference 3).
The results of the investigation are shown on Figure 2-9.
It is readily apparent that subsurface drainage is generally westward fs1281-1291a-09-72 2.2-1 Rev 0 8 91
toward the lake (see Profile A-A).
Minor variations; Le, flow toward surface streams, may exist but are not considered significant.
An average hydraulic gradient toward the lake of about 13 feet per mile.as obtained along Profile A-A as shown on Figure 2-9.
This gradient repre-sents only the upper surface of unconfined groundwater.
Water released on the surface would move toward the lake at an estimated rate of 650 feet per year (see Reference 3).
The nearest domestic wells to the site are located one-half mile to the east and south.
The data indicates that groundwater ia the vicinity of the eastern wells is flowing west toward the site.
Local groundwater in the area of the southern wells is also flowing west toward the lake, perpendLc-ular to the shoreline.
There are no major sources of groundwater withdrawal, eg, large-scale industrial or agricultural pumping, that might reverse the direction of groundwater flow and cause groundwater to flow from the Plant area toward any existing domestic wells.
Without such pumping, Lt is difficult to envision a condition which would cause sufficient groundwater lowering at any of the domestic wells such that the direction of flow might be reversed.
- 2. Plant Site Groundwater levels in the vicinity of the site are shown on Figure 2-9.
The water table generally slopes toward the lake.
During the site LnvestL-gations, groundwater elevations averaged 580 feet MSL beneath the building site.
This elevation corresponds to the approximate meaa level of Lake Michigan.
As shown by water levels measured during drilling,.groundwater levels rise to the east to approximately 604 feet MSL beneath the switch-yard and 601 feet MSL near the eastern site boundary (see Reference 3).
Field permeability tests performed during the 1965 exploratory drilling yielded values ranging from 30 to 1,720 feet per year in the site area, Table 2-11.
In Drill Hole 5, located approximately 500 feet northwest of the containment building, the permeability values ranged from 30.4 feet per year to 143 feet per year.
In Drill Hole 7, located approximately 650 feet south of the containment building, the permeability values ranged from 156 feet per year to 1,720 feet per year.
- 3. Groundwater Movement An unconfined aquifer is present in the dune area with groundwater levels controlled by the level of Lake Michigan. The rate of movement of ground-water downward into material underlying the dunes appears to be very slow.
Nine samples from Drill Hole 22 in the site area were tested for sodium ab-sorption ratio (SAR), Table 2-12. A high SAR indicates poor downward per-colation of water due to sodium deposition on and between soil particles.
At the Plant site, the SAR is considered to be high between elevations 596 and 566 feet MSL and low between 566 and 555 feet MSL (see Reference 3).
fs1281-1291a-09-72 2.2-2 Rev 0 8-1-91
Groundwater levels and permeability data from the sandy lake deposits underlying the dunes indicate a slow rate of discharge into Lake Mtchigan
- 4.
Conclusions
- a. Groundwater in the unconfined aquifer moves westerly from the Brandywine Creek basin to Lake Michigan.
- b. The hydraulic gradient is approximately 13 feet per mile and flow is essentially perpendicular to the shoreline.
- c. Water discharged on the ground surface at the Plant site will percolate downward at a slow rate and mix with groundwater moving toward Lake Michigan.
- d. Infiltration of surface water from the site to domestic wells offsite does not appear to be possible under present groundwater conditions.
2.2.2 GENERAL LAKE HYDROLOGY
- 1. Lake Levels N4 The level of Lake Michigan is cyclic and is expected to fluctuate with time and is dependent on long-term above-normal or below-normal amounts of pre-cipitation.
The highest monthly mean stage of Lake Michigan was 583.68 feet MSL in 1886.
Subsequent modifications in the St Clair River and the opening of the diversion out of the basin at Chicago have tended to reduce the maximum level attainable.
During the recent period of record (1900 to present), the highest recorded monthly mean stage was 582.6 feet MSL in July 1974, and the lowest monthly mean stage was 576.91 feet.MSL in March and April 1964 (see Reference 5).
Great Lakes levels are reported using International Great Lakes Datum which is converted to MSL at the Palisades site by adding 1.558 feet.
The 1.558-foot correction factor is taken from the reference point at St Joseph, Michigan.
Short-time variations in lake levels (seiches), caused by meteorological factors and measured in hours rather than days, occur occasionally.
The greatest level change of this type on record over a 105-year period in-volved a sudden'rise of 6 feet at Michigan City, Indiana (8:10 AM, June 26, 1954) and a rise of 8 feet at Montrose Harbor, Chicago (9:30 A.M on the same date) (see Reference 6).
These seiches were reported in the "Science" article by Ewing, Press and Donn (Vol 120, Page 684).
On passing into the shallow water at Michigan City, the wave was reflected and refracted to reach the Chicago shore of the lake.
The US Lake Survey gauge at Holland, Michigan, which is 30 miles north of the Palisades site and has similar lake geometry to the site, indicated no surge on June 26, 1954.
As part of the Systematic Evaluation Program (SEP Topic II-3.B), the maxi-mum probable surge elevation was reevaluated.
The offshore surge value was reevaluated to produce an onshore surge height of 10.9 feet.
The maximum monthly mean level was also reduced from 583.6 feet MSL to 582.6 feet MSL.
fs1281-1291a-09-72 2.2-3 Rev 0 8-1-91
This resulted Ln a probable maximum flood protection Level for the Palisades Plant of 593.5 feet MSL.
The service water pump motors at 394.7 feet MSL provide the basis for determining the minimum flood protection requirements for the Plant.
Therefore, the resultant wave surges from Lake Michigan do not present a problem at Palisades.
- 2. Water Movements Conclusions from a study of lake hydrology in the Palisades Park, Michigan area by Dr J L Hough (see Reference 6) indicate that surface currents gen-erated by wind conditions and modified by the earth's rotation and lake configuration will provide adequate mixing of Plant liquid effluents into the lake.
The study included actual measurements of lake water movement in the area near the Plant site, and water mixing where the Black River enters Lake Michigan at South Haven.
A summary of the study is as follows:
Lake water is almost constantly moving past the Palisades site, with an appreciable velocity of flow, under the influence of winds.
It is es-timated, on the basis of wind records, that an alongshore current flows northward about 33% of the time and an alongshore current flows south-ward about 23% of the time.
Offshore drift of surface water should occur about 38% of the time, according to frequency of offshore winds, but these would have a minimal effect close to shore, which is bordered by a high dune ridge.
It is likely, therefore, that the alongshore currents would tend to persist, once set up, while offshore winds were blowing.
Thus, the frequency of alongshore current flow is probably greater than the 33% and 23% based oa wind directions.
Under the procedure of taking water from a depth of about 20 feet, 3,500 feet offshore, raising its temperature as it is used for service water and dilution of cooling tower blowdown, and returning the efflu-ent to the lake near shore, the effluent water will almost always be warmer than the lake water into which it is discharged. This is because-a single Lake Michigan water mass is involved during most of the year.
When the effluent is warmer, it will tend to float at the surface, to drift with the surface current, and to be mixed by surface turbulence due to wave action.
On rare occasions, during the spring warming period when the upper layer of lake water is less than 20 feet deep, and during the summer when strong offshore winds cause a thinning of the normally deep surface mass to less than 20 feet, the intake water coming from a colder layer may not be warmed in the Plant suffic-iently to have a temperature higher than that of the surface lake water.
At such times, the effluent water will tend to sink to the thermocline and it will not be subject to vigorous turbulence caused by surface wave action.
It will tend to mix more slowly.
Surveys of the performance of Black River water, entering Lake Michigan at South Haven under various weather conditions, have indicated that fs1281-1291a-09-72 2.2-4 Rev 0 8-1-91
the river water is diluted rapidly, reaching a concentration oE abot only 1% in the lake within a mile of the river mouth.
The discharge of the Black River was evaluated because the rate was deter-mLned to be nearly the same as the discharge rate from the Palisades Plant with once-through cooling.
Since the Plant is now operated with cooling towers, the-discharge to the lake has been reduced to approximately 60,000 gpm or about 1/7 the original rate.
The mixing and dilution factors are considered to be as great as during the higher discharge periods and the discharge concentrations should be diluted at least 1,000 times by the time the discharge could reach the public water intake at South Haven, Michigan.
- 3. Conclusions
- a. The level of Lake Michigan is cyclic; however, the recorded high of 1886 is unlikely to be exceeded.
High lake levels are not expected to present a problem at the Plant site.
- b. There is no recorded evidence of short-time variations in lake levels (seiches) along the eastern shore of Lake Michigan which would be expected to affect the Plant site.
- c. Surface currents generated by wind conditions and modified by the earth's rotation and lake configuration will provide adequate mixing of Plant liquid effluents into the lake.
fs1281-1291a-09-72 2.2-5 Rev 0 8 91
A. -.C"H-."_:. i TABLE 2-11 FIELD PERMEABILITY TEST RESULTS Drill Hole Number 5
Elevation of Test 576 570 565 560 555 550 545 Flow "Q" Head "H" (Gpm)
(Feet)
Permeab i i t K"
(ft/Yr)
(cm
's) 0.0029
- 0. 0101 0.0088 0.0035 0.0136 0.0064
- 0. 0033 12.3 12.3 12.3 12.3 12.3 12.3 12.3 30.4 106.0 92.0 36.8 143.0 67.0 34.6
-4 0.3 x 10,
- 1. i A 10, 0.39 x 10,
0.36 x 10.
- 1. 4 x lo0 0.65 x 1° 4 0.34 x 104 Average 72.8 0.72 x t04 7
580 575 570 565 560 550 545 540 535 0.0303 0.0477 0.0588 0.0588 0.0834 0.3333 0.0677 0.2500 0.2000 25 25 25 25 25 25 25 25 25 156 246 303 303 430 1,720 350 1,290 1,035 1.5 x 10 4 2.4 x 10 4 2.9 x 10 4 2.9 X 10 _4
- 4. 2 X 10 4 16.7 x 10 4
- 3. 4 x 10 4 12.5 x 10 4 10.1 x 10 648 6.3 x 10 4 Average fsl281-1291n-09-72 Rev 0 8-1-91
TABLE 2-12 ANALYSES OF SOIL SAMPLES Sample No 1
2 3
4 5
6 7
8 9
Saturation Extract Values MiIlleguivalents per Liter
_H ECe Calcium
.YagnesLum Sodlum 8.25 1.2 0.5 Trace 11.7 8.4 1.4 0.5 Trace 13.0 8.3 1.3 0.5 Trace 12.3 8.45 1.4 0.5 Trace 14.4 8.5 1.5 0.5 0.1 14.8 8.3 1.5 0.5 Trace 14.8 8.5 1.3 0.5 0.05 12.7 8.2 0.5 3.0 0.4 1.1 8.1 0.6 3.4 0.7 2.4 SAR 23.5 26 24.5 29 27 29.5 24 1
1.5 Sample Descript':Dn DH 22 El 396 DH 22 El 391 DH 22 El 386 DH 22 El ;81 DH 22 Ei 576 DH 22 El 571 DH 22 El 566 DH 22 El 561 DH 22 El 55 ECe = Millimhos per centimeter SAR = Sodium adsorption ratio on saturation extract fs1281-1291o-09-72 Rev 0 8-1-91
I SIE PROPERIY BOUNDARY A EXCLUSION AREA POWER BLOCK PLAN CIRCULAIING WATER COOLING TOWERS SITE CONTOUR INTERVALS Of 20' IINCR"IIS 3300 foor INIAKE PIPE 10 CRIB DISChARGE STRUCTURE PARKING TOPOGRAPHIC ELEVATIONS IN ffEE ABOVE MEAN SEA LEVEL N.
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Lfl CONSUMERS POWER COMPANY PALISADES PLANT FSAR UPDATE SITE LAYOUT f IGURE N0 2-2 R1VISION N0 D
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( g consumers Power POWERING MICHIGAN'S PROGRESS General Offices: 1945 West Parnall FRoad, Jackson, Ml 49201 * (517) 788-0550 January 25, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
SUPPLEMENT - REQUEST TO RETAIN SOIL IN ACCORDANCE WITH 10CFR20.302 Consumers Power Company letter dated November 12, 1987, requested authori-zation to dispose of soil in place as specified by 10CFR20.302. The letter included the results of a survey and evaluation performed in August, 1987.
Following submittal of that letter, a cooling tower overflow on November 13, 1987 again flooded the South Radwaste Building. The flooding caused additional activity that necessitated Consumers Power Company to request placing our November 12, 1987 request on hold until further evaluation and surveys could be completed.
This letter includes the results of our evaluation and survey of the November 13, 1987 flooding incident and is intended to supplement our original November 12, 1987 request.
Following the cooling tower overflow, a survey indicated additional activity had been released from the building. The building was being maintained in a non-contaminated condition to prevent this type of occurrence; however, during this period a destructive testing program on waste packages was being con-ducted in a small area of the building. The survey clearly showed the release of activity from the building adjoining the testing area. The top six inches of soil from the sectors adjoining the building were removed and packaged (588 cubic feet) to prevent additional dispersion of radioactivity.
The area was then completely resurveyed.
An evaluation of the August 1987 and November 1987 (post packaging) activities is attached. The survey indicates a drop of 49% in activities between the August and November surveys. We propose the activities specified in the November 12, 1987 submittal be used as justification for the request because they are conservative. In addition since the November 13, 1987 flooding and following the most recent survey the area was subject to heavy rains which could have diluted some activities to below minimum detectable activity (MDA is nominally 1E-06 pCi/g).
Rev 0 OC0188-0018-NLO2 8-1-91
Nuclear Regulatory Commission 2
Palisades Nuclear Plant Retain Soil in Accord. with 10CFR20.302 January 25, 1988 The one non-conservative value from our August survey and evaluation is the maximum dose rate at 18 inches above the surface.
The November survey value from MICROSHIELD is 1.17 mR/hr as opposed to 1.02 mR/hr. This small increase only slightly changes the radiation workers' conservative dose estimate from 4.08 mR/year to 4.7 mR/year.
Following approval of this application, it is proposed to account for the most conservative values of activity, which was stated in the November 12, 1987 submittal, as an abnormal release in the semi-annual report.
In order to prevent recurrence of these releases to the environment, Consumers Power Company is also committing to transfer radwaste activities from this area, except for high level vault use which is not a potential flooding release problem.
Relocation of these activities to a new radwaste facility is currently scheduled to be completed in 1988.
A check in the amount of $150.00 was attached to our November 12, 1987 submittal pursuant to IOCFR170.12(c).
Thomas C Bordine (Signed)
Thomas C Bordine Administrator, Nuclear Licensing CC Administrator, Region III, NRC NRC Resident Inspector - Palisades Attachment Rev 0 8-1-91 OC0188-0018-NLO2
Attachment A Consumers Power Company Palisades Nuclear Plant Docket 50-255 Evaluation and Survey Results Comparison Post November Flood and Packaging Versus the November 12, 1987 Submittal Rev 0 8-1-91 MI0188-OOO1A-HP01
'A' Cooling Tower -
South Radwaste Flood In August of 1987, a resurvey was conducted of the soil at the South Radwaste Building and its adjacent areas.
The resurvey was conducted to verify the location of ground contamination and if any contamination migrated further into the ground since the 1986 survey.
After submittal of the 1987 soil results and request to retain, in accordance with 10CFR20.302, 'A' Cooling Tower Basin overflowed again flooding the South Radwaste Building and outlying areas.
Immediately following the occurrence, one liter sample was taken with no activity detected on the Multi-Channel Analyzer (MCA).
Then, another complete survey was conducted which included at least two surface samples and core samples in every sector (Figures 1 and 2).
Surface sample results-showed that activities have increased as well as new sectors contaminated.
The most heavily affected sectors were I-10, I-11, J-9 and K-9 (Figure 1A).
To keep these areas from spreading, the top 6" of each of these sectors was removed and placed in 6 LSA boxes (approx. 588 cu.ft) and stored for disposition at a later date.
After removal of soil, the sectors were resurveyed (Figure 1B) and core samples were taken in each sector in 6-inch increments.
Core samples were taken as far down as in the 1986 and August 1987 surveys, and in some instances even further in this survey.
Results showed that no activity was detected below 6 inches as shown in Figure 2.
Table 1 and Table 2 show comparisons between the August and November 1987 soil surveys. Table 1 compares the depths, the activities, the total cu.ft. and total pCi per sector. After the removal of soil, the November 1987 soil survey results showed approximately a 49.3% drop in total contaminated soil (cu.ft.)
and a 51.1% drop in total pCi in comparison to the August 1987 survey results.
In Table 2 the comparison is between sectors affected in each survey and the depth at which each of these sectors were sampled. No activity was detected past 6 inches in the November 1987 soil survey, in comparison to that of 18 inches detected in August 1987.
Direct dose to an individual working in the affected areas was calculated using the MICROSHIELD code. The activities from sectors H-9 and J-9 were used for a dose at 18" inches above the surface. The dose rates from H-9 and J-9 are 9.97E-6 R/hr and 1.17E-5 R/hr, respectively.
Therefore, a 50-hour occupancy in one week could result in a maximum exposure of.59 millirem. Normal occupancy of this area is on an "as needed basis" and averages less than 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s/week per individual in contact with the contaminated soil.
In reviewing the soil results between August and November 1987, the August 1987 soil survey remains more conservative based on the information shown on Tables 1 and 2. Therefore, the August 1987 soil survey is still valid in support of our request to retain the soil in accordance with 10CFR20.302.
Rev 0 8-1-91 M10188-000lA-HP01
Table I Comparison Table Between Total Cu. Ft. and Total UCi August 1987 Sector #
E-11 E-13 H-9 H-10 H-11 1-9 1-10 J-9 J-12 L-9 1-9 East Sq.ft.
375 375 625 625 625 527 275 450 200 150 98 4325 X
Depth 0.5 0.5 1.5 0.5 2.0 0.5 1.5 0.5 0.5 1.0 1.5
=
ft3 187.5 187.5 937.5 312.5 1250*
263.5 412.5 225 100 150 147 4173 X
q/ft3 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 X
uCi/q 2.07E-6 4.39E-6 4.79E-6 2.60E-6 3.75E-5 1.24E-5 5.39E-6 5.39E-6 6.39E-6 6.77E-6 1.40E-5 Total uCi 18.7 39.6 216.2 39.1 2256.8 157.3 107.0 58.4 30.8 48.9 99.1 3071.9
- 1250
- 2256.8 =
73.5% of total activity November 1987 E-11 H-9 H-10 T7 H-11 H-12 I-9 1-12 J-9 J-12 K-9 375 625 625 625 250 527 220 450 200 216 4113 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 187.5 312.5 312.5 312.5 125 263.5 110 225 100 108 2056.5 48144.
48144 48144 48144 48144 48144 48144 48144 48144 48144 1.80E-6 4.35E-5 3.20E-6 3.22E-5 2.20E-6 6.79E-6 3.OE-6 2.05E-5 2.60E-6 3.39E-6 16.25 654.46 48.14 484.45 13.24 86.14 15.89 222.06 12.52 17.63 1570.78 Rev 0 8-1-91 MI0188-000lA-HP01
Table 2 Survey Comparison Between August and November 1987 Soil Surveys August 1987 November 1987 Sector Surface 6"
12" 18" 24" 30" 36" 42" E-11 2.07E-6
<MDA
<MDA
<MDA 1.80E-6
<MDA
<MDA
<MDA E-13 4.39E-6
<MDA
<MDA
<MDA
<MDA N/A
<MDA
<MDA H-9 4.19E-6
<MDA 4.79E-6
<MDA
<MDA
<MDA 4.35E-5
<MDA
<MDA
<MDA
<MDA
<MDA H-10 2.60E-6
<MDA
<MDA 3.20E-6
<MDA
<MDA H-11 3.75E-5
<MDA
<MDA 8.45E-6
<MDA
<MDA
<MDA 3.22E-5
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA H-12
<MDA N/A 2.20E-6 N/A I-9 1.24E-5
<MDA
<MDA 6.79E-6
<MDA
<MDA I-10
<MDA
<MDA 5.39E-6
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA I-li
<MDA N/A
<MDA
<MDA
<MDA
<MDA I-12
<MDA N/A 3.OOE-6 N/A J-9 5.39E-6
<MDA
<MDA 2.05E-5
<MDA
<MDA J-12 6.39E-6
<MDA
<MDA 2.60E-6 N/A
<MDA K-9
<MDA N/A 3.39E-6
<MDA
<MDA
<MDA
<MDA
<MDA
<MDA L-9
<MDA 6.77E-6
<MDA
<MDA
<MDA N/A
<MDA
<MDA co I (D H
M10188-OOO1A-HP01
Mriro~uieid 3.02 (Consumer's Power Company -
+037)
Page
- I SOILI.MSH Run date: January 18, 1988 Run time: 4:17 p.m.
CASE: CONTAMINATED SOIL e H-9 LOCATION 6S INCHES DEEP)
GEOMETRY 11: Rectangular solid source -
slab shields Distance to detector....................,
X Source width.
W Source length.
L Rectangular solid, thickness toward dose pt TI Thickness of second shield.T2
- 60. 960 7G2.
762.
- 15. 240 45.720 CM.
Source Volume: 8.84901e+6 cubic centimeters MrATERIAL DENSITIES (g/cc):
Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium Source
.001220 Shield 2
.001220 1.70 Z.0 Rev 0 8-1-91
-: e: H;,
rL 1, ICH 1,~rMreE ti
<wje¢;7Ws
-'kx;HE-HzE-eUILOUP FACTOR: based on TAYLOR method.
Using the zharacteristics of the materials in shield I.
rNrEGRATrON PARAMETERS:
!iumber of lateral angle segments (Ntheta).....
Number of azimuthal angle segments (Npsi).....
Number of radial segments (Nradius)...........
5 5
S SOURCE NUCLIDES:
Ba-137m:
3.8493e-04 curies RESULTS:
Group 2
3 4
S 6
7 8
10 1 1 12 13 14 15 16 17 18 I19 Energy (MeV)
.664 fActivity (photons/sec) 1.282e+07 1.282e+07 Oose point flux MeV/(sq cm)/sec 4.808e+00 4.80Se+00 Dose rate (mr/hr) 9.969e-03 9.96ge-03 TOTALS:
Rev 0 8 91
tMicrosrneid 3.J2
-'onsumer's 3
wer Ccripany -
- 037) age Pile
- d30L2.1¶SH Run date: January 18, 1983 run time: 4:26 p.m.
CASE: CONTAMINATED SOIL @ J-' LOCATION (6 INCHES OEEP 4GEOMETRY 11: Rectangular solid source -
slab shields Distance to detector......................... X Source width.................................
U Source length................................ L Rectangular solid, thickness toward dose pt... Ti Thickness of second shield................... T2 60.960 762.
548.640 15.240 45.720 CM.
Source Volume: 6.37129e+6 cubic centimeters MATERIAL DENSITIES (g/cc):
Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Ti tanium Tungsten Urania Uranium Water Zirconium Source
.001220 Shield 2
____122_
.001220 1.70 1.0 Rev 0 8-1-91
?-age _
E:NEle: :,I-J7N NZE trCC
';NTAM[NATEL SOIL 8 1-9 ';LOCl7_7N
, 6 !NCHE-DEEF )
BUILDUP FCTORP: based on TAYLOR method.
Using the characterlstics of the materials in shield 1.
INTEGRATION PARPMETERS:
Number of lateral angle segments (Ntheta).....
Number of azimuthal angle segments (Npsi).....
Number of radial segments (Nradius)...........
S 5
5 SOURCE NUCLIDES:
Co-60:
8.9198e-05 curies RESULTS:
Group 1
2 3
4 5
6 7
8 9
Energy (Mev) 1.336 1.180
.695 Activity (photons/sec) 3.300e+06 3.300e+0s 5.383e+02 6.601e+06 Dose point flux MeV/(sq cm)/sec 3.411e+00 2.958e+00 2.889e-04 6.370e+00 Oose rate (mr/hr) 6.1SSe-03 S.497e-03 5.950e-07 1.165e-02 11 12 13 14 15 16 17 18 19 20 TOTALS:
Rev 0 8-1-91
FIGUhE IA SUR FACE LEGEND/
sIMIOMUM DE CTABLE ACTIVITI 71.OOE6 PCIAMIT SURFACE AREA UNSAl ASPHALT J
CONI SAMPLE UNdER FONDATO.
SWWA ACTIm OF SECTOR ASHALT AREAS "EXAMPLE: ONLY H
0~I
>- H a)
I Pi C
FIGUhei lB SUR FACE RES ULTS LEGEND 0 SURFACE I W
COME SAMI r
ASPALT
- EXAMPLI O
DETCTABLE ACTIV.ITI T1OOE-6 PCIIT EA WUNDER ASPHALT 4
PLE UNDER FOUNDATION.
ACTIVITY OF SECTOR AREAS
!: ONLY r-j 0 0\\
I mp co
F IGURE 2 DEPTH SAMPL RESULT LEGEND
~J L5T gICOW 7-WAga -,
AS4, AcmAS~nr In.
EN LEGU
_0OM LV9 XdAT*-Mt m-
-a MPAdMn
- aluM0wwm TV
top REGQt Docket No.
ODCM -
APPENDIX B REFERENCE 2 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 15, 1988 EAR I a r988 NUCLUAR UCENSING 50-255 Mr. Kenneth W. Berry Director, Nuclear Licensing Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201
Dear Mr. Berry:
SUBJECT:
PALISADES PLANT -
REQUEST TO RETAIN CONTAMINATED SOIL IN ACCORDANCE WITH 10 CFR 20.302 (TAC NO. 67408)
The subject request submitted by Consumers Power Company by letter dated November 12, 1987 and supplemented by information forwarded by letter dated January 25, 1988 contains detailed information evaluating the radiation doses via the liquid pathways for very low levels of contamination presently in areas of soil near the Palisades Plant South Radwaste Building.
Detailed evaluations are also presented of potential occupational doses from this contaminated soil.
One additional dose pathway should, however, be evaluated to complete the analysis of the impact viz., the inhalation pathway.
In your submittals, you have presented diagrams showing areas in which contamination has been detected.
It appears that for some of these areas 6" of soil has been removed, others are now covered by black top, and still others have not been disturbed.
In order for the staff to complete the evaluation under 10 CFR 20.302, we ask that you submit a diagram indicating all contaminated soil surface areas included in this request, the condition of this soil surface, and an evaluation of the radiation doses via the inhalation pathway associated with these soil surfaces.
The request in this letter affects fewer than ten respondents; therefore, OMB clearance is not required under PL 96-511.
Sincerely, Thomas V. Wambach, Project Manager Project Directorate III-1 Division of Reactor Projects - III, IV, V
& Special Projects cc: See Next Page D) L
+7 P-9k)f d6)
A4 XI, 34 Rev 0 8-1-91
Mr. Kenneth W. Berry Consumers Power Company cc:
M. I. Miller, Esquire Isham, Lincoln & Beale 51st Floor Three First National Plaza Chicago, Illinois 60602 Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Palisades Plant Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health P.O. Box 30035 Lansing, Michigan 48909 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Mr. David P. Hoffman Plant General Manager Palisades Plant 27780 Blue Star Memorial Hwy.
Covert, Michigan 49043 Resident Inspector c/o U.S. Nuclear Regulatory Commission Palisades Plant 27782 Blue Star Memorial Hwy.
Covert, Michigan 49043 Rev 0 8 91
%,)tA R EG( q 0
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D C. 20555 January 12, 1990 LCA 5 occ6 i).
Y. E D)
JARN 1 I190 I"$I DO III'FN~INQC Docket No. 50-255 Serial No. PAL 90-002 Mr. Kenneth W. Berry Director, Nuclear Licensing Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201
Dear Mr. Berry:
SUBJECT:
PALISADES PLANT - REQUEST TO RETAIN SOIL 10 CFR 20.302 (TAC NO. 67408)
IN ACCORDANCE WITH By letters dated November 12, 1987 and January 25, 1988, Consumers Power Company requested authorization under the provisions of 10 CFR 20.302 to dispose contaminated soil in place.
The NRC staff replied with a request for additional information which was forwarded to you on March.15, 1988.
By letter dated June 27, 1988, Consumers Power Company provided additional -
information in response to our request. However, in that response, CPCo expanded the original request to include the entire South Radwaste area as a contingency against future spread of contamination and to obviate the need for additional requests under 10 CFR 20.302. For the staff to complete its review of this request, additional specific information is required. This is because NRC approval under 10 CFR 20.302 is for the disposal of specifically identified and characterized slightly contaminated material by the applicant.
We request that you provide a revised submittal describing the licensed material for disposal and the analysis and evaluation called for under 10 CFR 20.302.
The attached request for additional information provides additional detail for the content of the revised submittal.
The reporting and/or recordkeeping requirements of this than ten respondents; therefore, OMB clearance under PL letter affect fewer 96-511 is not required.
Sincerely, Albert W. De Agazio, Manager Project Directorate III-1 Division of Reactor Projects - III, IV, V A Special Projects Office of Nuclear Reactor Regulation cc:
See next page 2LW3s2-Rev 0 8 91
Mr. Kenneth W. Berry Consumers Power Company Palisades Plant cc:
M. I. Miller, Esquire Sidley & Austin, 54th Floor One First National Plaza Chicago, Illinois 60603 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health P.O. Box 30035 Lansing, Michigan 48909 Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Comnission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Mr. Gerald B. Slade Plant General Manager Palisades Plant 27780 Blue Star Memorial Hwy.
Covert, Michigan 49043 Resident Inspector c/o U.S. Nuclear Regulatory Comnission Palisades Plant 27782 Blue Star Memorial Hwy.
Covert, Michigan 49043 Rev 0 8-1-9 1
ENCLOSURE SECOND REQUEST FOR ADDITIONAL INFORMATION (RAI)
ON THE CONSUMERS POWER COMPANY PALISADES PLANT REQUEST TO RETAIN SOIL IN ACCORDANCE WITH 10 CFR 20.302 The subject request submitted by Consumers Power Company (licensee) by letter dated November 12, 1987 and supplemented by information forwarded by letter dated January 25, 1988 contained detailed information evaluating the radiation doses via the liquid pathways for very low levels of contamination presently in areas of soil near the Palisades Plant South Radwaste Building.
Detailed evaluations were also presented of potential occupational doses from this contaminated soil.
Three significant questions arose during the staff evaluation of this request:
- 1.
The inhalation pathway for doses from the contaminated soil was not addressed.
- 2.
The proposals contain no delineation of the specific contaminated areas covered by the disposal request.
- 3.
The licensee's Technical Specifications for radiological environmental monitoring require an LLD of 2 x 10 pCi/gm for 137Cs determinations in sediment - yet all of the measurements reported in the request were made with equipment 5 to 10 times less sensitive for these gamma radiations.
Rev 0 8-1-91
By letter dated June 27, 1988 the licensee submitted additional information in response to the staff's RAI of March 15, 1988.
This submittal was unacceptable in that it addressed potentially contaminated areas and hypothetical maximum contamination parameters rather than measured licensed material to be disposed of under the regulations.
It is requested that the licensee submit a complete, revised 20.302 request incorporating the dose evaluation information of the measured contamination considered in the November 12, 1987 and January 25, 1988 submission and updated if appropriate with dose evaluations of the inhalation pathway based on the same measured contamination.
As part of the proposal the licensee should record exactly what areas of measured contamination are covered by the request for which disposal under 10 CFR 20.302 is proposed.
Rev 0 2
8 91
ODCM - APPENDIX B REFERENCE 3 Consumers Power POWERING MICHIGAN'S PROGRESS General Offices:
1945 West Parnall Road, Jackson, Ml 49201 * (517k-788-0550 June 27, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 -
LICENSE DPR PALISADES PLANT -
SUPPLEMENT - REQUEST TO RETAIN SOIL IN ACCORDANCE WITH IOCFR20.302 Consumers Power Company correspondence dated November 12, 1987 and January 25, 1988 requested authorization to dispose contaminated soil inplace as specified by 10 CFR 20.302.
The area, known as the South Radwaste Area, has been contaminated by numerous cooling tower overflows and redistributed by heavy rain showers. Although the majority of the radioactive material has been packaged for waste shipment, a large volume of very low activity-radioactive material remains.
This volume of material would be very expensive to ship as waste. The NRC, by letter of March 15, 1988 to Consumers Power Company, requested additional inhalation dose information and clarification of the contaminated area.
A generic inhalation dose evaluation is described in Attachment A. Conserva-tive assumptions have been made to get the maximum organ dose possible from inhalation of contaminated soil.
The inhalation doses are not significant.
Consumers Power Company requests to expand this 10 CFR 20.302 request to include the entire South Radwaste Area.
Periodic cooling tower overflows and occasional heavy rains have caused redistribution of radioactive material to areas which were below Lower Limits of Detection (LLD) during previous evalua-tions. Expanding the area would eliminate the need for a new 10 CFR 20.302 submittal if radioactive material is redistributed within the South Radwaste Area.
The South Radwaste Area is completely fenced and located directly South of the Plant South Security fence.
Area fence is shown in dark outline on Figure 1.
As described in our January 25, 1988, letter we intend to transfer the radwaste activities which caused the contamination of soil from the South Radwaste Area, except for the high level vault use which is not a potential flooding release problem.
Rev 0 8-1-91 OC0688-0049-NLO2
Nuclear Regulatory Commission 2
Palisades Nuclear Plant Retain Soil in Accordance w/10CFR20.302 June 27, 1988 Consumers Power Company requests approval to dispose of inplace low level radioactive materials which meet the following conditions without further 10 CFR 20.302 submittals.
- 1.
Material contained in the fenced area described as South Radwaste Area.
- 2.
Direct dose to a radiation worker would not exceed 5E-02 mRem/hour from contaminated soil.
- 3.
Average gross beta/gamma concentration not to exceed 5E-05 VCi/gm so inhalation doses to a radiation worker or at the site boundary would not exceed the values contained in Attachment 1.
- 4.
Additional radioactive material releases shall be identified in liquid Semi-Annual Effluents Reports as an 'Abnormal Release'.
Sampling, analyses and Semiannual Effluent Report inclusions of 'Abnormal Release' will be performed only when further flooding of the area occurs.
James L Kuemin (Signed)
James L Kuemin Staff Licensing Engineer CC Administrator, Region III, NRC NRC Resident Inspector -
Palisades Attachment Rev 0 OC0688-0049-NL02 8-1-91 The inhalation doses have been calculated on a generic worst case basis. A generic basis has been selected to compensate for the elevated Lower Limit of Detection (LLD) in analysis and also to address movement of radioactivity within the South Radwaste Area. The assumptions made are the worst case Dose Conversion Factor (DCF) used (see Table 3), a total average activity concentra-tion of 5E-05 pCi/gram and the entire area (500 mi
- 2) used instead of the indi-cated contaminated area (117 m2 ).
Increasing the area is self-explanatory. The total average activity concentra-tion is being used instead of actual to account for dose important isotopes which may be present near the analysis LLD of 1E-06 pCi/gm, but not detected.
The worst case DCF is used to demonstrate a maximum organ dose. A variation in isotope mixes could shift the maximum dose to a different organ but could not exceed the dose indicated.
Radworker and site boundary inhalation dose calculations are attached.
Rev 0 OC0688-0049-NL02 8-1-91
Table I Inhalation Dose From Contaminated Soil -
Adult Radiation Worker S
18 f14 f15 Ef
- DCFi Where:
Cs = concentration of waste:
5.OE04 pCi/Kg.
Ef - occupancy factor: 2080 worker hours - 8760 hrs/yr = 0.237 f
= areal mass available for resuspension (top 1 cm of soil):
16 18 Kg/rn2 f14= resuspension factor:
8.5E-9/m f15 = adult annual inhalation rate:
7300 m3 (RG 1.109)
DCF = 7.46E-04 mRem/(50 yr. pCi):
adult lung (RG 1.109)
Substituting:
DW = 8.78E-03 mRem/50 yr: maximum organ dose
Reference:
AIF/NESP-035 Evaluation of the Potential for De-Regulated Disposal of Very Low Level Wastes From Nuclear Power Plants Rev 0 OC0688-0049-NL02 8-1-91
Table 2 Inhalation Dose At Site Boundary -
Infant Most Limiting D
=C SB S
f18 f14 1l6
- X/Q p
- A
- DCF1 Where: Terms are identified in Table 1 and F
= 2045 m 3: infant annual breathing rate (RG 1.109) 16 X/Q -
1.4 E-6 sec/m 3: actual 5 year site average p = 3.8 m/sec average wind speed:
actual 1986 A = 500 m 2 : contaminated area DCFi = 3.22E-03 mR/(50 yr.
pCi):
infant lung (RG 1.109)
Substituting DSB = 1.19E-04 mRem/50 yr:
maximum organ dose Rev 0 8-1-91 OC0688-0049-NLO2
Table 3 Dose Conversion Factors for Inhalation:
Committed dose (mRem) over 50 years per pCi inhaled, per Regulatory Guide 1.109.
Onsite -
Radiation Worker Bone Liver Kidney Lung GI Cs-134 4.66E-05 1.06E-04 3.59E-05 1.22E-05 1.30E-06 Cs-137 5.98E-05 7.76E-05 2.78E-05 9.40E-06 1.05E-06 Ba-140+D 4.88E-06 6.13E-09 2.09E-09 1.59E-04 2.73E-05 Sr-90*
1.24E-02 0.0 0.0 1.20E-03 9.02E-05 Co-60**
0.0 1.44E-06 0.0 7.46E-04 3.56E-05
- Sr-90 is a factor of 5E-03 lower than Cs-137 based upon 10 CFR 61 sampling analysis and cannot be limiting.
Cs-137 was present in all samples where activity was identified.
Offsite -
Infant Most Limiting for Inhalation Cs-134 2.83E-04 5.02E-04 1.36E-04 5.69E-05 9.53E-07 Cs-137 3.92E-04 4.37E-04 1.23E-04 5.09E-05 9.53E-07 Ba-140 4.OOE-05 4.OOE-08 9.59E-09 1.14E-03 2.74E-05 Sr-90*
2.92E-02 0.0 0.0 8.03E-03 9.36E-05 Co-60**
0.0 5.73E-06 0.0 3.22E-03 2.28E-5
- Sr-90 is a factor of 5E-03 lower than Cs-137 based upon 10 CFR 61 sampling analysis and cannot be limiting.
Cs-137 was present in all samples where activity was identified.
Rev 0 OC0688-0049-NL02 8-1-91
/e FIGUhc 1 SURVEY GiRID LEGEND ASPHALT AREAS 25' x 25' SECTORS A RA Bea 3
AREA AD co w I
(D H
ODCM -
APPENDIX B REFERENCE 4 consumers Power G B Slade
'IC a
eneral lanager POWERING MICHIGAN'S PROGRESS Palisades Nuclear Plant 27780 Blue Star Memorial Highway, Covert, MI 49043 August 31, 1990 Nuclear Regulatory Commission Document Control Desk Washington, D C 20555 DOCKET 50-255 -
LICENSE DPR PALISADES PLANT SUPPLEMENT - REQUEST TO RETAIN SOIL IN ACCORDANCE WITH 10CFR 20.302 Consumers Power Company correspondence dated November 12, 1987 and January 25, 1988 requested authorization to dispose of contaminated soil in place as specified by 10CFR 20.302. The area known as the South Radwaste Area has been contaminated by numerous cooling tower overflows and contamination was redistributed by heavy rain showers. Although the majority of the radioactive material has been packed for waste shipment, a large volume of very low activity radioactive material remains. This volume of material would be very expensive to ship as waste. The NRC, by letter of March 15, 1988 to Consumers Power Company, requested additional inhalation dose information and clarification of the contaminated area.
After discussions with the NRC reviewer, a supplement was submitted on June 27, 1988 which was based on generic approval. It proposed that further submittals would not be required if flooding moved activity from an identified to an unidentified sector. Subsequently, on January 12, 1990, the NRC Staff requested additional information. The information request required the licensee to submit a revised proposal incorporating the dose evaluation information of the measured contamination considered in the November 12, 1987 and January 25, 1988 submittals and updated, if appropriate, with dose evaluations of the inhalation pathway based on the same measured contamination. As part of the proposal, the licensee was asked to record exactly what areas of measured contamination are covered by the request for which disposal under 10CFR 20.302 is proposed.
The attached material supplies the requested information.
The specific area contaminated is noted as Area B on the attached survey grid map.
The entire area is fenced and is about 12,000 sq ft of soil exposed with the remainder buildings and asphalt.
The inhalation pathway is for breathing suspended soil from this area. Table 2 addresses a radworker in Area B, and Table 3 addresses an infant on the site boundary. The radworker could receive 8.03E-04 mRem/50-year maximum organ (liver) dose and the infant could receive 3.16E-05 mRem/50-year maximum organ (liver) dose, both of which are insignif-icant.
Rev 0 8-1-91 OC0890-0074A-NLO3 A CMSY eVENGY COMPANY
4 When the flooding problem was discovered and planning for a formal survey was done, the environmental sediment LLD was not considered, as this was a nuclear plant site. We attempted to get the best LLD we could using our equipment and the number of samples we were going to have to run.
With the hundreds of samples run, we felt lE-06 uCi/gm was adequate.
To be conservative, we expanded the August and November 1987 surveys to use 1.OE-06 uci/gm Cs-137 in any sector which showed LLD. This will increase the radioactivity to 5,006 uCi from 4,643 uCi, an increase of 8%.
The activities are on Table 1 and 1A.
If this submittal is approved, we will add the released activity to the Liquid Semi-annual Effluent Report as an abnormal release, and the approval to retain the soil in place will be documented in the FSAR.
In summation, Consumers Power Company requests approval to dispose of in place the low-level radioactive materials which are contaminated soil contained in the fenced area described as South Radwaste Area (Area B).
Direct dose to a radiation worker would not exceed 1.7E-02 mRem/hour from this contaminated soil.
Inhalation doses to a radiation worker or at the site boundary would not exceed 8.03E-04 mRem/50-year. Tables 1 and 1A radioactive material release shall be identified in liquid Semi-annual Effluent Reports as an
'abnormal release'.
The disposal in place would be documented in the FSAR.
The radwaste activities which caused the contamination of the soil have been completely relocated to a new east radwaste area.
The South Building has been deconned and is being used for non-radwaste activities.
Some fixed contamination is present in floor cracks and vaults. This has been documented for plant decommissioning. No further contamination will be added to the south area from the South Radwaste Building.
Gerald B Slade (Signed)
Gerald B Slade General Manager pc: Administrator, Region III, USNRC NRC Resident Inspector, Palisades Rev 0 8-1-91 OC0890-0074A-NLO3
Attachment A Consumers Power Company Palisades Nuclear Plant Docket 50-255 Tables 1 and 1A, Survey Results Microshield Direct Dose Calculation Table 2 -
Radworker Inhalation Dose Table 3 -
Site Boundary Inhalation Dose Figure 1 -
Survey Grid Figure 2 -
Survey Results Rev 0 8-1-91 M10890-0074A-HP01
Table 1 August 1987-Survey Sector #
Sq.ft.
X Depth
=
E-11 375 0.5 E-13 375 0.5 H-9 625 1.5 H-10 625 0.5 H-li 625 2.0 1-9 527 0.5 1-10 275 1.5 J-9 450 0.5 J-12 200 0.5 L-9 150 1.0 1-9 East 98 1.5 Subtotals 4325 Et' 187.5 187.5 937.5 312.5 1250*
263.5 412.5 225 100 150 147 4173
- X g-ft3 48 144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 x
iCi/g 2.07E-6 4.39E-6 4.7 9E-6 2.60E-6 3.75E-5 1.24E-5 5.39E-6 5.39E-6 6.39E-6 6.77E-6 1.40E-5
=
Total wCi 18.7 39.6 216.2 39.1 2256.8 157.3 107.0 58.4 30.8 48.9 99.1 3071.9 Remainder c Section TOTAL 7613 0.5 3807 48144 1.OE-06 183 3,255.
11,938 Rev 0 8-1-91 M10890-0074A-HPOl
Table IA November 1987 Survey Sector #
B-9 C-9 C-10 D-9 D-10 D-11 D-12 E-10 E-11 E-12 E-13 F-12 F-13 G-12 C-13 H -9
[1-10 H -11 H-12 H-13 I-9 1-10 t-11 1-12 J -9 J-12 K-9 L-9 L-10 L-11 L-12 Sq. tt.
125 625 500 500 625 550 75 125 375 625 550 300 625 750 625 625 625 600 250 625 527 275 250 220 450 200 216 150 150 150 150 X
Depth 0.5 O.5 O.5 0.5 O.5 O.5 0.5 O.5 O.5 0.5 0.5 0.5 0.5 0.5 O.5 O.5 0.5 0.5 0.5 0.5 0.5 O.5 0.5 0.5 0.5 0.5 0.5 O.5 O.5 0.5 O.5
=Et, 62.5 312.5 250 250 312.5 275 37.5 62.5 187.5 312.5 275 150 312.5 125 312.5 312.5 312.5 300 125 312.5 263.5 137.5 125 110 225 100 108 75 75 75 75 X
g/ft' 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 48144 X
pci/g
=
1E-06 1E-06 1E-06 lE-06 1E-06 1E-06 1E-06 1E-06 1.8E-06
- 1. E-06 1E-06 1E-06 1E-06 1E-06 1E-06 4.4E-05 3.2E-06 3.2E-05 2.2E-06 1.OE-06 6.8E-06 1.OE-06 1.OE-06 3.OE-06
- 2. 1E-05 2.6E-06 3.4E-06 1.OE-06 1.OE-06 1.OE-06 1.OE-06 I
Total uCi 3
15 12 12 15 13 2
3 16 15 13 7
15 6
15 662 48 462 13 15 86 7
6 16 227 13 18 4
4 4
4 SubcotaLs 11,938 fta 5,969 ft 3 Maximum 4.4E-05 Average 6.1E-06 1751 Rev 0 8-1-91 M10890-0074A-HPOl
=====S=S=======ssa=s-l2onsuMer' 5 ~ower Copary -
- 037)
Page
- 1 Pile
- SOILI.MSH Run date: January 18, 1988 Run time: 4:17 p.m.
CASE: CONTAMINATED SOrL @ H-9 LOCATION (6 INCHES DEEP)
GEOMETRY 11: Rectangular solid source -
slab shields Distance to detector....................... '- X Source width...............,..I..,., W Source length...,,..,,,,,,,,,,,,,,,
L Rectangular solid, thicknes5 toward dose pt.. Ti Thickness of second shield.T2 so. 960 762.
762.
1S.240 45.7ZO cM.
Source Volume: 8.84901e+6 cubic centimeters MATERIAL DENSITIES (g/cc):
Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium water Zirconium Source
.001220 Shield 2
.001211-1.70 I.a Rev 0 8-1-9 1
Z.,!.
- 'l 3urLcUP AC'R
- nasa am r4YLR "etnocj.
J51mg t.ie ZaraCtar; t irs oN -ne
-ater:Ais IM 5h:e5: I.
- ,.4rE3Per[ON P4RAME7ERS:
N4uoer ai lteral angle segments CJtreta).....
^^c~er :
Z:-utdL a rgle segm ents os).....
umOeret of aal se;memts Nradjus)..........
S S5OURCE NUCLICES:
3a-137m :
3.a493e-04 curies RESUL r:
Actuall 6.51E-04 Ci (4.4E-05 uCi/gm)
Grauo 2
3 4
S 6
7 a
93 Erer-;y
.664 Activity onoters/see) 1.2842e+07 3se aDotnt tIuA MeV/(sq ct)/sec 4.808e+00 Dose rate (mr/hr)
- 8. 963e-03 I I 1 2 13 1 4 1 S I 6 1 7 18 I19 2Z I________
- 1. 282e+07 TOTALS:
- 9. 969e-03 6.51 Ci Ratio of 3.85 Ci = 1.69 x 9.969z-03 = 1.7E-02 r=R/hr Rev 0 8-1-91
Table 2 InhaLation Dose From Contaminated Soil -
Adult Radiation Worker DW = CS f18 f14 E15
- Ef
- DCFi Where:
CS = concentration ot waste -
4.4E04 pCi/Kg Cs-137 (actual max concentration)
Ef = occupancy factor -
2080 worker hours. 8760 hrs/yr = 0.237 f18 = a real/mass available for resuspension (top 1 cm of soil) -
16 Kg/m 2 E14 = resuspension factor -
8.5E-9/m f15 = adult annual inhalation rate -
7300 m3 (RC 1.109)
DCFi = 7.76E-05 mRem/50 yr
- pCi -
Cs-137 adult liver (RC 1.109) tibubstituting:
DW = 8.03E-04 mRem/50 yr -
maximum organ dose aference:
AIF/NESP-035 Evaluation of the Potential for De-regulated Disposal of Very Low Level Wastes From Nuclear Power Plants Rev 0 8-1-91 M10890-0074A-HP01
Table 3 InhaLation Dose At Site Boundary -
Infant Most LimLting DSB = CS E18 f14 t16
- X/Q
- p
- A DCFi Where: Terms are identified in Table 1 and F16 = 2045 m3 -
infant annual breathing rate (RC 1.109)
X/Q = 1.4 E-6 sec/rn actual 5 year site average p = 3.8 m/sec average wind speed -
actual 1986 A
1110 m2 contaminated area DCF; -
4.37E-04 mR/50 yr pCi -
Cs-137 infant liver (RC 1.109) stituting:
DSB = 3.16E-05 mRem/50 yr -
maximum organ dose Rev 0 8-1-91 MI0890-0074A-HPO1
FIGURE lB RES ULTS LEGEND G. All these sectors Assigned IE-06 uCi/gmm l MUM DE CTAEx ACd IT F
B M'
AfAtWR ASPHALTi LCORE SAMPL UNDER FOW/IOM.
E911 MFACE ACTwIMT OF SECTOR
/-
g
/
not ASPHLT AREASno EXAMPLE: ONLY H
o I
- co
FIGURE 1 SURVEY GRID LEGEND 3
ASPHALT AREAS 25' x 25' SECTORS Contami natoa Lil AREA B
3 AREA A/
0 0\\H
,UI
ODCM -
APPENDIX B Ae REFERENCE 6 aria consumers A e Power ri B Stly General Wanag PPOWERING MICHIGAN'S PROGRESS Pasades Nucleiar Plant 27780 Bue Star Memorial Highway Covert. Ml 49043 April 24, 1991 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
USE OF SOUTH STORAGE BUILDING AS AN INTERIM RADIOACTIVE WASTE STORAGE BUILDING On November 10, 1990, radioactive waste generators in the State of Michigan were banned from the three currently active burial sites.
As a result of this ban action must be taken to ensure that Palisades maintains the capability to store radioactive waste until such time as we are again able to gain access to the burial sites. We believe our actions are consistent with NRC guidance received in Generic Letter 90-09. Therefore, the South Storage Building will be utilized as an interim storage facility for low level radioactive waste (LLW).
The South Storage Building (then referred to as the South Radwaste Building) had been previously used for all processing and storing of radioactive waste produced at Palisades from 1976 to 1989.
During that period several cooling tower overflows occurred which resulted in flooding this building and spreading contamination from the processing area to the surrounding soil.
This spread in contamination resulted in NRC Open Items (85019-01 and 89025-
- 01) which required implementation of actions to prevent future flooding.
In 1988 it was decided to relocate the radwaste processing functions performed in the South Radwaste Building to a new addition at the East Radwaste Building to prevent the spread of contamination in the event of future cooling tower overflows.
All radwaste processing equipment was relocated to the East Radwaste Building and the South Radwaste Building was decontaminated.
The South Radwaste Building (then re-named the South Storage Building) has since been used for non-radiological material storage.
As a result of increases in radioactive waste, the South Storage Building is now needed to store low level radioactive waste (LLW).
This LLW, in the form of dry active waste (DAW) will be packaged in metal boxes and labelled, ready for future shipment to burial sites.
The DAW metal shipping boxes will be stored off the floor to prevent water damage. The metal shipping boxes are strong, tight containers designed to prevent any leakage of radioactive material during transportation.
Incidental water contact will not result in the spread of contamination.
Radioactive waste will not be processed in the Rev 0 A Ch(S 6VER GY COMPANY 8-1-91
ODCM - APPENDIX B REFERENCE 5 To BHolian From TPNeal Jft CONSUMERS POWER COMPANY Date October 23, 1990 Subject PALISADES PLANT-10CFR20.302 SOIL SUBMITTAL cc RLSmedley KMHaas TPN90*028 The following samples were obtained from sectors H-9 and J-9 on October 23, 1990.
Sample Location H9-1 H9-2 H9-3 H9-4 H9-5 J9-1 J9-2 J9-3 J9-4 J9-5 uci/gm 3.92E-6 3.70E-6 1.26E-6
<MDA
<MDA 1.90E-6
<MDA
<MDA 2.28E-6 5.86E-6 Sector H-9 was the highest reading in 1988 at 4.4E-05 uCi/gm and J-9, 2.1E-05.
Both areas are now showing a factor of 10 drop in activity.
Sector H-li could not be resampled because of equipment stored in this location. The data indicates direct dose would be less than 2E-03 mR/hr. Occupancy in this area should not exceed 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s/week or 100 weeks/year, which is less than 1 mR/year. Sample and analysis by MAWillers and GStama, review by TPNeal.
Rev 0 8-1-91
South Radwaste Building and the building will be maintained as a-normally clean (radiologically) area.
The current Palisades Radwaste Storage Plan requires low dose-rate DAW boxes to be placed adjacent to the walls of the South Radwaste Building to limit dose rates outside the building. All DAW boxes and the storage building will be inspected quarterly in accordance with Palisades Health Physics Procedure HP 6.27.
This procedure incorporates the storage and inventory guidelines contained in NRC Information Notice No. 90-09, "Extended Interim Storage of Low Level Radioactive Waste by Fuel Cycle and Material Licensee".
The same radiological and security controls currently in force at the East Radwaste Building will apply at the South Radwaste Building. The South Radwaste Building is surrounded by a locked fence and all building access doors will normally be locked, with keys controlled by Radiation Safety Department.
All access to the building will be controlled through the Radiation Safety Office and the Palisades RWP/Dosimetry System.. Building status sheets will be updated on a monthly basis or whenever radiological conditions change.
Any areas outside the building reaching 5mr/hr or greater shall be posted in accordance with current HP Procedures.
Since the South Radwaste Building will be used for the storage of low level radioactive waste and not for radioactive waste processing, it is believed that the public health and safety will not be adversely affected.
It is Palisades' intent to continue to use the South Radwaste Building to store low level radioactive waste (LLW) until such time when radwaste generators in Michigan are again allowed to ship radioactive waste to the burial sites.
Upon resumption of shipping to the burial sites the South Radwaste Building will be emptied, surveyed and returned to the plant for non-radiological material storage.
Gerald B Slade General Manager CC Administrator, Region III, USNRC Resident Inspector, Palisades Rev 0 8-1-91
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APPENDIX B REFERENCE 7 UNITED STATES 9
NUCLEAR REGULATORY COM TPNeal, Pal 5
0 9
WASHINGTON, D.C. 2056 041 June 7, 1991 Docket No. 50-255 JUN12t9re Mr. Gerald B. Slade Plant General Manager Palisades Plant Consumers Power Company 27780 Blue Star Memorial Highway Covert, Michigan 49043
Dear Mr. Slade:
SUBJECT:
REQUEST UNDER 10 CFR 20.302 TO RETAIN CONTAMINATED SOIL ONSITE AT PALISADES PLANT (TAC NO. 67408)
By letters dated November 12, 1987, and January 25, 1988, (Reference 1 of the enclosed Safety Evaluation (SE)), Consumers Power Company submitted a request pursuant to 10 CFR 20.302(a) for the disposal of contaminated soil onsite at the Palisades Plant. We have completed our review of the request and find your procedures (with commitments as documented in Reference 1) to-be acceptable.
This approa a
at tdpf th os.d Safety Evalyaton.atp Manual (ODCRI as an Append x Also, future mod1tic ri fIEe*E com~ft1ents shall be reported to the NRC in accordance with the applicable ODCM change protocol. We further find that the radiological -environmental impact of the proposed action meets the staff criteria as reflected in Reference 6 of the enclosed Safety Evaluation.
Sincerely, Brian Holian, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
As stated Rev o 8-1-91
Mr. Gerald B. Slade Consumers Power Company Palisades Plant cc:
M. I. Miller, Esquire Sidley & Austin 54th Floor One First National Plaza Chicago, Illinois 60603 Mr. Thomas A. McNish, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health P.O. Box 30035 Lansing, Michigan 48909 Gerald Charnoff, P.C.
Shaw, Pittman, Potts &
Trowbridge 2300 N. Street, N.W.
Washington, D.C.
20037 Mr. David L. Brannen Vice President Palisades Generating Company c/o Bechtel Power Corporation 15740 Shady Grove Road Gaithersburg, Maryland 20877 Jerry Sarno Township Supervisor Covert Township 36197 M-140 Highway Covert, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Roy W. Jones Manager, Strategic Program Development Westinghouse Electric Corporation 4350 Northern Pike Monroeville, Pennsylvania 15146 Mr. Patrick M. Donnelly Director, Safety and Licensing Palisades Plant 27780 Blue Star Memorial Hwy.
Covert, Michigan 49043 Resident Inspector c/o U.S. Nuclear Regulatory Commission Palisades Plant 27782 Blue Star Memorial Hwy.
Covert, Michigan 49043 Rev 0 8-1-91
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 2055S
- 0 SAFETY EVALUATION RELATED TO THE PALISADES NUCLEAR PLANT RETENTION OF CONTAMINATED SOIL ONSITE INTRODUCTION In reference (1), Consumers Power Company (CPCo) requested approval pursuant to Section 20.302 of Title 10 of the Code of Federal Regulations (CFR) for the disposal of licensed material not previously considered by the NRC in the Palisades Final Environmental Statement (FES), dated June 1972. The petition submitted contains a detailed description of the licensed material (i.e.,
contaminated soil) subject to this 10 CFR 20.302 request. The 6,000 cubic feet of onsite contaminated soil contains a total radionuclide inventory of 5.1 mCi, based on radioactive material that was deposited in the soil due to the flooding of the South Radwaste Building. The contaminated area is located inside the security fences, and is on company controlled land. This area (South Radwaste Area) is fenced in, within the plant's south security fences.
Thus it is inaccessible to the public (see Figures 1 and 2).
In the submittals (References 1-5), the licensee addressed specific information requested in accordance with 10 CFR 20.302(a), provided a detailed description of the licensed material, thoroughly analyzed and evaluated the environmental effects relative to retention of the contaminated soil onsite, and committed to follow specific procedures to minimize the risk of unexpected exposures.
Although the environmental impact of the proposed action is well within the dose criteria contained in the Commission's Below Regulatory Concern (BRC) Policy Statement, dated July 3, 1990, the licensee has not requested, and the staff has not considered, the actions describe& herein to be exempt from NRC regulation.
CPCo plans to dispose of the 6,000 cubic feet of contaminated soil onsite pursuant to 10 CFR'20.302. The area, known as the South Radwaste Area, has been contaminated by several cooling tower overflows (three times in an eight-year period), and has subsequently been redistributed by heavy rain showers.
The cooling tower overflows were caused by instrument failures that opened the cooling tower bypass valve during normal operation. This valve is now electri-cally isolated during cooling tower operation. The licensee conducted a soil survey because the South Radwaste Building was in the main path of the water overflows from the cooling tower.
Survey results indicated that radioactive material was deposited in the soil. Although the majority of the radioactive material has been packaged as radwaste and will be subsequently shipped offsite (16 boxes each having a volume of 98 cubic feet, containing 85X of the estimated activity), a large volume of low level contaminated-soil is contained in the fenced area described as the South Radwaste Area.
Rev 0 8-1-91 The specific area contaminated is noted as Area B on the survey grid map (see Figure 2).
The total activity of this area (5.1 mCi) is based on 6,000 cubic feet of soil contaminated with the spoils from the South Radwaste Building.
Table 1 lists the principal nuclides identified in the contaminated soil.
The activity in this table is based on measurements in 1987; see data from a recent submittal (Reference 5) shows that activity concentrations in the contaminated area have decreased by approximately 10 percent. The radionuclide half-lives, which are dominated by 30-year Cs-137, meet the staff's 10 CFR 20.302 guidelines (Reference 6, which applies to radionuclides with half-lives less than 35 years).
Table 1 Average Nuclide Concentration (pCi/g)
Total Activity (mCi)
Co-60 0.05 0.079 Cs-137 30 5.0 Total 379 RADIOLOGICAL IMPACTS The licensee has evaluated the following potential exposure pathways to members of the general public from the radionuclides in the contaminated soil:
(1) external exposure caused by direct radiation from radionuclides in the soil; and (2) internal exposure from inhalation of resuspended radionuclides. The staff has reviewed the licensee's calculational methods and assumptions and finds that they are consistent with NRC Regulatory Guide 1.'109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977.
The staff finds the assessment methodology acceptable. The dose assessments are based on the following:
- 1.
5.1 mCi of contaminated soil distributed over 12,000 square foot planar source having a thickness of 0.5 feet (6000 cubic feet source volume).
- 2.
Direct radiation exposure of 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year.
- 3.
Inhalation exposure based on 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year is minimized due to six-inch layer of gravel (which inhibits wind erosion.)
- 4.
Groundwater not considered because there are no domestic wells in the area down-gradient from the plant.
Doses calculated from these pathways are shown in Table 2. The total dose of 0.85 mrem per year is within the staff's guideline of 1 mrem per year (Reference 6).
Rev 0 8-1-91 Table 2 Whole Body Dose Received by Maximally Exposed Individual Pathway (mrem/year)
Groundshine 0.85 Inhalation 0.00081 Groundwater Ingestion 0.0 TOTAL The above doses are a small fraction of the 300 mrem received annually by members of the general public in the United States and Canada from sources of natural background radiation (Reference 7).
Based on our review of the proposed disposal of contaminated soil onsite, we conclude that:
(1) The radioactive material will be disposed in a manner such that it is unlikely that the material will be recycled; (2) Doses to the total whole body and any body organ of a maximally exposed individual (a member of the general public or a non-occupationally exposed member) from the probable pathways of exposure to the disposed material will be less than 1 mrem per year; (3) Doses to the total whole body and any body organ of an inadvertent intruder from the probable pathways of exposure will be less than 5 mrem per year since the burial location is on company-controlled land; (4) The radiation exposures to the nuclear station workers are small compared to the routine occupational exposures at the Palisades Plant; (5) The possible radiation risks to members of the general public as a result of such disposal are well below regulatory limits and small in comparison to the doses they receive each year from natural background radiation.
The licensee's procedures and commitments as documented in the submittal are acceptable, provided that they are permanently incorporated into the.licensee Offsite Dose Calculation Manual (ODCM) as an Appendix, and that future modifications be reported to NRC in accordance with the applicable ODCM change protocol.
Rev 0 8-1-91 REFERENCES (1) CPCo's letters, T. C. Bordine to NRC Document Control Desk, November 12, 1987 and January 25, 1988.
(2) Memorandum from L. J. Cunningham, DREP to T. R. Quay, T. V. Wambach, "Request for Additional Information (RAI)," March 15, 1988, April 7, 1989, and January 12, 1990.
(3) CPCo's supplement to Reference (1), J. L. Kuemin to NRC Document Control Desk, June 27, 1988.
(4) CPCo's supplement to References (1, 2), G. B. Slade to NRC Document Control Desk, August 31, 1990.
(5) CPCo's letter, T. P. Neal to B. Holian, October 13, 1990.
(6) E. F. Branagan, Jr. and F. J. Congel, "Disposal of Slightly Contaminated Radioactive Wastes from Nuclear Power Plants," presented at CONF-860203, Health Physics Considerations Decontamination Decommissioning, Knoxville, TN, February, 1986.
(7) National Council on Radiation Protection and Measurements, "Exposure of the Population in the.United States and Canada from Natural Background Radiation," NCRP Report No. 94, Bethesda, MD. December 30, 1987.
Principal Contributor: J. L. Minns Rev 0 8-1-9 1
---, f"%
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-1 LEGEND I
SITE PROPERTY BOUNDARY I EXCLUSION AREA 2
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